ML20138H590

Exam Rept 50-312/OL-85-02 on 851029-31.Exam Results:Six of Eight Reactor Operator Candidates Passed Written & Operating Exam,One Failed Both Exams & One Failed Written Exam.Senior Reactor Operator Candidate Passed

ML20138H590
Person / Time
Site:	Rancho Seco
Issue date:	12/10/1985
From:	Johnston G, Pate R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20138H576	List: ML20138H571 ML20138H590
References
50-312-OL-85-02, 50-312-OL-85-2, NUDOCS 8512170254
Download: ML20138H590 (150)

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s U.-S. NUCLEAR l REGULATORY COMMISSION

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> , -REGION V 3 , . ~ - ,- ,

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Report No'..

...; ; 50-312/0L-85-02 Docket No".' '50-312 ,

Facility Name: Rancho Seco Nuclear Generating Station

! Examinations Administered at: Rancho Seco NGS, Clay Station, California

'- ~from October 29-31, 1985 Chief Examiner: fjr* ' _ kr /a2f/a (f-G. W. Johnfton', OperatoVLicensing Examiner Date 4iigned Approved By: prt/ 6 /2 o/f'[

R. ~J. Pate, Claisf, Reactor Safety Branch Date $igned

. Summary:

Examinations on October'29-31, 1985 Written examinations were administered to eight Reactor Operator candidates and one Senior Operator Upgrade candidate. Six of the Reactor Operator candidates' passed the written and operating examinations, one candidate failed both the operating and written examinations, and one candidate failed the

. written examination. The Senior Reactor' candidate passed the written and operating examinations.

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. 6. ' Exit Meeting:

The Chief ~ Examiner met with.th'e facility representatives denoted in paragraph 3 above to discuss the examinati~on process.and -to indicate those candidates-who were clear passes of the Operating Examination.

~During the meeting the licensee representatives expressed concern with the Reactor Operator written examination. ~ Their concern was over the

' depth of some. questions, they felt that specific questions are not covered in their training program and questioned their appropriateness.

The Chief Examiner. indicated that the examination would undergo several reviews as part of the normal process and the comments that they provided would receive full consideration.

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IREPORTDETAILS 6

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PersonsrExamined: ' ' " [; '.

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l There were 8'c:and,idatesifor OperatorfLicense examinations and one

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. candidate for'a; Senior Operator Up' grade License examination.

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2 .' Examiners: m

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'i ' *G."Johnston,:RV. ,' .' '

.W./Appley,PNL- -

.B. Gore, PNL

• Chief Examiner-

3. Persons-Atte ng..the Exit. Meeting: >

-a 'NRC

--J.-Eckhardt, SRI' G. Johnston, RV '

B. Boger,.0LB:HQ

,B. Gore, PNL

, W. Appley, PNL SMUD F. Thompson,'SMUD,- Training Manager

.T. Hunter, SMUD, Training Coordinator J. Mau, SMUD

, .4. Written Examination and Facilit'y Review:

-Written examinations were administered as'follows:

Eight RO exams - October 29,~1985 One SRO Upgrade exam .0ctober 29, 1985

. At' the conclusion of the exam, the facility staff was given copies- of the Ltwo examinations. They'were instructed by the Chief Examiner to review the exams and provide th'e' examiners with the comments either prior to their departure or within five days after the completion of the O

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examinations. .This is to facilitate resolving their comments prior to-

grading the examinations.

i5. Operating Examinations:

Oral' exams and facility.walkthroughs were conducted October 30-31, 1985.

No general weaknesses were' identified by the examiners. One candidate;

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however, did display a lack of knowledge about the. waste gas header, in particular when~ queried about the method of purging the header. The SRO

. candidate and seven of the eight R0' candidates were identified during the exit as having been clear passes.

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' Resolution of Facility Comments a

Senior Operator Examination ,-

.1.01 Facility' Comment: Question 5.~12b

"This'~Lquestion is vague and difficult to correlate to operations. We have-two'-(2) rates of rod insertion at Rancho Seco, namely,"3" per minute Land.30" per minute. . We operat'e only at 30" per minute." -

- Resolution: .r '

The~e'xaminer was only seeking to. find whether the candidate understood

.the impact of the rate of positive reactivity addition that was occurring #

'when the reactor was brought back to 100% power and began to burn out the .

' Xenon at a high rate. The rate of insertion (i.e.,-bow fast the rods move) is immaterial. -

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-2.0 -Facility Comment: Question 6.01b .

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"The actual interlock looks at plus and minus.2" from nominal level -

.same as high and low level alarm.

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Also, additional interlock of no '86' lockout relay as indicated in the 1 control room by extra brightness in the stop BLPB. and s.t the 6900V.

~Switchgear by a flag."

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Resolution:

, p. c -

, Both comments accepted. -

3.0: Facility Connient: 6.03b  % v

'.'#6 means turbine in something other than ICS Auto; therefore turbine in

' manual or turbine in operator-auto would be. correct."

Resolution:

. - Will accept a response that indicates the turbine is in other than ICS -

= Auto.

4.0- Facility Commengs 6.04b v "A$ceptable response could?also be 'The Tivg of the loop that does not

,. have an activated low flow' switch '"' '

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5.0 Facility Comment: 6.08b

" Answer should be condensate storage tank, plant reservoir or Folsom South Canal or site water supply."

Resolution:

-< In th'e examiner's opinion the li'sted auxiliary systems are required to

, . . ~ ensure thatL the ' A' train.ca'n be made operable. Realize that the

. question.does not refer to just the 'A' DHR system,.but to the entire.

train of ECCS. . , <

6.0 FacildtyComment

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". s"We :normally' refer' to the.SFAS ' valves as modulating valves, not jogging

1. . valves., Also, normal. valves' fail open on loss of-instrument air."

Re'solutiion: - 's Thetermjogging'is-usedpurel)inagenericsense,andisnotusedto characterize the. type of-valves but to describe how the valves are controlled. As to.the failure of the-normal valves, it is true they fail

. open, how else,.in this case would a controlled cooldown be done? As an aside, the candidate was instructed to consider that the main feedwater

. pumps had both tripped.

' 7.0 Facility. Comment:' 7.02b

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-"The ' inadvertent' part of the resp ~onse is not asked for nor does it make any difference to the question asked. The valves stroking fully open with SFAS actuation and no control unless the operator operates the valve

.by taking. manual control at the SFAS panel."

Resolution:

. The examiner qualified the response as. being inadvertent only to make clear that if the actuation was necessitated, the response would have to be qualified because no concern would arise. The response sought'from the candidate is clearly that the valves would stroke fully open, partial credit will be apportioned on the basis of the candidates-assumptions and

. qualifications.

, 8.0 Facility Comment: 7.05a

" Time looks closer to 1 bour, poss'ible miscalculation on the key."

g Resolution:

Examiner agrees and will change key.

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. 9.0 Facility: Comment: r7.11

> "Thie answe' r kep[is incorrect for all' (3) guidelines. The guidelines were#

.put into'the'p'rocedure to give the operator guidance for indication of.

'subcooling margin; inadequate primary to secondary heat transfer and j excessive primary to secondary heat transfer. This provides the operacor

'with direction and ~ guides' him ~to the proper procedure - Steps 15,16 s.nd 1 17 in E.02." l 1

Resolution: 'l 4

The question, in the opinion of the examiner, seeks to elicit from the candidate his understanding of what is trying to be accomplished with the

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guidelines (i.e., what concern is being addressed by the specific

- guidelines) and not with what' guidance is being given. Further, the.

question-clearly asks for concerns and not for specific actions.

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Reactor Operator Examination Key,Chadge.by Region,V Reviewer:i. >

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Up'on[reviewbyLd. Region 1V'examideritwasdeterminedthatthisquestion s; did not 3 have~an.; unambiguous response among the four' possible responses.

f Item (d)'does apply.very,well to:a recirculating steam generator;

, however, for a once'through steam generator responses (a) and (d) do have somel applicability.' lBecause of the. ambiguity for this facility the

- examiner felt it was appropriate to delete this question.

iTh'e' resolution of codment's provided by the facility to Region V contract-g examiners are delineated"in the following pages. '

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OPERATOR LICENSING EXAMINATION, RANCHO SECO 1, OCTOBER 29, 1985 Responses to Facility Conssents:

-Question 1.1 One recognized knowledge area for which a complete operational feel is required is "subcritical multiplication". A sound knowledge of this phenomenon is crucial to safe reactor operations during startups, shutdowns, and fuel manipulations. Lack of operational feel for subcritical multiplication has resulted in inadvertent criticality events. This question is one way of determining if the candidate really understands the effects of subcritical multiplication and how it relates to delayed neutrons. The absolute minimum power level that the reactor would drop to given an infinite bank of negative reactivity being inserted upon reactor trip would be Beta times the original power level. This is simply the definition .of Beta Fraction. The fact that in an operating reactor, power only falls to about 7%

is due to the subcritical multiplication of delayed neutrons above the Beta fraction. Another way to think of this phenomenon is that a prompt drop is simply the rapid decrease of power to the power level resulting from the subcritical multiplication of the source neutrons plus the delayed neutrons.

'If a candidate does not understand this effect, then there is a documentable defficiency in his understanding of what a prompt drop really is.

Furthermore, there is probably some limitation .in the level of operational understanding of subcritical multiplication that could lead to misinterpretation of nuclear indications in other than the normal situation.-

The facility's concerns regarding the candidates use of actual approximations of total negative reactivity were considered sound, and full credit was given

.for answer "a." However, only three out of the eight candidates indicated an understanding of the effect that delayed neutrons play in response to rapid insertions of negative reactivity, and I suspect that there is a generic training deficiency in that area.

Question 1.2 The ' facility's comment is co'nsidered valid, and tne question was deleted.

Question 1.5 One of the key aspects associated with understanding subcritical multiplication is the understanding.of the inverse relationship between the equilibrium count rate and core reactivity. It is important to understand that'this relationship does not break down at criticality; rather, the source.

neutrons are simply multiplied linearly--i.e., count rate increases . linearly.

without bound (until some form of reactivity feedback effect-takes place in the-power range). This is an observable effect in any reactor, and to not' understand it implies a lack of understanding of what criticality really is and what subcritical multiplication really means to the operator at the controls. All too often the candidates' understanding of this phenomenon simply rests upon memorization of " magic" equations thrown at them during the training program rather than a true understanding of what is actually occurring p -

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in their reactor as a result of their actions at the control panel. Achieving criticality is the logical follow-on to subcritical multiplication...nothing mdgIC is happening at criticality.

The facililty comments regarding this question demonstrate a misunderstanding on behalf of the training department of this important concept. Despite this, haif of- the candidates were completely correct in their responses. I suspect that this is an area where the training department should investigate the technical content of their curriculum to ensure that the operators are in fact being given proper understanding of transient, low power, reactor behavior.

Question 1.9

'The facility is mistaken, the answer key is correct in showing that equilibrium is reached at 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> (actually about 41 hours4.74537e-4 days <br />0.0114 hours <br />6.779101e-5 weeks <br />1.56005e-5 months <br />). This is true because of the small ramo rate (2.5%/hr). At this ramp rate xenon is almost in constant equilibrium. However, we agree with the facility that xenon trends on power ramps are not cut and dry, therefore, wide latitude was given for responses to this. question. All candidates demonstrated some understanding of xenon behavior despite the fact that they probably had not considered this . exact situation previously.

Question 1.10 This question gets at the root of understandin9 what heat transfer under phase change.means. There is only one possible correct answer, and that is answer "d". Distractor "a" can be eliminated by considering examples of constant pressure processes where delta."Ts" are proportional...i.e., ANY subcooled heat exchanger. To consider "a" as a correct answer indicates a very severe misunderstanding of the heat transfer process undergoing in the UTSG. Distractor "b" is obviously in error. Distractor "c" is plausible, but rather than being the cause of.the lack of proportionality, it is more of an effect of the lack of proportionality. It is a very poor choice when considered against the correct answer, "d".

The mos't typical graphical rep'resentation of this phenomenon is a T-S diagram like the one shown in the facility reference provided in their response. The crucial point is that a constant temperature heat transfer process implies a change of phase, whereas ~a constant pressure heat transfer process does not imply a change of phase.

This question is considered a very good discriminator of whether or not the candidate truly possesses an operational understanding of the heat transfer process taking place in the steam generators. The facility's concern regarding the non-application of this question is not technically sound.

Water boils in the once through steam generator under the same physical laws that it boils in a teakettle. The. fact that steam superheating (a small heat transfer effect) is taking place in the upper tube area is irrelevant to either the question or the answer.

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3 Three of eight of the candidates gave the correct response. Four of the remaining eight gave "c" as their response. Only one candidate selected distractor "a". Judging from the response of the candidates and the comments received, operational thermodynamics may be a generic training deficiency at this facility.

No action was taken regarding this comment.

Question 1.12

-There has typically been a lot of resistance to this question, and upon pressing the issue, I have found that it is usually due to wide spread

' misconception of what Net Positive Suction Head is. The question requires no calculations; however, it requires an operational feel for Net Positive Suction Head in a situation that has direct operational relevance. In fact, the question could have been posed using the main condenser rather than a tank. This question gets to the root of what saturation is all about, and is

. considered very operationally relevant. Candidates that are unable to reason this problem out, in all probability suffer from severe missunderstanding of net positive suction head and saturation processes.

Judging from the performance of candidates, this is probably an area of generic training program weakness. No action was taken regarding this comment.

Question 1.13 All candidates selected the correct answer. No action was taken regard.ing this comment.

