L-86-382, Forwards Technical Comments Re Operator License Exams on 860908.Util Has Taken Steps to Improve Operator Training Program,Which Has Now Been Accredited by Natl Nuclear Accrediting Board

Forwards Technical Comments Re Operator License Exams on 860908.Util Has Taken Steps to Improve Operator Training Program,Which Has Now Been Accredited by Natl Nuclear Accrediting Board

ML20215D986
Person / Time
Site:	Turkey Point
Issue date:	09/19/1986
From:	Woody C FLORIDA POWER & LIGHT CO.
To:	Grace J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
References
L-86-382, NUDOCS 8610140381
Download: ML20215D986 (74)

Text

{{#Wiki_filter:- _ _ _ _ _ _ _. - _ __ ___ _ _______ _ __ t P. o. BoM 140C3, Juno I, ' 33408-0420 gy p. S'S SEPTEMBER 101986 L-86-382 Dr. J. Nelson Grace Regional Administrator, Region 11 U. S. Nuclear Regulatory Commission 101 Marietto St., N.W., Suite 2900 Atlanto, GA 30323

Dear Dr. Grace:

Re: Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Operator License Examinations This is to confirm the technical comments presented to your staff at the exit interview following the examinations administered to proposed and current operators at Turkey Point on September 8,1986. These previously discussed technical issues and several additional technical comments may bear significantly on your evoluotion of these examinations; accordingly, they are formally transmitted herewith for your consideration. Also, as your staff is aware, FPL has identified a number of concerns with respect to the examinations which go to their basic solidity and which we believe should be reflected in your final action.on the abose-mentioned examinations. We are prepared to discuss these matters with you at your earliest convenience. As you are also aware, FPL has previously taken steps to improve our operator training program and these programs have now been accredited by the National Nuclear Accrediting Board. These training improvements and the resources expended in support of these improvements cannot be totally effective if the testing process that our operators are subjected to is not valid. We look forward to discussing these issues with your staff. Very truly yours, [h C. O. Woo Group Vi resident Nuclear Energy COW /PLP/cvb g c/ Attachment '// cc: Harold F. Reis, Esquire i I: 8610140381 860919 PDR ADOCK 05000250 /, y PDR PEOPLE.. SERVING OFLE D

r TECHNICAL ISSUES PRESENTED AT EXIT INTERVIEW SEPTEMBER 12,1986 i I e

U3e 11:00 Page1 CROSS REFERENCE INDEX RO 1.03 (SRO 5.01, SRO Requal 5.02) RO 1.04 (SRO 5.02, SRO Requal 5.03, RO Requal 1.04) RO 1.12 (SRO 5.09, SRO Requal 5.06) RO 1.13 (SRO 5.10) RO 1.14 (SRO 5.11) RO 1.16 (SRO 5.13, SRO Requal 5.09, RO Requal 1.09) RO 1.19 (SRO 5.16, SRO Requal 5.07) RO 1.21 (SRO 5.18) RO 2.03 (SRO 6.02, SRO Requal 6.02, RO Requal 2.02) RO 2.11 (RO Requal 2.06) RO 2.12 (SRO 6.13, SRO Requal 6.10, RO Requal 2.09) RO 2.13 (RO Requal 2.08) RO 2.19 (SRO 6.17, SRO Requal 6.12, RO Requal 2.12) RO 2.22 (RO Requal 2.14) RO 3.08 RO 3.09 RO - 3.14 (SRO 6.20, SRO Requal 6.09, RO Requal 3.09) RO 3.15 (SRO 6.21, SRO Requal 6.08, RO Requal 3.08) RO 3.16 RO 3.18 (SRO 6.22, SRO Requal 6.07, RO Requal 3.05) RO 3.19 RO 4.04 (SRO 7.04) RO 4.16 (SRO 7.15) i RO 4.19 (SRO 7.20) ) RO 4.20 (SRO 7.21) RO 4.24 (SRO 7.25, SRO Requal 7.17, RO Requal 4.17) SRO 7.19 SRO 8.22 (SRO Requal 8.15) l

• Rt.5/cas

09/11/86 Page1 NRC Exam Section 1 QUESTION RO 1.03:(SRO 5.01, SRO Requal 5.02) 4 A reactor has been operating at full power for three months when a manual reactor trip occurs. All systems are operational and the steam dumps are immediately placed in the steam pressure control mode. Ten minutes after the reactor trip, all RCPs are tripped. Twenty minutes after the reactor trip, Loop 1 RCP is jogged momentarily. Which set of traces (a - c) on figure #174 most closely represents the previously described events?

RESPONSE

We request that answers "a" or "b" also be accepted or delete question. Answers "a" and "b" correctly show Te increasing at the beginning of the transient. In addition, the poor quality of the trace reproduction made it difficult tointerpretthe ATbetweenTe and thermocouple readings. Discussion with Westinghouse confirmed that none of the traces given in the question is truly representative of what should actually occur.

REFERENCES:

l Telecon with Westinghouse on September 10,1986

• R1-5/cas

- ~

09/11/86 Page 2 NRC Exam Section 1 QUESTION RO 1.04:(SRO 5.02, SRO Requal 5.03, RO Requal 1.04) For each of the following parameters compare a Core Thimble Cell to a Normal Core Cell and indicate which of the two has the higher value. (rated conditions, average values)

1. Total power production

2. Powerdensity

3. Enthalphy rise

4. Mass flow 5.DNBR

RESPONSE

We request that this question be deleted from Requal Exams. This topic was not covered in the Accelerated Requal Class. Lesson Plan in Accelerated Requal Class did not cover this topic as shown on reference.

REFERENCES:

Instructor Lesson Plan No. 0017-OL, Appendix A, Page 2, Rev. 0,03/31/86 i a

• RI-5/cas

,Pcga 2, REV. NO. 0 INSTRUCTOR LESSON PLAN NO.0017-OL, APPENDIX A HEATTRANSFER D TEttMINAL OHJECTIVE: Upon successful completion of this section the student will be to describ mechanisms of heat transfer in the PWR and the effect on heat when departure from nucleate boiling occurs in the core. Define convection, conduction and radiation and provide PWR EN ABLING OBJ ECTIVES: specific examples foreach Describe the significance of nucleate boiling to the heat transfer rate l from a fuel rod to the primary coolant Describe the interrelationship of reactor power, pellet centerline temperature, and temperature of the primary coolant Describe conditions under which radi[shve heat transf y; l the significant heat transfer mechaniam in the reactor Describ'e the overall heat transfer coemeient a heat exchanger and f evaluate heat transfer through different types of heat exchangen l Explain departure from nucleate boiling including its significance to l safe plantoperation Explain how the departure from nucleate boiling ratio is calculate ], and describe the DNBR varies with primary plant parameters and l conditions. 3 3 Reading Assignment 3 3-15,3 3-78, (-82 -$8$,13 13-24 I I i

09/11/86 Page 3 NRC Exam Section 1 QUESTION RO 1.12:(SRO 5.09, SRO Requal 5.06) Restrictions on F(Q)(T) limit include penalties due to uncover and reflood rates, post-LOCA. Give one other basis for core height restrictions on F(Q) (T) and explain where and why it is restrictive.

RESPONSE

We request the answer key to be expanded to accept answers that concern fuel temperature, clad temperature and DNB during SB LOCA and discussion of K(Z) curve. F(Q) was taught in Accelerated Requal as a limit on fuel temperature rise to protect the clad. I

• R1-5/cas n

09/11/86 Page 4 NRC Exam Section 1 QUESTION RO 1.13c:(SRO 5.10) Explain why curve "C"is valid at < 15% quality and not at > 15% quality. c. RESPONSE c: We request that the answer key be expanded to accept an explanation of analysis limitations to generate the curve.

REFERENCES:

Thermodynamic Principles II, page 13-54,55 and 56, Westinghouse copyright 1982. i

• RI-5/cas

---n-.-, 7-

680 3 .. t, s q s K N A rs '\\ UNACCEPTABLE = t \\ g OPERATION 6 \\ ~- - k. N %' 2400 PSIA i 660 -] %g ,\\ s .g-c r 2250 PSIA x -g s s. -- N 1 A 2000 PSIA ~ N 'g, 640 _ _ _g q q-- \\ yh g w A 3 s 6 \\ g 1860 PSIA .s 3 hE 620 ,yN l'..N B ,l i s n w x,x E % N\\ 's. N, 600 '9d 1 ACCEPTABLE OPERATION \\ --3 580 ~' 560 O 20 e 60 80 m j i POWER (PERCENT) FND-HT-100: T Yb' AV6 (REV. 1) 13-54 I l 0300C

Graph C limits the coolant exiting the core to less than 15 percent quality. This limit is necessary because the WR8-1 correlation used to predict DN8 only applies to fluid at less than eso - .1 l UNACCEPTA8LE f '2 C 2400 PSIA OPERATION .~N '80 Q NP'-,,,o,DQ ~ j .w w .g 1 N .i \\ [~ ,4, N. 2000 PSIA N .. '? N

1. _:.

N isso PslA N N N N N \\ q - - ~. N 1 ~' c i 3 l i l T - '1 - I ACCEPTABLE OPERATION sao 1 5 _:~. l 0 to 4o so 80 100 120 POWER (PERCENT) ) FIGURE SNP-RP-20: TYPICAL REACTOR CORE SAFETY LIMIT CURVE FOR FOUR OPERATING LOOPS (REV. 2) r 13-55 1 0394C

l Above 15 percent quality, the WR8-1 correlatieg 15 percent quality. At higher pressures, a higher Tavg is necessary f,,, is not valid. Furthermore, at relatively low pressures. Out 15 percent quality. (void fraction - 1.0) occurs at approximately 15 percent quality. higher pressures, the quality of water may be greater than 20 At percent before the void f raction is approximately equal to 1.0 and I DNE occurs. Figure SNP-RP-20 is the most conservative combination of graphs A, B, and C for various pressures. Only on the highest pressure ll curve does graph C further restrict allowed operating conditions, ll values permitted by the hot On the lower pressure curves, the T,,g { leg saturation curve and the DNS curve limit quality to belw 1$ i percent. Lowering the flow rate through the core also restricts Tavy At lower flow rates, the enthalpy rise of uith respect to power. each unit mass is increased for the same heat transfer rate. As a result, with reduced flow, coolant at the same Tavg may have a Limiting quality to 15 percent further res.tricts greater quality. the safety limit curves for three-loop operation. Figure FNO-HT-105 is the safety limit curve for normal f The graph below the safety limit curve is the operating pressure. l A comparison of these two graphs indicates normal operating curve. the margin available for temperature incre'ase during transients. An additional safety limit imposed by technical specifications In order to prevent is a restriction on primary system pressure. overpressurizing the primary system, system pressure is limited to 2735 psig. 13-56 -~

l 09/11,86 Page 5 NRC Exam Section 1 QUESTION RO 1.14c:(SRO 5.11) State three possible initiation signals which may initiate SI upon a steam c. line break accident. [0.75] RESPONSE c: We request that the answer key be expanded to accept Hi Steamline Flow with Lo Tave or Lo S/G Pressure

