ML20214K474
| ML20214K474 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 11/06/1986 |
| From: | Lawyer S, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20214K447 | List: |
| References | |
| 50-302-OL-86-01, 50-302-OL-86-1, NUDOCS 8612020317 | |
| Download: ML20214K474 (145) | |
Text
,
ENCLOSURE 1 EXAMINATION REPORT 302/0L-86-01 Facility Licensee: Florida Power Corporation Facility Name: Crystal River Unit 3 Facility Docket No.: 50-302 Written examinations were administered at Crystal River Training Center near Crystal River, Florida. Oral examinations were administered at Crystal River Unit 3 near Red Level, Florida.
Chief Examiner Wwffeg$W ///S/26 Date Signed
/gpifdy Lawyf f Approved by: / w[# ///4/E(.,
Date Signed c,Jopff F. MLK,ption Chief Summary:
Examinations on August 19-22, 1986.
Oral examinations were administered to seven RO candidates, all of whom passed.
Written examinations were administered to seven RO candidates, six of whom passed.
An oral and a written examination were administered to one SRO upgrade candidate, who passed. A written requalification examination was administered to one SR0 licensee, who passed.
Based on the results described above, six of seven R0s passed and two of two SR0s passed.
8612O20317 861113 PDR ADOCK 05000302 V PDR b -
- :. I." -
, 9'- ,.
REPORT-DETAILS 1l Facility Employees 1 Contacted:
- Ray Wittman, FPC Operation Support
- Bruce J. Hickle, FPC. Nuclear Plant Operations Manager
- Marsha Watson, FPC' Training
- James P. Owen, MEC/FPC' Training
'* Steve Blake,'FPC Training
- Paul McKee, FPC Director Nuclear Plant Operations
- Tom Stetka,-NRC SRI
- Attended Exit Meeting
- 2. Examiners
J. C. Huenefeld W. J. Apley
- Sandy Lawyer
- Chief Examiner-
- 3. Examination Review Meeting At the conclusion of the written examinations, the examiners provided Mr. Johnie Smith with _a copy of the written examination and answer key lfor review. The' comments .made by the facility reviewers are included as
' Enclosure 3 to this report, and the NRC resolutions to these comments are listed below,
- a. R0 Exam (1) Question'1.05 NRC Response:
' Agreed. The answer key was changed accordingly.
(2) Question 1.13 NRC Resolution:
If the proposed assumptions are clearly stated by_ the candidate, the proposed answer will be accepted. While neither of these cases is -specifically addressed by the reference, the general theory needed to analyze and explain this situation is covered by this and other references. No change required.
r -
2 (3) Question 1.14 NRC Resolution:
A reactor operator with an acceptable level of._ perception would
' easily ascertain the vertical line. No change required.
(4) Question 1.15 NRC Resolution:
A wording change was made in the answer key to better reflect utility terminology.
(5) Quest m 2.05 NRC Re=olution:
Based upon the reference material supplied, the additional answer,
" oxygen", was added to the answer key.
(6) Question 2.12 NRC Response:
The comment, technically speaking, is correct. However, none of the material supplied identified an alternate set of engine "name plate" ratings. Rather, most of the material supplied consists of a " summary of normal operating" data. We will, however, accept the additional answers "3250 Kw - 168 hour0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> maintenance interval" and "12,433 cu. in." as stated in the supplied material. The answer key was changed accordingly.
(7) -Question 2.14 NRC Response:
Facility comment accepted. The answer key and reference were changed accordingly. The facility material should be corrected prior to future use.
(8) Question 2.18 l NRC Response:
l The question as written strongly implies an AUTOMATIC Actuation.
l- The very act of questioning which feedwater valves go shut implies an AUTOMATIC main steam line actuation. A knowledgeable operator will have no problem inferring that the isolation is automatic.
No change required.
I l
i.
J
3 (9) Question 2.21 NRC Response:
Agreed. The answer key was changed accordingly.
(10 Question 2.23 NRC Response:
Agreed. The answer key was changed accordingly.
(11) Question'3.6 NRC Response:
The operation of control board switches is considered to be of paramount importance. No material was supplied to. support the -
facilities ~ contention. regarding knowledge -of control board switches. No change required.
(12) Question 3.19 NRC Response:
Agreed. The answer key was changed accordingly.
(13) Question 4.2 NRC Response:
Agreed. The reason was put in parentheses to indicate that it is not part of the required answer.
(14) Question 4.3 NRC Response:
Agreea. The answer key was changed accordingly. The original reference material should be updated to include opening MUV-103.
In addition, previous response #5 was put in parentheses since it is not required in order to feed.
(15) Question 4.3 NRC Response:
Agreed in so far as the additional limit is concerned. The answer l-key was changed to incorporate the 1000 mrem / week.
4 (16) Question 4.11 NRC Resolution:
It was noted prior to final examination grading that other-conceptually equivalent RCS leakage indicators are available to
'the operator. These were added to the answer key.
(17) Question 4.13 NRC Response:
The candidate was asked what the Verification Procedure states and therefore should respond as the answer predicts. However, as stated in the material supplied, 65% will be accepted. The answer key was changed accordingly. The procedural error should be corrected promptly.
(18) Question 4.11 and 4.13 NRC Response:
Memorization is not required. A conceptual knowledge of the intent of the emergency procedures is all that is required to answer these questions. No change _ required.
(19) Question 4.19 NRC Response The required knowledge in this case was tested (multiple choice question) utilizing the candidates cognitive faculties as opposed to memorization. Therefore, it is for testing purpose, moot as to where in the procedure the "important action" is located. No change required.
- b. SRO Exam (1) Question 5.02 NRC Response:
The utility " equal valid" arguments ignore the question's assumption that the generator is synchronized with the grid.
Under this condition, real load and output voltage are not proportional to excitation current. The facility attempts to explain that real load and cutput voltage are proportional by offering two explanations. Both explanations, as described below, are erroneous.
I l
V. . '; . .
5 I
-(a) .When' a sychronous machine is in parallel with a large grid, the - grid becomes the determining factor regarding voltage.
' Change in excitation, once paralleled, will have a reduced effect upon output /line voltage. Instead the excitation current will shift the amount of reactive load being supplied by or to the main generator to the other machines operating on the grid. Therefor output /line voltage is absolutely NOT proportional to reactive load.
(b) True, load power factor is essentially fixed and determined by the grid; however, .with one or more additional machines running in parallel on the grid the dynamics of reactive load sharing increase the complexity of the situation. In this situation (more than one machine in parallel) grid voltage will be more or less a constant. Attempts to change output voltage at one machine by changing excitation current will not change voltage. Rather, it will change the reactive load sharing between the paralleled machines. Changes in excitation will NOT cause appreciable changes tri real load.
Now, if real load is changed, in order to maintain a constant terminal voltage there must be an accompanying change _in excitation. This means that the converse of what the facility is saying is true, a change in real load will result in a change in reactive load. However, the ' facility's assertion is false; changes in reactive load may be made independently of real load. Additional reference can be found at Principles of Alternating Machinery, Lawrence &
Richards, McGraw Hill Book Co. , p. 366 and Electric Machinery by Fitzgerald, Kingsley and Kusho, We note that this question, in identical form, was utilized in the March 5, 1985 Crystal River SRO requalification examination without comment. No change required.
(2) Question 5.09 NRC Response:
The recommended response is equivalent to answer b on the answer sheet. This was clarified prior to grading.
(3) Question 5.22 NRC Response:
Strictly speaking, the utility proposed additional consideration is valid and a complete answer would have to include it in addition to the answer key answer. However, it is commonly accepted that the term " decay" (without reference to polarity of the charge) refers to the more common B minus decay rather than both. Therefore, the answer key answer (for B minus decay) will be accepted for full credit. No change required.
p.
6
.(4) Question 6.92 NRC Response:
The indicated conditions causing alarm C-3-12 are listed on page 8
-of AP-303. The R. C. Drain Tank Cooler is not one of the differ-ential' flows indicated. All three of the other differential flows are listed. . The utility submitted flow diagram indicates where flows are measured but does not indicate how the differential flows are arrived at. Since the newly submitted information does not refute AP-303 no change to the examination or answer key is required.
(5) Question 6.03
~The utility comment is correct and is additionally supported by 0P-603, p.2. The answer key was changed as recommended.
(6) Question 6.06 Choice b and the utility comment are incorrect in that the push -
button does not indicate loss of air to the E/P controller for MUV-51. We note that this question, in identical form, was utilized in the December 17, 1984 Crystal River RO examination without comment. No change required.
(7) Question 6.08 NRC Resolution:
Agreed. The answer key was changed accordingly.
(8) Question 6.18 NRC Resolution:
Agreed. The answer. key was changed accordingly. We note that a
'similar change needs to be made to Question Bank question 2.26.
(9) Question 6.21 NRC Resolution:
Agreed. The answer key was changed accordingly.
(10 Question 6.25 NRC Resolution:
Agreed that it is feedwater vice reactor power that is being changed. The answer key was changed accordingly.
r --
7' (11) Question 6.26 NRC Resolution:
Agreed. The answer key was changed accordingly.
(12) Question 6.28 NRC Resolution:
The utility comment was lacking in specificity. No particular examples of additional valve combinations were proposed. The additional material supplied was not specific enough to support either the comment or the recom'endation. No change required.
(13) Question 7.7 NRC Resolution:
The utility supplied reference gives >30 mr as a ~ suggestion.
RSP-101 -p9 states PIC's must be re-zerced prior to reaching 3/4 scale. The question asks for the requirement, not the suggestion.
Additionally, the procedure should be considered governing if there were a conflict between it and a GET lesson plan. No change required. -
(14) Question 7.8 NRC Resolution:
It is necessary that an SRO know these four indications that the frisker is operable. We agree that they need not be memorized.
Any wording which conveys the same meaning will be acceptable.
No change required.
(15) Question 7.11 NRC Resolution:
Agreed. The answer key was changed accordingly. ;
(16) Question 7.13 NRC Resolution:
We agree that operators should not be required to memorize procedure names and numbers verbatum. However, they must know some identifying feature of often used procedures by which they may communicate with their colleagues. Here, any description which uniquely identifies the Nuclear Services Cooling System operating procedure will be accepted. No change required.
t
7 , . .
8 '
-(17) Question 7.19 NRC Resolution:
Agreed. The question was deleted.
(18) Question 8.21b NRC Resolution:
During post examination review it was d'iscovered that an additional NSS responsibility was not originally included on the answer sheet. The. answer key was. changed accordingly.
- 4. Exit Meeting-At the conclusion of the site visit the chief examiner met with representa-tives of the plant staff to discuss.the results of the examination.
There were no generic weaknesses noted during the oral examination.
The cooperation given to the examiners and the effort to ensure an atmosphere in the control room conducive to oral examinations was also noted and appreciated.
The licensee did not identify as proprietary any of the material provided to or reviewed by the examiners.
4 f( &&& O k
, +. Mh.s un
. ,
- ee .
' .e--_e.Hq a
o r r.: .i . ' n.. o,-v
- s r.. n .v. y. c ccw 7. _ .
c, wc u., ys n: c. : r. r sn :. p r.ge,p:
i
- r. . .t. v. t. :. r_ Y r. ". T *w* T I 0 t?*
'a* r. T. '_ +. . .
4 r. N vrr*' .;ML 'L.%,'-
F *,. N - ' t %.2 ,i _
- . . ._ ;y2 ,. ' ^
, .* + r_ e, -n w .. v4 T e. r. : r r
- s g/./rq..n s J. .
y f..u - " .; . *.
_ r., -g y.r..: , :
- t , a, .
p w)on:'r: M A 577R v^ r. 4tlm. T.
- af e. -
. >s :_t r. T ] r.. y e.
. .. . . 9 4.,T. r
- Use s e r.,a r a t e .
, a r. e " *:r . ' - snswers. ' site ensue s c, one aide onl/. -
E q.e e,uestion sheet o r. top of the encuer sheets. c'oints fo- eet h c.uestion are indicated in .
,-entheses a rter the cuestien. The passinc .
etede
. i e c.u i r e s et least 70t ;" ecc5 cetecon - e r. C' r fine; evade c' r*
least 80%. Examinatio- .epers : ull'. be p ick ed up :" (6) hours afte*
th e ene.inction s t er-t s .
. ,, e CUEGORY % Or CANDI?A E'S CATEGO Y
- o. b- i .J.r l uTAt deu.raer_ v A ._ iJ L n T c_ r_.n e.s".
"3O'00 '
_-- _25*00 --__. - --___-_-- ---_-___ 5* 74r00Y - ~ Or Nur<eso onwee OPERATIONr rLUIDS, A N '-
- - - - u - ei AvT
- n -
T u e :,t. o~n y N g u. . t r. s2 g,g . .
., n , e' ' n
_ --_ --_-- ---_-- - f' Ct ~^NT CyCYrMC
.A
' - -- * . 7 O T r' N ,
< - - ' - a G gNT , g?'
og AND INSTRUMENTATION w
c.c. -a.oo oe
- pR0r ---
rn" r e --- W R'" A '- -- a N --D- ' " A ' * - -
r u,.r o.s-
_ . o_N r 'r. y Aun n. p A n T n' 0 r.> T. C Aa-r O N.T mo 0,.
n'- op ors *go g.
- A9aTy,3TeeT
' - - ' yr- -eqerenuqre,
- ~ ~ - --
/ / 8. 0 0 COND'TnNC- ann - -
I"77ATIONC er r e.
ro e .3,.
wfr*
- ne' G r- E d e All wo-W done en t 5 :. a e':.2 stion is t ,> own. I have neither :iven -
nCr reCe: Ved Cic.
_"a. p , r. *, r.' cw *e'e.
. C w 4.g.-*.- . -*.itr. c.
i 9
4 4 %
NRC RULES AND CUISELINES COR LICENSE EXAMINATIONS Durins the eds:nitt,etior of this eucnintt:ct; the followins rules apply:
- 1. Chea',ing an the enn r.a tion means a r- autenatic denial ch ycur application end could resu:t :n tore severe penelties.
- 2. Restroon trips are to be limited and only one candidate at a time may lecve. You must sveid cl2 conteets with enyone outside the enceinstion room to avoid even the appearance or possibility of cheating.
- 3. Use black ini- c- derk pencil en{y te facilitate leedble reproductions.
- 4. Print your n a n: s in the blanu. p-c"ided on the cover sheet of the examination.
- 5. Cill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
- 7. o rint your none in the upper right-hand corner o# the first page of each sectior c# the answer theet.
- 8. Consecutively rUnber each 2nswer,cheet, write 'End of Category __' as approprirte. etcrt each er:esory on e new page, write on2y on one side c# the paperr and write 'Lest Fase' on the lest answer sheet.
- 9. Number each answer as to category and numberr "or exampler 1.c r 6.3.
f10. Skip ct lerst three lines between each answer.
- 11. Separate :nswer sheets 'rce pad and place #inished answer sheets " ace down on your desk or teb2e,
- 12. Use abbrevictions only if they are cocmenly used in #acility literature.
- 13. The point vclue for.eceb questier, is indierted ;r perentheses efter the question and can be used ss a guide for the depth of :nswer required.
- 14. Show all calculetions, methods, or assumptions used te obtain en answer to mathematical problems whether indicated in the question or not.
- 15. Partici credit may be g i v e r. . Therefore, ANSWE; ALL PARTE Or THE QUESTION AND DO NOT LEAVE ANY ANSWER CLANK.
- 16. I# p e r t t, of the enca.instion are not c'ec*. as tc intent, ask questions c#
the exaniner only.
- 17. You tust cisn the statenent on the cover sheet that i r.d i c a t e s that the wo*k is your own end ye.. hrve not received or been siven E s t i s t t r, c e in completing the e : mination. Th;s must be done Ofter the exa't iration has been completed.
U
p P
g 4
- 28. '#.en you coeplete your e::eninetion, you shall:
- r. Assemb s your e::ca i net i er es follews: .--
t (1) T: 2n questions on top.
(2) E :: t - eidt - figures, tcblec. etc.
(?) Answer pases includine #iseres which I e part of the answer.
b, Tv n ie ycer copy of the e::stinction and e:: pages used te answer
- he er.cmirtation c'v e s t i o n .
- c. Tern in el2 screp ptpe - and the balence of the PEper that you dic not use for answering the questions.
- d. Leave the e>:cmination cret, es definec' b; the e::cmi ner . I# cfter leev:ns, you are found in this eres while the e:: amination is still ir progrest, your licenst: may be deniec' er revoked.
I e*,
r ,
J
- 5. THID9Y OF NUCLEA- '0WEr eLANT OPEPATICt!. FLUIDS, AND P t.C E 2 t
DUESTION 5.01 (1.00) .
Which c# the following sources ccn potentially introduce the largest (in standcrd ceb:e feet) amoent o# non-condensib2e gas 2nto the RCSi
. Zirc-water reaction,
- b. Core flood .enks.
- c. Pressurizer steam space.
- d. 100% friled fuel.
QUESTION 5.02 (1.00)
When the main generster is synchren::ed with the grid, excitatien cur-ent is proportionc! to:
- 3. reactive load (MVAR)
- b. ree! lect (MK)
- c. output voltage ( . V )
- d. generetor speed (RPM)
GUESTION 5.03 ,
(1.00)
Ouring a reactor startup, power is being raised above the point of adding heet (PGAM). Whi6h o' the following stetements is CORRECTO (Assume e linear reactor power increase to about 3% power)
- s. Since header pressure is 885 psig. Teve will not rise above the corresponding saturation temperature of 532 degF.
- b. Since the OTSGs are icw level limited end herder pressure i.s being mair.tained at 985 psis, Tave will rise and the steam temperature will tend to follow Th.
- c. With the header pressure being maintained at 985 psig, the OTSGs will r e tic i n et satureted conditions and no superhect will be added.
- d. Since the OlSGs are low level limited, the steam is superheated at zero power conditions and rises proportionally with power.
(****'* CATEGORY 05 CONTINUED ON NEXT DAGE *****)
=
- 5. THEORY OF NUCLEAR POWER CLANT CoERATIONr CLUIDE, AND DAGE 3 l
1MERM0 DYNAMICS
.- l
) '
90 eel!ON 5.04 f .00)
A seneral ule is often stated "doublins the ecunt rate halves the marsin to criticellity'. This is n,etheneticc11y stated by the equetion:
CR1/CR2 = (1-Meff2)/(1-Keff1).
Which one of the following striements is CORRECT concerning the cbove statement and equation?
- a. Both Me#f: cnd Keff2 heve to be less than 1.0.
~
- b. Equal changes in Keff result in equal changes in suberitical multiplication level.
- c. The equation only appronimates'the instantaneous change in count rate; once the equilibrium velve is reached, the count Pete will be higher,
- d. A cecend doubling o# the count rate will result in the "eactor beccming critical or supercriticci,
- e. The statement is a pp r o n i r.m t e l y correct but CR1 and CR2 are inverted.
QUESTION 5.05 (1.00)
Which one of the followins statements is CORRECT concerning the paralleling o# electrical systems?
3 Although it is desirable to have speed and phase position matched, it is much more importent to hcve volteses matched,
- b. If vcitages ,are not matched at the time the synchronizing switch is closed, there will be VAR flow from the lower voltage source to the hisher one.
l c. If the incomins machine is at synchronous speed but out of phase with the running bus when the br e el:.e r is closed, heavy currents will flow to either accelerate or retard the incomins machine.
- d. If the incoming machine is in phese but slightly faster then synchronous speed when paralleled, the system will tend to speed up the incomins machine t r. synchronous speed.
- e. If the resistances are not matched at the time the synchronizing switch is closed, heevy currents will #20. to tend to speed up the incomins machine to synchronous speed.
(rrrra CATEG00' 05 CONTINUED ON NEXT PAGE rrrrr) w
_d*-
. ./
p
- f
/
r A
u
^
- 5. THEORY Or NUCLEAR *0WER oLANT O P E r:A T I O N , FLUIDS. AND :' A G E *C 1
GUESTION 5.06 (1.00)
The reactor is at 70% power with the ICS in full auto. Power level is increesed to 80%. Choose the striement thc; be st describes whct happens to the shutdown margin. Assume sufficient rod worth to acecmodate the power changer cnd no baron chenses.
- a. Imnediately 'ollowins the power increase the shutdown margin will have ineteeseer appr o::i m e t ely five hours afte: tf.e transiente sh6tdown margin will be at its highest value and will itart decreasinb.
- b. Powe'- level has no effect on shutdown mersin.
- c. Immediately #cllowing the power increase the shutdown. margin will be unchanged, it will then decrease for appro;;inctely the first five hours following the transient, then start f r.cr e a s 1ns ,
- d. Imn.edirtely following the power inet ease ' the shuttewn mersin will.be unchsnsed, it will then increase for ap p r eniinately the first #im hours following the t r e r:t i e nt r then sttri decrectins.
QUESTION RS.07 (1.50)
Will the f ollowing paran eters increase. decrease or remain the saae'if you 3 ste-t with satureted steam and superheat iU isobarically?
- a. Enthalpy
- b. Tempereture
- c. Den ~..ty C. Pressure , ,
- e. Specific volone
- f. Specific weight QUESTION 5.08 (2.00)
Sescribe the heat transfer nechanism (type of-heat transfer) that is taV hs place et points : thru 6 on the attached 'eilios ; weter curve (fis 105').
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****'
w
^
s.
1* ?
t ge- 'e-4
- I. sTHEONY EO F NUCLIAR
- 50wER PLANT OPERATION, FLUIDS, AND PAGE 51
- 1HERMDDYNAMICE t
' 0 U E S ' I O t!' ' 5'.' 0 9 (1.00) -
List the two 'easons_why r the maximum linear heat rate is. limited.
. QUESTION e5.10 (1.00) .
According to the bases-for the linitins safety system settings, which of the'RPSLtrips will be reached first durins a slow reactivity insertion
~
accident from low-or high power?
LGUESTION 5.11 (2.00) .