Question 1.14 I disagree with the facility's criticism of the way that the question was worded. Candidates that do not readily understand the significance of this problem do not understand th0t the steam bubble will not readily collapse upon increasing pressure. In fact, at least one facility, the operators lifted primary reliefs trying to collapse a vapor bubble. 'The bubble will superheat upon compression rather than collapsing. All candidates answered this question correctly.

Question 1.15 No action was taken regarding this convaent.

Question 2.4 Comment addresses interpretation of question wording. Although wording could be improved, meaning is adequately clear. (No candidates misinterpreted it, either.)

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Question 2.5 Connent notes question theme and nonspecific indications are good. Grading retains these, deletes requirement for numerical values. An additional response was accepted: unbalanced HP flow indication.

1 Question 2.7 Deleted

-Question 2.9 Comment suggests a third answer. Accepted.

Question 2.11 Comment suggests an additional correct response. Accepted; not required for-full credit.

Question 2.14 Comment requests deletion of requirement for component numbers. Accepted.

Question 2.17 Comment notes alternate answer if auxiliary boilers are operating. Accepted if assumption stated.

-Question 2.18 Comment notes _second acceptable answer if breaker is assumed to be initially in the disconnect position. Accepted f assumption stated.

Question 2.19 Deleted Question 2.20 Comment suggests three additional observable sumptoms. Accepted.

Question 3.1 Connent corrects key due to incorrect training material. Accepted.

Question 3.3 Deleted

5 Question 3.4 Deleted Question 3.5 Connent suggests an additional correct answer. Accep ted.

Question 3.6 Connent. suggests a second set of responses resulting from wording

. interpretation. Accepted for part credit, (0.7 out of 1.0).

Question 3.7 Connent corrects key _ due to system modification'not incorporated in training material. Either answer accepted.

Question 3.15 Deleted Question 3.16 Comment words key (which was copied verbatim from training material) more precisely. Accepted.

Question 3.18 Connent corrects key due to incorrect training material. Accepted.

Question 4.2 The facility's comment was accepted, and credit was given for all answers.

Question 4.6 The facility's comment was accepted, a discussion of " wire drawing" resulted in full credit.

Question 4.14 The original facility reference material was in error. The question was deleted as per facility request.

Question 4.16 I agree with the facility, the question is not properly posed, and was therefore deleted.

1

m-U. S. NUCLEAR REGULATORY COMMISSION V

SENIOR REACTOR OPERATOR EXAMINATION Facility: Rancho seco Reactor Type: Babcock & Wilcor Date Administered:

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10/29/85 Examiner: Gary Johnston ,

Candidate: kest f(J?aferf k n

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INSTRUCTIONS TO CANDIDATE:

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

Staple question sheet on top of the answer sheet. Points for each question are indicated .in parenthesis 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.

Category  % of Candidate's  % of Value Total Score Cat. Value Catecory 25- 25 5. Theory of Nuclear Power Plant Operation, Fluids, and Thermodynamics 25 25

6. Plant Systens Design, C5ntrol and Instrumentation 75 ps 7. Procedures - Normal, Abnormal, Emergency and Radiological Centrol 25 _JJ__ 8. Administrative Procedures, Conditions, and Limitations ,

100 TOTALS Final Grade  %

all work done on this examination is my own; I have neither given 90r receiveo alc.

Cancicate's Signature 9

e EQUATION SHEET-s f = ma y = e/t-U w = mg 2 Cycle efficiency =_N Jork t) a=vt+ at E = mC a-= (vg - y )/t KE = my A = AN A=Ado

• v f=vo +a .

PE = mgh to = 8/t A = In 2/tg = 0.693/tg W ='vaP (ti )(t s)

AE = 931Am h (t +t)

= $CpAT 7 ,7 ,-Ix 9

Q = UAAT 7 ,7 ,-ux Pwr = W It -

f I=I 10

• P=P 10 SUR(t) TVL = 1.3/u P =.P tT e HVL = 0.693/u SUR = 26'.06/T T =.1.44.DT

./A o SCR = S/(1 - Kdf)

SUR = 26 i ' *f r )l CR = S/(1 - K gf )

( \g_j T=[(1*/p)+ {(f o)/1ett'

,,o] 1( eff)1 = CR2 (1 - K,gg)2 l=,"

T = 1*/ (o - py M = 1/(1 - K,ff) = CR /CR g 0 T = (3 - p)/'A o eff 7 M = (1 - K,ff)0 /(1 - K,gg)1 p'= (K /K eff eff -1)/Keff = AKeff SDM = (1 - Keff)/Keff gf)] g* = 1 x 10 seconds 9= [t*/TK,gg] + [E/(1 + A t

~l P = I4V/(3 x 10 0) A,gg = 0.1 seconds I = No -

Id11=Id22 WATER PARAMETERS Idg =Id2 1 gal. = 8.345 lbm 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 3 1 Curie = 3.7 x 10 dps 10 Density = 1 gm/cm i kg = 2.21 lbm Heat of vaiorization = 970 Etu/lbm 3 I hp = 2.54 x 10 -BTU /hr

. Heat of fusica = 144 Btu /lbm 6 l'Mw = 3.41 x 10 Btu /hr; I k 1 Atm.= 14,7 psi = 29.9 in. I'g. 1 Btu = 778 f t-lbf 1 ft. H 2O = 0.4333'lbf/in -1 inch = 2.54 cm F = 9/5 C + 32 C = 5/9 ( F - 32)

S Senior Operators Exam SECT ON 5 Theory of Nuclear Power Plant Operation, _

. Fluids. -ano Thermodynamics 5.01 ( 1 ~. 5 ) .

The- reactor 1s determineo to oe shutdown by 5'/. delta r :n-with indication.in the source range of 30 counts per seconc.

a)- What is the Keft when the reactor is shutdown - (0.5) by-57. delta K/K?

b):What would the count rate be if Keff is (0.5) increased to 0.98?

c)-What would the- count rate be if Keff is (0.5) increased to 0.99?

5.01' Answer:

delta K

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1 -K

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1 = K + 0.05(K) 1 = 1.05(K) a) KEff = 1/1.05 = 0.95 (0.5) 1 - 1C1 CR2 and 0.05 CR2

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1 - K2 CR1 0.02 CR1

= 75 cps (0,5) b) CR2 = (30) (2. 5)

Conversly CR3 = 0.05 30 0.01 c) CR3 = (30) (5) = 150 cps (0.5)

Reference:

Glastone and Sesonske, " Nuclear Engineering".

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5.02 (3.0) e Refer to FIGURE 5.1, a-sketch of a typical (nct necessartly Rancho Seco) au:< i l i ar y feedwater system utilizing two centrifugal pumps of similar characteristics and capacities.

.The plot of Volume Flow Rate versus Pressure shows the

-system with the "A" au::111 ary f etid pump in operation as the initial condition. ,

a) Snow, on the Fi gure provided , how the curve (s) (1.0) will change as the PORV opens and reduces 0TSG

= pressure by 507..

b) Show, on the Figure provided, how the curve (s) (1.0) will change when the discharge valve for the "A" pump is partialy shut. .

c) What effect will an increased temperature of (1.0) the water in the. storage tank have on the Net Positive Suction Head of the pump?

5.02 Answers a)~ attached (1.0) b) attached (1.0) c) Decreases. (1.0)

Reference:

Section III, Part B, of fac111ty manual on Heat Transfer and Fluid Flow.

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Refer to the. FIGURE 5.2 that follows this page. The figure is of " Heat Flux" versus " Temperature Difference between a Wall and the Bulk Fluid" for an operating reactor._ Note that there are two curves repr esented for two pressures ( P1 <

'P2 ). ..

a) What is the principle type of heat transfer is (0.5) occuring at pressure El and 129E4 RTyfHr;ft between the wall and the bulk fluid?

b) What is the principle type of heat transfer is (0.5) occuring at Pressure E2 and 139E3 DTyfHr:ft between the wall and the bulk fluid?

c) Wnat pressure will yield a Igwer fuel (0.5) centerline temperature at 129g3 RTyfbr:11?

d )' What pressure will yield a higher fuel (0.5) centerline temperature at _3 3 9Eh RTy/Ht:ft?

e) What type of heat transfer between the wall and (0.5) the bulk fluid is occuring at Pressure P1 and 33 955 plyfHr:fi?

'f ) Assuming bulk temperature well below (0.b) saturation, Will decreasing the pressure affect the bulk +1uid temperature at a heat f l u:: of 32 QED RI9fbtrit?

5.03 Answer: (0.5) for each.

a) Nucleat boiling.

b) Convection (other terms may be used).

c) Pressure 1.

d) Pressure 1.

e) Radiant heat transfer (steam blanketing).

f) The bulk fluid temperature remains constant (is independent of pressure below saturation.)

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

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Figure'S.3 shows three curves of 1/M plots. Curve 1 shows e condition where the source detector is too far from the source. Curve 2 shows an " ideal" source detector location.

Curve 3 shows a condition where the detector is too close to the source location.

a)'What' effect would'having the detector tgg fat (1 0) from' the source have on the prediction of criticality earl y in a fuel loading?

b) What effect would having the detector tgg cigse (1.Os to the source have on the prediction of criticality early in a f uel loading?

5.04' Answers a) If' the source is tco far away initial (1.0) predictions of criticality will be with too few fuel assemblies loaded.

b) If the source is too close initial predictions (1.0) of criticality will be with too many fuel assemblies loaded.

Reference:

Facility traing maunal, " Reactor Theory", Page RT-8.11 9

4 4

1.0 <

0.8

^

4/

0.6 ..

1/M Detector too far tway from Source.

0.2 0.0 -

0 2 4 6 8 10 Fuel Assemblies Loaded 1.0 <

0.8 Y 2, 0.6 IDEAL 1/M FUEL LOAD i

0.4 0.2 0.0 0 2 4 6 8 10 Fuel Assemblies Loaded 1.0 '

0.8 s, 0.6 Detector too close to Source.

1/M

~

0.4 O 0.2 0 2 4

, 6 8 10 Fuel Assemblies Loaded

• i-

a. Yb

-5.05 (1.0) w a) What will happen to the temperature differential between the fluid inside the tuoes and.the 'shell side of the tubes along the length of the tubes'in a counterflow heat' e::c hanger ?.

~

b) What will happen to the temperature differential between the inside of the tubes and the shell side of the tubes along the length of the tubes in a parallel flow heat e>: chang er ?

5.05 Answers a) remains the same (0.5) b) decreases (0,5)

Reference:

General Electric Morris Training Center,

" Thermodynamics Heat Transfer and Fluid Flow", Pages 8-28 through 8-30.

e e

~

5.06 (1.5) s The enclosed FIGURE 5.4 shows the profile of a typical

-ionization . curve for e gas filled detector. Match .the appropriate region with the indicated zone of the curve.

a) Proportional (Each response (0.25))

b) Ior i:ati on c)- Continuous Discharge d)' Limited Proportional-e) Geiger-Mueller f) Recombination The zones ares- 1, II, III, IV, V,.and VI as described-on FIGURE 5.4.

5.06 Answers- (Each response (0.25))

a) III

_ .b ) ' II c) VI d) IV e) V f) I

Reference:

LSystems Training Manual, " Nuclear Instrumentation",

.Page 34a-16.

4 e

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Curve 2: Radiation event of higher specific ionizatkn.

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5.07 (1.5)

FIGUPE 5. 5 shows a mariomet er connected to_a venturl, for tne following statements describe:

a)~ What side of the manometer will rise when a_ (0.5) fluid is flowing in the direction Indicated?

b) What rel ationship (i . e. equation) can be used (0,5) to determine the flow rate of the fluid in a pipe with a venturl or other device such as an orifice?

c) What other instrumentation is normally (0.5) provided in the plant to perform the function of the manometer?

5.07 Answer:

a) The leg connected to the restriction. . (0.5) b )" The Bernoulli equation (Flow rate is propor- (0.5) tional to the square root of the differential pressure.)

c) . Typically a Bourdon type transmitter (D/P (0.5) cell).

References . General Physics " heat Transfer ~ Thermodynamics and Fluid Flow Fundamentals".

k 0

. O e

7

e Venfuri -

Flem E

. 7 ~

Up str e o.vn Restricfion Leg Leg V

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5. 08 - ( 1. 5)

What - are. the~three major intrinste sources for- neutrons found within the core (not tne source ' assembly) ?

~

5.08 Answer:

a) .Spontaneacus fission. (0.04

2. 1 1 b) Photo-neutron.. (gamma +. H --> 'H+n ) (0.5) 1 1 0 18 21 1

c) Alpha-netron.

~

(; alpha + 0 --> Ne + n) (0.5) 8 10 0 Willl accept the formulae.

Reference:

Facility manual , " Reactor Theory", Page RT.5.5.

=

r 9

l e

f i

L

5.04 (1.0)'

Which statement is the r30st accurate?

The , Rhodium Self Powered Neutron Detectorn (SPND's)-are not

.used -for measurement of transient changes i r, neutron flux because:

a) The beta induced current is not proportional to +

the neutron flux.

b) The-~ charge built up on the center electrode is not' proportional to the neutron flux.

c) The time constant of the response of the detector is related to the half-lives of the Rhodium decay.

'd) The background gamma radiation causes an excess of emmited electrons that masks, during transients, the actual current change.

5.09 Answer:

c) is the most accurate. (1.0)

Reference:

Systems Training Manual, " Nuclear Instrumentation" Pages 34b-8 to 34b-13.

t 9

r

t-5.10~ (1.0) 4 For the following -indicate which is the mggt accurate statement:

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

1. Removing the latent heat of vaporization at ..

a constant temperature to allow the steam -

to condense.