REFERENCES:

Logic diagram 5610-T-L1, Sheet 19, Rev. 5

• R15/cas

e e g 41*5 O 43 CBCIEN Em te thEIR M35 3 63 fa4 .a.. 3 4Rs La MIB IWEA aERA88 se 3' 3 4 Wit Jasis tag a.. J 4 mES MEC /"" E O 3 /" I IN /" O I 9 N 3, 3 3 y y y v.. y n ^ n. = = ,R = =. = ,L g M Mk k k I 8Pt l 4 -C -C O 5;'h f-T l u a m ai 58 % it1r1 ST 84WI - =r M = se f'2B l

===5~M AL5A JuL ,e.g' u 57 R-Eams i Juut

== u m oir axe n g = g lf1P M "~" CS/4 1 e-aan 57tJe stafm aguse mas 14LIEL M 196 51 14k1 2 h t g met mW8 mat 3, 3 3 IERm313 LJB LDE IIBuft3 - Lag M f-sfMt R4 m,tg 3 LEB M 3 3, 3, 3 3 C5I et at WW 58 at et 38 31 se og m it tg tg R R ts tz R 81 tt LC 4 R 44 6 M N

1. 4 W

W W 44 6 W e -C -4 - 4 6- ^ n = = -o m aut u. = = T T. staans l La LDQ S % i $W RR dist13 M RW R&4 5 M M a:Je m na Rriry o e Watie 193 CDPVC i i i i i n.= icm um ca. ~c av a w ->w =- . ~ ,o m Q @ C M u. W TS D SHEIT ft _, _,,, a t i, i e le. r as ma.e== as.u is.. 4/f.L_/h '22:25E.%Hh r .3 _a u

09/11/86 Page 6 NRC Exam Section 1 QUESTION RO 1.16:(SRO 5.13, SRO Requal 5.09, RO Requal 1.09) Explain two reasons why the computed moderator-only power defect underestimates the actual defect, due to redistribution at BOL, assuming a ecleulational value of MTC of.4 pcm/deg. F. (HFP, Cb = 1200 ppm]

RESPONSE

We request the answer key be expanded to accept a general discussion of MTC, Power Defect and redistribution due to the large percentage of Operators who had difficulty in understanding the question. 4

• RI-5/cas 4

m n

09/11/86 Pcge 7 NRC Exam Section 1 QUESTION RO 1.19:(SRO 5.16, SRO Requal 5.07) Refer to figure # 363, discuss the two factors which afTect the Axial Power Distribution at EOL and result in a relatively flat power distribution.

RESPONSE

We request the answer key be expanded to accept the affects of redistribution and flux burnout in middle of core since Turkey Point is in Fuel Cycle X and our flux is slightly positive at BOL.

REFERENCES:

Telecon with Turkey Point Reactor Engineering on September 9,1986. G e i

• RI-5/cas

,.,y._ ,_p

09/11/86 Page 8 NRC Exam Section 1 QUESTION RO 1.21a:(SRO 5.18) a.Why is this action necessary during Natural Circulation Condition (.75)

RESPONSE

a. We request the answer key be expanded to accept responses that address non uniform mixing of the RCS in general, notjust the pressurizer. e G

• RI.5/cas

09/11/86 Page 9 NRC Exam Section 2 QUESTION RO 2.03:(SRO 6.02, SRO Requal 6.02, RO Requal 2.02) Which of the following statements correctly describes the RHR System lineup when HOT LEG RECIRCULATION is established? (assume both trains of RHR are available) a. Both trains are used to supply hot leg recirculation exclusively. b. One train is used to supply hot leg recirculation and the other train is used to continue cold leg recirculation. Both trains supply both hot and cold leg recirc simultaneously, c. d. One train is used for hot leg recire and the other train is put in standby (i.e., recirculates from RHR HXer outlet to RHR pump inlet).

RESPONSE

We request the answer key be revised to reflect"b" as the correct answer.

REFERENCES:

EOP-ES-1.4, Steps 1 and 2 dated 07/27/86

• R1-5/cas

,,,, 3 rt 7 3.EO P.ES.I.4 TRANSFER TO HOT LEG RECIRCULATION 7/27/86 l STEP l l ACTION / EXPECTED RESPONSE l RESPONSE NOT OBTAINED 3 NOTE l 1 l* Foldout page should be open. l In the event that a passive failure occurs and it is desired to isolate l l l the safety injection system. Attachment A should be performed to l obtain RHR hotleg recirculation. I i l* In the event the alternate RHR recirculation path is needed. l s. . _ perform steps of Attachmen t B. ....._._..............._.J 1 Align 51 Flow Path For Hot Leg Recirculation a. OPEN the hot leg injection header isolation valve MOV 3-869

b. CLOSE the following safety injection pump cold leg injection valves:

1) MOV-3-843A

2) MOV 3-8438 y._....._._..........._.....3 NOTE I

I If the second train of RHR cannot be made available, proceed to l alternately inject into the hot and cold legs using the single train l of high head recirculation. The period for each mode ofinjection l (hotleg or coldleg) should be 12 hours. Once the second train of l l Iow headinjection is available, the high headinjection path should l be returned to the hot legs and left there as long as low head i l injection to the coldlegsis maintained. l l

• Valves listed in Step 2a have a reach rod to auxiliary building l

i._._"o"d5"!^.=_"'dr._._._._._._._._._._._._._._.i 2 Estabiish iniection To coia togs utiiirin, Theidle RHR Train As Follows:

a. Dispatch operator to manually CLOSE the 3-759 valve on the RHR train supplying suction to the safetyinjection pumps

1) 3 759A 98

2) 3-7598 f{a z,o 3 emsswee

3.EOP ES 1.4 TRANSFER TO HOT LEG RECIRCULATION 7/27/86 lSTEPjl ACTION / EXPECTED RESPONSE ll RESPONSE NOT 08TAINED l 2 Establish injection To Cold Legs Utilizing The idle RHR Train As Follows: (Cont'd) b. CLOSE the $1 pump recirculation phase suction stop valve assocated with the idle RHR pump

1) MOV 3-863A 9.E

2) MOV 3-8638 c.

OPEN the following RHR Heat Exchanger Component Cochng Water Outlet Valve assocated with the idle RHR train.

1) MOV 3 749A OR

2) MOV 3 7498

d. OPEN the following RHR cold leg injection valves

~

1) MOV 3-744A l

2) MOV 3-7448 l

e. START the idle RHR pump i f. VERIFY RHR cold leg recirculation flow FI 605 .3 Retura to procedure An4 st.,in Aerect END OFTEXT e eM(84tidt { Q Q (b) }

g Q n ' Q F '. J..ii;i g Sy : A *n !!!{igl j hi j tfb'l P"N sibd "g%'3'ac's-li i 'i_.._d i4" i e s 1 ! ~ 1r I t eg is., Ifl;. e 1.. u : a, n dbEU h f& II- ~ ~ tidil__p_ _T.. --. ] nj 3 s, - -- a j i ~ h HI ,, d g<llg ig4_ 14 j4 [B;; li 7 q .g r" & lrll!j@i..g,Crin' l _ 4, i'jl=. I! i !l T'?hi i ? i;A.i l 1 'WeJ b'i a p!! .y'i ! --% 0:d i , M N, 1 r it ii h 4 !I!! tlasl hd 7 b! !!E di t i g i s n d eL c,',i li!!! p!!![#'S.p:!{'u i e (yP~LC 4 ',.) ii.,I. .a ,j ! k "ElI -y!$ 3z 3 ' {$ f i [

• I!5' l'i ii

[.. '. , %lff " {1 g gt, e Ia '! ; I 1 %i I w, y, g -4,e w i ! #4' ,l '6% W F,J.L a," g_y '. 7 5y p. ,a 't 3 ! .r --Il i 1 li 7- ,{g :gn [' "I ' I p d ~A! Icu 4! j g 1"

n

!,i k i ~ .f I i l@~{ ET J d 9 lejM shl' ! 4Hi I IH'" - aud1T g jg1 i! Iul 'I N I, .s h

:~__l,!'

k!?! e, ,.qd ilk 4 insi! d N"I' 4:!'. '-ar~~-$1I~i.liPi i 'iP ~~ 2 V Il 44'% !i "Y"b-._, sv., r s g= ygl g-s y ~ ,Y .!_.8._ _____ __j'! I-l i,i-@ g*,nl s,, i 1 I 9 1 E $..fI72. i E ec,2. o- ~~

09/11/86 Page 10 NRC Exam Section 2 QUESTION RO 2.11:(RO Requal2.06) List the 6 essential safety related loads served by the Component Cooling Water Syste'm (redundant components like CCW heat exchangers are one load).

RESPONSE

We request the answer key be revised to accept " Pass Heat Exchanger" and delete SFP Heat Exchanger

REFERENCES:

SD 040, Rev.17 b

• R15/cas w

a-'-v--e-m- - - - e- --v --=--- w v w -- --

MOTE: It would be possible to drain one or bot,h CCW systems (Unit 3 k 4) through a valving error at the $1 pumps or boric acid evaporators since at these points the two systems could be tied together. Parts of the CCW system are automatically isolated by accident situations. This isolation occurs for several reasons. One of the main reasons is most of this part of the CCW system lies inside the containment. Therefore during an accident situation, possible damage could occur to this part of the CCW Since the containment would possibly be inaccessible during certain system. accidents, safety injection with hi and hi-hi containment pressure for example, parts of the CCW systems are automatically isolated. In addition, the loads that are automatically isolated are unnecessary for plant operation during accident situations. Examples are the normal containment coolers, etc. These loads would cause an unnecessary heat load on the CCW system. The sequence of automatic isolation will be discussed in detail later in this system description. The essential safety-related heat loads served by the component cooling water system are: 1. Residual heat removal heat exchangers 2. Residual heat removal pump seal coolers 3. Safety injection pump oil coolers and seals 4. Centainment spray pump seal coolers 5. Emergency containment coolers tr-Component cooling water system design flow rates.

09/11/86 Page 11 NRC Exam Section 2 QUESTION RO 2.12a:(SRO 6.13, SRO Requal 6.10, RO Requal 2.09) a. Refer to figures #397.1 and #397.2 Reactor Coolant Pumps. List all interlocks which must be met to start an Oil Lift Pump (OLP). . [0.51

RESPONSE

a. Figure 397.1 requires breaker to be closed, but 397.2 does not. We request that no penalty be levied for a response that does not include " Breaker Closed".