Assume the followins. plant conditions enist' .-
Controlled cooldown of 210 desC/hr in progress usins both OTSGs RCS at 525 desF RMA-12 in.high riarm
- a. ' Determine the plant transient in progress. (0.5)
- b. List theEthree plant conditions, any of which make a cooldown. rate of 210 desF/hr acceptable durins-this transient. .1.5)
( o sc:es
~4 l' ^'
GUESTION 5.12 (1.001 The' limit imposed on' quantity of aluminum in the containment buildins is p based onlminimizins post-LOCA hydrogen seneration. What twe compounds.are-necessary in addition to aluminum to produce this hydrosen?
e GUEST' ION 5.12 (1.50)
. What temperature instrunent responses (instrument and direction) best
- indierte.the' presence. of natural circulation if OTSG pressure is decreased-according to plant verification procedure VP-580? Assume the reactor is in-hot standby with no reactor coolant pumps running.
L 4' '(***** CATEGDRY 05 CONTINUED ON NEXT PAGE *****)
1
=U l..
f
, a A
M -:
- .};~ ' C N'_,
~
is. .
i'g \ , '?, ,
z.,
,r- .
- - 'J
_4, f
j ; s,-
g m..
& W ,
- i
.t
- b. .
'5. THEORY 0F NUCLE AR POWER PLANT OPERATIONr FLUIDS, AND PAGE, 6 1
t 0UESTION .5.i4 -(2.00) y The' minimum' temperature for'eriticality LCO ensures that-the reactor will-Jnot'be'made eritical.with:the reactor coolant systet,everase-temperature
.less than 525 desF. WhatJare the four things ensured.by'this-limitation, -
- ie r what.is the basis-for this LCO?-
'00ESTION' 5.15' (: .50)' i 7
In the event lof a rod. ejection' accident, which will be the first reactivity coefficient to' insert negative. reactivity?
- 0UESTION 5.16 (1.00) :o 4 Why does OP-404 1 Decay Heat Removal System, require the operator:to open the. drop line'to the RB sump within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; after a LOCA? -Include;'in your treason'the RCS P-T condition'under which this{ would.be:rnost imPortant.-
~
DUESTION 5.'17 (1.50)
-About what. percentage of the total neutrons in the core at 'B'OL2a're' delayed
~'
ineutrons?- Explain how such e small'percentase'of neutronscang~f['ec.t'the-loperation of the reactor'so much. Assume-a~new core.
.- s ,,
. QUESTION 5.18 (1.00)
. Describervery-senerally and'qualitativelyr the difference between the.
,behavict - of xenon concentration and of samarium concentration when takins.
, .the.p:.cnt from shutdown to 1:ull-power followins an ext. ended full power run.-
QUESTION- -5.19 (1.00) <
Compare.a power change from 10% to 30% to a power change from..30%-to 50%.
'Is-the1samefac.ount of reactivity' required for both char,ses? Explain your answer.
(*****-CATEGORY 05' CONTINUED ON NEXT PAGE *****)
.g
\
/
4 Y
T
~..,+.-r-*e i--w e -,-v.e e w.,m.s-mire. m r - - + ~ - -, w w ----a, - - . , , , - -
t
+_ ,..
3s.
I1__ _____ __ _ $_ ~_ _ $_ _
THERMUDYNAMICS___ _ ____ _____ ____
i v .s
- GUESTION S.20 -(1,00)
- ec .
44f?,the plant.is heated up to normal operating temperature / pressure from the recctor-being in a. cold shutdown condition without any change in rod position'or baron concentration, what will. happen to the-shutdown mergin?
' Explain.- -
QUESTION _5.21 (1.00)
Describe the photoelectric ~effect as it occurs in certain radiation Lmonitorins instrumentation in terms.of the detected ray or ~ particle, how it
-interacts, what it interacts with, and what causes the subsequent ionication.
QUESTION 5.22 (1.00):
Describe the changes that occi in_the atomic number and'nass number.of an atom when it. decays by bete emission. Remember the atomic numbe.- is the
' number of protons and the mass number is the number of neutrons plus
. protons. .
. i,luj7 QUESTION M4. 23 (1.00)
E'xplain.the effect on.MTC and fuel temperature coefficient as the core ases (fuel-depletion). Note whether they become more or less negative and_the creason for'the change.
QUESTION 5.24 (2.00) '
- a. Sketch a trace of losarithmic neutron countrate versus time followins i a-reactor' trip. Label the axes of your trace, and clearly show the effects of prompt and delayed neutrons. Label this curve a.
- b. .$ The-trip in part-a inserts a value of reactivity much greater than Beta (delayed neutron fraction). Now, on the same sketch prepared for part a. . sketch a trace of eactor power versus time for the case where the_ reactor is made to be soberitical by an amount much less
-than Bete. Label this curve b.
s .
k'
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
?
5 th I
i J
a
- 5. THEORY OF HUC'_E AR POWER SLANT OPERATION, FLUIDS, AND PACE 8 t
DUESTION 5.25 (1.00) .
The speed o c : centrifusal pump is decreased to half of its ititial value.
Given the following initisi eenditions, what are the final conditions?
- a. Fluid horsepower 2 5 '-! P
- b. Flow 45 spt
- c. Head 250 psi, O
T p;; ;
- w. . t. .
(*rrxr END 07 CATEGORY 05 *****)
- t1
e - m i' .
1
'<*s
- 16.1 PLANT' SYSTEMS DESIGNr CONTROL,'AND INSTRUMENTATION PAGE 9
- 0UESTION 6.01. - ( 1. 0 0 ) I Which_of.the1.followins' statements about RB purse control is NOT true?
s.- Both purse supply fant must be operating'to permit start of1the exhaust. fans.
- b. Exhaust duct-temperature greater than 135 de3F will shut-down the exhaust fans.
- c. Dampers.D - 93 and 94 autom'atically adjust to maintain vent: flow rate about 50,000LCFM when' purse valves are open and fansnare operatins.
- .'d . : Purge valves'are. automatically closed by a HIGH radiation alarm, buti supply and exhaust fans continue ~operatins.
QUESTION 6.02 -(1.00)
Which=one of the followins is NOT.an indicated conditio.n of annuncistor
.C-3-12, . *NS CCC SYST.RB LEAM'. ,
- s. differential flow to AH units
- b. differential flow tc letdown coolers
!dk -c. differentia 1' flow to RC drain tank cooler b* , d . - differential flow to RCPs- '
QUESTION 6.03 (1 00)
Select the INCORRECT statemen't resarding the condensate injection system.
- a. Condensate injection is used for main turbine head. sprays in'the hish pressure turbiner pump seals in the feedwateer system, and valve steam sealing to prevent in-leakage of air to the condensers.
~
- b. When condensate pressure is above 220 psis the condensate pumps are supplyins seal and spray water.
- c. If the discharge pressure decreases to ~200 psis, the G.W.P. that has been selected will start automatically, d.~ The condensate injection system supplies water to the desuperheaters in the auxiliary steam system, steam supplies to both evaporators, and the sland steam system.
(***** CATEGORY 06 CONTINUED ON NEXT PAGE ***xx)
.m I
)
l
.]
,- w
-6.. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION DAGE 10 QUESTION 6.04 (1.00)1 5_
'Which of the.folloUins is NOT centrolled.by the *eedwater control.: subsystem
.in.the ICS?
- a. reedwater ratio distribdtion between the DTSGs.
b.- .Totel feedweter:floy besed on unit load,
- c. BTU availability to the OTSGs.
d.- ' Level coriditions-in the.0TSGs.
QUESTION 6.05 :(1.00)
The EFIC system will automatically initiate EFW vpon detection of'
- c. Loss'of both main feedwcter pumps and reactor' power->10%. ,
1 5.: Low level-(<24') in either.0TSG.
- c. Lossoof~cil RCPs.- I
- d. . Low pressure (<750 psis) in either OTSG,
~
QUESTION 6.06 ,
.(1.00)
Which one of the.followins is CORRECT concerning the ' air fail reset" pushbuttons for MUV-16, 31 and 51.
a- The pushbutton only indicates loss of air to the associated valve E/P controller,
- b. The pushbutton indicates loss of air to E/P controllers for MUV-16 and 51 andTalso loss of ct to the velve positioner for MUV-31.
- c. On loss of air supply, the solenoid valve supplying air to the air lock velve will'decnergire, causinS the effected valve (16, 31 or 51) 8 to close.
d.. When air pressure has increased, depressing the air fail reset pushbutton.will unlock MUV-16, 31 or 51.
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
- W
,. -i
- 6. . PLANT. SYSTEMS' DESIGN,' CONTROL, AND INSTRUMENTATION DAGE 11 lQUESTIONE 6.07 (1.00) i_
Which of_the following is1 the reason that the reactor contro1 subsystem low limitt reactor demand ete10%.
- a. ' Inability of.the centrol circuits to smoothly control reactor power due'to,the smell.masnitude of the: signals involved. ,
.b.: ~ Prevent the. reactor- control subsystem from ramping Tave automatically-and without the control board operetor's'knowledse.
- c. Due.to decay heat considerations.
d.- Prevents the reactor control subsystem from ramping lave automatically without the operator's knowledse and provides a minimum demand for decay hect considerations.
QUESTION 6.08 (1.00)
To start any circulatine water pump, four. start perr.issives.must be sctisfied. Name cl1 four and give their setpoints as appliccble.
QUESTION 6.09 (1.00)
.4, q.
List'eight, separate systems or components that receive their -L E}
the auniliary steam system. supplyfrostp[IRij JL4 QUESTI01 6.10 (1.00)
Identify the abnormal reactor condition which has occurred or is occurring' L as indicated by the PT trace on the etteched figure 1076.
QUESTION 6.11 (1.00)
Where, as a general ruler should the. speed droop control be set for a D-G unit when running alone?
QUESTION 6.12 (1.50)
List the five EFIC channel maintenance bypass function interlock features. [
(***** CATEGORY 06 CONTINUED ON NEXT PAGE xxxrr)
- t.5
s 1-
- ia .
i 6.- PLANT SYSTEMSfDESIGN, CONTROL,1AND INSTRUMENTATION PAGE~ 12 10UESTION '6.13 (1.00) i_
Name.three types of failure-mode fer which the fuel.handline<3rapple
'cechanisc. is protected when the-svapple.is loaded.
QUESTION 6 14 (1.00)
'What is.th'e primary fire protection-system protectins the emersency diesel .
genera, tors?
GUESTION 6.15 (1.00)
What are your immediate actions in the event of actuation of' a .- ~RMA-3?
be RMA-4?
OUESTION 6.16 (1.00) from where will the megalatt calibrating integral receive its input if the main turbine is in menval and all other ICS stations are in automatic?
- =ci t
-GUESTION H6.17
~
(1.00) -
/ f,ib
~What,is the normal method used for reducing RCS sas concentration?
GUESTION 6.18 (1.50)
The decay heat removal system is protected from overpressurization.by an automet'ic closure initiation (ACI). List the valves monitored by the ACI featurer the setpoint for actuation, and the automatic actions that will occur. .
QUESTION 6.19 (1.00)
What-are the only heater drain valves with control switches on the main '
control board?
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
= .'2
= '
n
- 3.-
- 6. PLANT SYSTEMS DESIGN,' CONTROL, AND INSTRUMENTATION PAGE 13
- QUESTION 6', 2 01 (1.00) i-List and~ explain the' difference between the.two types of monitors (two functional-ereas) that meke up,the loose parts monitorins system.
QUESTION 6.21. 1(1.00)
Discuss the arming of the.RBS by stating the armins signal and the automatic. actions,that occur to accomplish the arming.
DUESTION 6.22 - ( 1. 0 0 ) -
Describe the location.in the makeup and purification system where the letdownfline connections.to the' decay heatfremoval system connect.
QUESTION . 6. 23. (1.50)
.Whatfdo you1 expect the status of:the amber lamp (DG parallel block act) and
. breaker 3210 to be if breakers 3209,3205 and-3206 are closed? What can you say.of the' interlocking of breaker 3?10 under these conditions? What is the purpost of this interlock? See figure 859 attached.
- lt i ?
QUESTION '6.24 (1.00)
.f-.-
What=is'the. purpose of the clamp / clamp release switch on the diamond pane 19 OUESTION 6.25 (1.00)
Describe the control function of the ICS neutron cross limit.
QUESTION 6.26 (1.00)
Name and give the location of the power' supply for the post accident samplins system.
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
.e
_; % : T is. = -
'6.- PLANT SYSTEMS'DESIGNr CONTROL, AND. INSTRUMENTATION PAGE 14-4 QUESTION 6.27~ '(1~.50) cWhat reactor moder steamins' condition and time-limit' determine the'3.inimum -
condensate water:stcrase tank. volume?
.. QUESTION 6.28- (1.00)
. Under-what Positioning of' the sovernor and/or throttle valves are:the OTSG
' steam headers cross connected?
@Na .
i
' ;,: 1 5 .
, , k ' 'S
.it . V*,
l: -
(***** END OF CATEGORY 06 *****)
l l
l-I' I
=
.=
I l
?
i
e J .
s
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 15
~~~~R5656EU6EC5E~66UTR6E~~~~~~~~~~~~~~~~~~~~~~~~
t GUESTIDN 7.01 (1 00) .
Which of the followine conditions is a procedural requirement for manually trippins the reactor?
- a. Emergency feedwater actuates.
- b. Subcooling marsin dropt below 50 desF durins power operttion.
- c. Shutdown margin is determined to be less than 1.0% delta k/k.
- d. Feedwater flow is lost.
QUESTION 7.02 (1.00)
Accordins to the RCP operation procedure (DP-302), which of the followins statements it CORRECT?
- a. During cooldown, followins transfer to the decay heat (OH) system, the RCPs should be sequentielly stopped, about 5 seconds apart.
- b. RCFs may be operated in an emergency without seal injection flow provided NSCCCW is in service, a
- c. If RCP start permissives are bypassedr thej manimum allowsbile reactor power for startins the fourth RCP is 30Wi!4 i' -
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
- 7. PROCEDURES - NORMAL, A E:N O R M AL , EMERGENCY AND PAGE 16
~
~~~~EA65EEU5ECAL CUNTRUL~~~~~~~~~~~~~~~~~~~~~~~~
)
DUESTION 7.03 (*.00) -
Which one of the follouing is correctly stated regarding a permissible dose to an operator in e restricted area as specified in 10CFR20?
- a. Under non-accident conditions, the operator is permitted to receive no more than 7 1/2 rems per calendar quarter to each hand and to each foot. ,
- b. Under non-accident conditions, the operator is permitted to receive no more than i 1/4 rad of beta per calendar quarter to the lens of the eye.
- c. Under accident or emergency conditions, the operator is permitted to receive up to 25 rem once in a lifetime exposure.
- d. Under accident or emergency conditions, the operator is permitted to receive up to 100 rere once in a lifetime exposure.-
QUESTION 7.04 (2.00)
Answer the following TRUE or FALSE according to OP-412 ' Waste Gas Disposal System '
tR:);?
- o. An increase on a portable radiAdon detector (such as an E-102) is used to indicate that all, water has been drained from the waste cas surge tank drain pot,
- b. The " operator at the controls' is responsible for verification of the radiation monitor setpoints as they are specified on the release permit. ,
- c. If meteorological conditions show a delta temperature of zero or positive, you are NOT allowed to proceed with a gaseous release.
- d. If the flowrate nonitor is inoperable, tech specs allow continuing the gaseous release assuming the flow rate is estimated as specified.
QUESTION 7.05 (1.00)
According to OP-210, ' Reactor Startup', an RCS sample should be taken periodicaly during a certain evolution. What is the evolution and what is the sample analyzed for?
~
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
1 i
l l
j
p- .
I
' ^
a.
F I' .7. PROCEDURES - NORMAL, ABNORMAL,' EMERGENCY AND PAGE 17
~~~~RE6EUL5GIEst E5sTE6L t
GUESTION 7.06 '(1.00). -
Roactor coolant pumps have been lost because of a loss of offsite power.
Plant . control is beins maintained in accordance with the "Netural Circulation
- procedure, AP-530. According to this procedure an OTSG 1evel L
sotpoint of 95% should be selected under two conditions. . What are the two.
-conditions?
QUESTION 7 07' (1.00)
What are the three intervals (conditions) at which the PICS must be ceroed?
QUESTION 7 08 (1.00)
List the.four observations that " Basic Radiological Saf.ety In#ormation and Instructions for Radiation Workers', RSP-101, requires prior to friskins,.
t Refer to figure 1101 for the steps which immediately follow these observations.
QUESTION 7.09 (1 00) 110 '.
Qi y.. '
( According to the OSIM, dur ing ant"ne~r'sency , when is it permissible to use i the PORV to prevent a hish pressure trip? (2 conditions). ,
OUESTION 7.10 (1.00)
- Accordins to FP-203, what are the only two highly radioactive-components that are allowed to be temporarily stored above the seal plate during fuel or core internals handling operations?
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
f l:
l f
c ,
.. . 1
- 7. PROCEDURES -TNORMAL, ABNORMAL, EMERGENCY AND PAGE 18
~ ~~~~~~~~~~~~~~~~~~~~~~~~
~~~~RE656L6656dL 66NTRUL t
r
- QUESTION 7.11 '(3.00) '
. a '. . -List 1the immediate actions that are common-(identical) for the four runbackLAPs:
AP-540 Loss of booster pump AP-541 . Loss of main feed pump AP-542 Asymmetric rod- ,
AP-543 Loss.of one RCP
- b. List the additional immediate actions for AP-541 ' Loss of Main Feed
.Fump Runback'.
- c. Match the runback condition with the approximate power level-at which you expect the plant to stabilize i AP-540 a. 50%
ii AD-541 b. 55%
iii AP-542 c. 60% .
iv AP-543 d. 65%
- e. 70%
- f. 75%
- 11 '
.fjtiijf GUESTION 7.12 (1.00) TUT 4ite What are the 5 immediate operator actions in response to a confirmed fire in the plant according to AP-880?
QUESTION 7 13 (1.00)
Where would you expect.to find the correct procedure to transfer cooling water for the 3A and 3C makeup pumps from the DC system to the SW systemi te,-in what operating procedure (s)?
-QUESTION 7.14 (1.00)
List your immediate actions in response to an alarm from RM-L2.
GUESTIGN 7.15 (2.00)
List your immediate actions for AP-990, ' Shutdown from Outside Control Room *.
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) -
- M
c 27+ PROCEDURES - NORMAL,' ABNORMAL, EMERGENCY AND~ PAGE '19
~~~~Rd6E6E6EfEkE~E6UTEEE~~~~~~~~~~~~~~~~~~~~~~~~
t QUESTION 7.16 (1.00) -
According to OP-204, " Power Operation", if durins a' power-decrease ~, either OTSG.soes.on low level controle what is the appropriate operator action?
f-GUESTION 7.17 :(2.00)
.Ensineered!safesverds has actuated due to low RC pressure.-
a .. 'When are you' required to stop the reactor coolant pumps?
b .' under.what conditions (according to the remedial actions of AP-380) should the RCPs be kept. running?-
QUESTION 7.18 (1.00)
'Durins. plant heatup the fourth RCP is started at 500desF. What is the recson for waitins till we reach:this temperature (other then' waiting for the pump interlock to clear)?
' QUESTION 7.19 (1.00')M '
g .OP-203 cautions against '.
time.
ins both OTSG What is the reason for this' caution?
1evel transmittersiat, t'he same QUESTION 7.20 (1.00)
Why is the affected DTSG isolated at less than or equal to 540desF per.
ED-390, "0TSG tube rupture'? '
QUESTION 7.21- (1.00)
According to the ATOG svidelines, excessive use of HPI during a tube rupture can cause et least three problems with plant control. Name all three.
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
. n.
L
,. ~+
l7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY ~AND PAGE 20
~ ~~~~~~~~~~~~~~~~~~~~~~~~
~~~~Rd656L665CdL CENTR 6L i_
GUEST 10N 7.22 _ (1.00) -
What is the reason for'the caution ~'Use of temporary lea'd blankets must not be ellowec on any pipins or. components without. nuclear plant engineering
- cpproval' in CP-113, Handlins and Controllins Work Requests'and Work Packages?
QUESTION 7 23 (l'.00)
RSP-101 provides a list of items which identifies a **high radiation area"
(>100 mr/hr but < 1000 mr/hr) and another list which identifies a 'hish radiation area" (=or>1000 mr/hr). List the differences between these two lists and identify which cres'each applies to. Do not list any items which ove common-to the two lists.-
QUESTION 7.24 (2.00) ,
An actual plume release has occurred with a projected dose to the General population of 15 Rem whole body and 100 Rem thyroid. If the wind direction is constant from SSE, answer the followins for the general population. See fisurer 1105j[B,'andCattached.
- a. Between"Of di5 miles, which sectors should evacua'te?
- b. Between 5 and 10. miles, which-sectors should evacuate?
- c. Beyond 10 miles, which sectors should evacuate?'
- d. How lons will it take to evacuate handicapped, elderly and hospitalized individuals from within 5 miles con,sidering that normal conditions are present.
(***** END OF CATEGORY 07 *****)
- U
~ _ -
~
t 8.'_ ADMINISTRATIVE PROCEDURES, CONDITIONSr AND LIMITATIONS PAGE 21 QUESTION. 8.01- -(1.00) I_
While conducting a cooldown in Mode-5r if both diesel generators become inoperable
- c. no technical specification action statement is entered.
~b. technical specifications require that immediate actions should be
'taken'to establish Mode 6.
- c. technical' specifications require that positive reactivity changes be
. stopped.
-QUESTION 8.02 (1.00)
Which of the following may proceed given that a technical specification action statement has been entered requiti,ns you to ' suspend all core alterations'?-
D. Removin3 a neutron source from the core or positioning the auxiliary
,y neutron detector.
-..et
- b40jf usin3 the bridge in the core is allowed, provided that the low load EifMf61imit is jumpered out.
- c. Control rods and . burnable ' poison ' rods may be shuf fled ac long as Keff eor < 0.95,
- d. Completion of the movement of a component to a conservative position.
QUESTION 8.03 (1 00)
According to the technical specifications, while operating in modes 1 thru 3, the maximum level allowed in the OTSGs on the operating ranse is:
- a. 83%
- b. .87%
- c. 96%
- d. 98%
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
- *4.5, I
i
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS- PAGE. 22 4
QUESTION. 8.04 -(1.00) i Which of the followins statements is correct concernins-the Ovadrant Power Tilt-(DPT)?