2. Provi di ng 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 the saturation temperature.

b). Condensate depression ist (0.5)

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

2. Used to maintain constant temperature profile on the condenser tubesheets.

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

4. Used to maintain adequate Net Positive

. Suction Head for the main feedwater pumps.

5.10 Answers a) 1. (0.5) b) 3. (0.5)

Reference:

General Physics " Heat Transfer and Fluid .

Flow" Pages 10 I

l.

e.

5.11 (4.0)

• The Moderator . Temperature Coefficient (MTC) and Fuel Temperature Coefficient (FTC) are the major reactivity coefficients.

a) What effect would NTC have on core reactivity- (1.0) during a load change f rom 0 to 15% power?

b) What effect would MTC have on core reactivity ( 1 ". 0 )

during a load change from 15 to 100% power?

c) What effect would FTC have on core reactivity (1.0)

'during a load change from 0 to 15% power?

d) What effect would FTC have on core reactivity (1.0)

-during a load change from 15 to 100% power?

5.12 Answers a) Tavg ~

is increasing so MTC is adding negative (1.0)

' reactivity.

b) Above 15%-Tavg is constant, therefore there is (1.0) no appreciable contribution (delta T across core does go up).

c) Fuel centerline temperature is increasing over (1.0)

the entire range of O to 100% power. Therefore the FTC will be adding negative reactivity all the way up in power.

d) Same as c). (1.0)

Reference:

Reactor Theory handbook pages 12.2 to 12.9, and 13.2 to 13.0.

9 e

11

4 g, ;

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5.12 (4.0) ,

) _

Figure 5.7 is' a sketch of Reactor Power versus Time in hours. At- t=0 hours . reactor startup from Xenon free conditions to 100% power occurs. At t=50 hours a reactor trip occurs followed by a reactor st'artup to 100% power at, t=65 hours.

in the (1;6) a) Sketch the Xenon reactivity response ,

core from this power transient. (Inpicate approximate magnitude L 7. ade)ta k/k1 and duration.of each part of the' transient.)

b) What. is the time in hours that the manimum (0.5) rate of rod insertion will have to occur in order to overcome the Xenon transt'ent. (Assume constant Tavg and no boration or dilution.) ,

c) What are the productio.1 and removal mechanisms (1.0) ,'

for Xenon? -

d) What are the production and removal mechani sms. '

41.0) for Samarium? < =

5.12 Answer: '

,a) (Attached) ( 0. 6 % '

b).65-to 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> f , h. (0.5)

~ '

c) Xenon:

Production - lodine decay (0.25)

Direct yield f rom fi ssion (0.25)

Removal - Burnout by neutron', - (0.25)

Decay _

(0.25)-

l d) Samarium:

Production - Promethium decay (0.33) -

, Direct yield from fission (0.33) _

Removal - Burnout by neutrons (0.33)

Reference:

Procedure B.6 " Reactivity Balance Calculation",

page B.6-23. " Westinghouse Training Noteu" ,

' Introduction to PWR Control' and<' Fission ,

s .,

Product Poi soning ' . ,

END OF SECTION 5 O

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i "s. SECTION 6 Plant Systems Design, g

Control and Instrumentation ,

6.01 (3.0)

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( The Reactor Coolant Pumps can be started by pushing contr ol switch "RC Pump P-210C ' Start *". However some permissives

, must be met,before the pump will actually start.

(

the permissives is an interlock that (1.0)

) a) One of

, prevents the start of a fourth Reactor Coolant

/ Pump if the RCS Cold Leg temperature is less than 500 deg. 'F with three pumps running. What

> is the/ primary reason for this particular a interlock?

?

• b )' There are five other interlocks involved with (2.0) the start circuits of Reactor Coolant Pumps.

What are igut of those permissive interlocks (include setpoints)?

6.01 Answer

_ 'a) The interlock Hi s primarily to prevent core (1.0/

up1.tft at. low temperatures (<500 deg. F).

b) (Any four 0.5 points each) (2.0)

-'+ A

' / ?g 1. Reactor power <30%.

I 2. 011 lift pressure >1750 psig (+ or - 50 i f' '

psig).

,Y .

) y. Bearing oil sumps full. 'l ,

4. Seal injection flow >3 gpm.

f 5.' Component Cooling Water flow >307 gpm.

( ?)n y ' ' 1:

References- Chapter 2, Systems Training Manual,

,. ;c, y ~ d ?. d .-

Reactor Coolant

, System, pages 2-75 to 2-78.

/ ~

Iv ,

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? .

• - 6.02 (2.0)

Match the following core components with their functions.

a) Plenum Assembly -

(0.5) b) Core Barrel .. (0.5)

Assembly c) Lower Grid (0.5)

Assembly d) Core Support (0.5)

Shield Assembly

-FUNCTIONS:

1. Reactor vent valves assemblies are attached to this assembly.

2. Directs reactor. coolant flow f rom the fuel assemblies to the outlet nozzles.

3. Unif ormly distributes the primary coolant flow entering the reator core.

4. Channels primary coolant through the core and attenuates the neutron flux and gamma rays impingsn.g on the vessel.

6.02 Answer:

(0.5 points each) .

a) 2.

b) 4.

c) 3.

d) ~ 1.

Ref er ence: Chapter 1, Reactor Vessel Internals and Nuclear Fuel, pages la-1-to la-33.

2

p. -

6.403 (4. 0) .

Regarding'the Integrated Control System (ICS):

a) What is meant by "The 1CS is in Track"? (1.v) b) List EDi conditions that would cause the ICS (3.0) to be in " Track". ..

6.03 Answer:

a) Due to some abnormal condition the ICS will use (1.0)

-951921 9fD.eCpted mggggglis as the unit 1939 ggmang signal.

b)

(any six 0.5 points each)

1. The unit is operating under cross-limits.

2. The steam ~ generator / reactor master in manual.

3. Both feedwater loop masters in manual.

4. Di'amond control in manual.

5. Bailey rod control in manual.

6. Turbine control inmanual.[. k' '

)

7. Tripping of both generator breakers.

G. Reactor trip.

Reference Chapter 32, Systems Training Manual, Integrated Control System, pages 32-63 to 32-67.

W 4

r F.-

e.

6.04 (3.0)

Regarding the Integrated Control System:

Three Tavg signals are supplied to the AUTO / MANUAL TEMPERTURE TRANSFER SWITCH (HS-21045).

a) What are these three Tavg signals? --

(1.5) b) What determines which signal will be used in (1.5i AUTO?

6.04 Answers a) 1. Unit Tavg. (0.5)

2. Loop 'A' Tavg. (0.5)

3. Loop 'BTavg. (0.5) b): All four RCP's running - Unit'Tavg. (0.5)

Less than two pumps in a loop - . .(1.0) unaffected Loop Tavg (Candidate may respond with what happens for'.each loop.) j /f [g g .fg' [

Reference:

Chapter 32 rev 1., Systems Training Menual, Integrated Control System, pages 32-37 to

• / [

"'# 51 - ( ,

32-45.

O e

9 m

4

6.05-(2.0)

A n' Reactor Coolant System pressure decreases, what control functions, alarms, or activations occur at the following pressurizer pressure setpoints?

(0.33 points each) a) 2'145 psig -

b) 2135 psig

.c) 2120 psig d) 2055 pstg e) .1975 psig f) 1900 psig 6.05 Answers a) Modulating heaters f ullion (group 1). (0.33) b) Group II heaters on. (0.33) c) , Group III heaters on. (0.33)

-d)-Low pressure alarm. (0.33) e) Low-low alarm. (0.33) f)1 Reactor low pressure trip. (0.33)

Reference Chapter 3, Systems Training Manual, Pressurizer and Pressure Relief Systems, page 3-122.

f-e f

9 e

5

6.06 (2.5)

Regarding the Nuclear Service Bussect a) What two conditions will initiate the Nuclear (1,0)

Service Buss ' unloading / loading scheme other-than a Safety Features Actuation system signal?

b) On what three conditions will an interlock (1.5) prevent a start signal to the Emergency Diesel Generator following a SFAS start signal activation?

6.06 Answers a) Bus.undervoltage. (0.5)

Bus overvoltage. - (0.5) b)

"" -~-=n~

-.( (0.6)

Phase imbalance. d (0.5)

G -. =>utadvu. g4 (0.5)

Reference:

Chapter 45, Systems Training Manual, Vital Electrical Distribution 4160 V and Below, pages 43-49 to 43-57.

O M

S 6

6.07 (2.0)

There are four 120 VAC inverters in the Vital Electricat Distribution System. During troubleshooting of the AA inverter it is inadvertantly deactivated. What effect will this- loss of an inverter have on the following systems"?

(Include any subsequent effects.)

a) Safety Features Actuation System (SFAS) Analgg --

(1.0) channel 'A'.

b)-Safety Features Actuation System (SFAS) Digital (1.0) channel 'A'.

6.07 Answers a) Loss of the 1A inverter will de-energize the SFAS analog channel j93Dl. This results in a trip signal being sent to the (all)- SFAS digital channels 1QtSt. (No actuation will occur due to .the 2 out of 3 logic required however only one signal from the remaining analog channels would be needed to produce an SFAS actuation).

b) The loss of the 1A inverter will de-energize the SFAS digital channel 'A' j9351 The equ pment. controlled by this channel will not res ond to an automatic actuation signal for SFAS J92_51 References Chapter 43, Systems Training Manual, Vital' Electrical Distribution 4160 and Below, pages 43-84 and 43-85.

E

=

0 7

e' 6.08 (3.5)

Concerning the Auxiliary Feedwater System:

a) What. three signals will cause an automatic _ (1.5) start of the Auxiliary Feedwater System?

b) What are the two sources of supply of water to " (1.0)

-the Aux 212 ary Feedwater System?

c) What other two systems may be supplied water (l.Os f rom the Auxiliar y Feedwater System?

6.08 Answer 3 a) Low Main Feedwater epder pres.ure. (0.5)

SFAS signalSor low e D6f4 wis d%

Loss of all-four Reactor Coolant Pumps.

(0.5)

(0.5) l

.b) Condensate storage tank. (0.5)

'Plantresevoirf$4 ,,g[ g,4, (0.5) c) Main condensate through main condenser (0.5) recirculation path.

Nuclear Service Cooling water through (0.5) dedicated Itne.

Reference:

Chapter 29, Auxiliary Feedwater System, pages 29-37 through 29-43.

l l

i O

e 8

. 6.09 (2.0)

The Emergency Core Cooling System (ECCS) requires support systems to ensure that the ECCS will perform its design function, _

a) If train 'B' Decay Heat Removal pump has been ,.

(2.0) declared inoperable, what systems would have to .

be verified to be availiable for train 'A' to be considered operable per the Technical Specifications (provide only four of the six)?

6.09 Answer:

Any four (0.5) points each.

a) 1. Vital Electrical Distribution.

2. Non-vital Electrical Distribution.

3. Nuclear Service Raw. Water.

4. Nuclear Service Cooling Water.

5. Safety Features Actuation Systems.

6. Nitrogen System.-

Reference:

Chapter 27, Systems Training Manual, Emergency Core Cooling Systems, page 27-94.

O 9

e 9

r .

6.10 (1.0)

The Absolute' Position; Indication system provides indication of the CR4 s.

a) How much of a deviation from group average i s_ (0.5) required' to light an amber fault. light on the rod position indication (PI) pariel ?

b) What would the group average average be if a single CRA dropped to the bottom of the core from a group fully out position (Assume an 8 rod. group)?

6.10 Answer:

a) 7 inches (5%). (0.5) b) calculated 8(100)

---

= 100% Group Avg.

8 7(100)

---

= B7.5% Group Avg. with (0.5) 8 Dropped CRA.

Reference:

Systems Training Manual, Control Rod Drive Mechanisms and Control System, page 37a-19.

END OF SECTION 6' S

e e

10

Senior. Reactor Operators E>: amination SECTION 7 Procedures - Normal, Abnormal, Emergency and Radiological Control -

7.01 (1.0)

Procedure A.15 ' Makeup, Purification and Letdown System',

contains. a caution under section 5.4.5 "Boration During RCS Cool down" that addresses makeup with DI water tor contraction during an RCS cooldown, a) What two methods are used to assure that the (1.0) positive reactivity inserted.from the makeup using DI water is not a potential concern?

7.01 Answer:

a) Control rod safety-group 1 is withdrawn per (DP (0.5)

B.4 and T.S. 3.1.3.5.)

And one RCP or one DHR pump is. circulating (0.5) reactor coolant. (Will accept one pump for full credit).

Reference:

Procedure A.15 page 20.

O G

1

7.02 (2.0)

Regarding the Auxiliary Feedwater System:

The ' plant is being brought down after a long period of operations. During the later stages of lining up to cooldown to Cold Shutdown instrument air is inadvertantly isolated to the flow control valves FV-20527 and FV-20528, and cannot be returned to service.

a) Can a couldown proceed using only the Safety (1.0)

Features bypass valves SFV-20577 and SFV-20578

? Explain.

b) What concern arises if a Safety Feature (1.0)

Actuation System (SFAS) signal occurs while the Auxiliary Feedwater System is lined up through the the two Safety Features bypass valves SFV-20577 and SFV-2057B?

7.'02 Answers a) Yes. The control of flow to the OTSG's can be controlled through the valves described by jogging them to position to control the flow rate'jl 2 Ol.

b) The primary concern is f rom an inadvgrtant

-(and not from a necessitated) actuation of

'SFAS. This could cause the valves to stroke fully open (1.0).