REFERENCES:

Figures 397.1 and 397.2 I a

• RI.5/cas

09/11/86 Page 12 NRC Exam Section 2 QUESTION RO 2.13 a and b:(RO Requal 2.08) Refer to figure #398 Heater Power Distribution. State the purpose of the a. Current Transformers (CT's) wrapped on the heater supply lines. b. How are small electrical grounds (e.g. grounds which do not trip breakers) detected / indicated on Pressurizer Heater strings?

RESPONSE

a. We request the answer key to be expanded to accept the response "CT's supply any indication meter." b. We request the answer key be expanded to accept the response " annunciator in Control Room for 480 L.C. Ground." In addition, there is no ground detection instrument for pressurizer heater strings. We request that alternate answers for ground detection be accepted.

REFERENCE:

Training BriefNo.64,06/15/85 ~ i

• R15/cas

06/16/86 Page3 FLORIDA POWER & LIGHT COMPANY NUCLEARTRAINING DEPARTMENT TURKEY POINT NUCLEAR PLANT TRAINING BRIEF NO.64 Indication of pressurizer heater current is now provided on "3A", "3C" and "3D" 480 volt load centers for the control and backup group heaters. (See Figure 2) In addition, a phase selector switch was added. This will help operators quickly identify malfunctioning heater groups. An operator can go to the load centers a determine which heater group is malfunctioning by comparing current flow in er2ch group and each phase in each group by turning the selector switch to the de positions. T.E.Lightfoot PC/MTrainingCoordinator i l l l I Y ( n

06/15/85 Page 2 FLORIDA POWER & LIGHT COMPANY ~ NUCLEARTRAINING DEPARTMENT TURKEY POINT NUCLEAR PLANT TRAINING BRIEF NO.64 PC/M 83 96 " Pressurizer Hester Ammeters - Unit 3" t 1 j e r i l i l l l l s l l l I i l FIGURE 2 r f $b).I3 G{L i

4 .N <4 h.!! j 6 s v. Nit I; ( g.g # ,;,f, y c. y;! n.sJ-n ,r,.g e .(.s, "!4. 6 s;..l p:l l;, 3 2... Og.:M o o i m g - z. m..,. 11[{ hi,!'um v . h ) f.!"l 0 I l e. i gi ii 4 .. w,g,nl L j A,*e/ ,..,ll! wigu..;4Qn; i-em 1, S Wp r L i g H: a 4s.n..,. 7,. m... w .. g m. 1 ., r j! g n ny, i e j r,e.y 1.. ss m- ,7 r1 g J): ; L @r---M :pg,J, l x-( i. s!.,.,,. .,-.s m s: l g

i.

= 9 ?.

• g L),.,

.. ) .i m ] y y- +;.3.jg :..:ve g :1. n ,.a.:. n ,y i e w. g

4..,

.., );. <e-- w,-

ggy, 7../.; V; y c-

,4 , :,9 si <=' g., lI, g...,l! ip .:;y c i. %, j/ (- m U. 1! l il h lL M .:..y : q.g.A. I ,'k;

D,j. e

.1 , r-- g.i i 7------ g.i 3 3 eos v i Mo, i m yy. gap g...g. .nEg=m.to, y, e. gg1 j. : p. e e.emg x p l r 3 o. i m g mg 9 ;i u,,g g,. ];--u,n g j z gy:pp g n.,Ph?g; gig $i4 2 !it g k, M.O%.u.xy - i %..m... 1,,,e 4 :@ i w> i

g

l. x

. G,

,, i y ;

3 w E j '3g g ' L'- [ ',,5 '., [,

,}

,is 3 .. 7 j,: '

3.,r :. 3.pl l

q - U_ n ge fse n { N 'nN y{g [ i p.R.. ; '~& ; A *..l ~ ~- 7 (?,. 3 P i i $ 'QM - '"*j" ( ~{ i! i I ]! ji @d$'Ri o Jif*Q.. y,.j % A wwwl:-Q., p,.., g l i MQ n L .il u gik,X 2,, i. t. M01. T/3 j g 3_1 g .t f :. 3 1 -s

. s t,

s 31 It '1.* 't" ) !,: 8 i. .q

e jV.... }l;ip. -

a'l %m, ; I a cs 2. y g ;i Nhdb i k.' N)fJ r*** 'S j nwww. " {a nn a hk

• I

(! !! [ ? p@ !! !! :is .b.a ll MHM @w. I n, i l.. mm ,3.is i c o 4h o s ww n f.g;.&,f,1 ewg4;. }. #{y. V!:{t* g ag

9-c at

~ y .4 a1 p u i snwinnii i n amiwiiimn i a k% yf Nh }...... s.!g'gll!)Tj!ljjj!: g[ji jhhhhhi!ihffdfffftl 2 v. ('8.1

1. '~"Ifl*1!!II*iliii*itili'itiililit!,4

.p.;4 2,.9.-,4x. y, ,a ',I ~, ~. j': a ng.. 9,,:.,m g o a. m = ew,.. s

09/11/86 Pege 14 NRC Exam Section 2 QUESTION RO 2.22: (RO Requal2.14) Refer to figure #409 " Basic CVCS Flow Balance". For points 1 - 5 state the design values for flow (gpm) at each point.

RESPONSE

We request the answer key be expanded to accept 45 gpm to 120 gpm for point I due to varied LTDN orifice configurations possible. We request the answer key be expanded to accept the value of 25 gym for thermal barrier flow (point 5) which is the value of CCW flow through the thermai barrier. Students asked the proctor if they wanted the CCW flow and the proctor said "yes". I

REFERENCES:

Turkey Point FSAR, Table 9.2-2 l SD008, Rev.1-19 l l

• R t.5/cas

T. P. F5AR TABLE 9.2-2 NCBGNAL GENICAL AND VOLUME CONTROL SYSTEN FERFORMANCE* 40 Unit design life, years 24 seal water supply flow rate, spa ** 9 Seal water return flow rate, gym 60 Normal letdown flow rate, spa 120 aus letdown flow rate, spa J 69 Normal charging pump flow (one pump), gym 45 Normal charging line flow, spa Maximum rate of boration with one transfer and one charging pump, pps/ min. (from initial 23.8 acs concentration of 1800 ppe) Equivalent cooldown rate to above rate of 6.8 boration, F/ min Nam 4== rate of boron dilution (two charging pumps) pps/ hour (from initial RCS 350 concentration of 2500 ppa) Two-pump rate of boration, using refueling water, ppa / min (from initial RCS 6.2 concentration of 10 ppa) Equivalent cooldown rate to above rate of 1,7

boration, F/mia Temperature of reactor coolant entering system 555.0 at full power, F (design)

Tamperature of coolant return to Reactor 493.0 Coolant System at full power, F (design) Normal coolant discharge temperature to 127.0 boldup tanks, F Anount of 20,000 ppe boros solution required to meet cold shutdown requirements shortly after full power operation (Tabla 3.2.1-1. Line 37), sallcas 3080 (including consideration fcr osa stuck rod)

• Ranctor coolant water qml ty is gives in Table 4.2-4.

"Volmstric flow rates in spa are based os 130'F and 2350 peig. e

A motor operated isolation valve (626) is provided on the penetration outside containment. A flow element with local indication and a flow switch is installed downstream of the outside isolation valve. Normal CCW flow to each > themal barrier is 25 gpa. In the event of rupture of any thermal barrier a high flow condition would result. This condition will automatically close the outside isolation valve thereby shutting off thermal barrier CCW flow on all three pumps and actuate the RC PUMPS THERMAL BARRIER COOLING WATER HIGH FLOW annunciator on panel A, window 1/1 at 100 gpe. The RC PUMPS THERMAL i BARRIER COOLING WATER LOW FLOW annunciator pane) A, window 1/3 will be actuated at 70 gpa if cooling water flow is insufficient for thermal barrier heat exchanger operation. Relief valves are provided on each thermal barrier cooling water outlet line to prevent overpressure. The isolation valve can be manually opened or closed from the control room, however, if the isolation valve closes due to high flow the valve can be opened and kept open only by i holding the control switch in the OPEN position. A coolant pump can be operated without CCW to the thermal barrier provided that injection water is available. However, CCW must be supplied to the bearing oil coolers at all times that the pump is in operation in order to prevent damage to the motor. i l The themal barrier and oil coolers' component cooling water supply and return valves will be closed automatically on a phase 8 contairment isolation signal (P signal). Motor Lower 8 earing Heat Exchanger The lower bearing sump contains a one-pass, coiled-finned tube heat exchanger for cooling the bearing lube oil. Component cooling water passing through the tubes, at a flow rate of approximately 5 gpa acts as the cooling fluid. The pressure drop through the heat exchanger is approximately 2 psid. Motor Upper Bearing Heat Exchanger

• /131 50004 Rev.t 19

.a a w o s m a a + -,.. - .w a --.s COMPONENT COOLING WATER FLOW PATHS WW Ef5LY ; \\ Fim rte 1 08 S f F17 ] mase x x F1M Fies P e <r d A 4

E1' A 4

===. V" / t X w. 2 raes i I M ma ~

& C r,

J 43.n. 9 3 8eeru t I [ \\ f I , V x-4, L "a** os4 m 8 mv

• )

g('l:i p"m M mr n Krasa .s. mi is r= l maeo 4 3,,, g: l r.r X ... +: x ,,,,,, me i r, 4,O=rept. I i L 4 L. I I

BASIC CVCS FLOW BALANCE cy..- n-asm. DUECmON ~~~ p ,,,,,v V' An ~ A C (3 GPM)(3 GPM) ggAL ~ 8 LEAKOFF 'IS GPM)y if 9 GMt t if _j ( ,r

c. p 'S y <.Ga.g TMind.AL BARRWt RCP R

y y (CovFwhaC9puh a se opu CHAROMG PUMP FLOW AR.I FIGGE e ._----._.y m- - m- ____..--.v., ,,m..,..,, _.

09/11/86 Pcg215 NRC Exam Section 3 QUESTION RO 3.08d: Match the interlock descriptions in Column A with the appropriate logic required to cause rod withdrawal to be blocked in Column B. (column B items may be.used 2 more than once) COLUMN A COLUMN B a. Power Range High Flux @ 103% power

1. 1/2 b.

Overtemperature Delta T rod stop

2. 2/2 c.

Intermediate Range High Flux 3.1/3 d. Power Range Rod Drop

4. 2/3'

5. 1/4

6. 2/4

7. 3/4

RESPONSE

d. We request the answer key be revised to reflect "5" as the correct answer.