=a. If GPT exceeds the maximum limit of 20.0, the reactor must be immediately shutdown.
- b. If misalignment of a control rod caus'es the GPT to. exceed the transient limit, thermal power mu,st be reduced within 30 minutes.
- c. 'No action is required within one hour regardless of the OPT limit exceeded (steady state, transient or maximum).
- d. If OPT exceeds the steady state limit, but is~less than the transient limit, operation may continue indefinitely only up to.60% allowable for the RCP combination.
QUESTION 8.05 (1.00)
An event in process.which indicates a potential degradation of the level of cafety of the plant is classified as
_,._., an unusual event
.d'
.g- *gIlan alert
.- a. site emeP3ency
- d. a general emergency 9
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
- M
'8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 23 QUESTION 8.06 (1.00) A_
The attached figure 1132 is_the Crystal River technical specifications
' bases figure 2.1 Which one offthe,following statements is correct concerning these curves?-
- c. These curves represent the conditions at which a minimum DNBR of 1.30 is predicted at.the monimum'possible thermal power for the number of reactor coolant pumps in operation.
- b. For each of these curvest a pressure-temperature point below and to the right of the curve would result in a DNBR greater than 1.30.
~
- c. These curves include the potential effects of the ejected contro11 rod and reactor coolant pump locked rotor accidents.
- d. The DNBR curve for four pump operation is more restrictive than any other reactor coolant pump situation.
QUESTION 8.07 (1.00) .
Only one of the following conponents has a manimum baron concentration specified in technical specification for the operational mode indicated.
Which one?
- a. Pressuri=er - mode 2 j,i-
- b. Concentrated boric'pcid storage system - made 3
- c. Core flood tank - mode 4
- d. BWST - mode 5
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
.=
B. ADMINISTRATIVE oROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 24 GUESTION 8.08 (1.00) i_
DG Ar which supplies ES 4 KV bus A is inoperable. LPI pump B supplied by ES 4 MV but B is inoperable. Which statement below is correct concerning continued operation in mode l? The tech specs for ECCS subsysten.s and AC sources are attached as figures ll34A,B and C.
- c. The action statements for both the LPI pump and the DG are applied independentlyr each must be restored to operable in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
- b. Since the DG is required in mode 4 and the LPI pump is not, the unit must be taken to mode 4 within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
- c. TS 3.0.3 applies; it requires action to place the unit in a mode in which the specificaqtion does not apply within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
- d. TS 3.0.5 applies; it requires action to place the unit in a mode in which the specification does not apply within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
QUESTION 8.09 (2.00),
- c. What is the only condition under which the designated ' operator at the controls
- can leave the red-carpeted general area for non-emergency purposes?
- b. In the event of an emergency affecting the safety of operations, the
' operator at the controls
- may momentarily be absent from the general area in front of the control board provided ... (complete this sentence).
QUESTION 8.10 (1.00)
Consideration should be siven to shutting down the plant and invest 13ating the leakins OTSG tube (s) when primary to secondary leakage through the cteam Senerator exceeds the value specified by the OSIM. What is the value of this leakage and how lons does OSIh predict before tube rupture would occur?
GUESTION 8.11 (1.00)
State the requirements for correcting an error in the ' narrative los'.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
. -g
(
l l
l l
+
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 25 90ESTION 8.12 (1.00) I_
'Durins power operationr OP-204, the Power Operation precedure requires that
'the Chem Rad department be notified of power changes. What is the lower limit on power and time for which this notification must be made?
QUESTION- 8.13 (1.00)
What is the m,aximum number of main steam code safety valves.that may be inoperable in Mode 3? What actions must be taken if this many valves are inoperable?
DUESTION 8.14 (1.00)
OSIM lists seven systems requiring prior approval before performins maintenance on them such as RPS and ESAS. List the other five.
QUESTION 8.15 -
(1.00)
Durins a plant startup with the reactor at 2% powerr one power ranse channel, NI-6, is found to be inoperable. Whot are the only 2 restrictions on proceeding to 100% rated. thermal power? Refer to the attached TS on fi fisures 1129ArBr.C.' QUESTION 8.16 (1 00) - According to the OSIM, what is the preferred method of locating condenser J tubs leaks? Explain why. ,' QUESTION 8.17' (1.00) 5 When are the protective actions as specified in the protective action , suidelines warranted? ie, how do you judge whether or not to recommend ! the protective action be taken? j (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
- 8. ADMINISTRATIV
E. PROCEDURE
S, CONDITIONS, AND LIMITATIONS PAGE 26 DUESTION 8 18 (2.00) i_
.CP-115, "In plant Equipment Clearance and Switchins Orders'. states four conditions that-require PRC approval of a. clearance prior to issuance.
List these four conditions. QUESTION 8.19- (3.00)
$fter usins the PORV to prevent a reactor trip from 100% power on RCS pressure - high, indications lead you to suspect that the automatic reset of the PORV is out of calibration at 0200 hours. You request a calibration of the PORV lift setpoint and reset setpoints to verify PORV operability.
The calibration procedure requires removing power to.the PORV's solenoid valve which occurs at 0230 hours. Answer the following for this situation. See Technical Specifications attached. (figure 1120)
- a. By what clock time must the block valve be closed to avoid beins in Hot 7tandby within 6 hours? Explain your decision.
- b. Attempts to'close the block valve at 0245 hours are unsuccessful. IRC technicians determine that the pressure bistable n.ust be replaced which'may take up to 8 hours. Since the PORV is already closed and power removed from the solenoid, can operation in Mode 1 continue?
oi Explain. .! " ( *".y . i c. While in Hot Standby, the pressure bistable is replaced, thus ?j'"' returning the P08V to operability; however, the block valve remains inoperable in the open position. Can the unit return to power operations? Explain your answer. (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) 0 er
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 27-GUESTION 8.20 (2.00) '
Consider the table concerning RCS' leakage past CFV-1 during e reactor Startup and answer the questions below. Technical Specifications are Ottached. Figures-1121A,B and C. Time- Power CFV-1 leakage 010'O hours- 15% 0 spm
,0600 hours 50% 2 spm 1800 hours 80% No Measurement made 2300 hours 100% 4 spm
- a. Do the technicial specifications allow the unit to be'at 80 percent power at 1800 hours? E:: plain your answer.-
- b. During shift turnover, at 2400 hours, you are handed the 4 spm leskrate measurement. What actions are required by the technical specifications?
'0UESTION 8.21 (2.00)
When a reactor trip or plant shutdown occurst the Nuclear Shift Supervisor first ensures that the plant is placed in a safe condition by having the nqcessary operations performed in accordance with approved pr.6,c] es. a.- He then makes 5 notifications: List theni '4""
- b. Explain the NSS's responsibilities concerning form 912212, Reactor
. Trip and Shutdown Report.
QUESTION 8.22 (1.00) Dafine 'short term instruction
- as'used in AI-500, ' Conduct of Operations'.
When do they expire? Who har the responsibility to audit short term instructions? l GUESTION. 8.23 (1.00) Define an Immediate Temporary change as used in AI-401, ' Origination of and Revisions to P00AM Procedures't by comparison to an Interim change. (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) l
'M L
I { t.
r
.o 8o ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PA"E 28 QUESTION 8.24 (1 00) i_
In any case where a general emergency has been declared, what is the minimum protective action recommendation?
>g ; i
'thk
(***** END OF CATEGORY 08 *****) (************* END OF EXAMINATION ***************)
- L*:
Bt.ack001>% Ac7tJMTED 320b 8t.DCk 0 Detriq
}$C7uprCD 3208 LOCKClert-
/ CYtIn1CD 3 12 QE Ae4LLEl N OCA*$T.
s / s / s / g / 7g
/ s
-@-e s
/ s i s e, -
~
- b e. - Exe .
&n. - Ewn.
% ~ 3205 ~
~ 32O7
~
~ 321/ ~
~ 32D9 2
7 m '
- E un- - Bee _ bc - _ %ee.
~
~ 3209 -
-3205 -~3207 --3211 D e W saa. M -
we saa. A -
-- 3:? 10 : :320fr :~3206 ::4212 n
1 Se I 1 i S 2-:s =. 7.pe -- - l a; w-. ,. t i
- e l
l i l 5 ~ + *,
, <- + y k .s % .n om-m. .r. s e., w n.
.. I l, i I
H
- i
.t uGion + suioN lif640e/ Q
... \g)
Mk g , incuAsia cw - - - ons---- --- t _o L..___ ____ g NEAT ' FLUX. g g , ;
, gg5 -
9tv kr ft*
~
it III" 14 . . ; lij l .; 5 '
![ j gh A- MCION h fi,3 3
a ! I
>e REGIOe -
i 3 A S J ' i 2 10 8 le 2 3 13 10 10 v INCREASING JTgg - TMT) 'I j T 1 SMF stLev Tgay As WATER RIstS IN TCP.PERATURE l l *.* i l fig 4te /05 7 1
l .* a l l l
.' i.
~
! l ICOLD 1 o THDT h
}
l . t -\
. e '
! TURBINE l BYPASS g -
U lSETPDINT 2 _,- SATURATIDN nn . 3 EAR : ':5DBCOOLING )
' 4 ja'EARGIN LlulT
! ' j . [! i !
PRESSURE LIMIT ' P 9 Temperature - 4 4 i
.*n 1
. *g o
, l5 I
s 8 I NOTE: IF any problems are encountered, TEIK artify "Realth- ' t
- Physics
- personnel immediately.
3.3.2 conduct *m ala andy Prink"
\ :
307E3: 1) The distamos between the probe and sarface being l5 monitored shoald agt; estW 1/4*.
- 2) IF a significant increase (i.e, errester' thaa 50 l 5 '.
' spa' above background) la count rate is noted, l
. Tuss retara the probe to the' suspected, area of
" contamination". l5 ,
- p. 3) ,Ir the 'Alars' sounds, TEIN actify "Bealth Physics
- l
- . t -
j9 , fiM1 ismediately ame renais j;in 'present
. location until assistance arrives.
- 4) secoeTalemTIce Elst as coupscTun ' sNDER TEE
- estlancs or *EIRLTE PETSICs* DERaoIBIEL.
A 9 em Fip r e 110L = 2o ne-toi new. 05 '*S'
~'
< . % . - ww . , - , ,u 1. -,1 m. w e , 2 -r = - -mmmmw ets . .1 aLF- . -e': .-
** a' =~
- on. = 4 m . I. . - o, ** i am IDH35ENTUL WHUEDE UNUK FUWBEL ND IMEUL FUHAIEN CDCIT104 E090ED M.TIOi
- 1. G., - ,i, =
- n = iar i. 4".'lf4 W"J",.2lFor/eNa c cto,,.-
" t'" "'"* et.s. M'Tf4 Eb"'PiaAraff/" Shd s, curs.-
2 IJi"tnaP"<J:"5'E%T&g' B1Eit!.E' J E ct ry in conuin.a. -
~ ~
- 3. A Geners' r has decland.
f;a ma g gg a tad. M AS m
'1a putact irwencory in ce reent < GAS GAP. ,
sb'4vratEn tbially effected sectors.** I
- 5. A Geners' has diciarut. he utle, 360' tion.
nuent E$irwentory in containment, a E can f pd( l m abstanti ssion arrival . l NEAP. T$he's f gotentially effectai sectors.** pggggge - 1g;meprqe .- 1 av & fected stctors.** k d: > lism . l
- 1. m sinal release has occmid or < s in g' .
workers ea E"Mf3*Y#~ NN=A. fyl engosure of ahh* ra E'5/*g & tid 2 g.ngg*"llWich'iM:f'- % 'lll"" {2i,mmaanta.- gg cogo e y-V.: ggh thi tabgccubinstgf Table 5.1, sage 5.31, lley. 6#9 - Marual of Protective Actions foc.
**Affiscted sectors include, as a ulnisun, the danrind sector (s) ani acacent sectors. ,
~' ~~ ^^ ^ ~ ^ ^ ~ ^ ~ ~ ~ '^ ^^
^ ~. L ... ..'. .R. - - - - ---
4 ENCLOSURE 4 (Page 2 of 2) l
. 1 GUIDELINES FOR' REC 0peqENDED PROTECTIVE ACTIONS FOR GASEOUS PLUE EXPOSURE
~
EVACUATION TIE ESTIMATE TA8LES*
- i SECTORS -
Distance A-D E-H J-M N-R All (miles) Mr.-M n. Hr.-M n. Hr.-M n. Hr. -Mi n . Hr.-m n. GENERAL POPULATION (NORMAL CONDITIONS) - 0 - 2** 1 - 30 0.- 5 0 - 45 0 - 45 2 - 30 0-5 1 - 30 1 - 30 0 - 45** O - 45** 4 - 00 . 0 - 10 3 - 00 3 - 00 ' 0 - 45** O - 45** 6 - 30 g RAL POPULATION (ADVERSE CONDITIONS) 0 - 2** 3 - 00 1 - 00 1 - 00 1 - 00, 3 - 00 l25 0-5 3 - 00 3 - 00 1 - 00** 1 - 00** 8 - 00 0 - 10 e g 6 - 00 6 - 00 2 - 00*** 2 - 00** 13 - 00 yhaq j SPECIAL POPULATION JJ COMITIONS)*** i i! 0 - 2** N/A 'N/A N/A N/A , N/A 0-5 2 - 00 2 - 00 N/A N/A 5 - 00 t 0 - 10 3 - 00 3 - 00 N/A N/A
- 6 - 30 SPECIAL POPULATION (ADVERSE COWITIONS)*** ,
0 - 2** N/A N/A N/A N/A N/A 0-5 4 - 00 4 - 00 N/A N/A N/A 0 - 10 C - 00 6 - 00 N/A N/A' 13 - 00 -
*These are conservative estimates and are inclusive of warning times.
***These estimates are for personnel on FPC property. Evacuation time is . based on a staged / sequenced generating complex evacuation via the plant access road. .e
***This category includes the handicapped, elderly, and hospitalized individuals and is inclusive of the general population.
EM-202 Rev. 2 6 / ((h 8 Page 45
05-SITE FaerscTIVE ACTIONS Rectascas 6 ETec5&T10s Plantac gotas l25 N F g 4 or 4 .l mm asume nam , h h had tiseerd Smetata_Affanted (4) s 29-11 e aJa (3) M 13 33 SSW JE& (c) a wu su aLa (9). WE S?*?S WWW L hl 5 (R) R 79 101 W N3P (F) Ser 103 133 tuu 3P4 -g (e) et 134 146 m Pen (a) ses 147 164 puw 4aA (J) s 169-191 m aa3 - (E) $su 193-313 WE ABC (L) SW 3*4-336 m 5CD (N) Esw 337-3S4 But ttI (B) W 3S9 381 3 9RF (P) Wuw 303 303 Est ' BFS '] . WW 306 336 FSB (t) (R) WW 337 368 SE sat .): Se3 .o m ysA /u- ,
-<o- -
cafstax. atvan e== anno coeu:x 303 k - a qq 5
'6*
-as,e%
f . % ., y* K __ ll ,,
- ),
- M w. p, g 5 l f adetetstretten Parktes r
g, pyg Senes Posee Parktes Setts 1 4 Staff Adatetstretten Parties % .0,o I9 mite 9mit 3 Staff Adatatetretten Parktes
- ye Ostte 1 & 3 Coal Pile Parktes Watt 3 Warehouses and Temporary offLaos , p 23fe
,23e '
.. e
( ... c .
, ,e 1 , ,e "
?
35 tete of teetest for peroemmel assembly are provided betese BEE M eme , A Emite 4 & 5 Batte 4 & S Centret asse B/C EE3 Admietettettee 31d3 Pehlte Addrese System 5/C Berth Cool Tard Smits 4'& S Centrol Bessi - 3/3 CR3 Werebesee Area Corporate Seeerity Spectattet 9/8 Ene
- Spe festlere Serporate Security Specieltet
.p 3.F. S. and J emot Trete Area Wales 4 & S esotrek Been '
J. E. and L Seeth east ferd Entse 1 & 3 centret teen B
- Seite 1 & 3 Seite 1 & 2 esotrol Seen 1
C EM-202 Rev. 2 6 /g //0 [ age 50(LASTPAGE
.l
. REACTOR COOLANT SYSTEM POWER OPERATED RELEF VALVES lp CONDITION POR OPpRATION t
- 3.4.3.2 The power operated relief valve (PORV) and las OPERABLL --:=" 'block valve shah be - 'l APPLICABILITY: .q00E5 5,2, and O .
[ ACTION: _.
- a. With the' PORY ~ inoperable within I hour' either restore the PORY to OPERABLEstatus or close the associated block valve andTemove power from '
the block valves otherwise, be in at least flOT.5TANDBY within.the next 6 ~ hours and in COLD SHUTDOWN within the following 90 hours.- -
- b. With the block valve inoperahN within I hoor;etther restore the block valve to. l OPERABLE status.or close the block valveend remove power from -the block.
valve er.Tclose the PORY and remove power 'from the associatied splenoid valves ' otherwise, be>in at least HOT!57AND6Y within the next4 hours.and.in{0LD 5HUTDOWNiithin'the following 3G hours. ' ' ~
- c. The provisions of Specification S.0.4 are notappilcaldee SURVEILLANCE REQUIREMENTS
\l ,3 4.4.3.2.1 la addition to the requirements of Specifit:ations 4.0.5, the PORY shall be demonstrated OPERABLE at least once per 18 months by performance of a CHANNEL CALIBRATION. ,
4.4.3.2.2" The'idock val *shall, be demonstratedOPERABLE'at'least once per 92 days by . operating the valve through one completecycle of full. travel?: . , ' < ~ l l . t ei r l l Fip re WO
- CRYSTAL RIVER - UNIT 3 3/4 4-4a Amendments Nos. S$, $$, 77,77
o ,. REAC1CR COOLANT SYSTER , . . . A '
.. RELI(F YRVt3_.MRAY115 ,
- f g y yg,gg . , .. ,
- r. .
LIMITINE COISITION M K (FERATION
'.
- 7 ~ "- I ^
":'; :r. p ,-
- N 3.'4. 3.1 All pressurizer code safety- valves Dall' he M.'wlth e 11ft setting of'2500 4:13. + 15.- - -
APPLICABfLITY:',',.,NIEr.I, .t. Lamt.3.'., ; , . ACTION:~.... s. W1th one pressurtseritade. sa'fety valve' inoperable. either;restadi%e'M"E.Y inoperable .'velve' to :0PERABLE . status wtthfe 15 minutes;er'^- tel;tailWfl . . i SmlTDOWN wlthin.12. hour.s.t i' ~ * ' '
i
~
4 y
; . 4 i
$URVE!LLAhct' REQUIREMENTS i
. -sw.
.., li'N 4.4.3.1 No- additional + 5'neveillancei,:Repirements9.etterd 'then'%.i ,5 required by Speciffcation 4.0.5. ~
I I l ! , j 2 j s
.g l
CRYSTAL RIVER - UNIT 3 3/4 4-4 Amendment No. JF. 55 1 t a
REACTORCgA,NT_$YSTL f 1 OPERATIONAL LEAKAGE 1 LIMITING tounITION FOR OPERATION ' 3.4.6.1. Reactor Coolant System .leakaGa shall ha 11mitad tau .
. i s. No PRES $URE BOUNDARY LEAKAGE, . .
"b ;- O GPM UNIDENTIFIEF LEAKAGE,.
c.. 1 GPM total. primary-ta.marnada g leakage thi Q -staas'ges erstars,- -
- d. 10 GPM IDENTIFIED LEAKAGE,fromhthecReactor.:Coolantifystes',.
- e. ' 10 GPM CONTROLLED.1EAKAGE'.at a;AeacsrCoolant' System pressure of 2150".t,20~.pstys.and-- - -
- f. . ' Leakage:as'spectfled in Jable 3.4-2 for those r! Coolant.5ystem - '
Pressure Isolation-Valves identified-in Table. .
.APPLICAB1L1TY: PODES 1, 2, 3 ahe 4 .
ACYTONf
- a. With any PRESSURE"5OUNDARY:.LEAXAGE," be -in at. least HOT'5TANDBY witnin. 6 hours and in COLD SHUTOOWN withm the following 30 hours, t >.
- b. With any Reactor Coolant System leakage greater than any one" i- -
1
\
f)f h
. . of the.above limits, . excluding PRESSURE SOUNDARY LEAKAGE, .
reducaithekleakage . rate-to;within limits.'within-4thours or be- - - in at least:NOT t STANDEY!within the"next?6. hours and-in COLO:".' . ~
~
SHUTEWN:within the.fisilowing.00. hours. ' - + - - c.. W1th any Amector toolant.%stssrpressu't r Isolation-Valve-leakagepter.:'.' than therabove d4arit, reactor.operatton may : continue- provided-that at.. . least.two. val.veskirt.sech.htgh pressure.:line-having a man *fenctionativalvei: are in, and.Temairring the mode corresponding to the isolated condition.-- (Motor operated valv'es shall. be placed in the-closed post.tton.wuLpowera- ' supplies.dgenergtzed.t . i
- d. . The provisions xf Section 34.4 are not acclicable for entry into .fCDES:3 ?
and 4 for the purpose _of. tasting.the. isolation .cneck valves. fi(c gr6 ll1I A -
. .c CRYSTAL RIVER - UNIT 3 3/4 4 15 Orcer ote.4/zu/ul
~
r it4fTOR COOLANT SYSTEM SURVEILLANCE REQUIREM(NTS 4.'4. 5.2.1 Reactor Coolant'5vstem leakages shall be demonstrated to tw within each of the above limits by:
- a. Monitoring the containment atmosphere todine . rad osc.tivity monitor et least.once per 12 hours, -
- b. Monitortag- the containrent sump' inventory and discharge ht.
least .once.per 12 hours...