Reference:

Procedure A.51, Auxiliary Feedwater System, page A.51-12.

G t

2

7.03 (2.0)-

Regarding procedure B.1 " Pre-Startup Checks" . sect i ons 3.4 and~4.5:

contain certain (2.0) j a) Both .of these . sections l requirements that are to be verified, by the-Plant Superintendednt or his designated representative, to be- resolved that were -- ,

generated or performed during during shutdown.

What igtj t items must be so verified?

7.03 Answers a) 1. Non-conformance Reports (NCRs). (0.5)

2. Occurance Description Reports (ODRu). (0.5)

3. . Seri alized Work hequests. (0.5)

4. Surveillance Procedures. (0.5)

Reference:

Procedure B.1 " Pre-Startup Checks", page B.1-2.

e 9

7.04 (2.5) 3

'/.04 (2.5)-

Regarding' procedure C.10 " Loss of Steam Generator Feed":

a )' What.is the definition of a 'ory' OTSG? _

(0.5) b) How is feed initiated to a dry OlSG? (0.5) c)'What three 5cnditions would require a Reactor ( l '. 5 )

trip'to be manually initiated per the procedure C.10 " Loss of Steam Generator Feed"?

7.04 Answer:

a) Less. than or equal to 8 inches indicated (0,5)

Startup level.

ts ) Through the auxiliary feed no::les. (0.5) c) 1. Feedwater lost to both OTSGs. (0.5)

2. A confirmed low level in both OTSGs e::ists. (0.5)

(> 22% Rx power with both Operate range indications <10% and both OTSGs on low level l iini ts. )

3. Either OTSG has gone dry. (0.5)

Reference:

Procedure C.10 " Loss of Steam Generator Feea",

'page -C. 10-1 .

O 9

6 4

1

7.05 (2.0)

Regarding Procedure B.3 " Normal Operation":

The plant has been operating for a long period of time (175 EFPD) and for the last month has been at 100% power.

.a) It has been determined that the plant must come -- (1.0)

~

down to 30% power.for some critical maintenance that requires the plant to be brought down to that power level quickly. Using the enclosure 10.6 to B.3.(provided), how fast may tne plant be brought down in power in nours (show your work on the diagrams)?

~

b) Upon reaching 35% power it is determined that (1.0) two of the Circulating Water pumps are going to be shutdown. What considerations must be taken into account when the plant is operating on two pumps?

7.05 Answer:

a) Between 30 and 45 minutes (approx. 37 to 38 (1.0) min.) (0.5 points for each indication yon the diagrams, see enclosure 10.6)- , ,

t

.b ) Be aware of condenser vacuum. (0.5)

Two CW pumps in one loop lacks out turbine (0.5) byp' ass valves.

Reference:

Frocedure B.3 " Normal Oper ati ons" , pages B.3-12 and-E 3-13.

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7.06-(2.0)

Regarding Procedure C.6 "CRD Mal f unc ti on":

a) What are .igur of the five. defined condit1'ons (2.0) that would cause a control rod to be ' declared _

uto be Inoperable?

"7.06 Answers -

a) Any.four of the-five (O.5) each:

1. Excessive rod drop time.

2. 9 inches out of group average.

3. Improperly programmed'.

4. Can't be exercised.

5. Less than 2 position indications available.

c

Reference:

Pro'edure C.6 "CRD Malfunction", page C.6-1.

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6

-7.07 (3.5)

While you are standing watch.as the Shift Supervisor an Auxiliary Operator calls in to announce that he sees a fire in ,the Diesel. Generator room. You activate the Emergency Plan per the requirements of AP 520 of the Emergency Plan procedures.

a) As Shift Supervisor what is your function under (O"5) the Emergency-Plan?

b) How long do you fill that function? (0.5) c) You dispatch the Fire Brigade to the scene of (2.5) the fire. Who are the members of the Brigade, (include the number of personnel and their title or job description)?

7.07 Answer:

a) Emergency Coordinator. (0.5) b) Until properly relieved.' (0.5) c); 1. Senior Control Room Operator - Fire Brigade (0.5)

Leader.

2. Auxilliary Operator. (0.5)

3. Equipment Attendant / Power Plant Helper. (0,5)

4. Security. Officer. (0.5) 5.' Chemistry-Health Physics Technician. (0.5)

Reference:

' Emergency Plan Sections AP 520 " Fire" and'AP 501- "Rcognition and Classification of Emergency" W

9 7

7.08 (2.5)

You are touring in a' controlled area in the plant and you drop-yourdosimeter, you read tne dostmeter and it is reading

.offstale. _

a) What . is required of you as far as i mr. red i at e (1.0) actions? .

b) What will have'to be done before you will be (1.5) allowed to re-enter the controlled area?

7.08 Answer:

a) Immediately proceed to the exit. point 1QuGl.

Then Report to Radiation Protection Section IQtgl.

b) An estimate.of your exposure based on surveys of his work. area and route JO2 5). And of other workers in the area JQ351, if no reasonable estimate can be made your exposure is limited to 100 mrem for.the quarter IQtGl.

1R eference: AP 305-3 " Dosimeter Assignment and Use", and AP 517 "Radi ation Exposure".

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e 7.09 (2.0)

The plant is about to be brought critical with a calculated Estimated Critical Position (ECP) of 60% withdrawn on Group 5 rods.

a) It is determi ned during -the -approach to~ (2.0) criticality that.the reactor became critical 1%

delta ~ k/k before the projected ECP. _ At this ,

point, per the procedure, you reinsert the rods to a position of -0.8% delta k/k. What- two factors should you be concerned about when an ECP is missed?

7.09 Answer:

a) The' boron concentration and the calculated ECP (2.0) having-potential errors.

Reference:

Procedure B.2 " Plant Heatup and Startup", page B.2-19.

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7 10 (3.0)

• . The Radiation Protection Manual AP 305-4 " Radiation Work Permits" describes six requirements for when a RWP has to be issued.

a) What are three of those requirements? -

-(3. 0 )

7.10 Answer: --

Any three (1.0) points each:

a)' 1. Work in a~ Radiation Area except for routine walk-through surveillance.

2. . Work in or entry to areas of potential neutron exposure.

3. Work in or entry to contaminated areas or equipment.

4.. Work in or entry to High Radiation Areas, Secured Radiation Areas or Airborne Radioactivity Areas.

S'.- Work' involving opening, cutting or welding on Systems which contain radioactive contamination.

6. piork as deemed necessary by the Chem-Rad Group.

References Procedure AP 305-4 " Radiation Work Permits",

page AP.305-4-1.

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t-7.11 (1.5)

Emergency Procedure E.02 " Vital System Status Veri fication '

has three guidelines concerned with three primary concerns associated with the condition of the plant f ol l owi ng a Reactor trip or forced shutdown.

For each of the f ollowing Guidelines, what primary" concern is addressed by the Guideline?

1. Subcooling Margin. (0.5)

2. Inadequate Primary to Secondary Heat Transfer. (0.S)

3. Excessive Primary to Secondary Heat Transfer. (0.5) 7.11 Answer Guideline 1.:

To ensure adequate core cooling. (0.5)

Guideline 2.:

To ensure natural circulation is occuring. (0.5)

' Guideline 3.

To ensure cooldown is not exceeding DTT (0.5)

Concerns.

(Other answers, as appropriate, will be considered.)

Reference:

Procedure E.02 " Vital System Status Verification",

pages E. 02-8,-9,-10.

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, 7.'12 (1.0)

Regarding Rule 5 "LPI Control" in the Emergency Procedures:

a) Why must action be taken to throttle flow t o_ (0,5) 3000 gpm when.a 'DH High-Flow' occurs?

b) Why must. action be taken if a 'BWST Low Level' (0.5) alarm is received to rack in the Reactor Building sump valve breakers?

7.12 Answer:

a)' Throttle flow from Decay Heat pumps to less (0.51 than (3000 gpm) to prevent runout.

b) Reactor Building sump valve breakers. must be (0.5) racked in to prepare for recirculation phase.

Ref eren.ces . Rul e 5. - LPI Control, Emergency Procedures.

END OF SECTION 7 a

T 12

r._--

SECTION 8 e

Admini strati ve Procedures ,

Conditions, and Limitations 8.01 (2.0)

Technical Specification 3.1.3 specifies the minimum conditions required for criticality.

a) What are. fout of the five conditions (2.0) speci.fically required by T.S. 3.1.3 for criticality?

8.01' Answers a) 1. RCS temperature >525 deg. F. (0.5)

2. RCS temperature >DTT + 10 deg. F. (0.5)

3. If RCS temp. less than above, then Rx must (0.5) be'subcritical by > or = to reactivity inserted by depressurization.

4. Rx must have a bubble and level must be (0.5) between 10 and 316 inches in PZR.

5. Safety rods must be withdrawn and regulating (0.5) rod's within position limits.

Reference:

Tech. Spec. 3.1.3.

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8.02-(3.5) ,

e. Technical Specification 6.2.2 specifies the required staffing of the plant under various, conditions.-

a) What is the etniege shift crew cothplement when (2.5) the plant is in HOT SHUTDOWN ~(Specify job title-and number)?

b) Other than_the shift crew complement there are '

(1:07 two other manning requirements that must be met any time there is f uel in the Reactor..What are those twg requirements? '

8.O2 Answer: / '

a) 1 --SOL Shift Supervisor (0.5) 1 - OL Sr. Control. Room Operator (0.5)

(HOT SHUTDOWN is >1% delta k/k) ,,

~

1 - Au:< . Operator or Equip. Attendant (0.5) 1 - Equip. Attendant or Power Plant Helper (0.5)

_1 - STA 44 (0.5) b) 1 - Person qualified in Rad. Protection. (0.5) 5 - Fire Brigade s (0.5) t

Reference:

Tech. Spec. 6.2.2. ,

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r 8.03 (1.5)'

c Regarding Technical Specification 3.~1.4 Reactor Coolant System Activity.

a) The specification requires that the reactor- (0.5)

- coolant activity not exceed 43/E microcuries per gram whenever the reactor is critical. For ,.

what half lives of nuclides does this apply? -

b) What does the term E refer to? (1.0) 8.03 Answer:

a) > 30 minutes. (0.5) b) E = Average beta and gamma energies per (1.0) disintegration (MeV).

Reference:

Tech. Spec. 3.1.4.

8.

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o Regarding~10 CFR 55:

a) What does the term " Controls" mean in -the (1.0)

- context of.your duties as a licensed operator? -

b) Can an individual other than a licensed .- (1.0) operator. manipulate the " Controls"? Explain.

8.04 Answers a) As a licensed operator this means " apparatus (1.0) and mechanisms- the manipulation of which directly affect the reactivity or power level '

of.the reactor." -s s

b) Yes. (0.5)

.However, only under the direction of a licensed (O.Si-operator.

Reference:

10 CFR 55.4 and 55.9.

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8.05 (3.0)

Regarding Technical Specification 6.13 "High Radiation

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'a ) Specifically, what are the requirements per (2.0)

Technicalh Specification 6.13 for an area of between 100 mrem / hour and 1000 mrem / hour? -'

b) What, if any, additional requirements would be (1.0) required f or an area in excess of 1000 mrem / hour?

8.05 Answers a) 1. A barricade. (0.5)

,j 2. Posted as a High Radiation' Area. (0.5)

3. RWP required for entry. (0,5)

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4. Radiation instrument required for-entry and '(0.5)

• i to provide continuous monitoring.

gs 'b) 1. Door must be locked. ,

(0.5)

2. Key is under control of Shift Supervisor. (0.5)

Rhference: Tech. Spec. 6.13 "High Radiation Area".

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H8.06-(2.0)

After having operated for a full cycle the plant is placed in cold shutdown. During the conduct of Refueling interval surveillance it is discovered that both Decay Heat Removal pumps 'would not have performed their safety function on an SFAS actuation signal. ..

a) What two reports are required to be made to the ( 2.' O )

NRC7 8.06 Answers a ) 1 ~. 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report (non-emergency) (1.0)

2. LER (unanalyzed condition) (1.0)

Reference:

10 CFR'50.72 and 50.73.

~

6

8.07 '(3.0)

• 'Regarding AP .501 " Recognition and Classification of Emergency" has five duties that cannot be delegated by 'the Emergency Coordinator.

a) What are thrge of those duties? (3.0) 8.07 Answer .

^Any 3 (1.0) points each

.a) 1. Decision to notify'offstte agencies.

2. Making of Prot. Action recommendations to offsite emergency management agencies.

3. Classification of emergency.

4. Determining:the neccessity for assembling or evacuating personnel.

5. Authorization to exceed rad. exposure limits.

Ref erer;ce: AP 501 " Recognition and Classification of Emergency" 9

O 0

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7

a 8.08 (2.0)

An event' has occured that has caused power to be lost onsite during an emergency that has been classified as an ALERT.

You are informed, as Emergency Coordinator,' that the Security Computer is inoperable, and the Plant' Assembly Point survey instrument ind2 cates 10 mrem / hour. ,

a) What TWO actions, must _you as Emergency. (2.u)

' Coordinator, take to ensure plant personnel are safe and accounted for?

8.08 Answers a) 1. Proceed to evacute the site. (1.0)

2. Initiate a Manual Accountability Search. (1.0)

Reference AP 513 " Personnel Accountability".