REFERENCES:

i Operating Diagram 5610-TL-1, Sheet 17 l

• Rt.5/cas

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

q, 7 3 ,g o g gl ._ p xn u-M '# "E. 9 J L eo, ,as f t ,ns. ~c ~= ,q-a- ~. = ',at--O .r. r ' gr, ~ $ $' = = ,=r,=,. a m ,=,a =. l l J l r-p 1 c t

• i

,3. 1 4 i t + , A l, H-co a -_, =i=" = 1 a 1 c rs,l 36, , ra) y H a_i qa , p 4-g -61 , =. ? o' ', =. so r a -e ,1 1 ___t _g,as g'es ni L W F'" g @, ei L m - } as ~~" l ,_l ,l J C L f J U L ~ ~ ~ T I ~' 3 t N ,-o JU W UL i sa t t i b \\ f) 4,g 1 ype t -as: n==gggw;q,u=tt 3 L 4 , _,,a pg sGT17 ., wrg sm_ JM&"xs" 5610-TL-l

O F

/

1. d inn.1

,,, e. e s e- ) Ro 39V D ~. l l I

09/11/86 Page 16 NRC Exam Section 3 QUESTION RO 3.09a: For each set of plant conditions listed below STATE all functions (of 1,2 and/or 3) which would be generating a rod movement signal and STATE the respdnse of Bank D rods (in, out, no movement) to this signal. Bank D at 200 steps.

1. Bank Selector Sw.

2. In-Out-Hold Lever

3. Plant Condition a.

Manual IN Urgent Failure

RESPONSE

We request the answer be expanded to accept" Rods Do Not Move" since this a. l is true for an urgent failure in a logic cabinet.

REFERENCES:

Conto: System Diagram 5610-T-D-12A, Sheet 1 ofi

• R15/cas

. ~.

._ ~ j

=!=:=ljljjim !"l!=l !=L=i=l

~ ~' i ? J ,/ ll l i g i j 7......_............, E I g illl) l l 3' I!; N l[! ril j j

• I!

I!I! I ...1 [ ll l il ,,!lg!!! 13 j ti-ti-l - l 1( 1 'a 9i, t M,l I li;8 1.I8 __ l It' i il 1 .-1 ii g! 'lill.' il i. ligi; t i s. l Ill- - 8 hii

i.! '

l __1 : n

p. -

e h + l j , k I I en - > - = = s[ i.. i-lE + I g ~ I en . e h 'l +; i t i i. -lr i '1 ~ .jglI,1!i._._I i 1 'f,_)T t 1 i ij s,.- i... +p s l a l-. .. R, fh l l F

l. i H

j ,i i +, r -y lilli l ll I. II, -i I t: gli i sp t ljl'll, i i ii i i i s l l ti--- ~L[ + r l 5 - ) +1 . i e ,. ~

Wi

= '.!

, l; + - lie-- _--{5' '~ k . m i n_ c 2 - U U r'lill!I IlI 'I ~

! I r-:--

ll, I i.l t r Xu ll' ! 6 li Q'i[4""!! i i i i gi 'l \\'l l l l l s \\ til \\ ' /' I i .,!llljl{' d, 'I I t's l$ Ldji 1 ca-i 3 /}D I A i IE M 6 e I E8 ' ig ! E l g .g E I j'i j 5 I, B E L lj i l 1 4sdsmo._ b 9 i I s C 5 12 ri i 8 i i i_ i i 3 i A 3-0 4 E =

09/11/86 Page 17 NRC Exam Section 3 QUESTION RO 3.14a:(SRO 6.20, SRO Requal 6.00, RO Requal 3.09) Assume the following CCW pump alignments: B&C pumps are rackedin andin AUTO B pump is running A pumpis racked out and open I If'B' pump received an over current lockout what signal (s) would auto start a. 'C' pump AND at what time would it start?

RESPONSE

We request the Operators not be penalized for not including the low pressure setooint in the first part of the question. The setpoint appears in the answer key but was not asked forin the question. 4 1

• R15/cas

_.___-_--._...-_-_._.-,._.,-------__-_----m,_ y-,e___ -my

09/11/86 Pcge 18 NRC Exam Section 3 QUESTION RO 3.15:(SRO 6.21, SRO Requal 8.08, RO Requal3.08) a. For the Unit 3 Rod Position Indication (RPI) power supply, describe the supply to the DC power system, including both normal inverter supply and alternate supply. Include all automatic features associated with these two i sources. b. The Rod Deviation Monitor generates a ' Rod in Motion' signal. EXPLAIN the basis for this signal AND the action (s) which are initiated by the signal. 1

RESPONSE

We request the answer key be expanded to accept a drawing (sketch) of this a. system. It is our understanding that the Examiner stated that a drawing was acceptable. b. We request the answer key be expanded to accept " Alarm at 12 steps Stationary,24 steps moving" since this demonstrates the setpoint change as stated in answer key.

REFERENCE:

l SD6, Rev.0-12 1 .i l

• RI5/cas

1 b position, from the rod position indication system are compared to each other. Whenever predetermined levels are deviated from a bistable / relay circuit trips to actuate an annunciator and a local indicating light. The rod deviation monitor cabinet is located in i the control room, behind VPS, next to the RPI cabinets. See Figures 7 4 8. The relative positions of the control rods are continuously mont-tored while the rods are in motion and at rest. If any rod in a bank deviates in actual position relative to another rod in that bank by more than 12 steps (7.5') at rest and 24 steps (15') in i motion the SHUT 00WN R005 0FF T0P/R00 DEVIATION alars is triggered. These deviation limits ensure an acceptable power distribution. The shutdown rods are monitored by another drawer in the cabinet. l This drawer will generate the SHUTDOWN R005 0FF T0P/R00 DEVIATION i alarm on annunciator 8, window 9/3 when any shutdown rod drops below 217 steps. I l Operation f The input to the control rod comparator drawer is the 0-3.45,VDC l rod pos'ition signal from each RPl. Refer to Fleure 7. The inputs l enter the drawer through the Test /0perate switch. The inputs are l then conditioned and compared. The comparator circuit then pro-duces an output v11tage proportional to the algebraic difference between the high )nd low inputs. The comparator output voltage is amplified and then sent to a histable trip-relay driver circuit. When the voltage level to the histable reaches the setpoint the indicator and the SHUTDOWN R00 0FF T0P/R00 OEV!ATION alarm circuits are activated. The setpoint is adjustable between 6" and 24" and is set for 12 teps (7.5') When control rods are in motion the trip setpoint of the rod deviation histable is changed. This change prevents falso alarms from spurious signal inputs. The setpoint change is initiated by the rod cont ~rol system. Whenever rod motion is initiated a relay operates which changes the bias on the control rod deviation bi-stable. The setpoint for rod deviation is now adjustable between 15' and 36". The setpoint is set for 24 ste,ps (15') presently to I ensure an acceptable power distribution. g SD6-Rev.0-12 i

09/11/86 Pcge 19 NRC Exam Section 3 QUESTION RO 3.18b: b. Exactly how does Steam Generator level flow controller FC 478 operate the ' Main Feedwater Control Valve? Include motive force (s) and signal conditioningequipment.

RESPONSE

b. We request the answer key be expanded to accept instrument air as a motive force. Question was read as" motive force to move valve" I 1 o l l

• RI.5/cas

09/11/86 P:g220 NRC Exam Section 3 QUESTION RO 3.18b:(SRO 6.22, SRO Requal 6.07, RO Requal 3.05) i Unit 3 is operating at full power, all control systems in auto, Pressurizer level j I channel 1 (LT 459) is selected. Answer the following: l Would a common failure on LT 461 ~and LT 462 causing both to read a. maximum result in a reactor trip? (yes/no) b. The reference leg for LT 459 (PRZR level) develops a significant packing leak causing the reference leg to drain. State all the auto actions affecting the Pressurizer which would occur as a result of the failure. List all functions of pressurizer PI's 455,456 and 457. c.

RESPONSE

b. We request the answer key be expanded to accept " backup heaters on." In addition, we request the Operators not be penalized for not including the water solid condition written in the answer key since the question asked for auto-actions and did not refer to a transient description over time. We request the answer key be expanded to accept " reactor protection and c. SI" as stated in SD009 and also accept "Hi and Lo Pressure Trip, Lo Pressure SI and associated alarms."

REFERENCES:

SD-009, Rev. 0-40 5610-T-D-16A, Sheet 1 of1 i i

• R15/cas

high deviation alarm also turns on all of the back-up heaters to heat the insurging water so that on a subsequent outsurge water flash to steam and help maintain pressurizer 4 will be able to pressure. The speed of the charging pumps can be controlled automatically from the pressurizer level control system or manually by the One or operator at the " MANUAL-AUT0" controllers on the Console. The normal mode of more charging pumps can. operate in "AUT0". operation is automatic except for periods during plant startup and shutdown. 59 summarize the pressurizer level The table on pages 58 and control and protection settings (pressurizer level span 181" from l 0-1001,, narrow range indication.) PZR Pressure Control f The six pressurizer pressure transmitters use the three steam side See Figure 24. Three level taps to sense pressurizer pressure. of the six transmitters supply press'ure signals to the three protection channels. PT-455, PT-456, and PT.-457 supply channels ^ t I, II, and III respectively. Signals generated by these three channels are used for, reactor protection, safety injection, anticipatory alarms and indication on VPA. (See DWG 5610-T-D-16 sheet I for details.) The protection channels along with their various comparators and controllers are discussed in detail in 50-63, Reactor Protection System description. The last two transmitters, PT-444 and PT-445, are used for the See Figure 25_ and OWG pressurizer pressure control functions. 5610-T-0-16,, sheet 2. In addition, they provide indication, PI-444 and PI-445, on VPA. Either transmitter can be selected to a one pen recorder, PR-444 via a selector switch, both located on VPA. PT-444 also provides remote indication at the auxillary feedwater pump station, PI-444A. SD9-Rev.0-40 t -,w,-~~-m--v, w w. w e-m ww. ,--------,.--n------~w .e,-m.,-,.--_ -,,em,-cv-we-e,--

i 8.' k i i 2 e IIi ! g' 7"! 7-e g' 3 s a is ! s fg a <I a T-I ~ !}I .f !!!}{)ll j g ! j '-8 168 di.: , L __r---; i c. I rdIlrl } ._, 4 ':= '. 55 g,..................,- .g g;g dS8' ~ 3 .I sa' Sija =41.d_ 5 J. '. i.l o -i 2 L. j_. I;i.,. y b'l '32 ! 3 4' Eli *,Ei id{;,' -8 11' ll! l i Et I 5

ri 58; cI I"i

el-w - [ y -s y e g. g h, ~' l b.jg ..........s g

g... 7....