- c. Measurement .cfc the CONTROLLED. LEAXAGE freer the'reactorftoalant pumir seals when the Reactor Coolant System pressure':is 2150.+ 20 psig at'.least.once per:31 days;-
- d. . Perforsence:sf a Reactor'.Cholant Systus' water ' inventory balance at least once:per 72 hours',during steady state operatione -
4.4.6.2.'2 :~Each. Reactor-Coolant-System Pressure ' IsolationV # aive.specified M. Table 314-2 shall be.' individually demonstrated OPERABLE. prior'tofantering fCDE/2.;by: verifying-leakage to be;within-its111mit: ' '- -
- a. After each ' refueling outage," -
.b. Whenever the plant has been in COLD SHUTDOWN for 72 hours, or more, H\ if leakage testing has not been performed in the previous 9 months, and '
- c. n Prior to totur' ing'.the valveito . service following.mainten'ance, repair.or . replacement, work on:the val.ve. -
4.4,6.2.3.:Whenever fei; vrity;cf a' pres'surecisol'ation valv'eL11sted'M Table' 3.4-2 cannot he demonstrated, the. integrity.of the remaining valve.1rr eacF high pressareAine .having a leaking; valve shall be'detemined and: recorded daily. -In addition,:the position:sf-the other. closed / valve'locatedzint the. high pressure. piping shal.1'be.' recorded Aaily. - l l
, l
- I i
l I ['"f0, 4I6 !' r d Amendment No..,Xf. CRYSTAL RIVER - UNIT 3 3/4 4-16 Order ctd. 4/20/81
j
. . .- TABLE 3.4-2 REACTOR COOLANT SYSTEM PRES $URE !$0LAT!0N VALVES 5ystem Valve Maximum Allowable Leakane(s)(b)(c)
Decay Heat / Lee CFV.1 < 5.0 spa.
-Pressure Injection i DHV-2 < 5.0 gpm
< 5.0 sps.
CFV , DNV-1 < 5.0 gpm--
- Notes:2 .
D 'ximum Allowable Leakase-(each valve):
- 1. Leakage: rates less:.than~orcequallto';1.0. gym are considered:acceptabled
- 2. Leakage rates.: greater-than- 1.0 gperbut less- than ors:equalitof 5!0?gpm '
are considered acceptabledf the latest: measured. rate:Ms.aottenceeded , the rate"detamined by~the previousc-test 2y en.gnountrthattroduces'.the margin between measured leakage rate and :he maximum permissible, rate ; ;I of 5.0 gpm by 50", or greater. ?. .. b5 3. Leakage rates greateia than 1.0 gpm but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate dataruined'by.the; previous. test.hy an: amount that:: reduces 4the - margin between sionsured 1eekage rate and;the. maximum.9ermissiblerrate. 4 of.5.0.gpm.by.50%.or. greater.-
- 4. Leakage. rates- greater:than 5.0. gpm are. considered.. unacceptable.
I (b) To satisfy ALARA/requirementspleakage.may.be measur.ed .indirectif(as : .l from the prfoTwance of. pressure indicatorr).1f accompitshed.irt scenrdance u with approved procedures and supported by computations showing.that- the method is;rapable ef. demonstrating valve. compliance with the.deakage criteria.: . . (c) Minimum differential ~. test pressure.shall 'not be:1ess thaniSO.psid. FigqFe / /11 c - = CRYSTAL RIVER - Uti!T 3 3/4 4-18a Order dtd.4/20/81 -v,- ve,-,m--,,,w.-- . - - - - - - -
' T T A 81.E 3.3-1 . 0-
- U REACTOR PROTECTION SYSTEM DISTRUBAENTATION *b %
4 % n M TOTAL NO. CHANDE!L5 CHAfGELS AFFESCAELE r- - FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERAALE 8000E5_ W B ' .-
< I I,2 and
- 3 g I. Mandl Reactor Trip i I
- Nuclear Overpower 4 2 3 I, 2 2#
2.
~
RC5 Outlet Temperature - High 4 2 3 1, 2 f 3.
-e j w 4. Nuclear Overpower Based en RCS Flow and AXIAL POWER IMSALANCE 4 2(a) 3 I, 2 28 4 3. RCS Pressure - Low 4 2(a) 3 I, 2 W
- 6. RCS Pressure - High 4 2 3 I, 2 W
'7. VariaMe Low RC5 Pressure 4 2(a) 3 I,2 M Reactor Containment Pressure - High 2 3 I, 2 M i Y 8. 4 l y 9. Intermediate Range, Neutron Flux
$ w and Rate - 2 0 2 I,2 and
- 4 ;
~
i 10. Source Range Neutron Flus and Rate _ i A. Starty 2 0 2 2ff and
- 5
< 8. Shutdown 2 0 I 3,4 and S # I1. Control Rod Drive Trip Breakers 2 per trip I per trip 2 per I, 2 and
- 7e system system trip system
- 12. Reacter Trip Module 2 per trip i per trip 2 per I, 2 and *,
M system system trip system
- 13. Shutdown Bypass RC5 Pressure - High 4 2 3 2* * , 3* *, 6#
4**,See g 23
** Reactor Coolant Pump Power Monitois 1.from 2 2 per pump 1, t 14.
2 per pump l 7 or r.
., i p..p. (a,6)
* .3
.y
'~ ^ -
TABLE Abt (Continued) TABLE NOTATKW
*With the conttol rod drive trip breakers in the closed position and the control red drive systein capable.of rod withdrawal.
**When Shutdown Bypass is actuated.
)
#The provisions of 5pecification 3.0.4 are not applicable. .
)
##High voltage to detector may be de-enerS ized above 10-10 amps on both !
Intermediate Range channels. l
)
(a) Trip may be manually bypassed when RC5 pressure < 1720 psig by actuating Shutdown Bypass provided that: (1) The Nuclear overpower Trip 5etpoint is 13% of RATED THERMAL POWER, . i (2) 1he Shutdown Bypass RCS Pressure - High Trip 5etpoint of 11720 -i psis is imposed, and (3) The Shutdown Bypass is removed when RC5 pressure >l800 psig. ! (b) Trip may be manually bypassed when reactor power is less than or equn) to 2475 mt and 4 reactor coolant pumps sre operating. ACTION STATEMENT 5 , J; . ACTION 1 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and/or open the control rod drive trip breakers. ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Charjnels, STARTUP and/or POWER OPERATION may proceed provided all of the following conditions are satisfied
- a. The inoperable channel is placed in the tripped condition within one hour.
be The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours for swveillance testing per Specification 4.3.1.1. g ii19 E CRYSTAL RIVER -UNIT 3 3/4 b3 Amendment No.F.W. JJ,56 i
~
~
fABLt 3.301 (Continued) ACTION ET4TEMENTS (Continued) and the' inoperable thannel ebeve ser be by- ' passed for up to 30 minutes in any 14 hour period when necessary to test the trip breaker associates with me logic of me stannel being tasted per Specification 4.3.1.1. and ,
- c. Either. THERMAL POWER is restricted te N 75%
of RATED RATED THERMAL and the Nuclear Iver. poner Trip Setpoint is reduced to y,855 of' RATED THERMAL POWER Mthin 4 hours ter the OuADRANT POWER TILT .s senttered at least once per 12 hours. ACTICN 3 - With the numer of OPERA 8LE channels one less than the Total huseer of Channels STARTUP and POWER OPERATION may proceed provided noth of the following consistens are satisfied:
- a. The inoperable channel is placed in De tripped canditten within one hour.
- b. The Minimum Channels OPERABLE requirement is met; homever, one additional channel may.be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1, and the i inoperable channel above may he bypassed for up l
to 30 minutes in any 24 hour period when necessary to test the trip Dreaker associated with me 1ogic of the channel being tested per Specification 4.3.1.1. Action 4 - With the numer of channels OPERASLE ene less Den reevices by the Minimum Channels OPERABLE requirement and with the THERMAL Power level:
- a. < 5% of RATED THERMAL POWER restore the inoperable channel to IPERABLE status prior to increasing THERMAL POWER above SE of RATED THElW14L POWER.
- b. >5% of RATED THERMAL POWER, POWER OPERATION may continue.
b kN
~
ll18 C 4 3/4 F4 Amendment No. 55 l CRYSTAL R:YER - UNIT 3 l
...- - I
~
toDC y t cunyt 't - 3 Pump .,
. ~ _
1230
$ I
. I / ~
3 : : : l/ In = i y , , cumvt e - I
. : ;. .
- Pump ,
- m. .
i ; l w . I ' I gggg :
?
I : . .: lu - i
- i.
m
;// ;
t' l- i { 1t00
//
l -
,l .
. l
- i 580 000 42C 440 asseren ou?g,z: iguessa ung. p REACTCR CQQLANT FLOW PUMPS CPERATING Flow !% D ESIGN) POWER (RTS) (TYPE gF ([Mj?)
Cuevt 135.7 s IC' (106.5%) 113.05 % 4 PUMPS (ONER) . 1 104.4 x ids ( 79.6 %) 90.84 % 3 PUMPS (DNBR) l BASE 3 FIGURE 2.1
- 333 Amendment Ilg. X, Jf; 41 CRYSTAL RIVER =* UNIT 3 R pure Il32
, - . - - - - ,,. , ..e --,,,e . , . - - - - - , - - .--r-.- - , , - - . , - - - - - - . . , , - . _ , _ , . . , - , . - - - - - - - . , - . . . - - , - --- - . .
f .
.- , l i
EMERGENCY CORE COOLING SYiTDt$ ECCe SUB5Vaivii - 7T ;;
- 200*E LIMITING CONDITION FOR OptRATION 3.5.2 Two independent ECC5 subsystems shall be OPEblE with each subsystem comprised of-
- a. OneOPERABLEhighpressureinjection(HPI) pump,
- b. One OPERABLE low pressure injection (Lp!) pump. l l
- c. One OPERABLE decay heat cooler, and
- d. An 0PEPl.8LE flow path capable of taking suction from the borated water storage tank (BWST) on a safety injection signal and manually transferring suction to the containment sump duriep '
the recirculation phase of operation. APPLICABILITY: MODES 1. 2 and 3. ACTION: i
- a. With one-ECCS subsystem inoperable, restore the inoperable subsystem to QPERA8LE status within 72 hours or be in HOT SHUIDOWN within the next 12 hours,
- b. 'In the event.the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the' actuation and the total accumulatpd actuation cycles to date.
l l
. ~g l
CRYSTAL RIVER - UNIT 3 3/4 5-3 F /p @6 Il.3 '//)
3/4.8 ELECTRICAL POWER SYSTDIS 3/4.8.1 A.C._$0URCES OPERATING' LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERA 8LE: ,
- s. Two physically independent circuits between the offsite trans-mission network and the onsite Class 1E distribution system, eM
- b. Two separate and independent diesel generators each with:
- 1. A separate day fuel tank containing a minimum volume of 400 gallons of fuel,
- 2. , A separate fuel storage system containing a minimum volume of 20,300 gallons of fuel, and
- 3. Asekratefueltransferpump.
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: -
- a. With either an offsite circuit o'r diesel generator of the above required A.C. electrical power sources inoperable, demonstrate ,-
the OPERABILITY of the remaining A.C. sources by performing ' Surveillance Requirements 4.8.1.1.1.a and 4.8.1.1.2.a.4 within one hour and at least c ce per 8 hours thereafter; restore at least two offsite circuts and two diesel generators to OPERA 8LE status within 72 hours or be in at least HOT STAND 8Y within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. -
- b. . With one offsite circuit and one diesel generate )f the above l
i required A.C. electrical power sources inoperat , demonstrate the OPERA 8ILITY of the remainira A.C. sources by performing Surveillance Requirements 4.8.1. 1.a and 4.8.1.1.2.a.4 within one hour and at least once per 8 hours thereafter; restore at least one of the inoperable sources to OPERABLE status within l 12 hours or be. in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours. Restore at least two offsite circuits and two diesel generators to OPERABLE status within 72 hours from the time of initial loss or be in at = least H0T STAND 8Y within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. CRYSTAL RIVER - UNIT 3 3/4 8-1 gpg //Jq $
_ _ -_ - - =_ _
/ .
.? -
l ELECTRICAL POWER SYSTDt3 ACTION (Continued) !
- c. With two of the above required offsite A.C. circuiks inoperable, demonstrate the OPERABILITY of two diesel generators by perfomin; Surveillance Requirement 4.8.1.1.2.a.4 within one hour and at least once per 8 hours thereafter, unless the diesel generators are alrea @ operating; restore at least one of the inoperable offsite sources to OPERA 8LE status within 24 hours or be'in at least HDT STAND 8Y within the next 6 hours.
With only one offsite source restored, restore at least two offsite circuits to OPERA 8LE status within 72 hours from time . of initial loss or be in at least HDT STAND 8Y within the next 6 hours and in COLD SMJTDOWN within the following 30 hours. 1
- d. With tivo ~o'f the above requ' ired diesel generators inoperable. (
demonstrate the OPERASILITY of two offsite A.C. circuits by i perfoming Surveillance Requirement 4.8.1.1.1.a within one hour and at least once per 8 hours thereafter; restore at least one of the inoperable diesel generators to OPERABLE . status within 2 hours or be in at least HDT STAND 8Y within the i next 6 hours and.in COLD SHUTDOWN within the following 30 hours. Restore at least two diesel generators to OPERABLE ;
' status within 72 hours frem time of initial loss or be in at least NOT STAND 8Y within the next 6 hours and in COLD SRITDOWN 1 within the following 30 hours.
SURVEILLANCE REQUIREMENTS l . 4.8.1.1.1 Each independent circuit between the offsite transmission network and the onsite Class 1E distribution system shall be: .
- a. Determined OPERA 8LE at least once per 7 days by verifying; .
- 1. Correct breaker alignments and indicated power availability, and
- 2. That the sump peps in the tunnel containing the DC i control feeds to the 230kv switchgear are OPERA 8LE.
- b. Demonsitrated OPERABLE at least once per 18 months during ~
shutdown by transferring unit power supply from the normal circuit to the alternate circuit. , l l CRYSTAL RIVER - UNIT 3 3/4 8-2 f Qf"f_ l / 3 k C-
i ! ' f/Wh L. NbkKEbc? A5 75K L.
)g?' ,
j' - 1
- 5. THEOPY 0 "- ' '.' f. E A R POWEP cLA>U O C'E R Al lO N , ~L' JIM , AND PAGE 29
-------__----- . s ,.+
ANSWERS -- CRYSTAL R I;)E R -86/08/1E-LAWYEP, SAUD* 7 i ANSWER 5.0". II.00) . ._ ( f
/
(a) - c20 000stft 26,000 scf(b); 140 scf(c) 19320 scf(d) RE ERENCE CR lesson '7 9 n d a r.i e n + a l s o# Natural C : r e .' l a t i o n ' t C9- :0T 3-ic, 7 2 M 302 02/05/95 02 00 05 00 0 '/ -0.c29 -0.c29 A N S L' E :' 5.02 (1.00'r
'2)
RECERE*TE C :' , OP-203 Re", c7, p 20. e
\
l 201 03/05/95 03 00 01 06 10 +0.800 +0.800 302 08/18/06 , o
)i t.
ANSWER 5.03 (1.00) (b) REFERENCE DC OP 1102/01, enc 1 4.3, p 3; CR, STM 504. pill; OC, IC-ICS, p 85; OC CM-SC, :- 19. CR '-IT F r , pp 148-156. 035/010 K1.0c ' 269 8/19/85 12 00 04 00 +0.500 +0.500 I
=
1 5 THEORY Or Nyctggg e0WE9 *LANT OPERATIONr CLUISSr AND PAGE 30 ANSWERS -- CRYSTAL RIVEP -86/08/18-LAWYERr S6UDY ANSWER 5.0c (1.00i (a) REFERENCE DC FNRE, p 120; OCr NETRO, 12.1-ai CR, NETR0r 12.1-a. 001/010 K 5.16 (2.9/3.5) 269 8/19/05 09 03 03 01 16 +0 250 +0.250 ANSWEP 5.05 (1.00) (c) f REFERENCE OCr OP 1106/01, p 10. i BKrGP-03, p 19. CR3r power sytten ops., pp 22-23. BKr 20-2-C4r p 37. 62/0 A4.03 (2.8/2.9)
- 269 8/1?/85 10 02 03 01 16 +0.375 +0.375 'R:]' 325 5/19/86 06 02 00 00 08 +0.500 +0.a17 302 8/18/86 ANSWER 5.06 (1.00) ,
(c) REFERENCE CRr Reactivity Balance Calculations SP-421. ANSWER 5.07 (1.50)
- c. Increase
- b. Increase
- c. Decrease
- d. Remain the ser:s
- o. Increase
- f. Decrease
.w
S. THEORY OF NUCLEAo r o w E:- o' ANT DoERATION. CLUlDS> AND oAGE 31 ANEWERS -- CRYSTAL RIVER -86/08/18-LAWYER, SAUDY RErgpENCE CR, ROT 2-2, p 78. ANSWER 5.08 (2.00)
- 1. Subcooled water receiving heat
- 2. Bubbles collapse in subcooled utter
- 3. Bubbles remain out into uster channel
- 4. C' e r t i c ! bubbles, portie. unstable filr.:
- 5. Conplete vapor *il-
- 6. Dryout REFERENCE CR. NETPP, p S.2-3; CR, 001.16.
ANSWER -5.09 (1.00)
- 3. To prevent centerline +' v e l melt
- b. T c- ensure thzt clad temperaturps ren.cin lett th r- or equcl to 2200 dest on worst case LOCA.(fr A'd" #w m DM6M2/.3 -
REFERENCE ', I CR, TS, pp B2-2, B 3/4 2-1. ' ANSWER , 5.10 (1.00'. RCS pressure high setpoint REFERENCE CR, TSr p B 2-6 A N S L!E R 5.11 (2.00)
- 3. stea- generator tube rupture (0.5)
- b. n.cin condenser not aveilcble (1.5)
HPI required to maintain pressur:rer level RCPs not operating REFERENCE CR, ROT 2-8, p SGTR14.
- N
- 5. THEORY OF NUCLEAR 00WER PLAN 1 OPERATION, CLUIDSr AND SAGE 22 ANSWERS -- CRYSTAL RIVER -G6/08/18-LAWYERr SqhDY ANSWER 5.12 (1.00)
- a. Sodium Pydroxide
- b. Water REFERENCE CR, ROT 3-14, p 4.
ANSWER 5.13 (1.50)
- s. Te decrease
- b. T5 decrease
- e. Incores decrecte REFERENCE CR, VF-580 RS. p 7.
ANSWER 5.14 (2.00)
- 1. the moderator temperature coefficient is within its analyzed temperature renser .< ;
- 2. the protective instronentation is within i t's a; nor n. a l operatins ranser
- 3. the pressurl:er is capable o' being in an operable status with e steam bubbler and
- 4. the reactor pressure vessel is above its minimum RTndt temperature.
RECERENCE CR, TS, p 93/4 1-2. ANSWER 5.15 ( .50) Doppler coefficient. REFERENCE CR, NETPO, p 8.2-2. ANSWER 5.16 (1.00) To avoid baron precipitation in the event the RCS is not subcooled.
e t
- 5. . THEORY Oc NUCLEAR'p0WER PLANT OPERATION, FLUIDS, AND PAGE. 33' ANSWERS:-- CRYSTAL RIVER' -86/08/18-LAWYER, SANDY.
-REFERENCE ~
CR, ROT 3-6,_ p 7. ANSWER L5.17 (1.50)
- a. 0.7% .
(0.5)
.b. Deleyed neutrons have z seneration time of 12.7 seconds.while prompt'
. neutrons have a generation tine of 10E-5 seconds. 'If we use weighted
-averases, we can see thet the averase generation time is reduced.to
.about 0.1 second. This factor provides better reactor control. '(1 0)
REFERENCE CR, NETRO,-pp 5.3-1,5.4-7~. l ANSWER 5.18 (1.00) , 1 Xenon concentrations may increase or decrease but samarium will always
' decrease during_this transient.
' REFERENCE CR FNRE, 1 pp 162,168.d4 T'l 7
ANSWER 5.19 (1 00) No. 1; larger amount of reactivity is required to' increase power from 10%
- to'30% because Tave and Tfuel will be required to' increase. This will' add negative reactivity from both. A power. increase from 30% to 50% will add recctivity only.due to the. change in Tfuel.
REFERENCE
.CR, STM 504, p 177.
ANSWER. 5.20 (1.00).
. Increase. The MTC will introduce nesative reactivity due to the temperature increase.
REFERENCE
~CR, NETRO, p 8.4-3.
- 19
n; .. s I 5.- . THEORY--OF NUCLEAR POWER PLAN'T OPERATION, CLUIDS, AND PAGE 34 ANSWERS,-- CRYSTAL RIVER -86/08/18-LAWYER, S$40Y
! ANSWER ~ 5.21 !(1.00)
A gammajis completely asbsorbed by.an orbital electront. the electronRis
- cejected from the stome ecuting the detector gases to ionine.
^ REFERENCE CR, 'NETBNCr p 8.2-3) CR, GB l'.99.
. ANSWER - 5 .' 2 2 (1.00)
.The new element will have increased in atomic number by one and_the mass number will remain the same as the original-element.
-REFERENCE CRr NETBNCr'p 7.3-1.
ANSWER 5.23 (1.00)
- 1. The fuel temperature coefficient becomes more negative as Pu-240 builds Lup causing-increased resonance capture.
2.-The'MTC becomes more. negative.due to the reduction of the soluble poison
'effect!as temperature increases.
REFERENCE , CR,'NETRO,-p.11.4-4. 1 ANSWER ~5.24 (2.00)
-See Answer figure 1055 attached.
REFERENCE-OCr FNRE, .pp 97r 105; CRr Principles of NRE, GlasstonerS., p 238) CR, NETR0r~ fig 6.6-2. DC :07/14/86 _ C R; 08/18/86
*M
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PACE 35 ANSWERS -- CRYSTAL RIVER -86/08/18-LAWYER, SASDY ANSWER 5.25 (1.00)
- a. 3.125 HP
- b. 22.5_ spm
- c. 62.5 psi REFERENCE CRr NETPP, p 6.2-3.
K 302 8/18/86 bl[ 1 :; n . 6 e
e
- 6. PLANT' SYSTEMS DESIGN, CONTROL,~AND INSTROMENTATION PAGE- 36 ANSWERS - CRYSTAL RIVER -86/08/18-LAWYER, SANDY.
1 ANSWER 6.01 (1.00)
- (s)
REFERENCE STM-22, p22-25.