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8.09-(2.5)

Regarding AP 2- " Review, Approval ~ and Maintenance of Procedures":

a) What is the definition of " Intent" as regards " (1.0) making a temporary change.to a procedure?

b) A change. in the intent of a procedure is one (1.0) item that will not allow a temporary change to a procedure. What are the other two requirements that must be met if a temporary change to a procedure is to be made?

c).Who determines how long may the temporary (0.5) change remain in ef f ect?

8.09. Answer:

a) Intent is defined as obtaining the requirements (1.0) of the acceptance criteria without exceeding the Limits and Precautions.

b) 1. Must be approved by a S.S. and S.C.O. or_ (0.5)

Plant engineer.

2. ,Must be forwarded to PRC and Plant Super. (0.5) for review and appoval within 7 days.

. c) FRC. (0.5)

Reference:

AP 2,." Review, Approval and Maintenance of Procedures" 9

0 9

.o.

1

.0 8.~ 10 (2.0)

Each High Reactcr Coolant Pressure channel is required by Technical Specifications to undergo a' channel check on a shiftly basis (at least once every 12 hours).

Some extensions of the basic interval are allowed by the Technical Specifications.

Records show that this was done on:

October 27 at.OOOO

. October 27 at 1500 October 28 at 0000 October 28 at 1400 October 29 at 0500 a) What is the maximum allowable interval between (1.0)

- channel check surveillances7 b) When- is the next channel check surveillance (1.0) due?

8.10 Answers a) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> + 25% = 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> (1.0)

-b) October 29 at 1500 (1.0) 0500 + 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> is 2000 but man interval for 3 surveillances is 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> + 257. of 12 which is 39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br /> total. From 10/28 9 0000 to 10/29 9 OOOO-is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> only 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> of.the 39 remain.

, References Technical Specifications 1.9 " Time Periods" e

10

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t 8.11 (1.5)

= Technical Specification 16 . 7 "Saf ety Limit. Violation" describes _ the actions that shall be taken'in the . event' a

~ Safety Limit is violated. -

.a) ~ What are three of those actions?

8.11 Answers a) Any 3 (0.5) each.

1. Plant is shutdown per. (10 CFR 50.36).

2. The violation is reported to (the Plant Superintendent, Manager of Nuclear Ops, Chairman of. the MSRC and) the Commission immediately.

3. A Safety Limit Violation Report prepared.

4. The report is submitted to the NRC, (MSRC Manager. of Nuclear . Ops. and Plant superintendent) within ten days.

Reference:

Tech. Spec. 6.7 " Safety Limit Violation" END OF SECTION 8 END OF EXAMINATION e

6 11

p  ;

MeAUe -

bei c-U. S. NUCLEAR REGULATORY COM41SS10N REAC10R OPERATOR LICENSE EXAMINATION j

Facility: Rancho Seco 1 Reactor Type: PWR Date Administered: October 29, 1985 Examiner: J. C. Huenefeld/B. F. Gore l

Candidate: ANSWER KEY i

INS 1 RUCTIONS 10 CANDIDATE:

Use ' separate paper for the answers. Write answers on Staple question sheet on top of the answer sheets. Points forone side only.

each question are indicated in parentheses af ter 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 af ter the examination starts.

Category  % of Candtdate's  % of Value Total Score Cat. Value Ca tegory I.2g 25. 1

1. Principles of Nuclear Power Plant Operation, lhermodynamics, A4.9 Heat Transfer and Fluid Flow MAS N 2. Plant Design Including Safety.

and Emergency' Systems

_ d AA

3. Instruments and Controls 2543.g _ 25.i

4. Procedures: Normal, Abnormal.

Emergency, and Radfological -

Control l

1ptl 9).[ TOTALS Final Grade  %

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

UiiidlTate's 51gnature b

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1 Rancho Seco October 29, 1985 1.0 PRINCIPLES OF NUCLEAR POWER PLANT -

( b QPERATION. THERMODYNAMICS. HEAT ,

TRANSFER AND FLUID FLOW Points Available OUESTION 1.1

a. Idhat is the lowest possible neutron power level (i.e., prompt drop) that the reactor could fall to immediately after a trip from 100% full reactivity? power (Select (FP))that inserted maximum negative (1.0) one.

, approximately 7% FP

. approximately 0.7% FP

. approximately 0.07% FP

. approximately 0.007% FP

b. Exclain why you won't see a drop to this lowest possible value upon a reactor trip in your reactor? (1.5)

ANSWER 1.1 a.- (g) b (Beta is approximately 0.007) .-> '7 -

r). 7 E

b. Depending on the shutdown margin, there will be sufficient reactivity in the core following a reactor trip such that subcritical multiplication will amplify the effect of delayed neutrons.

REFERENCE (S1 Rx Theory Manual, pp. RT-8.2 and RT-9.4.

en p .v-[ '+1 fa Il c. y L.$ n% WM 3,Jx np NW &*

f t'e,yp t lt%p \ 1 mina u to ( s t ':  ?= ff g_ g>

-Section 1 Continued on Next Page-

2 Rancho Seco October 29, 1985 Points Vailable 00ESTION .2

% long doe ft take before change in 1 #

. changes in AP c nt 111ng group pos nswffibe appreciated? on ) (1.0)

8. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 3 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> c 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> d 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ANSWER 1,7 d.

REFFDFNCE(S)

% N 1 Operation, 8.3, P. 8.3-10,

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~Section 1 Continued on Next Page.

I

b 3- Rancho Seco October 29, 1985 Points Available QUESTION 1.3-

,a) . Define " Shutdown Margin." (1.5)

- b)' Define " Actual Shutdown." (1.5)

ANSWER 1.3 a) Shutdown Margin: The amount of negative reactivity held out of the core in the CRAs over and above the reactivity required to compensate for doppler and moderator temperature effects in

. bringing the reactor from any power level to hot shutdown. This includes the conservative assumption that the stuck rod will not be inserted in the event of a reactor trip. It must be demonstrated that the shutdown margin is greater than 1.0% Ak/k, subcritical.

The axial power sheping rods do not change position during a trip,

-therefore, no consideration is given to their worth in determining the shutdown margin.

. b) Actual' Shutdown: The amount of net reactivity in the core. This is a " realistic" calculation which does not assume that.the " Stuck Rod" is~" Stuck Out."

REFERENCEfS1 Reactivity Balance Calculations, B.6, p. B.6-1.

-Section 1 Continued on Next Page-

4 Rancho Seco October 29, 1985 Points Available

~

QUESTION 1.4 IRUE or FALSE. Running two (2) HPI pumps essentially doubles the flow rate provided by one (1) HPI pump. (0.5)

ANSWER 1.4 FALSE REFERENCE (S)

Loss of Subcooling, E.03, p. E.03-12.

i

-Section 1 Continued on Next Page-

,-,,--,-e+, , , ,--,.---e,- , -, , . -- ,,,.-,-,-..-w,,v_-,-n. e,,,,.-,,--,., , , , , , , _ ,-----.,--,-,-.w,.--,, -, . . , . - -

5 Rancho Seco October 29, 1985 Points Available QUESTION 1.5 The graph below assumes a source strength of 100 n/ generation and shows the result of changes in k-eff up to a valkue of k = 0.i5.

What would the graph look like if exact criticality were achieved (i.e., k = 1)? (1.5) 2000-1500-N, n/ gen- = 0.95 K= 0.8 5

~ K = 0.9 500 i. .

K = 0.8 .

~

i i i i i 0 60 120 180 240 Generations, n

-Section 1 Continued on Next Page-

6 Rancho Seco October 29, 1985

.. Points j Avail-able ANSWER 1.5 f t < A ,- in c r A 4.

d /W hf gu

%, ,s m ort. Y U d h t"> v N 2"/ pan j

f.reer,n REFERENCE (S)

Reactor Theory Book, Chapter 8.

-Section 1 Continued on Next Page- '

9 7 Rancho Seco October 29, 1985 Pof'1ts Available OUESTION 1.6 ~

hihat are the components of reactivity that make up the net reactivity in the reactor? (1.5)

ANSidER 1.6

1. net " #f uel + #CRA + #b oron + # Xe * # mod * # power + F Sm REFERENCE (S)

Reactivity Balance Calculation,'B.6, p. B.6-2.

i n

-Section 1 Continued on Next Page- '.

8 Rancho Seco October 29, 1985

- Points Available QUESTION 1.7 Imagine two (2) cases: Case A, the reactor is critical at 10-8 amps and a reactor trip occurs; case B, the reactor is critical at 10-8 amps and -0.05% Ak/k reactivity is-inserted. For the following items, compare the difference in response between Case A and Case B.

a. size of the prompt drop .

b. magnitude of the stable negative SUR

c. final power level .

ANSWER 1.7

a. Case A Case B

b. Case A Case B, i.e., A ~ -0.33 B ~ -0.13

c. Case B Case A REFERENCE (S)

Reactor Theory Manual, Chapter 11, p.11-4.

-Section 1 Continued on Next Page-

9 Rancho Seco October 29, 1985 Points Available QUESTION 1.8

a. Sketch the shape of the reactivity worth curve for Group 8. (1.5)

b. 11 the curve drawn in (a.) above an " integral" or a

" differential" rod worth curve? (0.5)

ANSWER 1.8

a. See attachment

b. Integral REFERENCE (S)

Reactivity Balance Calculations, B.6, p. B. 6-1.

E r

-Section 1 Continued on Next Page-

~7 E;d. A-5 7.2 (C d.EJ -

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'I WITHDRAWN R E'

'k' "k Note: Use the above curve to predict APSR worths '

at all power levels fi;;ne' 5 RANCHO SECO UNIT 1 APSR HORT!I CYCLE 7 VERSUS POS:'!

B 3

11 - Rancho Seco October 29, 1985 Points Available OUESTION 1.9 Assume that the reactor is started up and power is increased at a constant slow rate such that power increases from 0% FP to 100% FP in 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />. Further, assume that the reactor runs for 40 more hours at 100% FP and then trips. Sketch the Xe .

concentration for this power history. (2.0)

ANSWER 1.9 l.

i l Xe f o ..e<.et.. / l s, ,

_/ - _

'x N

'IO s* Y 50 %

REFERENCE (S)

Reactor Theory, Chapter 15.

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-Section 1 Continued on Next Page-

12 Rancho Seco October 29, 1985 Points Available QUESTION 1.10 The energy being transferred at the steam generator is proportional to the primary Delta T (i.e., Th-T ), but not proportional to the secondary Delta T (i.e., T -

is because: feed). The reason for this (Selectone.) steam (1.0) a) the energy transfer taking place within the steam generator is essentially a constant pressure process, b) the secondary flow rate is greater than the primary flow rate.

c) the secondary flow rate is lati than the primary flow rate.

d) the energy transfer taking place within the steam generator. is essentially a constant temperature process.

ANSWER 1.10 d)

REFERENCE (S)

Heat Transfer and Thermodynamics, Chapter 1 Heat Cycles.

-Section 1 Continued on Next Page-

7 13 Rancho Seco October 29, 1985 Points Available QUESTION 1.11 ~

In the condenser energy is being transferred to the circ-water.

If the circ-water flowrate were reduced slightly while holding generated megawatts constant, the most probable result would be:

(Select one.) (1.0)

-a) that the average temperature of the circ-water will increase slightly, b) that the amount of energy transferred at the condenser will decrease slightly.

c) that the saturation pressure within the condenser will decrease slightly.

d) that 'the condenser delta T will decrease slightly.

ANSWER 1.11 a)

REFERENCE (S)

Heat Transfer and Thermodynamics, Part B, Chapter 2, Heat Exchanger.

-Section 1 Continued on Next Page-

p v 14 Rancho Seco October 29, 1985 Points Available OUESTION 1.12 Giran a large vented tank 30 ft. In diameter and 60 ft. high with a centrifugal pump taking a suction from its base. The pump is located at a vertical elevation corresponding to the bottom of the tank and it requires 5 ft. of net positive suction head (NPSH) to prevent cavitation.-

and .is maintained at 60,The tank is almost F by heaters. entirely The tank full of water is designed such that it could withstand 15 psi differential pressure in either direction.

Assume the vent becomes totally clogged while the pump is in operation. Further assume that the pump is of relatively low capacity such that equilibrium conditions are maintained inside the tank.

Answer the following questions:

a) What is the lowest pressure that the tank will drop to as the pump continues to remove water from the tank? Exnlain. (1.5) b) Mill the pump lose NPSH and begin.to cavitate prior to reach-ing a level of 5 ft. In the tank? Exnlain. (italgany assumptions.) (1.0) c) Could the pump continue to pump water at a level below 5 ft.

without cavitation if the vent were open? Frnlain. (Assumeno vortexing.) (1.0)

ANSWER 1.12 a) Thelowestpressurethattgetankcoulddroptowouldbethe saturation pressure for 60 F which is 0.256 psia.

b) Assuming head loss due to flow is negligible, the answer is no.

Cavitation would not begin until the level drops below 5 ft. in the tank.

c) Yes. The added pressure of 14.7 psia at the pump suction would allow all of the water to be removed.

REFERENCE (S)

Fluid Flow Applications, Pumps, p,. 319.

-Section 1 Continued on Next Page-

15 Rancho Seco October 29, 1985 Points Available

~

OUESTION 1.13 The feedwater loop demand stations, the reactor demand station, and the S/G/Rx master are in " hand" with reactor power stable at 15%

power. The operator conngnces a power escalation by bumping rods out and T-ave goes to 580 F. Idhich of the following describeg the required operator course of action to being T-ave back to 579 F7 (Selectone.) (1.0) a) The operator must insert rods slightly, but not as far as they were withdrawn. He must request the I&C technicians to adjust the low level limits.

b) The operator must increase feed flow, restoring OTSG level, and lower the steam header setpoint slightly.

c) The operator must over-feed the OTSG slightly, increasing OTSG inventory, then stabilize at a higher feed flow and a higher OTSG level.

d) The operator does not negd to take any action. Doppler feedback will return T-ave to 579 F.