11,: .. q

g la

l'-

a = 31 I I e_ '. 8' gh h . f, MN 'I* ri : I,'e ' -i 8 l,. i in e :. = I J li; I -a q i j. lisii l 8 d ts 4. sh.

m...: I 1,,. I r-------

tl8i

---

~~

l '3' ge ll .n e tt. .I egt . );gyI Q gg r-------------- g I .. y y _e{., s e.:

gJ

! L __.r--',I i. r= 1... 4;........L se 8.. = @r--- -- ! : 41 i l, 3 l-i i.....= i. II =......... -4I.........r-i,q 8 I t I sy y I .,a I f li. r-.!: i aj .!,?. ql g. ....... 3 g.e I _ ! @= a'..s.................=...4..'.yI.,; gl 8 sls _ j, s $1 3 i..- I 5. 4 .........c.,!, l, pels 3s e" d [ I ,__r salees I l 8 a 4 , L_ _'r - i_ r: l _I u 6.39, lil=II. = - r v : ql,..........1....i..iQ- ------- :-- ' l - Al -- tu; i r 3 I..............a l g { L.A I 4' I I 8 I l._sa{ l 3

...:............... 4 3.I, l,

a 8.. .i-E/ 3l qg s,, g,q I si' sist '...............4.....'...r.Is I 3 Y l lil s! Ilg{in II Ie83:3 -t il i I t t t = l g ', 1

09/11/86 Pcg2 21 QUESTION RO 3.19 Refer to figures 82.1,.2: n. What is the purpose of the Dropped Rod Normal / Reset / Bypass or Switch on the Power Range A drawer? b. What automatic actions (if any) would occur as a result of a " loss of detector volt alarm"on the Power Range A drawer? State the purpose of the gain switch on the Power Range B drawer. c.

RESPONSE

b. We request the answer key to accept 1) Blocks Auto Rod Withdrawal 2) Initiates aTurbine Runback Signal Students were led to believe that the P.R. Detector lost its hi voltage supply, which will cause the detectors output to fail low, therefore tripping the NIS 5% pwr decrease 5 seconds bistables which generate the above listed signals.

REFERENCES:

TP System Description #004 Pg. #46 & #47 TP Fact Sheet #012 TP Fact Sheet #003 TP Drawing 5610 T-Li Sh. #21

• RI 5/cas

.high isolation output, also amplify the low level detector current signal input by a factor of four. Bistable Relay Drivers The O to +10 volt DC signal from the summing and level amplifier is applied simultaneously to six bistable relay drivers. These assemblies produce an indication when the input voltage exceeds a preset value. When tripped, the bistable removes an AC control signal to relays located in the reactor protection racks and the miscellaneous relay racks. The modules are: 1. Overpower rod stop signal and alarm 2. Power above P-8 permissive signal (45%) 3. Low range high level trip signal (25%) 4. High range high level trip signal (108%) 5. Power above P-10 permissive signal (10%) 6. Dropped rod circuit An additional relay driver in the Power Range is: 1. Detector voltage alarm All seven relay drivers have indicating lights on the front panel Of the Power Range A drawer assembly for each channel. The lights when illuminated, each indicate their respective relay driver is in the tripped condition. Dropped Rod Circuit (Power Range A Drawer) The dropped rod circuit is a sensing circuit which monitors fast rates of change in reactor power. If a rapid negative change in power occurs as a result of a rod dropping into the core, a signal is generated to trip the dropped rod bistable relay driver. Operation of this bistable differs from those previously discussed only in that it must be manually reset after it is tripped. The dropped rod reode selector switch, located on the front panel of the Power Range A drawer assembly is used to accomplish the reset. The dropped rod assembly basically consists of a differential amplifier and a bistable relay driver. The differential amplifier is designed so that the

• /ec S0004-Rev.1-46

/ I and some is zero for a steady-state condition I output of the ampliffer The differential ~ asitive voltage for a momentary decrease in power level. aplifier receives the 0 to +10 volt DC signal from the sunning and leve This signal is applied to both inputs of the differential ampli-amplifier. fier, but one input is connected directly to the amplifier and the second This adjustable time input is first passed through an adjustable time delay. delay, between 1 and 5 seconds, is set by means of the delay adiust octentiometer located inside the drawer assembly. L Any abrupt., dec hactor' power will trip't'hi droppInd ' rod bis'tiable wPiich will p rod withdrawal, initiate a turbine runback, and triggers the POWER RANGE DROP AUTO TURB RUMBACK AUTO ROD WITHORAW STOP alarm on a M Low Voltage Power Supplies (Power Range B Drawer) +25 volt DC power supplies provide the regulated OC voltages for the operation of the circuit assemblies. High Voltage Power Supply (Power Range B Drawer) The high voltage power supply develops an adjustable, regulated +300 to The high voltage DC level output is adjusted.by ~ v'olts DC for the detector. potentiometer located inside the drawer -

r. high voltage adjust means - of A low voltage DC sig.ial, proportional to the high voltage output,.

assembly. is appl.ied to histable relay driver which will trigger the MIS POWER-RA 65 1 . LOSS 0F DETECTOR V0t.TAGE alarm on annunciatpr 8, window. i:t-Test-Calibrate Module ti.- - Individually adjustable test signals can be injected ~ independently-or simultaneously at the input of either anneter-shunt assembly of the ~A a An Operation Selector switch on the 8 drawer. i-detector section signals. provides means of selecting either A or 8 or both detectors for injec The test signals are continuously adjustable by means of the test signal. controls with calibrated dials. In all cases, the l two front panel mounted Bi-stables' test signals are superimposed upon the nomal detector signal. they which will be affected during channel test do not require bypasses since ,-, -_, _ _ ^ ^ - - - - ' - - - - - - - _ _. _,,. _ _, _ _ _ _ _

Page 1 FLORIDA POWER & LIGHT COMPANY RKEY POINT NUCLEAR TRAINING DEP ARTMENT FACT SHEET NO.012 ROD WITHDRAWAL STOPS l CONDITION SIGNAL ACTION Intermediate Range 1/2 @ AMPS approx. Auto & Manual Rod i High Flux 20% PWR Stop , Power Range High 2/4 @ 103% PWR Auto & Manual Rod SWp Mux Overtemperature

• Variable Setpoint Auto & Manual Rod Stop l

Overpower Delta " Variable Setpoint Auto & Manual Rod i I Temp. Stop P.I. System Rod le45 @ _< 20 Steps Auto Rod Stop Only j rop Auto Rod Stop Odly, P-2 1st STG. < 15% Load Overpower conditions block automatic and manual rod withdrawal, i.e., conditions 1 thru 4 above. i rod withdrawal stop when actual aT is within 2.50forsetpoint. AT rod withdrawal stop when actual AT is within 1.Sofofsetpoint. l l u

wieu E Page 1 I i FLORIDA FOWER & LIGHT COMPANY TURKEY POINT NUCLE AR TRAINING DEPARTMENT FACT SHEET NO.003 Turbine Runback (1) OPAT 2/3 C H Cyclic Runback on Governor 5% shot untilsignal clears (2) OTAT 2/3 CH Cyclic Runback on Governor 5% shot untilsignalclears (3) Steam Generator 1/2 Pump Runs back to 60% load on governor Feed Pump >60% Load L . :... fcKitf}6T/47EOMor'- (4) ENISUrop-Rod M & a..... p I N '"*5 >70% Load Runs back to 70% load on loadlimit. (5) RPI Rod Any rod on Runs back to 70% load on load limit. bottom. (Other rodsin One shot 30% runback on governor for bank must be any power level ~~ > 35 steps) > 70% load. All have a defeat switch that can be used to block the 5 SGF1, NIS, (6) RPI runback signals. Has a selector switch which allows you to select NIS or (7) NIS & RPI RPIsignal to initiate the runback.

References:

(1) 5610-T-L1. Logic Sheet 21 (2) 5610-T-D-14 f.U -E. 'f Qu. k

7 im - 48 OWEMM3f- ~ ot a mw g _ist sar g 4 y ,m, -- en mm M M g, TC IC N

• esa eeft egg 4323 IPCI c'es W

' I/i }[ u u CUenGL1 N A a - r, - -r .a 5" = 1 Elk A NU L.i k dC L e EuuL [ L. t I en=.= m cuts ewiu a man WIL MT s BRT SIGIAL CLEAA5 .puumastEB) TERE Penramaan 1g.3 l l 't I)L ~O- + 1 ~~ ( ~ l tw t=n nis

=

mai u.

=

m t i=== t..t=1 y 7 Co o WR se StaN efe.SL.N. TO MREE LW.I a M 15 %,at.t o.r.. s 1 t EP%.m e.CF.OSS LDAO Of ises iauma WINT 3 Cr - = __ -s S=== esi=== aw.- c AI = = * 'u.su.is FLORIDA POWER & LIGHT CO. o._I = =,- -... u ,gr 4

• TLRKEY POINT PLANT omne a M

,m. aw/rv ano maa ecaac/w as-se K 'FM a are/ as-esti sta asp-tw Ls .5ed (e n..m un%M,*:ra* '*

• • 3 2 Y

$ $$op* 5610-T-L u ra s (*" ",( [ (, ~ , TURSINE UNBACK SNEET 21 -I b

• #f7

e l A r - el gefW. K-g f un .n om a 3.,,. 8 WEE 888EE TC II stiS Aasst 1757e El ftW ,C t.aa,s s.e.t asrv.3. att fu.R 48K 'IE

• II 8

vs v -M -W -d PC44GS 1/4 .fT. wi ,11 L1,x-

m. a. n=

.uip yg .v. = ^ ?,,,m,,,.,, WW V w. ~ ~,m. =m,

=

g,g= F.o A m vwi ..n -M.. t,a .u.i== e STP uJ c t . ) l t u ( t r 95 l l ,.,,. a ~~ = ;; = = = ~_..., v.n, .... a i .i-a-miu. on ,. c

09/11/86 Page 22 QUESTION RO 4.04 (SRO 7.04) Which of the following statements describes the correct usage of the Critical Safety Function (CSF) Status Trees while performing EOP-E-0 " Reactor Trip or Safety Injection"? a. The CSF Status Trees are ONLY monitored when EOP-E-0 directs. f Awareness of Red Path Criteria is required at all times, but the CSF Status Trees are monitored ONLY after it is determined that SI can NOT be terminated. c. Monitoring of the CSF Status trees commences as soon as the immediate action steps are completed. d. CSF Status Trees are required to be monitored as soon as the procedure is entered and a valid SIis determined to have occurred.

RESPONSE

We request the answer key be expanded to accept"a" or"b" The question is worded "while performing EOP-E-0" which makes "a" correct also.