,240E0000 K
302' -03/05/85 _0 0 _ 03 05 _02 10 -0.'600 -0.600 ANSWER 6.02 (1.00)' (c)
. REFERENCE CR, AD-303, p 8.
'K 302 '12/17/84 04 '00 00 00 04 +1.000 +1.000 z 2 6.03 (1.00)
~ ANSWER ; .
u > Ath) y
. REFERENCE CR, STM 25-13/14.
K 302.' 12/17/84 03- 01 00 00 04 +0.500 +0.500 ANSWER : 6~.04 (1*.00) (c) REFERENCE CR, STM-504, s 1.3.4, p 7. r l t
- , .-,, - - - - . _ . . . . , . . - - . - , _ , . , - - = - - , . . . . . . . .. - . - . . _ - - . . . - _ .
PLANT ~ SYSTEMS DESIGN, ___________________ ______' CONTROL, AND INSTRUMENTATION 6. PAGE 37-ANSWERS'-- CRYSTAL RIVER -86/08/18-LAWYER, SANDY i ANSWER 6.05 (1.00) - (c) REFERENCE CR, ROT 4-15, p 14. 302 08/18/86 - ANSWER 6.06 (1.00)
-(d)
REFERENCE CR, STM 17-17/18; CR,' ROT 4-1, p 71. K 302 12/17/84 03 01 00 00 04 +0.500 +0.500 ANSWER 6.07 (1.00) (c) gr,*
*1 1 ' i '
REFERENCE CR, STM-504, s 4.1, (QB); CR, STM 504, p 124. - k ANSWER 6.08 (1.00) 15 pg '4h q (.ra.psc),
- 1. : : r. ' r. ;; - ;;;12 - ^#
st leret ?7i :che: "5 .m c u i, o L .' 4 2 l m J .
- 2. Lube water flow =or> q sen (+or- [spm).
- 3. Water box has been printed to =or> 1 5 feet (+or- 15 feet).
- 4. Pump trip permit thirt seconds ha elapsed since pump last ran.
(+or-3 sec). REFERENCE CR, STM 35 ~7
.s
r-
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 30 ANSWERS -- CRYSTAL RIVER -86/08/18-LAWYERr SANDY ANSWER' 6,00 (1.00) -
- 1. Gland sealins steam system 7. Deaerator sparser no::les
- 2. RW evaporator 8. EFP-2
- 3. RC evaporator 9. Ultrasonic sink
- 4. CBAST 10. Dearator dome (tray section)
- 5. CWST 11. Decon pit
- 6. MFWPs (any 8 0 0.125 each)
REFERENCE
~CR, STM 606, pp 7,8 (GB)
ANSWER 6.10 (1.00) Normal PT trace followins a reactor trip. REFERENCE CR, Generic ATOG suideliner. figure 10-4; CR, ROT 4-17. ANSWER 6.11 (1.00) It should be set on zero. REFERENCE CR, STM 10, pp 36,37; C,R , ROT 4-6, p 19. ANSWER 6.12 (1.50)
- a. If a NI/RPS channel is in channel bypass only the corresponeins EFIC channel may.be bypassed'.
- b. If an EFIC channel is in maintenance bypasse and any other but the corresdponding NI/RPS channel is placed in bypass, the bypass for the EFIC channel will be removed.
- c. Placing a NI/RPS channel in bypass will cause the corresponding EFIC channel's maintenance bypass indication to actuate and the bypass condition for the EFIC channel will be annunciated, even though the EFIC channel is not actually in maintenance bypass,
- d. Only one EFIC channel may be bypassed at a time.
- e. The EFIC maintenance bypass feature does not bypass the EFW initiation from ES-HPI.
=
L ._
~ f.: :P [6. . PLANT SYSTEMS DESIGNr CONTROL, AND-INSTRUMENTATION PAGE 39
.ANSHERS --~ CRYSTAL-RIVER -86/08/18-LAWYER, SANDY-i REFERENCE-CRr ROT 14-15r p 21. .
ANSWER' 6.13 E(1.00) .
- a. operator error
-b. electric failure
,c. hydraulic failure REFERENCE CR, Requal fuel handlins Llesson plan rev 1r p 13; CR, FP 601, p6.
ANSWER 6.14 (1.00) Preaction sprinkler system (dry-pipe) REFERENCE C,Rr,Trainins_ letter TRA.0013. ANSWER 6.15 (1.00) a..RMA-3 - ensure stopped AHF-11A and 118. .. (0.66) (b.,RNA-4 -ensure' stopped AHF-10. ~ it{j r (0.33)'
' $%F * ', . REFERENCE CR, AP-243 and AP-244i p 2.
ANSWER 6.16 (1.00) Trom pressure error-REFERENCE
'CRr STM 504, p 82.
ANSWER 6.17 (1.00) , Venting the PZR to the makeup tank.
+t
t,
^
60- - PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUMENTATION PAGE 40
. ANSWERS -- CRYSTAL RIVER -86/08/18-LAWYER, SANDY
)
REFERENCE CRr OP-301, (GB). - ANSWER 6.18 (1.50) DHV-3,4,41 and 91 are monitored. (0.5) 284(+or-15 psi) j (0.5) D w atically close,(if DHV-41 or 91 are- open F (0.5) , REFERENCE CRr ROT o-ir p 77, (GB). ANSWER 6.19 (1.00) High level dumps for the HP flash tanks. REFERENCE' CR, STM 628, p 15. - ANSWER 6.20 (1.00)
- 1. Active - these channels are contino'u.s'ly monitored.
2.--Passive-thesechannelsmustbejlsgallyconfigured.tobe monitored. , REFERENCE- . CR, ROT 4-2.,, p (8) (0B). gR 6.21 (1.00) 3 A RB pressure of 4 psi arms the RB spray system by opening the NaOH tank
-outlet valves (BSV-11 a 12) and by openin3 the RB spyay pump suction valves g & -4) to the 1550'spm pos
/
o c4 500 RCS W e7pition - AoLo4. h L}-p s ig. m M r ES Q p
.B. M Q ,
p 89j SOI 4~#3'# M CR, STM-5, pp 1-4; CRr ROT 4-1, r > >
.a
. , Jt' '..
- e. ,.,
, ;.1. g n.
6
~
-6.' PLANT SYSTEMS-DESIGNr CONTROL, AND INSTRUMENTATION \ PAGE 41' ANSWERS -- CRYSTAL RIVER -86/08/18-LAWYER, SAhDY s
' ANSWER 6.22 (1.00) -
Frior.;to the prefilters and after the block ~ orifice and check valve 252. REFERENCE-CRr;STM 17, pp 4rSi CR, ROT 4-1, fis 9. ANSWER- 6.23 .(1.50) The' amber lamp will be lit and-breaker 3210 will"bn openfond cannot'be m closed. (1.0) This is the cross tie interlock provided to prevent paralleling the DGs. (0.5) RECERENCE
- CRr.STM-10, pp 56,57; CR, ROT 4-3r p.'17.
4
~ ANSWER ~ 6.24 (1.00)
It is used ,to. cross c'onnect the auxiliary' power supply and DC hold bus.
!CR,REFERENCE STM 504rf 10B') /]e.'.:[.^
' ANSWER 6.25 (1.00) .
If a neutron demand (reactor demand vs actual neutron power) of >5% is senerated, a signal will be senerated to modify feedwater demand either up or .down as appropriat,e to match [ actual: neutron power (thTeedwaterg(r5%) _ _ - _ y REFERENCE CR, STM 504r'p 79, (GB). 4
-- ' ANSWER 6.26 (1.00) p. O
~
An uninterruptable power supply located in 'A' switchgear room.
-) A s
*M s
I e
.e --
.q o
1
~ 6 ~. - PLANT' SYSTEMS.DESIGNr CONTROL., AND INSTRUMENTATION- PAGE ~ 42 a
' ANSWERS -- CRYSTAL' RIVER -86/08/18-LAWYER, SANDY REFERENCE CR, dws -302-693, (09)~. .
ANSWER 6.27 (1.50)
- a. hot standby- . .
- b. . steam'discherse to etmo' sphere
- c. 24 hours. ,
REFERENCE CRr.TS,'r B'3/4 7-2. .
' ANSWER 6.28 .(1 00) ,
k When'both. throttle valves-in one steam chest are open. REFERENCE 4 CR, NA0 83, p 1; CR, NAO 83, dws FD-302-011. g. 3r'"4/et i e e e 4 t
3' (_ 8
?, i t *,
- 17. ~ PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND . -,- PAGE 43-c p'
' RADIOLOGICAL. CONTROL 1 3_
. ANSWERS - -CRYSTAL RIVER -86/08/18-L AWYER e' SIVIDY.
s t-
.)
-ANSWER- 7.01 .(1.00) .
, u w
..(d)
.d
' l';
1 l REFERENCE CR,'OP-204, Rev.'42, p 8. '
', i i
~ ANSWER 7.02 -(1.00)
~
v' s (b) REFERENCE' . CR, 0P-302,.p 2. ANSWER 7 03 (1.00) & ) er (b) REFERENCE ' 10CRF20.101 R 1/1/86, p 258. .; ,
, .- $(
- h .
- - ANSWER 7.04 (2.00) # J '; * -
- a. True . ,
- b. Felse
- c. False
- d. True- - '
REFERENCE CR,-OP-412, pp 8,11,14,16. j g ANSWER' 7.05 (1.00) a.-criticality by deboration
+
-b. boron concentration ,
REFERENCE - t 1 CR,-OP 210, p 10. > fr
. ~;
- 7. PROCEDURES - NORMALr ABNORMAL, EMERGENCY AND PAGE 44
~~~~ ~ ------------------------
RA656LUEECEL CUNTR5t ANSWERS -- CRYSTAL RIVER -86/08/18-LAWYER, S$RDY ANSWEP 7.06 (1.00)
-a. Less than adequate subcoolins marsin
- b. Less then-2 HPI pumps are available REFERENCE
.CR , AP-530, p 3. -
ANSWER 7.07. (1.00)
'a. As directed by ' health physics' personr el
- b. .As specified by appropriate 'RWP"
- c. Prior to reachins 3/4 scale (ier normally 150 mt)
REFERENCE CR, RSP-101 rev 5, p 9. ,
. ANSWER 7.08 (1.00)
- c. Observe knob (labeled "0FF, BATTe X100, X10, X1') is set on 'X1".
- b. Observe ' Response' switch is.i4 *gthe ' slow' position.
- c. Observe meter movement foru f 'le deflection.
d.~ Listen for audible response. , REFERENCE CRr RSP-101 rev 5, pp 19,20. ANSWER 7.09 (1.00) -
- o. The PORV block valve is operable (capable of cl_ sing).
- b. The operator on the switch does not leave the PORV switch unattended (until he assures that the PORV is closed and there is no flow thru it).
REFERENCE CRr OSIM, p V-14.
*5
e
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 45 RADIOLOGICAL CONTROL ANSWERS -- CRYSTAL RIVER -86/08/18-LAWYER, S4sDY ANSWER 7.10 (1.00)
- a. LBPR retainer
- b. Source retainer REFERENCE CR, FP-203 rev 17, p 9. '
ANSWER 7.11 (3.00)
- c. 1) ensure turbine runback (1.0)
- 2) ensure RC pressure stable
- b. ensure 1) affected MFP tripped (1.0)
- 2) open FWV-28 (cross tie)
- 3) closed FWV-29
- 4) control rods insertins -
- c. i' 55% (*. 0) ii 55%
iii,60% i [h g' M 7gcuSM-OHj j F. A ,. ANSWER 7.12 (1.00) ,
- a. notify plant personnel of the fire location over the PA twice.
- b. depress the " fire eierm' pushbuttons.
- c. ensure operation of autonatic fire protection system.
- d. establish continous monitorins of PL-1 arid portable radios,
- e. verify fire brigade is responding to notifications.
REFERENCE CR, AP-880 R3, p 2. ANSWER 7.13 (1.00) OP-408, Nuclear Services Coolins Systen.. REFERENCE CR, OP-408 R44, p 9.
==
,n .
- 7. PROCEDURES - NORMAL, ABNORMAL,' EMERGENCY AND PAGE 46-
~ ~~~~~~~~~~~~~~~~~~~~~~~~
~~~~R5656LU6565L 66 TR6L
-ANSWERS - CRYSTAL RIVER -86/08/18-LAWYER, SAADY'
-ANSWER' 7.14 (1.00)
~ Notif y th.e - Au:: building operator'to ensure closed
- WDV-891rsnd 892.,
REFERENCE CR, AP-272 R0r p :2. ANSWER. 7.15 (2.00)
- 1. .Announce over. the PA. system that the control room is.being evacuated.~ '
- 2. Depress.' reactor trip' push button.
- 3. Ensure the tvrbine trips
- 4. Actuate emergency feed flow
- 5. Assure OTSC levels are beins controlled
- 6. Actuate ~ main feed isolation
.7. Actuate main steam isolation
.8. Close MUV-49 9.~ Ensure power is available to all ES buses 10.Close RCV-11 (PORV. block valve). (10-2 0 20 ea)
REFERENCE
- CR, AP-990 R01r,pp 213.
ANSWER 7.16 (1.00) Place the Te load ratio control station in ' hand'. REFERENCE CRr OP-204, p19. ANSHER 7.17 (2.00)
- c. If subcoolin3 marsin is. inadequate - >1500 psis 20 desF i
=or<1500 psis 50 desF.
- b. If.all RCPs have not been stopped within two minutest they must be kept running.
REFERENCE CRr AP-380 R6, p 4.
- 7. PROCEDURES - NORMALr ADNORMAL, EMERGENCY AND PAGE 47
~ ~~~~~~~~~~~~~~~~~~~~~~~~
~~~~R5656LUGEddL UUUTR6L ANSWERS -- CRYSTAL RIVER -86/08/18-LAWYER, SANDY ANSWER 7.16 (1.00)
This is a core hydraulic lift consideration. REFERENCE CR, OP-202 R70, p 44a.(GB) ANSWER 7.19 (1.00) g This could result in an EFW actuation. REFERENCE CRr OP-203, (DB). ANSWER 7.20 (1.00) To prevent challensins the secondary safety valves. REFERENCE CR, EP-390, p 8, (09). , Kir 427jO'r ANSWER 7.21 (1.00) .
- a. lifting secondary safety valves
- b. flooding the steam lines
- c. excessive cooldown causing a PTS event REFERENCE CR, ATOG svidelines, (GB).
ANSWER 7.22 (1.00) The additional weight and its distribution may negate the seismic design of the pipins. REFERENCE CR, CP-113 R49, p 32.
*t
,s.
" * :" :.. :9
' 7.; PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND . PAGE 48
~
'~~~~Rd656E65EddL EUNTREE~~~~~~~~~~~~~~~~~~~~~~~~
ANSWERS -- CRYSTAL RIVER- .-86/08/18-LAWYER, SARDY
'~-ANSWER- 7.23- (l'.00)
/>1000 mr/hr - 1. flashing light 2.11ocked gate or guard .
REFERENCE
=CR, RSP-101 RS, p 17.-
- J ANSWER -7.24 (2.00) ,
ie. all
- b. ORA.
- c. 'none d.15'hourt REFERENCE '
CRr -EM-202 R26, pp 44,45,50. . ill
- hhf - *
' j!yf]4 .
? ,1 e
e S s
't
- r. -
8, ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 49 ANSWERS -- CRYSTAL RIVER -86/08/18-LAWYER, SANDY t ANSWER -8.01 (1.00) . (c) REFERENCE
'CR, .TS'3 8.1.2, p 3/4 B-6.
-ANSWER 8.02 (1.00) -
-(d)
REFERENCE CR, FP-203, p14. ANSWER 8.03 ( 1. 00 )- (c) REFERENCE . . CR, TS 3.4.5,'p 3/4 4-6 and fis 3.4-5. AN$ WEE. 8.04 (1.00)
.w 4.
gg, r -.- REFERENCE . CR, TG 3.2.4, p 3/4 2-8. ANSWER 8.05 (1.00) (a) REFERENCE CR, EM-202 R19, p 3. ANSWER 8.06 (1.00) (a)
*2
em
- 8. ADMINISTRATIVE tROCEDURES, CONDITIONS, AND LIMITATIONS DAGE 50 ANSWERS -- CRYSTAL RIVER -86'08/1E-LAWYERr SANDY t
REFERENCE CR, TS bases figure 2.1, p B 2-8; CR, TS, pp B 3/4 2-2,3; - ANSWER 8.07 (1.00) (b) REFERENCE CR, TS, pp 4-5, 1-16, 5-1, 1-14. ANSWER 8.08 (1.00) (d) REFERENCE CR, TS, p 3/4 0-1. ANSWER 8.09 - (2.00)
- a. Must obtain a qualified relief operator at the controls.
- b. He remains within the confines of the control center (.5) and maintains an unobstructed view of the operational control panels (.5).
REFERENCE CR, AI-500, p 6~. ' ANSWER 8.10. (1.00) 0.3 spn (+or- 0.1); up to one week (+or- 2 days) REFERENCE ' CR, OSIM R47, p VI-3.
).
ANSWER 8.11 (1.00)
- 1. Single line thru the incorrect entry
.2. Enter correct information adjacent to or in a space aveilable with reference to deleted entry
- 3. Individual making correction shall initial and date error correction. .
=%
GI . 1
;8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND~ LIMITATIONS lPAGE 51 ANSWERS --' CRYSTAL RIVER -86/08/18-LAWYER, SANDY t
- REFERENCE.
CRr AI-500 R 53' P 9' - ANSWER 8.12 (1.00)
.15%'in 1 hour.
REFERENCE-CR, OP-204 R42,.p'8. ANSWER 8.13 (1.00) . Threet provided the RPS setpoints are loweredi to 67.5% RTP. (0.33 ea) REFERENCE CR, TS 3.7.1.1 R77, pp 3/4 7-1,2. ANSWER' 8.14 (1.00)
'1.
Turbine EH. control system
- 2. 500-KV protective relaying ,,
3.' 4160V ES and unit protective relayinc 11 .
- 4. . Generator and exciter protective relaying ], Ijj~ ;j.
i
- 5. Steam line rupture matrix j REFERENCE CR, OSIM-R47, p VI-2.
-ANSWER 8.15 (1.00)
- 1. NI-6 be placed in a tripped condition
- 2. QPT be monitored at least once per 12 hours.
REFERENCE CR, TS 3.3.1.1, pp 3-3,4. C
r ,
- 8. -ADMINISTRATIVE PROCEDURES, CONDITIONSr AND LIMITATIONS PAGE 52 ANSWERS -- CRYSTAL RIVER -86/08/18-LAWYER, SANDY t
ANSWER 8 16 (1.00) . a.. Helium lea!t detection
- b. It has steater sensitivity, allowin3 a lower limit to be the criteria.
for water box removal for tube repairs. ANSWER 8.17 (1.00) , They are warranted if the recuction in individual dose txpected to be achieved by carryin3'out the protective action is not offset by. excessive risks to individual safety in takin3 the protective action. REFERENCE CRr RERP RS, p 4-2. ANSWER 8.18 (2.00)
- a. The clearance is to be issued for an unusual, non routine or abnormal evolution.
- b. The clearance to be issued cannot meet the double valve protection guidelines of 500 psis-a.nd/or 200 desF and =or> 1/2 inch diameter openin3+ /4Ie'dannot meet the ES train separation' criteria.
- c. The clearance to be i's' >
d.-The clearance to be issued gennot meet the limiting contitions for voluntarily entering a destaded mode of operation. REFERENCE CR, CP-115. ANSWER 8.19 (3.00)
- c. 1) If the SS declares the PORV inoperable 0 0200, then 0300. or
- 2) If the SS does not declare the PORV inoperable G 0200, then 0330.
- b. Hot Standby within 7 hours from the time stated in part a.
- c. Yes. Provisionss of 3.0.4 are not applicable.
REFERENCE CR, TS 3.4.3 2 R77, p 3/4 4-4a. 1 4
=Y
(
~
FE = 85 ADMINIETRATIV
E. PROCEDURE
S,. CONDITIONS, AND LIMITATIONS PAGE 53
. ANSWERS -- CRYSTAL RIVER -86/08/18-LAWYER, SANDY t
ANSWER- B.20 (2.00) -
- s. Yes - Note 2, . Table 3.4-2, 2 spm.is less than 2.5 spm 50% of band.
b, Note 3,' Table 3.4-2, reduce to within-limits within 4 hours or'be in' hot standby within the following 6 hours. REFERENCE CR,-TS R4/20/81, pp13/4 4-159.16,10a. ANSWER. 8.21 (2.00)
- a. SOTA Nuclear Operations Superintendent .
Person on call NRC (red phone) NRC resident inspector
- b. Completes steps 1 - 9. -
Assigns nent1 consecutive report number agi t 1 Operations Superintendent for disposition.
; REFER NCE~
JilW ,7 AI-500 R53, p 17. ' ipskh- < ANSWER- 8.22 (1 00) < - c..Short term instructions are any miscellaneous instructions that may arise and shall be used for routine maintenance and personnel instruction where plant safety is not affected. (0.5)
- b. 90 days .
(0.25)
- c. Nuclear shift supervisor (0.25)
REFERENCE CR, AI-500 R53, p 14. ANSWER 8.23 (1.00) Immediate Temporary changes (ITCs) are normally of an utsent nature such
'that time and/or plant conditions necessitate implementation of changes prior to the time required by Interim Changes or Permanent revisions.
t1
e 1
...o
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 54
TWERS -- CRYSTAL RIVER -86/08/18-LAWYER, SANDY t
ANSWER.s g,24 (3,00)
- a. Evacuate all people within a 2 mile radius, and
- b. Shelter ell people w: thin 5 miler in the potentially affected sectors.
REFERENCE CR, EM-202 R26, p 28.