ANS1 DER 1.13 c)

REFERENCE (S)

ICS Training material. -

-Section 1 Continued on Next Page-L

1 16 Rancho Seco October 29, 1985 Points Available OUESTION 1.14 Assume that a steam bubble has formed in the reactor vessel head during a RCS natural circulation cooldown. 11 the collapsing of that bubble a relatively fast process or a relatively slow process?

Exnlain. (2.0)

ANSWER 1.14 This process is a relatively slow one compared to how fast a bubble can form. Because there is no spray available in the vessel head, there fs no fast way to cool the steam space. Rapid compression of the bubble results in it becoming superheated.

REFERENCE (S)

Heat Transfer and Thermodynamics, Part A Basic Concepts.

- Section 1 Continued on Next Page -

17 Rancho Seco October 29, 1985

~

Points Available QUESTION 1.15 The reason that it takes longer and longer for startup rate (SUR) to reach zero as the reactor nears criticality is:

(Selectone.) (1.0) a)' because the effect of the delayed neutrons is becoming more dominant than prompt neutrons, b)' because the decay of SUR is becoming dominated by the decay of the longest lived precursors, c) because as K effective nears one (1) the effect of the previous generations of neutrons is becoming more dominant in the current neutron population.

d) because as the rods are withdrawn from the core they are providing less shielding of the source range detectors.

ANSWER 1.15 c)

REFERENCEfS)

Reactor Theory, p. RT 8.2 - RT 3.11.

-End of Section 1-6 O

9 0

18 Rancho Selo OctcDer 29, 19E5 2.0 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS _ (f.d)

'E Points .

,' ' Available OUESTION 2.1 Briefly exnlain ham excessive RCS leakage into the DHR system .

(1.0) through check valve RCS-001 during normal plant operation is -

detected. '

ANSWER 2.1 -

CFT level and pressure will increase due to a small line frca the DHR outlet piping with a check valve allowing fle,w into the CFT outlet line. > -

REFERENCE (S)

STM 10, DHR, p. 63. '

b l -

\

\

. s

-Section 2 Continped on ?! ext.Page-

\

. 19 Rancho Seco c October 29, 1985

_ Points Available OUESTION 2.2 Drax a one-line diagram showing the 4160 V inputs to buses (2.0) 4A, 48, 4A2 and 4B2. Include all four diesel generators and identify supply transformers. Show circuit breakers and indicate whether normally open or closed. You do not need to show electrical loads.

SNSWER 2.2 3.U # 1 S.U. # 2 i

&m&% & &%

6.9 hv 4.18 k' fin $e ve.

7 14-NV P )

_ f4AtI) 4Al. *I)48_(V  !)4ll2 V I I .-

<) I) I) I) I)

, .) ,) I) h M 43At h M 434 438% M 438th M '

) ')

. 3At . . . l) 3 A 38 l) l _ 382 _

I I I l) I) i) 1) 1) ,) ,)

1) i t i PA3 283 2At 282 S4 WM 240/tt0VAC IA3 240/tto v4C 883 kmVRE i REFERENCE (S)

OD 24 C 0107 MOD 40 - Electrical Distribution, p. 1.

STM 43, Vital Electrical Distribution, p. 4.

-Section 2 Continued on Next Page-

20 Rancho Seco October 29,-1985-Points Available i ..

QUESTION 2.3 State the normal source of air for the Emergency Diesel Generator (1.0) air-start compressors and the reason why it is' preferred over the alternate source.

ANSWER 2.3 Instrument air is used to reduce moisture that might condense and freeze in the starting air motors.

REFERENCEfS)

STM 45, EDG, p. 152.

4

)

)

, e

-Section 2 Continued on Next Page-l

t i 21 Rancho Seco October 29, 1985

~~

Points Available QUESTION 2.4 List thr'ee ' interlocks affecting condensate pump mini-flow recirculation line valve FV-35109 and state how they position the valve. (1.0)

ANSWER 2.4

1. Opens on startup of any condensate pump.

2. Closes when all condensate pumps are stopped.'

3. . Closes when total condensate flow exceeds 5000 gpm.

REFERENCE (S)

,STM 20a, Condensate System, p. 9.

e i ,

s b

~'

_ s b

i e

> /

4 i

i

~

A l i .

-Section 2 Continued on Next Page- 1 1

22 Rancho Seco October 29, 1985

_ Points '

Available

^

'00ESTION'2 5 .

List three (3) symptoms that indicate that a large LOCA which (1.0) has occurred is due to a CF tank line break.

ANSWER 2.5

1. One, but not both, CF tank blows down immediately.

2. LPI flow to affected nozzle (none to other) with RCS pressure Q240psig.)

3. RCS pressure remains > 600 psig for approximatel 2.5 minutes)

(no.CFT flow in unaffected leg) or, also acceptable, RCS pressure remains > 240 psig for approximately(7 minutes) (no LPI flow in unaffecte loop).

REFERENCE (Si b b N'

~

CP.101-1.

{ptL16 tLAA0

• 5 f

-Section 2 Continued on Next Page-

l 23 Rancho Seco October 29, 1985 l l

Points Available QUESTION 2.6 IRME or FALSE. At any RCS pressure, HPI flow is greater with (1.0) three (3) umps than with two (2) pumps, which is greater than

- with one ( ) pump. However, the flow difference between one (1) and two (2 pum between two (2)p operation and three~(3)ispump greater than the flow difference operation.

ANSWER 2.6 TRUE

~ REFERENCE (S)

E.07-19, Figure 3.

4 4 a

-Section 2 Continued on Next Page-

24 Rancho Seco October 29, 1985 Points Available

~

QUESTION 2.7 Briefly exclain how the impulse steam trap shown in the attached Figure 2.7 operates to expel small quantities of condensate from the main steam system. (1.0)

ANSidER 2.7 There is a continuous flow through the control orifice. Flow throttling past the flange of the valve disk causes reduced pressure above it and the disk lifts allowing condensate to blow past the seat. When the condensate becomes hot enough (before steam actually enters the valve) it will flash to steam when throttling past the disk, reducing the pressure drop. The valve will close because the disk will be forced down on the seat since there is a greater area above the disk than below it for the equalized pressure to act upon.

REFERENCE (S1 STM 14, Main Steam,'p. 66.

O g.. -

n] (

p 'v-

-Section 2 Continued on Next Page-

25 Rancho Seco October 29, 1985 Points Available QUESTION 2.7 (Continued)

CYLINDER STEM

/ /

. CONTROL ORtFICE CYLINDER g VALVE CIRCtLAR ' I h

V p$*

9  % = /)

/A ,,s=-

//__a. gxxxxxxxxxx1g--

...A. m g%$_\.Y_q;m- p 1 3

20i?E

• " ^ S

(, j VALVE DISK 3 }LaANcE VALVE SEAT

}

/

NRC EXAM FIGURE FIGURE 2.7 eigge e s a +cie IMPULSE STEAM TRAP l

-Section 2 Continued on Next Page-

26 Rancho Seco October 29, 1985 Points Available OUESTION 2.8 Containment isolation on receipt of a SFAS signal involves use of two (2) isolation valves in series in each line. What types of valve actuation are used, how are they located, and in what position will they fail: (2.0)

a. when one is inside'and one is outside the RB7

b. when both are outside the RB7 ANSWER 2.8 a b fail Motor operated inside nearest RB as is Air operated outside second from RB closed REFERENCE (S)

STM 4, RB and Support Systems, p. 44.

-Section 2 Continued on Next Page-

27 Rancho Seco October 29, 1985 Points

-~

Available

~

QUESTION 2.9 State two (2) ways to depressurize the RCS during a natural (1.0) circulation cooldown.

ANSWER 2.9

1. Use auxiliary spray from the HPI system if spray water to pressurizer AT ( 410 F.

2. Open EMOV

$,. O&l YM V REFERENCE (S)

CP.102-1.

v/F-At f

-Section 2 Continued on Next Page-

k 28 Rancho Seco October 29, 1985 Points Available QUESTION 2.10 State the normal and alternate power supplies to the Makeup Pump (2.0)

P-236 and discuss the interlock between these power supplies.

ANSWER 2.10 48 normal and 4A alternate. System consists of two (2) locally mounted, manually operated, key interlock switches. There is only .

one (1) key available. Close permissive and interlock switch prevents closure of the breaker unless key is in and switch is in the CLOSE-PERMISSIVE position. Key cannot be removed unless in the OPEN position.

REFERENCE (S)

STM 5, MVP, p.70.

W

-Section 2 Continued on Next Page-l

29 Rancho Seco October 29, 1985-Points Available

~

QUESTION 2.11 State the normal and alternate destinations of water from steam generator blowdown and state during which phases of plant operation SG blowdown is performed? (1.0)

ANSWER 2.11 g (g, ec, h, ha Blowdown is directed to the HP condenser (normal) and th polishingandmakeupdemineralizerareasump(alternate)f , during plant heatup, low power -((15% FP) operation, shutdown and cooldown.

~

REFERENCE (S)

STM 14, Main Steam, p. 6.

L. Gm% Se,:1p- Duc y A l- 53 L-l l

-Section 2 Continued on Next Page- i

_ _ _ l

30 Rancho Seto October 29, 1985 Points Available

-~

~

QUESTION 2.12 What cools the auxiliary feedwater pump turbine bearings and in where is the heat transported? (1.0)

ANSWER 2.12 Water from the pump's first stage cools the bearing lube oil and is returned to the pump suction.

REFERENCE (S)

STM 29, AFW, p. 11.

-Section 2 Continued on Next Page-

31 Rancho Seco October 29, 1985

~

Points Available QUESTION 2.13 Exolain why increased letdown is required following termination (1.0) of an overcooling transient during which HPI was initiated.

ANSWER 2.13 Reheating of the RCS by decay heat will cause the primary coolant to swell, possibly causing a solid pressurizer and/or overpressure condition.

REFERENCE (S)

E.05-4.

Section 2 Continued on Next Page-T r - -- r r .w -

32 Rancho Seco October 29, 1985 Points Available OUESTION 2.14 -

Wha.t are the auxiliary feedwater flow paths and flow controls after leaving the AFW pump during the following conditions:

(Note: identify components by number for parts a and b.) (2.0) a.

b.

SFAS not actuated?

SFAS actuated?

@ [ W h ct[ b T' f p

c. Minimum flow recirculation?

ANSWER 2.14 a and'b. pumpsk318/319 to SG A/B with cross-tie valve open

a. throughICSauto/manualcontrolledvalves(FV-20527/8

b. bypassingcontrolvalvesviaSFASvalves(SFV-20577/8) a and b.

{throughqheckvalves(FWS-061/2andisolationvalves FWS-063/4)intoSGs

c. from each pump discharge to the low pressure condenser via flow orifices, check valves and locked open valves REFERENCE (S)

STM 29, AFS, p. 7.

' Q,, a n osrf nw+s G % au b u t 14

~

-Section 2 Continued on Next Page-

l i

i 33 Rancho Seco October 29, 1985 Points Available QUESTION 2.15

~

What are the normal and backup sources of RC pump-lift oil pressure during startup/ shutdown and normal operation? (1.0)

ANSWER 2.15 Startup/ shutdown: AC and backup DC motor operated pump.

Normal operation: The RCP pumps its own oil using thrust bearing skimmers.

REFERENCE (S)

~STM 2, RCS, p. 33.

- Section 2 Continued on Next Page-e c --- ,_

34 Rancho Seco October 29, 1985

~

Points Available ,

QUESTION 2.16 Briefly exnlain how the cooling water (condensate) flowrates through the air ejector condensers and gland steam ~ condenser are maintained constant regardless of variations in condensate-flow. (1.0)

ANSWER 2.16

~

Af4ps[)deltapismaintainedacrosseachbyconnectingthemin paTallel with a bypass line whose flow is adjusted to maintain the dp.

REFERENCE (S)

STM 20, Condensate System, p. 9.

-Section 2 Continued on Next Page-

35 Rancho Seco October 29, 1985 Points 4

Available

~

QUESTION 2.17 What is the effect on turbine bypass valve operation of shutting (2.0) main steam to auxiliary steam valves HV-20560 and HV-205657 '

Exolain.

ANSWER 2.17 TBVs will fail shut. This is caused by loss of condenser vacuum when air ejectors and turbine sealing steam is lost.

N0 YWNAW WA9,& &

REFERENCE (S) c&,

E.05-2.

Aw s% CyL, cmx yrs t

w

' ~

, -Section 2 Continued on Next Page-

36 Rancho Seco October 29, 1985 Points Available OUESTION 2.18 -

Which of the following statements about an electrically operated (1.0) 4160 VAC circuit breaker are TRUE?

a. When-the manual trip push button is pushed, it activates an electrical circuit that opens the breaker contacts,

b. To charge the closing springs during an emergency, the charging lever must be used with the manual charging handle,

c. When racking in the breaker using the racking crank, it will automatically stop and lock only in the TEST and CONNECTED positions.

d.. When withdrawing the breaker from the switchboard,- the closing springs must be manually discharged to prevent accidental discharge.

ANSWER 2.18 b.

REFERENCE (S)

STM 43, Vital Electrical Distribution, p. 28.