REFERENCE:

" Users Guide for T.P. EOP's and Background Documents" 19635:4 Pg. #16

• R1-5/cas

09/11/86 Pcge 23 NRC Exam Section 4 QUESTION RO 4.16:(SRO Question 7.15) What plant condition is assumed in the valve / breaker " Normal" position designation for the Operating Procedure alignment sheets?

RESPONSE

We request the answer key be expanded to accept " Mode 1,"" Mode 2" or " Power Operation." The question did not state as per 0-ADM-201 and the Operators are experienced in the use of alignment sheets for power operation.

REFERENCES:

OP-0202.2, Table 1,5/1/86

• R15/cas

OPERATING PROCEDURE 0202.2, PAbt J4 UNIT STARTUP - HOT STANDBY TO POWER OPERATION TABLE 1 LOC-VALVE / POSITION / IITIALS SYSTEM ATION INDICATOR FLOW OESCRIPTION_ RHR OC FCV *-605 CLOSED RHR HX Bypass RHR OC

• -745A OPEN RHR Recire. FI-608 RHR.

OC

• ~745B OPEN RHR Rect rc. FI-608 RHR OC
• -752A LO "A" RHR PP SUCT RHR OC
• -7528 LO "B" RHR PP SUCT RHR OC
• -754A LO "A" RHR PP DISCH RHR OC
• -754B LO "B" RHR PP DISCH RHR OC
• -756A OPEN "A" RHR PP DISCH PC-601 RHR OC
• -7480 OPEN "A" RHR PP OISCH'PC-601 RHR OC
• -766A OPEN "A" RHR PP Seal Leakoff Telltale RHR OC
• -766C CLOSED "A" RHR PP Vent for Mechanical Seal RHR OC
• -766B OPEN "B" RHR PP Seal Leakoff Telltale RHR OC
• -7660 CLOSED "B" RHR PP Vent for Mechanical Seal RHR OC
• -756B OPEN "B" RHR PP DISCH PC-600 RHR OC
• -748F OPEN "B" RHR PP DISCH PC-600 RHR OC
• -757A LO "A" RHR HX Inlet RHR OC
• -7578.

LO "B" RHR HX Inlet RHR OC

• -757C CLOSED RHR PP DISCH Bypass Around Hx RHR OC
• -7570 CLOSED RHR PP DISCH Bypass Around Hx RHR OC HCV *-758 NOTE 1 RHR HX Outlet RHR OC MOV *-860A CLOSED RECIRC Sump to RHR Suction RHR OC MOV *-8608 CLOSED RECIRC Sump to RHR Suction RHR OC MOV *-861A CLOSED RECIRC Sump to RHR Suction i

RHR OC MOV *-8618 CLOSED RECIRC Sump to RHR Suction RHR OC

• -759A LO RHR HX Outlet RHR OC
• -7598 LO RHR HX Outlet RHR OC
• -761A OPEN RHR Flow FT-605 RHR OC
• -7618 OPEN RHR Flow FT-605 RHR OC MOV *-862A LO RWST to RHR RHR OC MOV *-8628 LO RWST to RHR RHR OC MOV *-863A LC RHR HX to Alternate LHSI RHR OC MOV *-8638 LC RHR HX to Alternate LHSI RHR OC MOV *-872 CLOSED RHR to Alternate LHSI

vr w. mu e m...... ....,.,m. UNIT STARTUP - HOT STAN08Y TO POWER OPERATION TA8tf 1_(Cont'd) LOC-VALVE / POSITION / NITIALS SYSTEM _ ATION INDICATOR FLOW DESCRIPTION _ SIS OC MOV *-843A CLOSED BIT Discharge SIS OC MOV *-8438 CLOSED BIT Discharge SIS OC

• -845A LO SIS PP DISCH Crossconnect SIS OC
• -8458 LO SIS PP OISCH Crossconnect SIS OC
• -845C LO SIS PP DISCH Crossconnect SIS OC
• -8450 LO SIS PP DISCH Crossconnect SIS OC MOV *-864A LO RWST Outlet to SI and RHR SIS OC MOV *-864B LO RWST Outlet to SI and RHR SIS OC
• -864C LO RWST Cross Tie SIS OC MOV *-867A LO SI PP DISCH to BIT SIS OC MOV *-8678 LO SI PP OISCH to BIT SIS OC MOV *-869 CLOSED Hi Head SI to Hot Leg SIS OC
• -870A CLOSED Sectionalizing Valve - SI PP SUCT SIS OC
• -8708 CLOSED Sectionalizing Valve - SI PP SUCT SIS OC MOV *-878A OPEN SI PP DISCH Header-Sectionalizing SIS OC MOV *-8788 OPEN SI PP DISCH Header-Sectionalizing SIS-OC
• -886A LO SI PP'Suctfon SIS OC
• -8868 LO SI PP Suction SIS OC
• -886C L0 SI PP Suction SIS OC
• -8860 LO SI PP Suction SIS OC
• -888A LO SI PP Discharge SIS OC
• -8888 L0 SI PP Discharge SIS OC
• -888C LO ~

SI PP Discharge SIS OC

• -8880 L0 SI PP 01scharge SIS OC
• -895D OPEN Hot Leg SI PRESS PT-940 SIS OC
• -895E OPEN Cold Leg SI PRESS PT-943 SIS OC
• -895F OPEN Hot Leg SI Flow FT-940 SIS OC
• -895G OPEN Hot leg SI Flow FT-940 SIS OC
• -895H OPEN Cold Leg SI Flow FT-943 SIS OC
• -895J OPEN Cold Leg SI Flow FT-943 SIS OC
• -897A L0 RWST LT *-6583A SIS OC
• -8978 LO RWST LT *-65838 SIS OC
• -898J OPEN SIS Suction Pressure SIS OC
• -898K OPEN SIS Suction Pressure SIS OC
• -898L OPEN SIS Suction Pressure l

SIS OC

• -898M OPEN SIS Suction Pressure-Qg Q(

OPERATING PROCEDURE 0202.2 PAGE 34 UNIT STARTUP. HOT STAND 8Y TO POWER OPERATION TABLE 1 (Cont'd) LOC-VALVE / POSITION / .41TIALS SYSTEM ATION INDICATOR FLOW DESCRIPTION l MISC. SI. - CONT. SPRAY (C.S.)l CS OC

• -844A LO CNTMT Spray PP Suction CS OC
• -8448 L0 CNTMT Spray PP Suction CS OC MOV *-880A CLOSED CNTMT Spray PP Discharge CS OC MOV *-880B CLOSED CNTMT Spray PP Discharge CS OC
• -891A LO CNTMT Spray PP Dwischarge CS OC
• -8918 L0 CNTMT Spray PP Discharge CABLE CONTAINMENT PENETRATION PRESSURE (CP)

ROOM (CPR) CP CPR

• -11-2054 LO PT-63068, PT-64258, PS-2008,PS-2057 CP CPR
• -11-2055 OPEN PT60368, PT-64258 CP CPR,
• -11-2057 OPEN PT-64258 CP CPR
• -11-2059 L0 PS-2009 PS-2058 CP CPR
• -11-2063 L0 PT-6306A, PT-6425A, PS-2007, PS-2056 CP CPR
• -11-2061 OPEN PT-6306A, PT6425A CP CPR
• -11-2065 OPEN PT-6425A 8

l l Ro 9.g

OPERATING PROCEDURE 0202.2, PAGE 35 UNIT STARTUP - HOT STAND 8Y TO POWER OPERATION TABLE 1_ (Cont'd) LOC-VALVE / POSITION / IITIALS SYSTEM ATION I!DICATOR FLOW OESCRIPTION CCW Cont. Rm. FI *-1470 > 100 gpm CCW To Emer. Coolers CCW Cont. Rm. FI *-1471 > 200 gpm CCW To Emer. Coolers CCW Cont. Rm. FI *-613A > 100 gpm CCW Header "A" CCW Cont. Rm. FI *-6139 > 100 gpm CCW Header "B" CCW Local FIC *-637 9 gpm '8" RHR Pp. Seals CCW Local FIC *-638 9 gpm "A" RHR Pp. Seals CCW Local FIC *-657 > 10 gpm "A" CSP Seals CCW Local FIC *-659 > 10 gpm "B" C3P Seals CCW Local FIC *-658A > 30 gpm SI Pp. Seals CCW Local

• -701X OPEN RT Viv. to PI 612 CCW Local
• -7040 OPEN RT Viv. to PC 611 and PI 6:2 CCW Local
• -701G OPEN RT Vly. to PC 611 CCW S.F.P.
• -708A OPEN CCW Surge Tk Level LT-614 CCW S.F.P.
• -7088 OPEN CCW Surge Tk level LT-614 CCW CCW Pp Rm
• -714A OPEN Ndr. Flow FT-613A CCW.

CCW Pp Rm

• -7148 OPEN Ndr. Flow FT-613A CCW CCW Pp Rm
• -714C OPEN Hdr. Flow FT-6138 CCW CCW Pp Rm
• -7140 OPEN Mdr. Flow FT-6138 CCW Below P/V Rm.
• -10-738 OPEN FT-1464 Emer. Cooler Flow CCW Below P/V Rm.
• -10-739 OPEN FT-1464 Emer. Cooler Flow CCW Below P/V Rs.
• -10-740 OPEN FT-1465 Emer. Cooler Flow CCW Below P/V Rs.
• -10-741 OPEN FT-1465 Emer. Cooler Flow CCW P/V Rs.

MOV *-749A CLOSED CCW From RHR Hx CCW P/V Rs. MOV *-7498 CLOSED CCW From RHR Hx CCW Local

• -746A OPEN CCW To RHR Hx CCW Local
• -7468 OPEN CCW To RHR Hx CCW P/V Rs.
• -748A LO 305 CCW From RHR Hx CCW P/V Rm.
• -7488 L0 305 CCW From RHR Hx e

QC 4X

UNIT ST kYb b Hbf hfkNh8h h5 PhWER hPERATION TA8LE1(Cont'd) LOC-VALVE / POSITION / .TIALS SYSTEM ATI'ON INDICATOR FLOW DESCRIPTION ICW CCW Pp. Rm.

• -358 OPEN FI-1409 Flow through CCW ~Hz ICW CCW Pp. Rm.
• -359 OPEN F1-1409 Flow through CCW Mx ICW CCW Pp. Rm.
• -368 OPEN F1-1408 Flow through CCW Mx ICW CCW Pp. Rm.
• -369 OPEN FI-1408 Flow through CCW Mx ICW CCW Pp. Rm.
• -378 OPEN FI-1407 Flow through CCW Hx ICW CCW Pp. Rm.
• -379 OPEN FI-1407 Flow through CCW Hx ICW CCW Pp. Rm.