.t 11%
e I f U
1
- b. ~
< -80 see period Log CR [ d _
N .. _
,80 see period +
e Time - one e ...,e.w j 5 I . i 05 Answer R J ure r
-m
-+em6 + - enw e
~i
_A mehuu a. D ed
> F.m'hu h{}}JtME D t)P) 9 6
MAsnL U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: CRYSTAL RIVER g REACTOR TYPE: PWR-B&W177 DATE ADMINISTERED: 86/08/19 EXAMINER: HUENEFELD, J. CANDIDATE: ANSWER KEY " INSTRUCTIONS TO CANDIDATE: Use separate paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question a're indicated in parentheses after the question. The passing grada requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 30.00 25.00 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 30.00 25.00 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 30.00 25.00 3. INSTRUMENTS AND CONTROLS 30.00 25.00 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 120.00 . Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.
Candidate's Signature
e-3
.o g- ,
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: 1.. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave.. You must avoid all. contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil only to facilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the.date on the cover sheet of the examination (if necessary).
~
. 6. Use'only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet.
- 9. Number each answer as to category and number, for example,1.4, 6.3.
- 10. Skip at least three lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer l
to mathematical problems whether indicated in the question or not. i
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
.~
- 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has -
been completed. 9
L t *
- 18. When you complete your examination, you shall:
- a. Assemble your examination as follows: -
(1) Exam questions on top. (2) Exam aids - figures, tables, etc. (3) Answer pages including figures which are part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions. -
l
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
- d. ' Leave the examination area, as defined b'y the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
O l 1 l
a t 1.~o PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2 THERMODYNAMICS, HEAT-TRANSFER AND FLUID FLOW QUESTION 1.01 (.50) When at normal operating temperature, H0W.many inches will PZR -
; level change for one deg. F change in T-ave? (0.5)
- 'AVESTION 1.02 -'(1.50)
'At-higher RCS temperatures, will a GREATER or LOWER letdown volumetric flowrate rate? be required
-PROVIDE a one sentence to maintainexplanation. pressurizer level for the same heatup(1.5)'
QUESTION 1.03 (1.50) If a Reactor Coolant Pump is secured, WHY won't the seal cooler system provide sufficient cooling? (1.5) QUESTION 1.04 (2.00) in
' SPECIFY DESCRIBE and EXPLAIN the relative values of feedwater the ICS flow, response T to fouling (c)'one OTSG.
s,T(h)'s, and OTSG levels. (2.0)
. QUESTION 1.05 (1.50)
The reactor is operating at 75% power with ICS when one RCP in the
- "A" loop trips. When the transient is over, WHAT fraction of feedwater flow will go to the "A" loop and WHAT fraction of the .
a total flow will go to the "B" loop? (1.5) .. i i
' QUESTION 1.06 (1.50) i STATE three (3) reasons why the initiation of Emergency Feedwater provides for an increased rate of heat transfer over the use of Main Feedwater. (1.5) !
l (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
---wm-,n--,---~vgev--a am_ --ww.w--e+ ~ - -~g,w - - + - -,, e g ----,=,,--m, ,,-o.wm,c-n.,-m-nv, -,,-,.,--,,-c.--we---,,~ , w-, ,,eweee -e-re-eey--m, , , - .w-m-- wm,
E 1
'1 PRINCIPLES ~OF NUCLEAR POWER PLANT OPERATION,
.t PAGE 3
' THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
.-QUESTION 1.07. (1.00)
WHAT parameter is used by EFIC to determine the actual ramp rate _for controlling the rate at which OTSG level is increased? ~(1.0) i QUESTION 1.08 (1.00) i
'The actual value of Reactor Coolant System pressure is decreasing rapidly. The actual values of Pressurizer Level and Makeup Tank Level are remaining the same. WHAT is the cause?
(1.0) , -QUESTION -1.09 (1.50) IS imbalance narmally kept on the positive side or the negative side of 07 WHY? (1.5) QUESTION 1.10 (1.00) , Should method preferred a xenon for oscillation dampening the start, WHICH oscillation? one of the (SELECTone. following (1.0)
)is the
. (a.) Make a correction when the oscillation fs at its average value by driving imbalance to one of the operational extremes.
I' (b.) Make a correcticn 1 to 2 hours before the peak value by driving imbalance toward its average value. (c.) Make a correction 1 to 2 hours after the peak value by driving imbalance toward its average value. (d.) Make corrections continuously to hold imbalance at "0".
~
- QUESTION 1.11 (1.50)
- DEFINE Shutdown Margin.
(1.5) l l I 4 (***** CATEGORY 01CONTINUEDONNEXTPAGE*****)
.,.,,.7m .,++,.4.,--.,,y,._, -.,_,_,,,..._.__m._,.,..,__,,,m_.,. ..,,-,_my.-,, . _ . , ,_,,_,__-,_,,%,m.,-...--..4--,--
~'.
t 1.
- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 4 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.12 (2.00)
STATE the four (4) criterion that are used in accordance with VP-580 to-ensure that Natural Circulation is occurring. (2.0)
~ ~
QUESTION 1.13 (1.50) WHAT special and unique reactivity concern would be associated with an event involving an OTSG tube rupture in an OTSG that is experiencing a steam leak inside the reactor. building? (1.5) 1. 4 (***** CATEGORY 01CONTINUEDONNEXTPAGE*****)
----r, - ,, , e- . . _, . , .,- . ,,, -, ,. ..- , - .-- - , - , . , . . _ , , . , . , . _ . , , , , - . , - . , - . .
4 v 1.
- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION2 PAGE 5 l
- THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.14 (3.00)
- a. CONSTRUCT a plot of the concentration of xenon for the following power history. (2.0) 1'~ ' .' --
,, ',, [' %
"* $ o 40 80 120 IGO TIME (HR)
, , g 150 f ,
- d w .
8 :.
~ '
100 ,, , x* - so I t f I O 40 80 120 160 TIME (HR)
- b. EXPLAIN the shape of the xenon trace you have drawn for the power reduction from 100% to 50%. (1.0)
(***** CATEGORY 01CONTINUEDONNEXTPAGE*****)
, 1
- PRINCIPLES OF NUCLEAR POWER PLANT OP'ERATION, PAGE 6 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
- QUESTION 1.15 (3.00) 1 t
^
,,a,0 ! . . .
'A
- o
- ullE Key: Dotted line is startup rate.
Solid line is neutron countrate. Above is a graph of simulated reactor response to two (2) simple reactivity manipulations. The first is a rod withdrawal from criticality. The second is a rod insertion. Assuming no reactivity - feedback and negligible source effects, ANSWER the following questions.
- a. On the graph itself, INDT.CATE the periods of time and direction of rod motion. (1.0)
- b. Are the amounts of reactivity in:erted the same for both the insertion and the withdrawal? (0.5)
- c. EXPLAIN the shape of the plot of startup rate. (1.5) j QUESTION 1.16 (1.00) .
The reactor is made to be subcritical by a very small amount of reactivity (less than the value of beta). WILL the resultant stable negative startup rate approach -1/3 dpm? EXPLAIN. (1.0) J (***** CATEGORY 01CONTINUEDONNEXTPAGE*****)
-y _ - ,.,.- __m, r ,,__,.- - - , s,,_-y, y~- __--y. , . _ _ , - . . , _ . - _ _ _ , _ . - _ _ _ _ . _ - . , , . _ . _ , . _ _ _ _ _ _ _ _p,, , , . ., y._% ,_, ,
-s
;y .1.* PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 7 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.17 (1.00)
Over core life, the reactor becomes more responsive for a given
-reactivity change due to: (SELECTone) (1.0)
- a. Pu-239 causing an increase in Beff,
- b. Pu-239 causing a decrease in Beff. _
- c. Pu-240 causing an increase in Beff. ~
- d. Pu-240 causing a decrease in Beff.
QUESTION 1.18 (1.00)
. During fuel loading, WHICH of the following will have NO effect on the shape of a 1/M plot? (1.0).
(a.) The location of the neutron sources in the core. (b.) The strength of the neutron sources in the core. (c.) The location of the neutron detectors around the core. i (d.) The order of placement of fuel assemblies provided the
- proper enrichments are placed in their proper location.
QUESTION 1.19 (1.00) OP-210, " Reactor Startup," requires that the critical rod position i
- be taken at 10**-8 amps on the intermediate range. If, during a xenon free reactor startup at MOL, the operator " overshot" 10**-8 amps and instead leveled off at 10**-7 amps, WHICH of the following
, statements is CORRECT? (1.0) l (a.) At 10**-7 amps, there are.little or no effects from nuclear heat but since the reactor is a decade higher in power, the j critical rod position would be higher. -
'(b.) At 10**-7 amps, there are little or no effects from nuclear l heat; therefore, the critical rod position should be the same as at 10**-8 amps.
(c.) At 10**-7 amps there are substantial effects from nuclear I heat; therefore, the critical rod positions will be higher - than at 10**-8 amps. (d.) At 10**-7 amps, nuclear heat, xenon and the decade higher in power level will result in a higher critical rod position. (***** CATEGORY 01CONTINUEDONNEXTPAGE*****)
--,-...e--rwr-ew,.e.e -
.,,,.e-n-,~,,,nn.en,..
')g 1. , PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 8 THERMODYNAMICS. HEAT TRANSFER AND FLU 10 FLOW QUESTION 1.20 (1.00)
Concerning-the behavior of samarium-149, in the reactor, WHICH of the following statements is CORRECT 7 (1.0) (a.) Most of the Sm produced comes directly from fission. (b.) Most of the removal of Sm is by radtoactive decay. ~ (c.) Sm reactivity is independent of flux once it has reached equilibrium. (d.) Equilibrium Sm is reached about 40 hours' after the initial startup of the reactor. QUESTION 1.21 (1.00) WHICH of the following statements about pump Net Positive Suction Head (NPSH) is CORRECT? (1.0) (a.) NPSH is the amount by which the saturation pressure is greater than the suction pressure for the water being pumped. (b.) When a pump is started, the NPSH will decrease by the amount of the pressure drop in'the suction piping. (c.) NPSH is essential for operation of centrifugal pumps but not for positive displacement pumps. (d.) NPSH can be calculated by subtracting the suction pressure from the discharge pressure. (***** END OF CATEGORY 01 *****)
,---,,-,,e.m --
- - - - - , -y -,-- ,,- - -- - --- ----r--,w,- - , -.--- - - - , - - - - - - - - - - , - , - - - - , -
i
)
l I si 2. OPLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 9 I l QUESTION 2.01 (1.00) During normal ' operations,,about HOW MUCH of the pressurizer is full c of liquid water: (SELECTone) (1.0).
. 1/4
- . 1/3 i . 1/2 ~ ~
. 3/4 QUESTION 2.02 (.50) j ANSWER TRUE or FALSE.
^ DHV-41, the Decay Heat Removal suction line containment isolation valve, is rated for full system temperature and pressure. (0.5) i QUESTION _2.03 (1.00) 4 WHAT is the potential inadvertent consequence of improper operation
- of the suction valves to the Decay Heat Removal pumps from the Reactor Building sump (DHV-42, and 43)? (1.0)' -
QUESTION 2.04 (1.00) STATE the shut-off head for the LPI pumps. (1.0) l t QUESTION 2.05 (1.00) - WHAT is the principle gas of concern present in the Low Pressure . i vent header? (1.0) 4 E (***** CATEGORY 02CONTINUEDONNEXTPAGE*****) 1 l.,,,-.----.,-..... . . - - , - . . . . , - . - - - . , . - - - - - - - . - - - - - - - . . - - - - . - - - - - - - - - - - - - - - - . I
- e 2. ' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
- PAGE 10.
i QUESTION 2.06 (1.50)-
- a. - WHY must a minimum discharge pressure of the' waste gas compressors be maintained? (1.0)
- b. WHAT cooling water system provides cooling to the waste gas compressor. heat exchangers? (0.5)
QUESTION 2.07 ('.50) ANSWER TRUE or FALSE.. When the Waste Gas Decay Tank Sequencer is used to manually shift waste gas tanks, the on-coming waste gas tank inlet valve will open before the off-going tank's inlet valve closes. (0.5) QUESTION 2.08 (1.00) 4 The RCS Hot Leg vents contain orifices that limit flow through the vent lines to: (SELECTone) (1.0) ' (a.) 59,000 lbm/sec saturated steam and 100,000 lbm/sec saturated liquid (b.) 59,000 lbm/sec saturated liquid and 100,000 lbm/sec saturated
~
steam-
.(c.) 59,000 lbm/hr saturated steam and 100,000 lbm/hr saturated liquid
-(d.) 59,000 lbm/hr saturated liquid and 100,000 lbm/hr saturated steam QUESTION 2.09 (.50)
~
WilAT is the design capacity of the PORV? (0.5) . 4 QUESTION 2.10 (1.50) STATE the normal and alternate cooling water suppites to each of the three (3) makeup pumps. (1.5) f (***** CATEGORY 02CONTINUEDONNEXTPAGE*****)
, 2.. PLANT DESIGN-INCLUDING SAFETY AND EMERGENCY SYSTEMS- PAGE 11 ' QUESTION 2.11- (.50)
HOW MANY design start attempts will the diesel air start reservoirs support? (0.5) QUESTION 2.12 (1.00) _ _ STATE one (1) of the-three (3) nameplate ratings for the diesel engine. (1.0) QUESTION 2.13 (1.00) WHAT two (2) things will have to be done to the EFW suction valves to the CST (EFV-3 and 4) in the event that they are to be. operated to align the EFW system with the Hotwell? (1.0) QUESTION 2.14 (1.00) WHICH RCPs are supplied from WHICH 6900 VAC busses? (1.0) QUESTION 2.15 (1.50) WHAT do the Low Load control valves do when the Main Block valves open during a nonnal power escalation? WHY? (1.5) QUESTION 2.16 (3.00) a.
~
The "B" makeup pump may be powered from either of the 4160 VAC . safety features busses. Basically DESCRIBE the design features that make this transfer of power possible without electrically paralleling the two busses. (2.0)
- b. If the "B" makeup pump is to be selected as an ES HPI pump, WHAT must the ES switch lineup be at the 4160 VAC busses? (1.0)
(***** CATEGORY 02CONTINUEDONNEXTPAGE*****)
. ~
-o 2.
- PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS . PAGE 12 QUESTION 2.17 (1.00)
During normal operation, from WHERE does the power for the two (2) 4160 VAC busses come? (1.0)~
~
i
~ ~
QUESTION 2.18 (2.50) LIST those Main Feedwater valves that will be closed by EFIC wheni a Main Steam Line Isolation occurs on one side. (2.5) QUESTION 2.19 (1.50) Each Main Steam Isolation valve has an external bypass assembly consisting of an exterior pipeline running from the underseat area of the valve body to the area underneath the bonnet with an in-line stop check valve. WHAT is the purpose of this bypass ' assembly? INCLUDE the stop check valve in your. discussion. (1.5) s.s . no QUESTION 2.20 (1.00) x WHERE are the main steam line radiation monitors relative to the main steam safety valves and the branches to the atmospheric dump valves? (1.0) s QUESTION 2.21 (.50) WHAT is the shutoff head of the feedwater booster pumps? (0.5) QUESTION 2.22 (1.00) , There are two (2) "C" breakers and two (2) "D" breakers in the CRD system. WHAT do each supply? (1.0) -
'~
(***** CATEGORY 02 CONTINUED ON NEXT PAGE'*****). %- r , y r -, - ----..-,-w, . - - - - . , e+w -- -----, ,-r-,- , ,. -- e - ,-e- ,- . - - - - , - - .- - , - - - . . - -
Le 2.
- PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 13
'_ QUESTION 2.23 (1.00)
WHICH one of the following statements is CORRECT regarding the hazards of letting an OTSG boil dry? (1.0) (a.) The shell becomes hotter than the tubes which can lead to tube buckling. (b.) The shell becomes hotter than the tubes which'can lead to exceeding yield stress in the tubes. (c.) The tubes become hotter than the shell which can lead to . tube buckling. (d.) The tubes become hotter than the shell which can lead to exceeding yield stress in the tubes. QUESTION 2.24 (1.00) WHICH one of the following components of the EHC control oil
. system are provided to maintain system pressure at approximately 1250 psi? (1.0)
. unloader valves
. relief valves
. high pressure accumulators
. pressure switches QUESTION 2.25 (1.00) 4 WHICH one of the following plant ventilation fans is operated from its local control station (as opposed to operation from the control room)? (1.0)
- a. turbine building ventilation supply fan
- b. steam generator compartment fan (AHF-4A)(AHF-25A)
- c. spent fuel pool supply fan (AHF-23A) ..
- d. Intermediate building ventilation exhaust fan (AHF-29A) .
1 (***** CATEGORY 02CONTINUEDONNEXTPAGE*****)
, , -,n.=...- -
e , - - - - - --, --- ,, ~ . , . ,., ,- a.. ,,,-., - ,, _ --. .. - , - , , - , , . , . , , --- ,n, ,.-----..,-a- - ~ , . . _ . , _ . - - -
+ ,- , 2. . PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 14 i
i QUESTION 2.26 (1.00)
~
WHICH one of the following makcup pumpilube and gear oil system pumps has NO auto start provision? . (1.0)
- a. main lube oil pump >
- b. main gear oil pump
- c. backup gear oil pump
- d. backup lube oil pump - ~
QUESTION 2.27 (1.03) , WHICH of the following trip conditions is common to both a main feedwater pump and a feedwater booster pump? (Remember: trip condition,notsignal.) (1.0)
- a. overspeed
- b. low deaerated level
- c. suction valves not full open
- d. main steam line rupture matrix actuation 1 ,
i l (*****ENDOFCATEGORY02*****)
- 3. . INSTRUMENTS AND CONTROLS PAGE 15 QUESTION 3.01 (.50)
ANSWER TRUE.or FALSE. The pressurizer spray valve may be controlled, in the manual mode, from its MCC. (0.5)
~ ~
QUESTION 3.02 (1.00) The pressurizer heaters are interlocked to deenergize upon reaching WHAT level setpoint: (SELECTone) (1,0)
. 100 inches
. 80 inches
. 40 inches
. 20 inches QUESTION 3.03 (1.50)
STATE five (5) of the seven (7) RCP interlocks. Setpoints are NOT required. (1.5) QUESTION 3.04 (1.50) L. WHAT action does the Decay Heat Removal System Automatic Closure Initiation cause? (1.0) b. HOW may it be bypassed? (0.5)
-QUESTION 3.05 (1.00)
GIVE a basic (two sentence) description of the principle of operation ' cf the RCS High Point Vent flow detector. (1.0) (***** CATEGORY 03CONTINUEDLNhEXTPAGE*****)
- , - - , _ _ , . . - - - . , - - , - . . - - - - _y _ .. ., -- ,.,---.c., -,- .- -. . , vv., . . . -.-, ,4#- , , . . - - - , , _ - , . , - - - - - - - - , - - . _ . _ -
c > t
. S.' INSTRUMENTS AND CONTROLS PAGE 16 QUESTION 3.06 (1.00)
WHICH, if any, of the following switches are NOT spring returned to the automatic or normal position: (SELECTtheappropriateanswer(s) or answer "none".) (1.0)
- a. The PORV control switch
- b. The pressurizer spray block valve control switch
- c. The pressurizer spray valve control iwitch
- d. The Decay Heat Spray line isolation valve control switch.
QUESTION 3.07 (2.00) STATE the four (4) ICS runbacks, their rates and their limits. (2.0) QUESTION 3.08 (1.00) WHAT three (3) conditions will prevent feedwater from taking T-ave control when reactor demand is taken to hand? (1.0) QUESTION 3.09 (1.00) WHICH of the following conditions will cause a reactor trip? (SELECT one.) (1,0) (a.) One RPS channel trips, another is in channel bypass (b.) One RPS channel trips on high RCS pressure, another trips on high RCS temperature (c.) One RPS cabinet is deenergized (d.) MFP "A" and MFP "B" test switches are both placed in test for RPS channel "A" . QUESTION 3.10 (.50) A reactor trip on high temperature occurs. After the reactor trip, - RCS temperature rapidly drops below the trip setpoint. When the operator looks inside the RPS cabinet at the high temperature trip modules, WHAT lights will be brightly lit on that module? (0.5) (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
- 3.* INSTRUMENTS AND CONTROLS PAGE 17 QUESTION 3.11 (1.50)
When the shutdown bypass switch is activated, WHICH trips are bypassed and WHICH single trip is substituted? (1.5) 4
~ ~
QUESTION 3.12 (1.00) WHAT happens if an HPI auto or manual actuation signal were to occur 4 while in a degraded voltage condition? (1.0) QUESTION 3.13 (1.00) ABOUT HOW long after the onset of a degraded voltage condition does it take before the Emergency Diesel Generator output breaker closes, reenergizing the safeguards busses? (SELECTone.) (1.0)
. 3 seconds
. 10 seconds
- c. 25 seconds '
- d. 50 seconds QUESTION 3.14 (2.00)
Besides miscellaneous valves, WHAT components are included in
" Block One"? (2.0)
, . QUESTION 3.15 (1.00) Under normal standby conditions HOW should the diesel governor speed
- droop be set? WHY? (1.0)
QUESTION 3.16 (1.00) . STATE two (2) ways that the EDG may be stopped locally, even with - an ES actuation present. (1.0) (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
,- - --n, e.-- - - - - --- -- ,r-,--,nn,-r,- ,,w--, -
, 3.* INSTRUMENTS AND CONTROLS PAGE 18 QUESTION 3.17 (1.50)
- a. WHAT fault actuates the EDG 86 Lockout? (1.0) b.- WHERE must this lockout be reset? (0.5)
QUESTION 3.18 (1.50) WHAT five (5) conditions will cause the EFIC system to initiate Emergency Feedwater? . (1.5) QUESTION 3.19 (1.00) WHY should you place the EFW control valves in " Hand" prior to resetting the EFIC logic? (1,0) QUESTION 3.20 (1.00) WHAT are the automatic actions that occur in the event that setpoint is reached on RMA-11 (Waste gas)? (1.0) QUESTION 3.21 (.50) At WHICH alarm level, ALERT or HIGH, will the interlocks for the Radiation Monitoring system activate? (0.5) QUESTION 3.22 (1.00) WHAT happens to the main feedwater block valves if the main feedwater crossconnect (FWV-28) is opened while they are open? (1.0) (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
o 3,-' INSTRUMENTS AND CONTROLS PAGE 19 QUESTION 3.23 (1.00) If both of the lights on a Bailey meter hand / auto control station are lit:- (1.0) (a.) the station will control in either auto or manual, whichever was last selected. (b.) the station will not control in either auto or manual" (c.) the station will control in auto since auto overrides when both are selected. (d.) the station will control in manual since manual overrides when both are selected. QUESTION 3.24 (1.00) WHICH one of the following statements concerning the Control Rod Drive Position Indication System is CORRECT 7 (1.0) (a.) The zero (0) percent switch is located 1.5 inches above the in-limit switch. (b.) The 100 percent switch is located 1.5 inches above the out-limit switch. (c.) The first rod in any group to reach the 100 percent switch will stop further travel of all rods in that group. (d.) When actuated, the out-limit switch will generate an out-inhibit condition on the Diamond Panel. I l l (***** CATEGORY 03CONTINUEDONNEXTPAGE*****)
. 3. INSTRUMENTS AND CONTROLS PAGE 20 QUESTION 3.25 (1.00)
A synchroscope moving slowly in the SLOW direct (counter-clockwise)
-indicates WHICH of the following? (1.0)
(a.) Machine frequency higher than bus frequency, phases matched. (b.) Machine frequency lower than bus frequency, phases not ~ matched. _(c.) Machine voltage higher than bus voltage, currents not in phase. (d.) Mach.ine voltate lower than bus voltage,' currents not in phase. QUESTION 3.26 (1.00) During a " rapid bus transfer," WHICH one of the following feeder circuit breaker operations will occur (within 6 cycles)? (1.0) (a.) both the outgoing and the incoming will close (b.) both the outgoing and the incoming will trip (c.) the outgoing will trip and the incoming will close simultaneously (d.) the outgoing will trip before the incoming will close (e.) the outgoing will trip after the incoming closes l l (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
. 3.* INSTRUMENTS AND CONTROLS PAGE 21 QUESTION 3.27 (1.00)
WHICH one of the following is CORRECT concerning the " Sequence - Inhibit lamp (amber)"? (1.0) (a.) When it comes on, it also generates an "Out-Inhibit" signal. (b.) This indication will be on until all safety groups are withdrawn to the out-limit. ~ ~ (c.) Sequence monitor number one (1) provides a fault indication when the groups are inserting or withdrawing out of sequence in the manual mode.