-Section 2 Continued on Next Page- '

37 Rancho Seco October 29, 1985 Points Available OllESTION 2.19 In the low pressure condenser, condensate dropping from the tubes is reheated to saturation by falling through a

' band of steam on its way to the hotwell. Idhat is the source of this heating steam, and what other function is provided by its supply? (1.0)

{pi5cw ,u ptud0")

Cross tie from HP condenser which also equalizes pressure between HP and LP condensers.

REFERENCE (S)

STM 20, Condensate System,'p. 1.

i b N (: h

(

c + f)

J..) >

I v

,/

-Section 2 Continued on Next Page-i

~

38 Rancho Seco October 29, 1985 Points Available

~

QUESTION 2.2Q.

List four symptoms observable locally and/or in the control room which indicate a tube rupture in a high pressure feed-water heater. (1.0)

ANSWER 2.20

1. Annunciator "Feedwater Heater Level Hi-Lo"

2. Alarm on computer "Feedwater Heater Level High"

3. High-level indication at local gauge glass

4. Extraction non-return valve closed for affected heater REFERENCE (S)

STM 22, MFW, p. 84.

C. I 5 - l

-End of Section 2-E.EcA1:Lud ca"Nc4 cd4 M 9k i,3p Kdspu A-usyn

1. feck et h L y oid % q ()< A . E j, O

e

39 Rancho Seco October 29, 1985 3.0 INSTRUMENTS AND CONTROLS _ 0)

Points Available 00ESTION 3.1 A Decay Heat Removal pump can be started at any time during plant (1.0) operation except during one (1) condition. State what that condition is, and exclain the time sequencing required for normal and ESFAS startup of the pump once the condition has been cleared.

ANSidER 3.1 Startup is blocked if either DHR suction block valve (HV-20001 or 2) is off its fully open seat. After'etther valve leaves its fully open seat, DHR pump startup is blocked for 3 minutes-even W/eg 41 fully open status is regained. ESFAS overrides the time delay.

(Pu+ talk [n Ph.A Leaky n duvth)

REFERENCE (S)

STM 10, DHR, p. 50.

W O *' E - 703, S h. 3

-Section 3 Continued on Next Page-

40 Rancho Seco October 29, 1985

_ Points Available

~

QUESTION 3.2 Channel A of the RPS is in a tripped state due to a defective power supply. Briefly exclain what ynn must do to allow testing of Channel B components without tripping the reactor and what this action accomplishes. Be sure to include any effects on other channels. (1.0)

, ANSWER 3.2 A key operated bypass switch on the reactor trip module must be turned to the Manual Bypass position. This supplies power holding the Channel B trip relay closed despite interruption of.the normal trip string continuity. It also opens contacts in Channels A, C, and D preventing manual bypassing of any of these other channels.

REFERENCE (S)

STM 36, RPS, p. 13.

-Section 3 Continued on Next Page-

41 Rancho Seco October 29, 1985 Points Available

~

QUESTION 3.3 Briefly exolain what the " crowbar" circuit 1.n a 24 V power (1.0) supply does.and why.

ANS1 DER 3.3 fiv4eyeefIS)

It short circuits the power supply output if output voltage exceeds a preset level to protect delicate NNI loads.

REFERENCE (S)

STM 33, NNI, p. 9..

?

Y

-Section 3 Continued on Next Page-

42 Rancho Seco October 29, 1985

- Points Available QUESTION 3.4 Which of the following statements comparing thennocouples and (1.0) resistance temperature sensors are TRUE7

a. Thermocouples provide a higher output voltage.

b. Thennocouples are better suited for high temperature measurements.

c. -Thermocouples are more sensitive to small temperature changes.

d. Thermocouple calibration accuracy and stability can be better by a factor of 10 in moderate temperature ranges.

ANSWER 3.4 b.

REFERENCE (S)

~

STM 33,.NNI, p. 28.

L .

-Section 3 Continued on Next Page-

43 Rancho Seco October 29, 1985.

_ Points Available QUESTION 3.5 Liit the two (2) Emergency Diesel Generator Shutdown Circuits (1.0) that remain in service following an SFAS signal, and the four (4) that are isolated.

ANSWER 3.5~

Not blocked: Generator differential and ground relay trip

. rk Engine

%t Ts overspeed d re swttrip tr lr 0K abo Block'ed: Low oil pressure High crankcase pressure High water temperature Manual pushbutton REFERENCE (S)

STM 45, EDG, p. 116.

$06 O'bf tw G r- G 19 2 - E~- y ~p p y,

-Section 3 Continued on Next Page-

44 Rancho Seco October 29, 1985

_ Points Available l 1

OUESTION 3.6 An SFAS digital . subsystem will receive a trip signal from the logic buffer of an analog subsystem under what four (4) conditions? (1.0)

ANSWER 3.6

1. A' trip signal is sent from the analog subsystem bistable.

'2. Any module in the analog subsystem is removed.

3.. Any module in the analog subsystem is placed in test.

4. The DC power supply to the pressure transmitter for the analog subsystem is lost.

REFERENCE (S)

STM 35, SFAS, p. 26.

~

l & fCAldW Q 4; (0.))

? AfSi i li

~7 :;o yS l t: 'l 4 %rof Si fCS ,

h4(c Mb / "# 'l

-Section 3 Continued on Next Page-

l 45 Rancho Seco October 29, 1985 Points Available QUESTION 3.7 Briefly exolain how the ' main steam line failure logic control (1.0) ensures interlock actuation on an actual main steam line rupture yet prevents it on failure of a single pressure switch or rupture of a single sensing line.

ANSidER 3.7 Each SG outlet line is monitored by two (2) low-pressure actuated switches on a single sensing line. One (1) switch Chr] O v e r inputs to Channel A and the other to Channel B. Trip of g &7 b' either channel actuates interlocks. However, for either channel to trip requires low-pressure inputs from both SG [A d i / g/.4 d I outlet lines, so failure of one (1) sensor will not actuate interlocks. Failure of a single sensing line will input only.one (1) low pressure input into each channel, so inter- f<h b e locks will not be actuated.

REFERENCE (S) ACghb U^64 #

M '

STM 14, Main Steam, p. 88. L %[ 7>

~

Tw we cay v- m-s30 M ' $ 5 0aAf

& ,vj w am ce &&

! (LL

-Section 3 Continued on Next Page-

. n

+

'. 46 Rancho Seco m

~' '

October 29, 1985' Points Available.

~

~

OUESTION 3.8 i IRUE or FALSE. Isolation of Auxiliary, Steam from the Auxiliary (1.0)

Building occurs upon actuation of any one (1) of the six (6) temperature switches located in tae HPland makeup' pump rooms. ,

ANSWER 3.8 TRUE u

REFERENCE (S) 00 24 C 0120, Auxiliary Steam to Auxiliary Building Isolation.

~

s s

.=*

T

(

% s.

4 4

\

1

-Section 3 Continued or. Next Page-en ,

s

• x g i

m A 4

i 47 Rancho Seco October 29, 1985

~

Points Available 00ESTION 3.9 y (

Which of the following statements about OTSG startup range level instrumentation are TRUE7 (1.0)

a. Indication is most accurate at low SG temperature.

Output of the level transmitter not selected for b.

control room indication and control is used to provide the HI-LO alarm signal.

(* c. The maximum level indicated b./ the SUR meter corresponds to 266 inches abo're the lower tube sheet.

w: "

d. The plant process computer receives a signal from the transmitter selected for control room indication and control.

% '-[ ANSWER 3.9 c.

REFERENCE (S)

STM 33, NNI, p. 87.

' ~

-Section 3 Continued on Next Page-s 9

1& < -, , ,

o 1

-1 48 , Rancho Seco October 29, 1985

_ Points

. Available QUESTION 3.10.

/

IRUE or. FALSE. The CRD Breaker Shunt Trip Circuit causes.a '

(1.0) shunt trip of the A and/or B breakers when the test switch is placed in the " Shunt Trip De-energized" position.

ANSWER 3.10 .,

FALSE REFER'ENCE(S)

OD 24 C 0115, ATWS-Shunt Trip for CRD Breakers.

-Sect ion 3 Continued on Next Page-

7 (4

49 Rancho Seco October 29, 1985

_ Points Available i

t ,_/ ..c : ,~ -

~

Ik l OUESTION 3.11 State the conditions that will cause actuation of the AFW

{

> pump runout alarm for the steam and motor driven AFW pumps. (1.0)

T

- ANSWER 3.11 s

Pump discharge pressure less than 650 psig in combination with closed AFWP breaker (motor driven) or open AFWP steam inlet valve (steam driven) after a time delay to allow pump startup (120 sec-not required)..

s T

REFERENCE (S)

'00 24 C 0112 AFW Pump Runout Alarm.

j-ts t k' .?

, 1

! -Section 3 Continued on Next Page-4 ~

t I

p .-

i 50 Rancho Seco October 29, 1985

_ Points Available QUESTION 3.12 Which of the following statements concerning signal isolation (1.0) cabinets H4SIA and H4 SIB ARE TRUE7

.. Class IE incore thermocouple instrument loops are served by these cabinets.

b. H4SIA is located in the East Switchgear Room.

c. The Class lE side of H4SIA is powered from SIA2-1.

d. Upon loss of one (1) cabinet, the other will provide redundant signals.

ANSWER 3.12 a,c,d REFERENCE (S)

OD 24 C 0102, Signal Isolation Cabinets.

f f

-Section 3 Continued on Next Page-i

51 Rancho Seco October 29, 1985

_ Points Available QUESTION 3.13 Briefly exolain han source range NI indication isolated from the control room may be obtained. (2.0)

ANSWER 3.13 A portable source range instrument cabinet.is provided near the containment penetration.- RPS Channel A is disconnected from the control room and connected to the portable cabinet. Cabinet power. is provided by an extension cord located in the~ area.

REFERENCEfS1 OD 24 C 0103, Appendix R Modifications.

3

-Section 3 Continued on Next Page-

- - _ , - , - - . , w ,

52 Rancho Seco October 29, 1985

~

Points Available QUESTION 3.14 Feedwater temperature is used in the Feedwater Demand Calculator and Btu limits sections of the ICS. Describe the effect that

' failure of the feedwater RTD (high) would have on the plant when

' operating at 100% power in full ICS Auto. Assume no operator intervention. (2.0)

ANSWER 3.14 FW demand calculator would increase FW demand. T-ave would decrease, causing Rx demand increase. Rx demand high limit would prevent plant trip on high flux. Plant would trip on low pressure or variable temperature-pressure. Stu limit has no effect.

REFERENCE (S)

Ch 32, ICS System Description.

4

-Section 3 Continued on Next Page-

53 Rancho Seco October 29, 1985 Points Available OUESTION 3.15 Which of the following statements about variable area flow (1.0) meters (Rotameters)-are TRUE7 ,

1

a. The metering element is a float in a tapered tube l which is pushed up by the flow,

b. .They are used to measure RCP seal return flow.

c. Meter output is proportional to the square root of the flow rate.

d. Metering element motion is converted to an electrical signal by mechanical and magnetic coupling to an electric coil.

ANSWER 3.15 a,b,d REFERENCE (S1 STM 33, NNI, p. 39.

1 f '

i l ,

-Section 3 Continued on Next Page-

r s

54 Rancho Seco October 29, 1985 Points Available OUESTION 3.16 State what is indicated on the meter of the auxiliary feedwater valve  !

- HAND / AUTO station when the meter is switched to the " Meas. Var."

and then to the "POS" indication. (1.0) l ANSWER 3.16 g

" Meas. Var."-FWjlevel[s'etpointinSG(0to100%ofsetpoint transmitterrange)

"POS" - EFW valve position demand (0 to 100% demand)

REFERENCE (S)

STM 32, ICS, p. 102.

~

-Section 3 Continued on Next Page-

a 55 Rancho Seco October 29, 1985

- Points Available QUESTION 3.17 Briefly exolain under what conditions and why the subcooling

~

(" margin to saturation") meters may not accurately indicate true RCS subcooling conditions. (1.0)

ANSWER 3.17 Under voided or low flow conditions, the RTDs may not be measuring a representative sample of RCS water. There may be hotter regions within the system.

REFERENCE (S)

STM 33, NNI.

--Section 3 Continued on Next Page-J' k

1

-c--- , -

, .,, -- , . . - - + ,-.e -- ------ ,e. -- - .. - - - - - , - ,-

56 Rancho Seco October 29, 1985 Points Available QUESTION 3.18 Upcn receipt of a SFAS initiation signal by the RB spray system, several system backlighted pushbutton control switches start flashing. Identify which switches flash and exnlain the significance of this indication. (2.0)

ANSWER 3.18 ,

RB spray pump d STOP.

RB spray pump discharge valves SFV-29107/8, , 0 and CLOSE.

Pump start and valve opening are delayed for 5 minutes to allow the operator to take manual control if RB spray is not necessary.

REFERENCE (S)

STM 28, RBS, p. 23.

SO5 Q4& Y f-203 6]41 f s Lu+ 55 l'y % i a 4

-Section 3 Continued on Next Page-4

-.y-- ,__ . . . . .

57 Rancho Seco October 29, 1985 Points Available

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OUESTION 3.19 List the types of releases monitored by the RB purge stack (1.0) accident rad monitor (R-15044) during normal plant operations.

ANSWER 3.19 4 [ Tf r h nY 6. e 5 tA $

noble gases ,C 5 P u v 9 <.s iodines g g /,"

7 t,q Q/use particulates Md MM4 k W4/ W REFERENCE (S)

OD 24 C 0118, RB Purge Stack Accident Rad Monitor Modification.