FI *-1407 TOTAL ICW Flow through CCW Hx ICW CCW Pp. Rm. FI *-1408 FLOW ICW Flow through CCW Hx ICW CCW Pp. Rm. FI *-1409 >6000 gpm ICW Flow through CCW Hx NOTE 1: HCV *-758 OPEN with air supply Locked CLOSED. NOTE 2: Some flow indicators are in-line and have no instrument root valves. Once system is lined up by procedure, they will automatically readout. (i.e., FIC-658A and 8) m i I { d.l(, __-,-.,..._.,__,.,,-.._,_..m. -__.___,,,_m-- ,m, r- _-,.-,m._-,,e

09/11/86 Page 24 NRC Exam Section 4 QUESTION RO 4.19:(SRO 7.20) List all the valves which reposition as a result of a Phase B ("P") isolation signal.

RESPONSE

We request that the answer key be expanded to include " Containment Spray Pump Discharge Valves"

REFERENCES:

~ Drawing # 5610-T-E-4510 Sh.1 of 2 I

• RI-5/cas

1 l i ~ I+ -it.'.f I..=+n E=Q=Yg.E.=,l . - '1%. n:m.,,.r,:N.< .. I iM *J SiTA's"6. u@ h.-;,q. - ,;;;.;;, O t-s:

• I I E,% !

[1 IN b; G l'* D. = a=IM 3-p.,.. y-.

==i-.-- = H i..i ]a. '. $ ,n;=- i i-- e eu,i4 y,' l. T t = = =.. g =g -'T,a ', m.i.a 4 I c z. =w =, +. .as. $n .g..t._. i :.e-. e I t i.,,. . = = =. =ei u .-. ue,.. o g.,.i, i 1, -.g a.d ({.,t a. 3. ~ :. 4... N, u: , - !- *. * *1" l j e,,, l p i 4. i. u= ~ 2 - 3. a. o$ "$, in I.'.'. ,.. E. I-= !.. ::t_.!8 / l ,,. a.s m ;; u .. I 4,I. g,'. i [ aMpha.,,.j i f.'.*~ = !*'.Yf -_i. i,,,, ...x,.@ c =. %. I, e,) . 'It t'.. r g g.

g..

1 '= .a u s I@ %.@.--y. e ..i 1 v. . ;.t. i ,.=. u X 3.,.u m; r 3 l = - n.e -u.ug m = ~ - """3 ~[ f-d 5--@ -6@t @-2 .m. ^a'i M '.'.-=.m' -I., h ,,,,,1 ~ 'O i. u-. %.'li ~ a.T w tar mmW8r;;- o "'E I l trit. f**! ? sr= s- = T.*"!S "::T=-- - j'r i.. } l' grJR:r, j

• • 'b"4 -. yea >iq'y.w l

l L-a sF5 o s a t ~ - O, c..- ? "P i.-an -" ~ _ -L-- @., m. sa:..,_ ie . L' 's,. l u .s*

m..

u== ,u;.a..C @4g.E.' S1 gM

n~=

W.J 1 ,; : N.M*

== p'+-=.=,=.iA Ds,M,' $ t.2.-.- q o 9 = l f.seessi.rw<w ' I ",,='st"" ~ = l .. f n W .= m-,@a *:n i k. I- = ~n ,,z n .s ;.-..v..., l-.. am@e

n 2

z.,ar. .L ~~-- d. ,f

r. g-c.

i

c. u.+ t.-}-

.:,:-. -l,'; ~.- .g e ,,,,, f-r i -a one. 4 u .= p. ~,

== ~ Qn "" ',y, g-f. $"A, I R' ,] e.- ~~ c-rM -s + s. - - ~ - ' .~ T -..w M RO

==Mm? 2.J{,n.M. e,e,,, I i e ch as.La " ' =

i 09/11/86 Page 25 NRC Exam Section 4 QUESTION RO 4.20:(SRO 7.21) With rod control in automatic while at 80% power, what are all the immediate opeiator actions if a single rod drops (assume no rx trip) as stated in ONOP 1608.1," Full Length RCC Malfunction"?

RESPONSE

We request that the answer key be expanded to include " Verify Turbine Runback" as per ONOP 028.

REFERENCES:

ONOP-028, step 4.5.2,03/28/86

• RI-5/cas

4F LQ9toselonts. 4,0NOP.028 Recctor Centrol System Malfunctlan 3/28/84 4.0 IMMEDIATE ACTIONS 4.1 RCC Misalignment 4.1.1 Place the Rod Motion Control Selector to the MAN position. 4.1.2 Maintain steady sta'.e conditions as follows: 1. Borate! dilute AND/OR change turbine load to maintain Tavg equal to Tref. 2. If possible, avoid insertion of the control rods. 3. DO NOT withdraw any control bank until the RCC(s) have been aligned. 4.2 Immovable RCC 4.2.1 Place the Rod Motion Control Eelector to the MAN position. ~ 4.2.2 Maintain steady state conditions as follows: 1. Borate / dilute AND/OR change turbine' load to maintain Tavg equal to Tref. 2. If possible, avoi.d insertion of the control rods. 3. DO NOT withdraw any control banks until4.he RCC(s) have been aligned. 4.3 Failure of an RCC Control Bank to Insert with Reactor Controlin Automatic 4.3.1 Place the Rod Motion Control Selector switch to the MAN position. 4.3.2 Manually position the RCC control bank to restore steady state conditions. l_.the RCC control bank will still not move THEN go to step 4.2. F 1. 4.4 RCC Position Indication Malfunction 4.4.1 IF the RCC position indication malfunction caused a turbine runback,THEN go to 4.ONOP-089, Turbine Runback. 4.5 Dropped RCC 4.5.1 IF_ three or more rods have dropped,THEN manually trip the reactor and turbine. IDELETEDI 4.5.2 IF one or two rods have drop xd.THEN verify turbine runback and concurrently perform 4-ONO?-089, Turbine Runback. ~ ffo ti. 26 ~~

09/11/86 Page 26 NRC Exam Section 4 QUESTION RO 4.24:(SRO 7.25, SRO Requal 4.17, RO Requal4.17) Define "Inside the Target Band" in accordance with ONOP-1008.6, ' Operations outside the Target Band'.

RESPONSE

Expand Answer Key to include "i S% of Target"

REFERENCES:

ONOP 1008.6 Pg.1 Sect. 3.2.2,01/29/82 e l l

• RI-5/cas w.,

. - - ~..

I I I' FLORIDA POWER AND LIGHT COMPANY TURKEY POINT UNITS 3 Am 4 0FF-NORMAL OPERATING PROCEDURE 1008.6 JANUARY 29. 1982 1.0 T1,tl 3,: OPERATION OUTSIDE THE AXIAL FLUX DIFFERENCE TARGET BAND $5 Approval and List of Effective Pages:

2.1 Aporoval_

Change Dated 1/29/82 Reviewed by Plant Nuclear Safety Comittee: 82-09 and Approved by Plant Manager - Nuclear: 1/29/82 2.2 List of Effective Pages: _ Pag, Date Page Date h Data Page Date a , '.. i ' c. ,1 1/29/82 2 1/29/82 3 1/29/82 4 1/29/82 - A '. 0 Purpose and Discussion: s c.

3.1 Purpose

v Qpg *:._wv ~s @w - -1his procedure lists actions to be taken if the reactor is outside the axial . '.TY ' - flux difference target band. .m.; _. ;;. w-- n ~. . ~. - - ' '" ~ Discussion: T~ ~ ~ ~. C.,., 3. 2 ., ' 3.2.1 If the indicated axial flux difference at a power level greater than p!~lO.P J,', l ?Q ?' 905 of rated power deviates from its target band, Technical j l T r'.1. N M ".'- " Specifications require that either the deviation be eliminated W ]Q[IF immediately or the reactor power be reduced to a level no greater.. , pg*II-{, than 905 of rated power. 4 l % 4.* u. I 3.2.2 A' a power level no greater than 905 of rated power, Technical Specifiations allow the indicated axial flux difference to deviate 5ce from its tS5 target band for a maximum of 60 effective minutes ,',q$P.l.,- (cumulative) in any 24 hour period provided the flux difference does -;-lpg -A not exceed an envelope bounded by -115 and +115 (of the target flux) g,y,;./,Iy.:l, at 905 of rated powsr and increasing by -15 and +15 for each 25 of l c l , 3..- rated power below 905 rated power. Effective time out of the target band is defined as the sum of the time out of the target band at power level above 505 plus one half the time out of the target hand ~ at power of 505 and below. l 3.2.3 At a power level no greater than 50% of rated power Technical !g7 Specifications allows the indicated axial flux difference to deviate 'q from target band. l i 70 9 t4

tyt o

• ycna L

~< ~s a - N NCN-Gat.iAT10NAL N REGION 73 y4g ONE SfX.R '-'~ REGION 11 T ARGET 9 ANO -. q I _._.11 ~ ~ ~ - - - _ _.to / / N i N 19 S.4r ~ G N .h. N .to i i i u. I T ARGET 9 ANO q I i I 5- -11 i g / in i i i f 21 h f I / N i .- N +19 a y.43: l i N i N .1q l i e i h: TAMUL15%N> 9 5 i i i I I I I -19 21 5I l l 1 l / l 1 -- i l I i l i / -21 i i X 3 l 6 I + / l i ! l N i i i A i +11 i N ^t; i I 1 l I N N +17 ~- i i l i M*

• I '3 " >-

i 9 l i i -!9 8 l C / I i / -11 i g / l f i t ~ I 10 40 50 60 70 90 90 111 [ cgpaEDDOCUMENT % REACTOR POWER R15:1

8. n k. ; u c

l l e

09/11/86 Page 27 NRC Exam Section 7 QUESTION SRO 7.19 Besides required notifications, what are the immediate operator actions if you are on shift in the control room and the refueling supervisor in the containment reports they dropped a spent fuel element in containment?

RESPONSE

Request that the Answer Key be expanded to accept the following as an additional answer:

1. Sound ContainmentEvacuation Alarm

2. Initiate Containment Ventilation Isolation

REFERENCES:

E-0, Page 6, Step 12 i i l

• R1-5/cas

OULLL .~.- 3.EO P.E.0 REACTORTRIP OR SAFETY INJECTION 8/15,86 1 STEP l l ACTION / EXPECTED RESPONSE l RESPONSE NOT O8TAINED l C A U TIO N Only one (1) CCW pump A or C and one (1) ICW pump 8 or C e should be powered from a single 4KV bus when the bus is powered from an EDG. In the event one CCW heat exchanger is out of service, one CCW e pump should be placed in the pull to.fo<k position. 9 Verify 2 of 3 CCW Pumps - RUNNING: Manually start pumps. Verify RCP seal cooling water outlet

a. OPEN FCV 3 626 a.