~
(d.) This lamp is controlled by two sequence monitors, utilizing Groups 5, 6 and 7 Absolute Position Indication (API) average signals. A I (***** END OF CATEGORY 03 *****)
*^
. 4.* PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE' 22 RADIOLOGICAL CONTROL QUESTION 4.01 (1.00)
WHAT administrative controls are applied to operating the DHR pumps in the recirculation mode? (1.0)
~ ~
QUESTION 4.02 (1.00) Whenever the reactor vessel head is removed, WHAT is done with DHV-3, DHV-4, and DHV-417 (1,0) QUESTION' 4.03 (2.00) WHAT five-(5) actions must the operator take to feed from the BAST to the MUT7 (2.0) QUESTION 4.04 (.50) Both the "A" low load block valve and the "A" startup valve are in
. hand. The "A" load block valve is shut. WHICH of these stations must be placed into auto first? (0.5)
QUESTION 4.05 (1.00) h Some facilities have experienced Rx trips caused by improper switching of the main lube oil coolers. Basically, HOW should this switch be accompitshed? (1.0)
~
.~
QUESTION 4.06 (1.00) The OSIM states two (2) criteria for determining whether or not it
.is permissible to utilize the PORV to control high pressure during an emergency. WHAT are those two (2) criteria? (1.0) -
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
..s_, ____ . _ _ _ _ _ _ . _ _ _ . - - - -
.. j
- 4.* PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 23 I RADIOLOGICAL CONTROL QUESTION 4.07 (1.00)
STATE the two (2) administrative radiation dose limits for an
- individual who does NOT have prior authorization from the Nuclear Plant Manager.
(1.0) QUESTION 4.08 (1.00) WHICH one of the following conditions does NOT constitute grounds for an immediate manual reactor trip: (SELECTone) (1.0) (a.) Two (2) main steam isolation valves on different loops have been inadvertently shut.
'(b.) Reactor power is (15% and main feedwater is lost.
(c.) Pressurizer level is 295 inches and decreasing. (d.) All- four safety groups drop into the core while at 100% power. QUESTION 4.09 (.50) ANSWER TRUE or FALSE. The Diamond station must be in Auto prior to placing the Reactor Demand li. ?uto. (0.5) QUESTION 4.10 (1.00) WHAT parameter are you procedurally directed to observe to independently verify that generator output voltage is, in fact, - 22 kV7 (1.0) . QUESTION 4.11 (1.50) The Runback Verification VP-540 states six (6) parameters that should ' be observed to check for increased RCS leakage. LIST those six (6) parameters. (1.5) (***** CATEGORY 04CONTINUEDONNEXTPAGE*****)
, . , , - . . , - , , , - . - - - - - - - -. - - ~ , ,n. _, . - - - . , , , - . . ~ .
W
. 4.* PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 24 RADIOLOGICAL CONTROL QUESTION 4.12 (1.00)
According to the Plant Safety Verification Procedure (VP-580), WHAT ' constitutes adequate subcooling margin? (1.0) QUESTION 4.13 (2.00) STATE the required OTSG levels for the following situations as specified in VP-580:
- a. RCPs operating. .
- b. No RCPs, adequate subcooling. .
- c. Less than required subcooling. .
- d. Less than 2 HPI pumps. 0.5)
QUESTION -4.14- (2.50) - i a. STATE.the emergency cooldown limits to be used in the event of an OTSG tube rupture, including the OTSG tube-to-shell limits. (2.0)
- b. ARE you required to maintain the fuel pins in compression limits during this emergency? (0.5)
QUESTION 4.15 (1.00)
- a. At WHAT flowrate must HPI recirc be established? (0.5)
- b. WHAT is the maximum allowable flow for an HPI pump? (0.5)
QUESTION 4.16 (1.00) HOW is tube-to-shell delta T calculated? (1.0) (***** CATEGORY 04CONTINUEDONNEXTPAGE*****)
-e- --w . y -w- - ---,,x , , , , . -
, , - - - w- --m-r-- --. - , - - - -w - -- ---- - , , , , , - - - - - - - - -
o
. -4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 25 RADIOLOGICAL CONTROL QUESTION 4.17 (3.00)
STATE Subcooling(margin the three is lost as a result of a small break LOCA.3) major actions, automatic initiated. (3.0) QUESTION 4.18 (3.00)
- a. WHAT condition must be satisfied prior to throttling HPI subsequent to a valid manual or automatic initiation? (1.0) b.- WHAT conditions allow securing HPI subsequent to a valid manual or automatic initiation? (2.0)
QUESTION 4.19 (1.00) WHICH of the following is an immediate action required in AP-320,
" Loose Parts monitoring system." (1.0)
(a.) Select " DISABLE" on "HIGH ALARM" toggle switch OR "Lo ALARM" toggle switch. (b.) Notify I&C technician to remove the LPMS tape deck. (c.) Adjust "FS range" switch on affected channel signal amplifier until alarm clears. (d.) Select the alarming channels to " audio matrix" position. (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
, - - - ,- ,,,,,en,- -- - - - - - -.--, --, - - - -
. 4.* PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 26 RADIOLOGICAL CONTROL QUESTION .4.20 (1.00)
The nuclear services cooling water to the CRD stators should be secured under WHICH of the following conditions. (1.0) (a.) Less than three stators are energized and the reactor is not critical. (b.) RC system. temperature is between 200 degrees F and 300 degrees F. (c.) -Air temperature around the drives in the service structure is 1.ess than 150 degrees F. (d.) Any time all CRDs are deenergized for an extended period of time. QUESTION 4.21 (1.00) WHICH one (1) of the following is an appropriate "immediate' action according to EP-120, " Inadequate Shutdown Valve?" (1.0)
- a. Stop cleanup.
- b. Establish letdown flow to MUT greater than 40 gpm.
- c. Start boric acid pump CAP-3A or 38.
- d. Verify reactor trip, go to VP-580.
QUESTION 4.22 (1.00) WHICH one of the following provisions is the responsibility of a reactor operator leaving the 95 inch elevation control complex RCA as part of the first step in his whole body frisk? (1.0) (a.) Ensure the equipment is turned on and operating on the highest scale.
.~
-(b.) Ensure proper operation of the instrument by noting that the instrument has been source checked.
(c.) The background reading should be less than 300 dpm/100 cm**2. (d.) Ensure the beta window control is in the open position. - (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) _ , , _ , _ _ . , - - - -w --
o 4. ' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 27 RADIOLOGICAL CONTROL QUESTION 4.23 (1.09) According to EP-290, " Inadequate Core Cooling," the hot leg vents should be used under two (2) separate conditions. WHICH of the following choices states those conditions? (See attached figure 4.23.) (1.0) (a.) Pressurizer press greater than 1500 ~psig; press and temperature in Region 2. (b.) Pressurizer press greater than 1500 psig; subcooling margin less than 50 degrees F. , (c.) Pressurizer press greater than 1500 psig; press and temperature in Region 3. (d.) Pressurizer press greater than 2300 psig; subcooling margin less than 50 degrees F. (e.) Pressurizer press greater than 2300 psig; press and temperature in Region 3. b (***** END OF CATEGORY 04 *****) (*************ENDOFEXAMINATION***************) ,
Crystal Rivar Page 27a QUESTI0ft4.23(contd) CORE EXIT FLUID TEMPERATURE FOR-INADEQUATE CORE COOLING 26oc , , , , , , , , 2400 - SAT- LINE _ 2I00 ~ REGION ~ 2000 - - TCLAO DI400
~
C REGION
~
2 1600 - , g1400 - REGION kI200 - 3 M 1000 _ REGION , 4 800 - , . - 600 - - TCLAD >l800 400 - - - 20C 400 600 800 1000 12 0'O CORE-TpERNOCOUPLE TEMP (*fra . FIGURE 4.23 '- l
- --- , ,, - - -- , , , - - - - - . - -- - , - ,, - - - -- - - , - , - - , - 9 - , .- --.-----,.-,w,-- y - v.
- 1.* PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 28 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- CRYSTAL RIVER -86/08/19-HUENEFELD, J.
ANSWER 1.01 (.50) 5 inches [+0.5] REFERENCE
- 1. Crystal River: STM, Ch. 419, p. 15.
ANSWER 1.02 (1.50) Greater [+0.5]. This is due to the increase in the expansion rate of water at increased temperatures. [+1.0] REFERENCE
- 1. Crystal River: STM, Ch. 419, p. 45.
ANSWER 1.03 (1.50) Because, with the RCP secured, the recirculation impeller will no longer be supplying any flow through the Seal Cooling heat exchanger. [+1.5] REFERENCE
- 1. Crystal River: STM, Ch. 419. Figure 1.4-6.
ANSWER 1.04 (2.00) The ICS Delta T(c) circuit will carry out its function to maintain T(c)'s in both loops equal. Since both T(h)'s are equal, and RCS .- flow has not been affected, the amount of energy being transferred . by both OTSG,must be equal. This implies equal feed flow to both OTSG's. But with one of the OTSG's fouled, there must be a compensation to ensure e That compensation is produced by the Delta T(qual heat transfer.c) circuit that allows transient reratioing of feedwater to create and maintain a proportionate difference in OTSG levels. [+2.0] - REFERENCE
- 1. Crystal River: STM, Ch. 504, p. 22.
- - - , , ~ , . ,-m, . _-----m- - - - -
s 1.' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 29 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- CRYSTAL RIVER -86/08/19-HUENEFELD, J. ANSWER 1.05 (1.50) In the RCS, the flows redistribute such that there -is about 42% flow in the "A" loop and 105% in the "B" loop. T-hi cr::te3 th;
- d te r;t;v f;;i::ter '!cr t- dir et prepertion.- In-the 5tc;dy-
. stat , :Scet 10% of-the re;*2? red >!ce . ill we to th; "A" efde and 400- t: the "B" :fd . At 7;; powcr thi; c::ald b; abcut 20% tetel-fc;dWeter d:r2ad te the "A" .ide end 45% tet:1 feedrater demend tv-b ibw:^ iho """ fdc_ [+1.5] /' .d.ve e $w 6e,%.5 t %,.he w;A &c.
r M iersta : 4 3,, g,u M R FERENCE' Apr:
"- -'~ b'#0+64# M 45e.c. : 2J.9- 9, Bre:
- 1. Crystal River: STM, Ch. 504, p. 24. 8sw : $3.lo %
ANSWER 1.06 (1.50)
- 1. The initiation of EFW lowers the saturation pressure of the OTSG, and therefore the saturation temperature. [+0.5]
- 2. The EFW is sprayed directly on the OTSG tubes high in the steam generator, therefore increasing the effective area for heat transfer. [+0. 5]
- 3. The EFW is essentially at ambient temperature, as compared to main feedwater which has been heated. [+0.5]
REFERENCE
- 1. Crystal River: ROT-4-15.
ANSWER 1.07 (1.00) OTSL outlet pressure [+1.0] - REFEREliCE
- 1. Crystal River: ROT-4-15, p. 15.
. . . - , . - - , -- . - - - , . - - - - . . - - , . . , , - . - . , , , - , n. -,., -,
- 1.* PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 30
-IHERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- CRYSTAL RIVER -86/08/19-HUENEFELD, J.
ANSWER 1.08 (1.00) The pressurizer spray valve is open. [+1.0] REFERENCE 1.. Crystal River: Thermodynamics and Heat Transfer Training Material. ANSWER 1.09 (1.50) Negative side. [+0.5] There is more operational latitude with imbalance negative. This is largely due to the greater DNBR in the lower regions of the core. [v1.0] REFERENCE
- 1. Crystal Rive,r: OP-204, p. 17.
ANSWER 1.10 (1.00) (b.) [+1.0] REFERENCE
- 1. Crystal River: OP-204, p. 17.
ANSWER 1.11 (1.50) Shutdown Margin - the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present .
. . condition assuming: .
No change in axial power shaping rod position and all control rod assemblies (safety and regulating) are fully inserted except for the single rod assembly of highest reactivity worth which is assumed to be fully withdrawn. [+1.5] 7 REFERENCE i 1. Crystal River: Technical Specifications, p.1-3. I e
-m a,,-
1." PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 31 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- CRYSTAL RIVER -86/08/19-HUENEFELD, J. ANSWER 1.12 (2.00)
- 1. Tc is approaching T-sat of its OTSG
- 2. Core delta T develops and stabilizes _ ._
- 3. AVG of 5 highest incore thermocouples follows Th within 10 deg F
- 4. When OTSG pressure is lowered, then Ti:, .Tc and incore thermocouples lower
[+0.5] each REFERENCE
- 1. Crystal River: VP-580, p. 7.
ANSWER 1.13 (1.50) The RCS will be losing borated water to the OTSG. The steam escaping from the OTSG will be relatively boron free. This will end up causing a dilution of the RCS water in the RB sump as this steam is condensed. [+1.5] REFERENCE
- 1. Crystal River: EP-390, General Effects.
a b f W
- 1.
- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 32 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- CRYSTAL RIVER -86/08/19-HUFNEFELD, J.
ANSWER 1.14 (3.00) a. Lar:e.> . M :.s.u..ry . - y{v.'.n. - e k cg ;M' -:.lf.h s.G.'.r;::l,y.e M: 9to :.r.c, .~, &
. ? % .'sso Q .H.). O d:1Q
>.b
.g.f.;.t. :r,- . . .c.c: r e. ..w
.e. ;.s.
. . y . . . . . ;. ,
. ... .c ., .- . . ;.
. t -. t c > p .t.,;.z< . .. .,. . . L. i %
- . :.:..O :,. ..e . ;, . .g ..
* '.;.t.{s.... 100
'~{*
s
*.s .- '"' e .'. . . 'I'**.'
, Nae, % . , .
- A.... . . ,
OF MAX. 3.; s-
**l.,,..,
s-
,. c * (, ,
. s: ,.
. . ,6. .;- .
4 :.;v;iln'
. < . . < ~ < . '. 60 _.- . . . .. . -
. i... .
.,-t'...
'I.lj. : :. , .
.x : ..t.* .: . . :.s . > ...
,i.. *'
_ ,.- . . ' . =. ,- ;
... ,.,.s.y ; r ...-
..;.,s - ,- _ ,
e- r. , 80
.. ,40 ,
120 ~ , .i, 160 TIME (HRS)' [+2.0]
- b. When the power level is reduced to 50%, the Xe-135 burnout is reduced by a factor of 2 while the formation from the decay of I-135 is at the 100% power value. This causes a peak as shown and then decay and burnout down to the new equilibrium N(Xe) for 50% power. [+1.0]
REFERENCE
- 1. NUS Training Manual, Module 3, Ch. 10.
e. e
. 1.
- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 33 THERMODYNAMICS, HEAT TRANSFER AND FLUIG FLOW ANSWERS -- CRYSTAL RIVER -86/08/19-HUENEFELD, J.
ANSWER 1.15 (3.00) a. w+A %4 _ r,...-e u., I
. -.{ ^]
tire [+1.0].
- b. No (the insertion leaves the reactor on a stable positive period). [+0.5]
- c. During the withdrawal there is a pronounced response due to prompt neutrons. When the withdrawal stops, the response of cer.t..... pj_. .. ..juickly promptneutronstothergtipi,g:gs,gj,onleaving ethe diesenoff, reactor
.! d-governed by delayed neutrons. The rod insertion again causes a pronounced response due to prompt neutrons. The prompt negative response overcomes the positive effect of delayed neutrons. When the inse.'. ion. stops, the period increases slightly as the delayed .
neutron population comes to a new equilibrium at the new reactivity. [+1.51 REFERENCE
- 1. NUS Reactor Kinetics, 'Jnit 6.
b
-+
1.* PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.- PAGE 34
- THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- CRYSTAL RIVER' -
86/08/19-HUENEFELD, J. ANSWER 1.16 (1.00) No, it won't. Delayed neutrons behave as a source themselves. When a subcritical reactor has sufficient reactivity, the subcritical multiplication of neutrons becomes significant. .The large number of generations associated wi'th this multiplicstion effectively holds up the decay rate. The negative SUR becomes
>- 1/3 dpm. [+1.0]
REFERENCE
- 1. NUS Reactor Kinetics, Unit 6.
4 ANSWER 1.17 (1.00)
'(b.) [+1.0]
REFERENCE -
- 1. NETRO 11.4-3.
ANSWER 1.18' (1.00) (b.) [+1.0] REFERENCE
- 1. Duke Power Company, FNRE; pp. 126-128.
ANSWER 1.19 (1.00) (b.) [+1.0] .. REFERENCE
- 1. Crystal River: OP-210, Rev. 16, p. 8.
- 2. NUS, NETRO, Unit 6. -
- 3. Westinghouse Reactor Physics, Section 3 and 5.
O e -g--, ,u - .--,- -g-,n- ~,-r.e- - - ---n---- , - - - , - - - - , . - , - , w s a , - , , , . , , - r w--r-
r i 1.
- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 35 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS -- CRYSTAL RIVER -86/08/19-HUENEFELD, J.
ANSWER 1.20 (1.00) (c.)- [+1.0] REFERENCE ,
- 1. Westinghouse NTO, p. I-5.77.
- 2. NETRO, 10.5-2.
- 3. Duke. Power Company, FNRE, p. 169.
ANSWER _ 1.21 (1.00) (b.) [+1.0] REFERENCE
- 1. NETPP, pp. 6.5-1 to 6.5-3.
~
D
i .*
+-
'2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE . 36 -
. ANSWERS -- CRYSTAL RIVER -
86/08/19-HUENEFELD, J. ANSWER. 2.01 (1.00) (c.) [+1.0] REFERENCE Crystal River: Lesson No. ROT-4-1, p. 15. 1.
. ANSWER 2.02 (.50)
False - [+0.5] REFERENCE
- 1. - Crystal River: System Training Manual, Ch. 404, p. 4.
ANSWER 2.03 (1.00) Draining of the BWST to the RB sump. [+1.0] - REFERENCE
- 1. Crystal River: System Training Manual, Ch. 404, p. 4.
ANSWER' 2.04 (1.00) About'200 psig. [+1.0] REFERENCE
- 1. Crystal River: System Training Manual, Ch. 404, Figure 6-9.
ANSWER 2.05 -(1.00) hydrogen [+1.0] or, o"Yt " ik hyd'*{ * b 're4 REFERENCE
/ us t . -
- 1. Crystal River: System Training Manual, Ch. 412, p. 9.
e- -.-,--. .-- ,- =---,,---.---,v , - - - --,r- - - . . ,,,--------.m ,- - w-- - - - , - - 7.--- . ---v - --,rk.,-- -- -m,-t - ----- m.- - , - , -
- 2.
- PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 37 ANSWERS -- CRYSTAL RIVER -86/08/19-HUENEFELD, J.
ANSWER 2.06 (1.50)
- a. The minimum pressure ensures that there will be sufficient seal water flow through the compressor heat exchanger and back to the compressor suction for cooling and lubrication. [+1.0]
~
-b. Nuclear Services Closed Cycle Cooling. [+0.5]
REFERENCE
- 1. Crystal River: System Training Manual, Ch. 412, p. 40.
ANSWER 2.07 (.50) True [+0.5] REFERENCE
- 1. Crystal River: STM, Ch. 412, p. 51.
ANSWER 2.08 (1.00) (c.) [+1.0] REFERENCE
- 1. Crystal River: STM, Ch. 419, p. 8.
ANSWER 2.09 (.50) 100,000 lbm/hr [+0.5] E REFERENCE . l 1. Crystal River: STM, ch. 419, p. 13. l
- 2. ' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS ~ PAGE 38 )
i ANSWERS -- CRYSTAL RIVER -86/08/19-HUENEFELD, J. l ANSWER 2.10 (1.50) MVP 3A - SW (normal) DC-A (alternate) [+0.5] MVP 3B - SW onl [+0.5] MVP 3C - DC-B (ynormal) SW (alternate) [+0.5] REFERENCE
- 1. Crystal River: ROT-4-1, p. 53.
ANSWER 2.11 (.50) They will support si) starting attempts. [+0.5] REFERENCE
- 1. Crystal River: ROT-4-6, p. 7.