Section 3 Continued on Next Page-

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'58 Rancho Seco October 29,-1985 Points

" Available I

OUESTION 3.20 What instrumentation other than out-of-core detector response, is available to determine if there is a steam bubble or voiding in the vessel head and het would you use the infonnation pro-vided by-these-instruments? (2.0)

' ANSWER 3.20

1. pressurizer level

' 2. makeup tank level

3. system pressure

4. incore.T/Cs Pressurizer. level changes will not correspond to change in temperature of volume. The level changes will be more rapid than could occur due to changes in temperature or_ volume.

Spray into the pressurizer will ca'use level to increase and makeup into the RCS will cause the level to decrease.

Pressurizer level could increase with the RCS pressure constant or even decreasing. If the plant is supposedly solid, the pressure will'not respond to temperature or volume

~

changes, as expected in a solid condition. Makeup tank level changes will not correspond to pressurizer level changes. The incore T/Cs could be used in conjunction with system pressure and the steam tables to determine margin to saturation in core area. This indication should be used with verification from the previously mentioned indications.

REFERENCE (S1 B.4, Plant Operations Manual.

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-Section 3 Continued on Next Page-6 i

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59 Rancho Seco October 29, 1985

- Points Available QUESTION 3.21 Other than alarms, what ' indication is available in the control room of Fire Protection System status? (1.0)

ANSWER 3.21 Backlighted start /stop buttons (on panel H2X) for the motor--

driven and diesel-driven fire pumps.

REFERNNCE(S)

STM 24, Fire. Protection, pp. 44, 77, 78.

- End of Section 3 -

O 1

i 60 Rancho Seco October 29, 1985

4.0 PROCEDURES

NORMAL. ABNORMAL. EMERGENCY. AND _ 4 3.(

RADIOLOGICAL (&510) i

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Points Available OUESTION~4.1 IRME or FALSE. According to the Safety Parameter Display System Procedure,. A.76, the control room operator should not perform

.11 agnostic checks on the SPDS. (0.5)

ANSWER 4.1 TRUE REFERENCE (S)

Safety Parameter Display System, A.76, p. A.76-1, Rev. 2.

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-Section 4 Continued on Next Page-f

61 Rancho Seco November 5, 1985 Points Available QUESTION 4.2 State the limiting values for the following operational parameters:

a. Maximum allowable AT between RCS and OTSG average shell temperature. (0.5)

b. Minimum temperature for criticality (0.5)

c. Level below which an OTSG must be considered dry and fed through auxiliary feedwater nozzles. .(0.5)

ANSWER 4.2

a. 60g U

Ref: B.2,p.B.2-2)

b. 525 F Ref: Tech. Spec.3.1.3,p.3-6) ,

c. 8 in. Ref: B.2, p. B.2-2)

REFERENCE (S1 Plant Heatup and Startup, B.2.

Technical Specifications.

l

-Section 4 Continued on Next Page-

e e p -- - J A - - _ 4ma__

62 Rancho Seco November 5, 1985 Points Available QUESTION 4.3 Early in the Plant Heatup and Startup Procedure, B.2, the operator is directed to draw a vacuum in the OTSGs. For what two (2) reasons is this action required? (2.0)

ANSWER 4.3

1. Produce steam at low OTSG water temperature which will condense on the OTSG shell and warm it.

2. . Remove noncondensible gases from the OTSG.

REFERENCE (S)

Plant Heatup and Startup, B.2, p. B.2-9.

-Section 4 Cont inued on Next Pace-

4 63 Rancho Seco November 5, 1985 Points Available QUESTION 4.4 TRUE or FALSE. - When a reactor coolant pump is secured while at power, seal injection flow'is NOT required provided component cooling water flow is maintained. (0.5)

ANSWER 4.4 FALSE. (Without pump operation, the auxiliary impeller will not provide flow through the seal cooler.)

REFERENCE (S)

Reactor Coolant Pump System, A.2, p. A.2-8.

1 l

. -Section 4 Continued on Next Page-i

64 Rancho Seco November 5, 1985

_ Points Available 00ESTION 4.5 When performing a reactor plant heatup with Group 1. Safety Rods withdrawn, the Plant Heatup and Startup Procedure, 8.2, -requires that a 1% Ak/k shutdown margin be maintained. Dags that mean that the reactor must be maintained at 1% shutdown even with Group 1 withdrawn? Exolain. (2.0)

ANSWER 4.5 Yes. The explanation should indicate knowledge of the fact that withdrawing Group 1 Safety Rods will reduce the amount that the reactor will be subcritical; therefore, requiring sufficient boron concentration for a shutdown. Alternatively an understanding of the difference between the term " shutdown margin" when applied to a critical reactor or a shutdown reactor.

REFERENCE (S1 Plant Heatup and Startup, B.2, p. B.2-3.

Technical Specification 3.1.3, p. 3-6.

Reactivity Balance Calculation, B.6, p. B.6-1.

Soluble Boron Concentration Control, 8.9, p. B.9-1.

4 -Section 4 Continued on Next Page-

65 Rancho Seco November 5, 1985 Points Available

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QUESTION 4.6 When maintaining pressurizer (PZR) level during the early phases of an RCS heatup, the level control valve is not to be placed in auto. Why? (1.0)

ANSWER 4.6 The controller is not to be placed in auto with excessive delta pressure across the valve.

1 REFERENCE (Sl Plant Heatup and Startup, B.2, p. B.2-10.

J

-Section 4 Continued on Next Pace-

66 Rancho Seco November 5, 1985 Points Available QUESTION 4.7 Which one (1) of the following statements is IRUE regarding the procedure for stopping a reactor. coolant pump (RCP) in a non-emergency situation (A.2)? (Selectone.) (1.0)

a. The RCP may be tripped without manually starting the high pressure oil lift pump because the oil lift pump picks up automatically, b.- The RCP should be tripped while simultaneously manually starting the high pressure oil life pump.

c. The high pressure oil life pump should be started prior to securing the RCP.

d. Operation of the high pressure oil lift pump is not required for stopping an RCP.

ANSWER 4.7 c.

REFERENCE (S)

Reactor Coolant Pump System, A.2, p. A.2-8.

t

-Sectic: 4 Continued on Next ."ags-

67 Rancho Seco November 5, 1985

_ Points Available OUESTION 4.8 Which one (1) of the following requires the implementation of containment integrity? (Selectone.) (1.0)

a. Increasing pressurizer temperature > 200'

b. Increasing k,ff to ) 0.99

c. Placing one (1) train of decay heat removal out of service

d. Removing the reactor vessel head with boron concen-tration at 1900 ppm ANSWER 4.8 b.

REFERENCE (S1 Technical Specifications, 3.6, p. 3-39 and 3.8, p. 3-44.

1

-Section 4 Continued on Next Page-

68 Rancho Seco November 5, 1985 Points Available QUESTION 4.9 The reactor is in " hot shutdown" with all control rods fully inserted. The axial power shaping rods (APSRs) are to be repositioned. According to the, Plant Heatup and Startup Procedure, B.2, WhAt must be done prior to repositioning the APSRs? (1,0)

ANSWER 4.9 All safety groups must be withdrawn to 100%.

REFERENCE ($1 Plant Heatup and Startup, 8.2, p. B.2-16.

D e

O e

>- -Section 4 Continued on Next Page-

69 Rancho Seco November 5, 1985 Points Available 00ESTION 4.10' If neither SPOS display is available and no RCPs are available, what indications are used to detemine subcooling margin and what constitutes adequate subcooling margin? (1.5)

ANSWER 4.10 RCS pressure Average of five highest incore thermocouples 0 Adequate subcooling margin exists when RCS > 50 subcooled REFERENCE (S)

Vital System Status Verification, E.02,. Guideline 1, p. E.02-8.

O 9

l 0

-Section 4 Contiriued on Next Page-

70 Rancho Seco November 5, 1985 Points Available QUESTION 4.11 State tne two (2) operator actions that must be performed upon loss of subcoolin Procedure, E.03)nogmatter (in accordance what the cause withwas.

the Loss of Subcooling (2.0)

ANSWER 4.11 Trip all RCPs.

Initiate and control HPI.

9 REFERENCE (S)

Loss of Subcooling, E.03, E.03-1.

S 9

9 9

-Secliun 4 Cunlinueu un Next Page-

71 Rancho Seco November 5, 1985 Points Available QUESTION 4.12 A reactor trip occurs conincident with a loss of power to all reactor coolant pumps. State two (2) indications that natural circulation has been established as given in the Plant Shutdown and Cooldown Procedure, B.4. (2.0)

ANSWER 4.12 Verify Natural Circulation

1. Th and incore T/Cs' rise about 35 F above T cand commence to track T in 15 to 30 minutes. (These values are indicative of a fuil-power trip with both OTSGs operable.)

2. Tc remains steady or decreases slightly and is at or slightly greater than saturate temperature for OTSG secondary pressure.

3. Tetracks OTSG secondary side T sat'

4. Th and incore T/C's track changes in OTSG secondary side pressure.

REFERENCEfS)

Plant Shutdown and Cooldown, B.4, p. B.4-14.

ALTERNATIVELY:

~

Verify Natural Circulation

1. OTSG T sat approximately the same as Tc -

2. Lowering OTSG pressure causes a corresponding decrease in T c and Th /incores.

3. Incore temperature within 10 F of Th-REFERENCE (S)

Vital System Status Verification, E.02, p. E.02-2, Rev.1.

-Section 4 Continued on next Page-

72 Rancho Seco November 5, 1985 Points Available

-00ESTION 4.13 IRUE or FALSE. The RCS total volume will not be circulated or mixed when on decay heat. (0.5)

ANS1 DER 4.13 TRUE REFERENCE (S)

Plant Heatup and Startup, B.2, p. B.2-2.

-Scction 4 Ocntinusd on Next Page-

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73 Rancho Seco November 5, 1985 Points Available IDN 4.14 The Plant tdown and Cooldown Procedure, B.4, states that both generator (0CB-220 and OCB-230) and the generator field breaker should be o or to tripping the turbine.

Idhy? (1.0)

ANSWER 4.14 -

If all three (3) breakers.are not open, a reactor trip will result.

Cl2 g '

REFERENCE (S)

Plant Shutdown and Cooldown, B.4, p. B.4-4.

-Section 4 Continued on Next Page-

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.., x v ss-74 Rancho Seco November 5, 1985

-- Points '

AvailabJa, s:

s N

OUESTION 4.15

• When the reactor is shutdown, "RCS overpressure protection" is established. Glyn a basic description of what constitutes RCS overpressure protection. Your answer should include four (4) c separate elements. (2.0) .

Ib...

.s-ANSWER 4.15 -% '

tm s The EMOV block valve caution tag open.

EMOV . key lock selector in " low" position. _- %

' A Place clearance on HPI injection valves. s s Place clearance on HPI pumps.  ?.

1 REFERENCE (S) ,

Plant Shutdown and Cooldown, B.4, p. B.4-10.

i

.- 't

.s es t i, N 1 v 'u '

t

\,-

t ,

l s  !

l 1

l L. j 1

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-Section 4 Continued on Next Pace- N l l

r --

75 Rancho Seco November 5, 1985 t-s s

Points Available

~~

w .

J QUESTION 4.16l

\

IRUE or Maccordance with the Loss of Subcooling Procedure, E.03, th~e' preferred temperature indication to be used for determining whether the R superheated or not is the average of the five (5) hig ore thermocouples. (0.5)

~

ANSWER 4.lfi

'^' '

' FALSE .

ar-

) ~C s ( <l . (1-,-

REFERENCE (S)

Loss of Subcooling, E.03, p. E.03-4.

s 4t 4

s, '

-Section 4 Continued on Next Page ~

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76 Rancho Seco November 5, 1985

- Points Available QUESTION 4.17 Describe the actions that must be taken in accordance with the Loss of Heat Transfer Procedure, E.04, to initiate "HPI cooling." (1.5)

ANSWER 4.17 The answer must indicate knowledge of the general flow path, i.e., in via HPI, out the PORV. The full credit answer must state that SFAS channels IA and IB only are to be actuated with immediate manual control of RCP seals.

REFERENCE (S)

Loss of Heat Transfer, E.04, p. E.04-3.

-Section 4 Continued on Next Page-

77 Rancho Seco November 5, 1985

_ Points Available ,

QUESTION 4.18 What is the meaning of " bump" an RCP? (1.5)

ANSWE' L.lfi Start the pump. Allow starting current to die off. Allow the pump to run for 10 seconds and then stop the pump.

REFERENCE (S)

Loss of Heat Transfer, E.04, p. E.04-6.

-Section 4 Continued on Next Page-

78 Rancho Seco November 5, 1985 Points Available

^

- 00ESTION 4.19 What must be done prior to throttling HPI flow below 105 gpm per pump? (1.0)

ANSWER 4.19 The HPI miniflow valves must be opened.

REFERENCE (S)

Loss of Heat Transfer, E.04, p. E.04-10.

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-Section 4 Continued on Next Page-L

y, i

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79 Rancho Seco November 5, 1985 Points Available QUESTION 4.20 The guidelines for terminating HPI given in " Rule 2 - HPI Flow control" of the emergency procedures does NOT apply for one (1) specific situation. Enr what situation does this guideline not apply? (1.0)

ANSWER 4.20 When core cooling is being provided by HPI cooling.

REFERENCE (SI Rule 2 - HPI Flow Control, Emergency Procedures.

. - End of Section 4 -

- END OF' EXAM -

-Section 4 Continued on Next Page-