FCV 3 626 is OPEN I jQ Verify 2 of 3 Intake Cooling Water Pumps - Manually start pumps. RUNNING

a. IF only one pump is available -

THEN Dispatch personnel to isolate CV 3-2201 TPCW Hx Outlet Control Valve by dosng the following manual isolation valves:

1) 3 50-401

2) 3 50-403 11 verify : of 3 Emers.ncy Containment -

Manoaily start 2 coerers and fiiter fans. Cooler And Filter Fans RUNNING 12 verify containment ventilatioa isolatica: a. Containment purge and supply fans - a. Manually stop fans OFF

b. Purgevalves CLOSED

b. Manually close purge valves.

c. Instrument air bleed valves-CLOSED c. Manually close valves.

d. Verify control room ventilation

d. Place control room ventilation in isolation emergency recirculation mode.

f 13 Che* if usia steamlines should se isolated: a. Go to Step 14. If, the following status lights on VP8 a. are ON

1) High steam flow

2) Low Tavg OR low 5/G pressure 9.6

3) Hi Hi containment pressure (3

THEN

b. Verifymainsteamisolation

b. Manually close valves.

emtannms -r ,r

09/11/86 Pcge 28 NRC Exam Section 8 QUESTION SRO S.22(SRO Requal 8.15) List the MINIMUM requirements for each of the necessary electrical systems listed below in order to conduct a reactor startup on UNIT 3, assuming Unit 4 is shutdown. a) Unit 3 and 4 4160 KV buses b) Unit 3 480 VAC Load Centers, c) Battery Chargers

RESPONSE

We request that the Answer Key be expanded to accept "4 of 4 4160 Volt Buses" as an additional answer to meet the requirements ofTech Spec 3.4.1.A.4

REFERENCES:

Tech Spec, Page 3.4-1 4-OP-062, Pages 4 and 24 l

• R1-5/cas

,,----,c-.,

!CI;F.IRO SAICT TMW3 ).& Applies to the operattat status of the Easincered Safety 3eolicability: Features. To define those limits 3 :or.ditions for operation that giective: to et=ove decay heat f rom the core are nettssary: (1) in energency or normal shutdown situations. (2) 50 re-is normal operating and move heat from containcent to remote airborne iodine amergency situations, and (3) from the containeent at=esphere in the avant of a Maximum Hypothetical Accident. $AFETT INJECTION R 3 MS DUAL 8. EAT RE40 VAL S 1. Scacif' cation: The reactor shall not be made critical, 4Ecept for a. low power physics tests, unless the following conditions are =et: The refueling ta er tank shall contain not less 4 1. than 320,000 gal. of water '.eith a boron con-centration :f at tagst 1950 pp=. The boron injection tanir, shall contain not less 2. I than 900 gel. of a 20,000 to 12.500 ppe boros "he selution in the tank, and in solution. isolated porti:ns of the inlet and outlet piping.-shall he saistained at a camperature of at least 1&$F. TWO channels of heat tracing shall be operable for the flow path. Each accumulat r shall be pressurized to at 3. least 600 psig and contain 875-891 fe of lassc veter with a bar:n concancration of at 1950 ppe, and shall not be isolated. TOUR saf ety inf ection pur.ps shall be operable. 4

• See reference (11) on Page 53.4-2 Amendments 78 & 72 3A-1 e

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

r era.eure.o.

• ee.:

4 4 OP.062 Sity Inbetica 01M I .0 PURPOSE This procedure provides the prerequisites, precaution / limitations and instructional guidance for abgning the Safety Injection System valves, breakers and switches prior to a unit startup.

2.0 REFERENCES

2.1 References 2.1.1 Technical Specifications Section 3.4.1, Safety Injection and Residual Heat RemovalSystems 2.1.2 FSAR Section 6.2, Safety Injection System 2.1.3 Operating. Diagrams 1. 5610-T-E-4510, Sh 1 - SI and RHR System Outside i Containment 2. 5610-T-E-4510, Sh 2 - SI and RHR System Inside Containment 2.2 Records Required 2.2.1 The date, time and section started and the date, time and section completed shall be logged in the Reactor Control Operator (RCO) logbook. Also, any problems encountered while performing the procedure should be logged (i.e., malfunctioning equipment, delays due to changes in plant conditions, etc.). Com'hment($) constitute Quality Assurance records and, ther leted 'es of the below listed section(s), enclosure (s) and/or 2.2.2 attac shall be transmitted to Document Control and be retained for a minimum of 5 years in accordance with Quality Assurance records requirements: 1. Section 5.0 2. Attmehmenti

3.

4.

5.

6.

2.2.3 A copy of the completed procedure shall be filed and maintained by the Plant Supervisor - Nuclear until the next performance of the procedure. ,,.v -.- - =, -,-- -.-- ~- -' ' ' - ' ~ ' " ' '

• m: 24 n=wwa r m

... e 'TOtNhl Safety injection 4.OP 062 ATTACHMENT 5 (Page 2 of 2) SI BREAKER ALIGNMENT Comgonent Component' Description f,0'",*[ N',f Iy((,','[ g RACnD 3AA13 SIPump 3A g RACMD 3AB12 SIPump 3B g MCMD 4AA13 SIPump4A g MCED 4AB12 SIPump 4B g OFF 40712 MOV-3-864A OFF 40605 MOV-3-864B f ON 0831 MOV-878A ON 40708 MOV-878B OFF 40737 MOV4869 OFF 40738 MOV-4-843A OFF - 40622 MOV4843B OFF 40732 MOV4866A OFF 40621 MOV4866B r ~ FINAL PAGE l } NPMM J T

ADDITIONAL TECHNICAL ISSUES SEPTEMBER 19,1986 l / l t A I i l } } 1 l

TESTITEM ANALYSIS SHEET Question # / Point Value: RO 1.10 (SRO S.071/1.5 pts. Question: A reactor which has been operating at full power tEOLi trips due to a Loss of All OITsite Power. Natural circulation conditions have been established and steam d. ump to conden.<er is operational. Steam Generator levels / pressures are at no-load stable values. Core decay heat generation however, is exceeding the free convection heat transfer rate capability. EXPLAIN the thermodynauie principles which will cause fuel element temperatures to eventually reduce to lower values. (assume a relatively constant heat generation rate within the fuel) Answer: Since free convection heat transfer rate under subcooled conditions increases as a function of the available temperature difference between wall and bulk Guid (delta T) [0.25l the heat transfer rate to the Rx coolant in the core area will increase [0.25] causing an increase in core exit temperature [0.251 increasing the available natural circulation driving head [0.25) which will cause an increase in mass now [0.25] thereby increasing the convective heat transfer capability of the RCS. The higher mass Gow will also cause [0.25] a corrective decrease in fuel element temperature.

Reference:

TPT CNTO Volli 14-26 K/A EK 1.013.7/4.2 Inaccurate information stated in stem orquestion. l Technical Review: 1) Steam dump to condenser can not be operational during loss of all offsite power. CW pumps will trip, condenser vacuum will decrease, locking out the condenser steam l l dumps. 2) Increasing steaming rate will also increase driving head. 3) liigher mass now resulting solely from increased coolant temp will only stabilize fuel tem'peratures, not cause a " corrective decrease" Construction Review: No comments Relevancy Review: No comments I K&A Review: Insufficient K/A reference 1 i Recommendation: Delete question. l l RUll 10

• RI.id.4p 09/IS!S6

TESTITEM ANALYSISSHEET Question # / Point Value: RO 2.16/ 0.50 pts. Question: Refer to figure #401 Charging Pumps. STATE the forcets) which maintain seal water flow through the packing of the charging pumps. Answer: (thermally induced) natural circulation and head of water form seal tank [0.25 eachl

Reference:

TPT SD 12 Terhnical Review: Seal water flow through the oacking of the pump is maint~ained by the static head of the head tank and has nothing to do with thermally induced natural circulation. Construction Review: Answer not relevant to question; answer key does not answer question. ES-107 C1, (pg. I of 3), Wood (pg. 46), Thorndike & Hagen, (pg 79), ES-202 E10,(pg. 4 of 6). (FIUn Relevancy Review: No comments. t K&A Review: No K/A reference i Recommendation: Accept seal water static head as answer.

• RI-5/dsp.09.1& S6 Rail.n

TESTITEM ANALYSIS SHEET Question #I Point Value: RO t.13 cSRO 5.106 /1.5 pts. Question: Refe'r to figure e J57.*Tavg vs. Pow er* and answer the following: On the craph curve 'A' bounds the maumum operating loop a. temperature at a given system pressure for a particular reutter power. Esplain the basis for graph *A*and w hy it has a negative slope. b. Graph *B* ens 2res that two adverse conditions are not exceeded. State these two conditions. Explain why curve 'C' is valid at <l5T quality and not at.>15+ c. quality. Ariswer: a. This curve is based upon preventing hot leg temperature from reaching a saturation temperature as power increases.the difTerence between Th & Te increases. (0.251 In or der to keep Th below saturation. Tavg must decrease with power (0.251. b. To maintain minimum DNBR10.25l Ensures that the avg. enthalpy at the vessel exit does not exceed the enthalpy of saturated liquid. (0.251 At higher pressure. a higher Tavg is necessary for 15% quality [.251 at c. low er pressures DNS occurs at t 5%. At higher pressures, the quality of water may be >20% before DNFloccurs.10.251

Reference:

cNTo votir i3 53 Technical Review: Ro i.t3c: The answer key does not address the question. Curve *c"is valid at < t 5% quality and not at > t 5% quality because the WRB. I correlation used to predict DNB only applies to fluid at less than 15% quality. Above 15% quality.the WRB.1 correlation is not valid. l Construction Review: Part "a' is misleading in that it refers to

• maximum operating loop temperature" but s:raph iabet is Tm vs. pow er.

Part "c' Figure doesn't apply to PTN < ES-201 I spg 5 of 8P. Westinghouse told trainees concept has no applicability to PTN; double jeopardy question ES-202 E l3.f pg 4 of 61. Thorndike and Hagen t pg 96l. Relevancy Revlew: Not relevant for Ro candidate. NRc Examiner standard ES.202. Scope of Written Examinations Administered to Reactor Operators. Power Reactors *. Rev.1.10/1/84. Part E.1 states that Tech Spec questions for reactor operators should be conceptual in nature, which does not include specific basis explanation. Page 3 of 6.ES-2021. K&A Review: No lUA reference Expand Answer Key (per previous submittal) or Recommendation: Delete entire question rrom Ro esams. Delete part "c' from SRo exam. I 1 l l

• Ria d,p.t19/l&S6 Roll 13

__. _ _. _ _}}