ANSWER 2.12 (1.00)
- 1. 2750 kW at 0.8 power factor w/ maintenance periods
- 2. 3000 kW at 0.8 power factor - 2000 hours no maintenance
- 3. 3300 kW at 0.8 power factor - not more than 30 minutes if . 3 %Te Nw- IGt.Ry nn*&-=+ 1 :h ad Any one (1) [+1.0] r,11.j 4 93 A. .
REFERENCE
- 1. Crystal River: ROT-4-6, p. 14. -
ANSWER 2.13 (1.00) They will have to be unlocked [+0.5] and manually [+0.5] shut. REFERENCE
- 1. Crystal River: ROT-4-15, p. 6.
t
~~ ~
- 2.
- PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 39
-ANSWERS -- CRYSTAL RIVER. -86/08/19-HUENEFELD, J.. j ANSWER: 2.14 (1.00) 3 g .o 3Al and SAS are powered from Aux Bus 3A '+0.5' iME and 3B2 are powered from Aur Bus 3B l+0.5
-p D FERENCE
[ 1. . Crystal River: STM, Ch. 420, p. 17. i ANSWER 2.15 (1.50) -) They " freeze" in position [+0.5]. This provides a relatively accurate i starting point when a power decrease reaches the point where the ; main block valves go closed .[+1.0]. ! REFERENCE
- 1. Crystal River: STM,-Ch. 504, p. 104.
ANSWER- 2.16 (3.00)
- a. There is a breaker cubical for the "B" MVP in both safeguards busses. There is, however, only one breaker. The breaker may be transferred between the two busses. Additionally, there is
! a disconnect switch at the pump for selecting which buss will 1 supply power. There is an elaborate system of kirk-key interlocks under load,that prevent inadvertently or paralleling operating2.0] the two ES busses. [+the disconnect
- b. The selector switch on the "A" MVP breaker should be selected to " Pump 3A" and the selector switch on the "C" MVP breaker should be selected to " Pump 38." [+1.0]
l l REFERENCE t,
- 1. Crystal River: ROT-4-1, p. 57.
ANSWER 2.17 (1.00) From the 230 kV transmission circuits through the Unit 3 startup transformer. [+1.0] O
c 2.
- PLANT DESIGN INCLUDING SAFETY- AND EMERGENCY SYSTEMS PAGE 40 ANSWERS -- CRYSTAL RIVER -86/08/19-HUENEFELD, J.
1 REFERENCE Crystal P.tver: STM-15-3.
- 1. l
. ANSWER 2.18 (2.50)
~ ~
main block valve , startup block valve ' low load block valve .! MFWP suction valve MFW cross connect valve [+0.5] each REFERENCE
- 1. Crystal River: ROT-4-15, p. 22.
ANSWER 2.19 (1.50) This bypass vents the valve bonnet to the downstream side of the valve, therefore minimizing the pressure needed to maintain valve position [+0.75]. The check valve prevents leakage pr.:.t the valve disk piston rings from.being allowed to flow directly from the bonnet
'to the underseat area [+0.75].
REFERENCE
- 1. Crystal River: NA0-83, p. 6.
ANSWER 2.20 (1.00) ! The monitors are installed on each main steam line and are located b in the intermediate building just outside of the reactor building . 1 penetrations. All steam flowing out of the atmospheric dump valves or the main steam safeties would pass by these monitors. REFERENCE l
- 1. Crystal River: NA0-83, p. 13. '
**-a ---*v*~' F
- 12. ' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 41 ANSWERS -- CRYSTAL RIVER -86/08/19-HUENEFELD,J.
ANSWER 2.21 (.50) About (+- 50 psi) b psig. '[+0.5]
' REFERENCE
' Crystal River; NAO-96, p. 4.
~
1.
~
I l 2.22 ANSWER- (1.00) One "C" breaker and one "D" breaker supply power to groups 1 and 2. One "C" breaker and one "D" breaker supply power to groups 3 and 4. [+0.5) s!.ch REFERE14Cn.
- 1. Crystal kiver: STM, Ch. 510, Figure 9.
ANSWER 2.23 (1.00) c (d.) [+1.0] REFERENCE
- 1. STM2-11.
ANSWER 2.24 (1.00) (c.) [+1.0] REFERENCE
- 1. Crystal River: STM 28-4.
l- ANSWER 2.25 '(1.00)
.(a.) [+1.0]
*- 2. ' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 42 ANSWERS -- CRYSTAL RIVER -86/08/19-HUENEFELD, J.
REFERENCE
- 1. Crystal ~ River: STM 22-1 and 44.
ANSWER 2.26 (1.00) (a.) - [+1.0] ~ REFERENCE
- 1. Crystal River: STM 17-10 and 12. -
ANSWER 2.27 (1.00) (c-) [+1.0] REFERENCE -
- 1. Crystal River: OP-605, Rev. 28, pp. 5 and 6.
v- - - + , , . . , , ,. ,y. , . - , - --w- , y w -- . ,- w-
-3. ' INSTRUMENTS AND CONTROLS PAGE 43 ANSWERS -- CRYSTAL RIVER -86/08/19-HUENEFELD, J.
ANSWER 3.01 (.50) True [+0.5] REFERENCE
~
- 1. Crystal River: STM, Ch. 419, p. 30. ~ '
ANSWER 3.02 (1.00) (c.) [+1.0] REFERENCE
- 1. Crystal River: STM, Ch. 419, p. 34.
ANSWER 3.03 (1.50) 1.- Lift oil pressure
- 2. SW flow
- 3. Reservoir oil level
- 4. Seal injection flow
- 5. Reactor power
- 6. Fourth pump only
- 7. Controlled bleedoff valve open Any five (5) [+0.3] each, +1.5 maximum ;
REFERENCE
- 1. Crystal River: STM, Ch. 420, p. 16.
ANSWER 3.04 (1.50) .,
- a. It automatically isolates the Decay Heat dropline by shutting DHV-3 and 4. [+1.0]
- b. By using key operated switches located in the ES Test Cabinets.
[+0.5] - REFERENCE
- 1. Crystal River: System Training Manual, Ch. 404, p. 9.
.-, - , , , - , - ~ ,-
E
- 3. ' INSTRUMENTS AND CONTROLS PAGE 44 ANSWERS -- CRYSTAL RIVER. -86/08/19-HUENEFELD, J.
ANSWER 3.05 (1.00) The flow is passed over two RTDs, one with an attached low power heating element. When the two RTDs are indicating a no temperature difference, that is an indication of flow. [+1.0]
~ ~
REFERENCE 1.- Crystal River: STM, Ch. 419, p. 26.
-ANSWER 3.06 (1.00)
(a.) and (d.) [+1.0] REFERENCE
- 1. Crystal River: STM, Ch. 419, pp. 30-33.
ANSWER 3.07 (2.00) RCS flow 20%/ min 1.1 x flow RCP's 50%/ min 75/45% FWP's/FWBP's 50%/ min 55% Assym. rod 30%/ min 60% [+0.3] for runback, [+0.1] for rate, [+0.1] for limit REFERENCE
- 1. Crystal _ River: STM, Ch. 504, p. 12.
ANSWER 3.08 (1.00) ;
- 1. BTU limit
- 2. both OTSGs on level limit l
- 3. both feedwater loop demands in manual l
\
[+0.33] each REFERENCE
- 1. Crystal River: STM, Ch. 504, p. 84. j i
, , ., -, y .r~-, ,-- --, , s - , -. , , - - , -
g =
~*
~ '
*D 3.
- INSTRUMENTS AND CONTROLS PAGE 45 ANSWERS -- CRYSTAL RIVER -86 /
08 /19-HUENEFELD, J. ANSWER 3.09 (1.00) (b.) . [+1.0]
' REFERENCE Crystal River: STM, ch. 510.
~ ~
1. ANSWER 3.10 (.50) ,
.Both the output state lamp and the memory lamp. [+0.5]
REFERENCE
- 1. Crystal River: STM, Ch. 510, p. 23.
ANSWER 3.11 .(1.50) Bypassed--power / imbalance / flow, RCPPM, low pressure, and variable low pressure.. [+1.0] Substituted--a reduced low pressure trip. [+0.5] REFERENCE
- 1. Crystal River: STM, Ch. 510, p. 46 and Figure 2.
ANSWER 3.12 (1.00) The normal bus undervoltage relays will immediately actuate (no time delay) leaving only the ES Block one load's breakers shut, the bus will be stripped and following a 3-second time delay, the EDG .. output breaker will be enabled to close~. [+1.0] REFERENCE l
- 1. Crystal River: ES Electrical Distribution, Bus Relaying and Feeder Breaker Interlocks, p.14. -
- , e-,, e, w y . , - ,~ w- w.n ,y__., _-., na m n . , , - - . .-- , . , . , , .c-. , - ,-=- m -. - , , ., ,
7~
3.
- INSTRUMENTS AND CONTROLS PAGE '46
= ANSWERS -- CRYSTAL RIVER -86/08/19-HUENEFELD, J.
ANSWER 3.13 (1.00)
- (c.) [+1.0]
REFERENCE
- 1. Crystal River: Attached figure from IS Electrical Distribution, STM.
ANSWER 3.14- (2.00) HPI pump DHR pump emergency lighting inverters control complex lighting battery chargers [+0.33] each REFERENCE
- 1. Crystal River: Electrical Systems training manual, ROT-4-1,3,6,10, p. 8-29.
ANSWER 3.15 (1.00) ! .The speed droop control will be set to "0". This is done to ensure highly regulated speed control regardless of load. [+1.0] ! REFERENCE l .
\
l'
- 1. Crystal River: ROT-4-6, p. 19.
l ANSWER 3.16 (1.00) l l Pressing the Emergency Stop push button. [+0.5] l Tripping the fuel racks. [+0.5] l REFERENCE i
- 1. Crystal River: ROT-4-6, p. 29.
t i
A m F C Degraded Voltage 7, r Undervoltage Relags flotuate
] Diesel Start Time ,
e r [ $'EDG Breaker Closes a m d. r 1 , I .Ii , , 1 E. 11-1111111122222222 o 123456789012345678901234567 T0
- d' Time Chart in Second l Bus Degraded Voltage i-2 I
a r
* ,e 9
. *:13.
- INSTRUMENTS AND CONTROLS PAGE 47
-ANSWERS -- CRYSTAL RIVER- -86/08/19-HUENEFELD, J.
ANSWER 3.17 (1.50)
.a. differential current [+1.0] i
- b. at the local generator control panel [+0.5]
REFERENCE
- 1. Crystal River: ROT-4-6, Figure 5.
ANSWER 3.18 (1.50)
- 1. Loss of both main feedwater pumps and reactor power >20%
- 2. Low level ((6") in either OTSG
- 3. Loss of all RCPs
- 4. Low pressure ((600 psig) in either'0TSG
- 5. HPI actuation on Both A and B ESAS channels
[+0.3] each REFERENCE
- 1. Crystal River: ROT-4-15, p. 14.
ANSWER 3.19 (1.00) Because hitting the RESET button causes all of the EFW control valves to come open if they are in 4wwwd. [+1.0] REFERENCE A"UD ~
- 1. Crystal River: ROT-4-15, p. 19.
ANSWER 3.20 (1.00) . The recycle isolation valves shut. [+0.5] The Waste gas release isolation valve shuts. [+0.5] REFERENCE
- 1. Crystal River: STM, Ch. 505, Table 5.
~ ' . ..
' l3.' INSTRUMENTS AND CONTROLS PAGE 48
' ANSWERS -- CRYSTAL RIVER -86/08/19-HUENEFELD, J.
4 ANSWER 3.21' (.50) The HIGH alarm level. [+0.5] l REFERENCE
- 1. ' Crystal River: STM, Ch. 505, p. 34.~ ~
ANSWER 3.22 (1.00) The main feedwater block valves are interlocked to close. [+1.0] REFERENCE'
- 1. Crystal River: NAO-96, p. 18.
ANSWER 3.23 (1.00)
'(b.) [+1.0]
REFERENCE
- 1. Crystal River: STM 13-2.
ANSWER 3.24 -(1.00) (a.) [+1.0] REFERENCE
- 1. Crystal River: STM 12-11.
l l< 2. Crystal River: STM 12-13. - ANSWER 3.25 (1.00) (b.) [+1.01 - REFERENCE
- 1. Power Systems Operation, R.H. Miller, pp. 22-24.
- 3.
- INSTRUMENTS AND CONTROLS PAGE 49
-ANSWERS -- CRYSTAL RIVER -86/08/19-HUENEFELD, J.
j' . ANSWER 3.26 (1.00) (c.) [+1.0] . REFERENCE
- 1. Crystal River: STM 15-10. ~ -
ANSWER 3.27 (1.00) (b.) [+1'. 0] REFERENCE 1.- Crystal River: STM 12-14. 4
,y .- - - - ,, w,- ,-m. , ,--, , , - , . - , - - - + , - - -
p , ,
-s 4C' PROCEDURES'- NORMAL, ABNORMAL, EMERGENCY AND PAGE 50 RADIOLOGICAL CONTROL ANSWERS -- CRYSTAL RIVER -86/08/19-HUENEFELD, J.
l ANSWER 4.01 (1.00) Both continuous and cumulative run time are restricted. (Toensure that limits-are not exceeded, run time in this mode is carefully
-logged by the reactor operator.) [+1.0] _ _
REFERENCE
- 1. Crystal River: System Training Manual, Ch. 404, p. 14.
ANSWER 4.02 (1.00) They are to be locked out(to prevent an isolation of the D' ystem upon spurious ACI). [+1.0j REFERENCE
- 1. Crystal River: System Training Manual, Ch. 404, p.15.
ANSWER 4.03 (2.00)
- 1. select the flow path on the Feed Select Switch
- 2. ensure that the number on the batch size thumbwheels is larger t than the number on the totalizer. -(MU&P Feed Permit)
- 3. manually open CAV-57 the supply valve to the MU&P system (complete the WD Feed Permit)
+ 4 op e.n Mu&lO3
( 4. manually start the boric acid pump gj (control feed flow rate with the valve knob on the batch 1. controllerY - '. [+0.4] each j REFERENCE ! 1. Crystal River: OP-403. -
- 2. Crystal River: ROT-4-1, p. 67.
,,-ee -, - , - - - - , - , , - , - e-, .,y,--. - . . . - - - - . v..-.,e, ,-, - - , .y-.4 -,.,,-w +.w,-,.,.,...,,.w.. - - - . - - , . . . . . -
- .~
- f.4.
- PROCEDURES - NORMAL, ABNORMAL,- EMERGENCY AND PAGE 51 RADIOLOGICAL CONTROL ANSWERS -- CRYSTAL RIVER -86/08/19-HUENEFELD, J.
' ANSWER 4.04 (:.50)
The startup station must be placed into auto first. [+0.5]
' REFERENCE
- 1. Crystal River: OP-504, p. 15.
. ANSWER 4.05 (1.00)-
First the standby cooler must be verified to be supplied with cooling I water and full of oil. It is verified full of oil by opening the interchange valve and observing flow at the sight glass at the top of the reservoir. Then the swap-is accomplished using the transfer valve. [+1.0] REFERENCE
- 1. Crystal River: OP-602, p. 11.
ANSWER 4.06 (1.00)
- 1. The PORV block valve must be operable. [+0.5]
< 2. The operator on the switch does not leave the PORV switch unattended until he assures that the PORV is closed and there is no flow through it. [+0.5]
- REFERENCE
- 1. Crystal River: OSIM, p. V-14.
ANSWER 4.07 (1.00) 200 mrem / week,"not to exceed 1250 mrem / quarter. [+1.0] A ! REFERENCE #f b 20 are/ytu(
- 1. Crystal River: RSP-101, p. 3.
I O
)
f .,
$' 4'.
- PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 52-
-RADIOLOGICAL CONTROL ANSWERS -- CRYSTAL RIVER -86/08/19-HUENEFELD, J.
. ANSWER 4.08 (1.00)
(b.) . [+1.0] REFERENCE
- 1. Crystal River: OP-203, p. 3.
- 2. . Crystal River: OP-502, p. 4.
~
ANSWER 4.09 (.50) True [+0.5] - REFERENCE'
- 1. Crystal River: OP-203, p. 12.
ANSWER 4.10 (1.00) Exciter current. [+1.0] REFERENCE
- 1. Crystal River: OP-203, p. 20.
ANSWER 4.11 (1.50)
- 1. RCP seals and dumpsters
- 2. PZR level
- 3. RCDT
- 4. MUT -
- 5. RB sump .
- 6. . Relief valve tailpipe TEMP y one (h
[+0.25] each C,d:b gim he ,co - or et sd *tt:Ediq M REFERENCE
# "' W96 Mox b.ailoUw,,
y 6.g.) 5g 1
- 1. Crystal River: VP-540, p. 4. (
O
. . . _ _ _ . , ~ y . +-. - . _ , . - . , _ . , , _ , _ . _ _ , , , ._,-,,_.r. ., _ _.
f; ' 1 ~4.
- PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 53 RADIOLbGICAL CONTROL ANSWERS -- CRYSTAL-RIVER -86/08/19-HUENEFELD, J.
ANSWER 4.12 (1.00)-
- 1. > 1500 psig-20 deg F
- 2. ( or = 1500 psig--50 deg F[+0.5] [+0.5] , ,
REFERENCE
- 1. Crystal River: VP-580, p. 5.
ANSWER 4.13 (2.00)
- a. LLL . .
- b. 50% on the OR er 4 f */, 6 y f/$6
- c. 95% on the OR.
- d. 95% on the OR
[+0.5] each REFERENCE
- 1. Crystal River: VP-580, p. 6.
ANSWER 4.14 (2.50)
- a. When RC TEMP > 500 deg F--cooldown at ( or = 240 deg F/hr. [+0.5]
When RC TEMP ( or = 500 deg F--cooldown at ( or = 100 deg F/hr. [+0.5] Maintain tube-to-shell delta T ( or = 150 deg F7hc, [+1.0]
- b. No. [+0.5]
REFERENCE -
- 1. Crystal River: EP-390, p. 6.
ANSWER 4.15 (1.00)
- a. When total flow is ( or = 100 gpm. [+0.5]
- b. 540 gpm. [+0.5]
..v. + . . - , - . ~ - , ,....-.,_am ., - - - , _ _ . . , . . , _ . . . - , , - , , . , , . . - , , - , . . ,
=
. o
- 4'.
- PROCEDURES - NORMAL, A8 NORMAL, EMERGENCY AND PAGE 54 RADIOLOGICAL CONTROL
' ANSWERS -- CRYSTAL RIVER -86/08/19-HUENEFELD, J.
REFERENCE
- 1. Crystal River: AP-380, p. 7.
ANSWER- 4.16 (1.00) _ _ tube-to-shell delta T = AVG of 5 shell TCs - Tc [+1.0] REFERENCE Crys.tal River: AP-380, p. 17. 1. ANSWER 4.17 (3.00)
- 1. Initiate full UPI.
- 2. Stop all Reactor Coolant Pumps.
- 3. Feed up OTSG's to 95% on the OR.
[+1.0] each REFERENCE
- 1. Crystal River: AP-380, Babcock and Wilcox Technical Bases Document.
ANSWER 4.18 (3.00)
- a. Adequate subcooling margin must exist. [+1.0]
- b. Adequate subcooling margin, AND PZR level ) 50 inches with normal makeup, AND secondary heat transfer is established [+1.0],
OR LPI flow > 1000 gpm in each train for > 20 minutes [+1.0]. L REFERENCE .. 1.- Crystal River: AP-380, pp. 7,12, and 16. ANSWER 4.19 (1.00) - (a.) [+1.0] h s
-. _-,-.-nn , , - - - ----,,.----------~,--,----~,--~~w,,*,<--r-~~~ -'~--~'-~s " " ' = ' * =r ^ ' ~ " ^ ^ * ' ~ ' ^ ~ ~ ~ ' ' " " ' ' ~ ' ' " ' '
(,----
, ~ - ;e s 4s
- PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 55 RADIOLOGICAL CONTROL-ANSWERS - _ CRYSTAL RIVER -86/08/19-HUENEFELD, J.
REFERENCE
- 1. Crystal River: AP-320, Rev. 00, p. 2.
ANSWER 4.20 (1.00) _ ,, (d.) [+1.0] REFERENCE
- 1. Crystal River: OP-502, Rev. 13, p. 4.
ANSWER 4.21 (1.00) (a.) [+1.0] REFERENCE
- 1. Crystal River: EP-120, Rev. 00, p. 2.
ANSWER 4.22 (1.00) (b.) [+1.0] REFERENCE
- 1. Crystal River: RP-101, Rev. 19, p. 23.
ANSWER 4.23 (1.00) (e.) [+1.0] REFERENCE
- 1. Crystal River: EP-290, Rev. 02.
1 l l
t c
..~.
- s...
EQUATION SHEET Where mi = m2 (density)i(velocity)i(area)1 = (density)2(velocity)2(area)2 2 KE = mv PE = mgh PE + KE +PgV1 = PE +KE +P Y22 where V = specific 2 i i 2 2 volume P = Pressure Q = mcp (Tout-Tin) Q = UA (T,y,-Tstm) Q = m(hi -h2 ) P = P010(SUR)(t) P = P0 et /T SUR = 26.06 T = (B-p)t T p delta K = (K,ff-1) CR1 (1-Keff1) = CR 2 (1-Keff2) CR=S/(1-K,ff) (1-K,ffi) (1-K,7f) x 100% SDM = M = (1-Keff2) K eff
, decay constant = in (2) " 0.693 A1 = Ag e-(decay constant)x(t) t t 1/2 1/2 Water Parameters Miscellaneous Conversions 1 gallon = 8.345 lbs 1 Curie = 3.7 x 10 10 dps 1 gallon = 3.78 liters 1 kg = 2.21 lbs 1 ft 3= 7.48 gallons 1 hp = 2.54 x 103 Btu /hr 3 6 '
Density = 62.4 lbg/ft 1 MW = 3.41 x 10 Btu /hr Density = 1 gm/cm 1 Btu = 778 ft-lbf Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8 x Degrees C) + 32 Heat of Fusion = 144 Btu /lbm 1 inch = 2.54 centimeters 1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 ft-1bm/lbf-sec 2 - 9}}