ML19220B478

Examination Rept on Senior Operator License Issued for TMI-2.Written Examination Encl.Operating Test Administered 770914

ML19220B478
Person / Time
Site:	Crane
Issue date:	10/17/1977
From:	Collins P Office of Nuclear Reactor Regulation
To:	METROPOLITAN EDISON CO.
References
NUDOCS 7904260295
Download: ML19220B478 (48)

Text

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i e ] e E 1 ^ OPER ATING AND ORAL EXAMINATION SUMM ARY REPORT EvALuATLCN chcan t's

• • "GL 19 an SRO aconcant is unsatisfxt::ry at ene SRO Level, comotere cne RO column to evaivate ac NUVOER FC9 SRO RO aaequacy at the RC Level.
• For each unsattsfactory ("U"). list t!ae page number (s) of tree coerating oral enamonation notes on shoch th "U"

unsatisfactory responses are emotained.

1. CPERATit4G DEMONSTR ATICN 5

Prestar*up and instramer1 Checks 1.1 12 Console Coeratson xlQ

f., j, f [,,,,

V f

a. Manipulations

b. Undentanding

2. FACILITY ECUIPstENT 5

a. Major b

b. Auxiliary Engineered Saieguards Svstems c.

3. INSTRUMENTATICN 5

a. Nuclear b

13. Process b

4. REACTOR PROTECBON f.

PROCEDURES 5 a. Normal 5

b. Abnorma4 S

c. Emegency

) REACTIVITY EFFECTS (Except C:nscle Operation) si 5

7. ACMINISTRATIVE RECUlREMENTS h

8. EMERGENCY PLAN

)

9. RAD. ATICN PROTECTION

)

10. RESPONSIS!LlTIES CONtM E N TS V :D A *34 e a)

GA psg. : NRC FORM Yb7 lifL 7%) O

i ^ s ~ U. S. NUCLEAR REGULATCRY CCMMISSICN SEN?0R REACTOR OPERATOR LICENSE EXAMINATION Three Mile Island - Unit 2 Facility: Reactor Type: 'an f"a Cate Administered: c/ /b / e ' Examiner: J. J. cuzy Appli c an t' INSTRUCTICNS TO APPLICANT: Use separate paper for the answers. Staple question sheet or, top of the answer sheets. Points for each questien are indicated in parenthesis after tne question. Category

of Applicant's of Value Total Score Cat. Value Ca teco ry la 20 k!.3 kk H.

Raactor Theory 14 20 O.( (Th I. Radioactive Materials Handling, Disposal and Hazards 14 20 - J. Specific Cperating Characteristics il 7 ]G "/ K. Fuel Handling & Core Parameters la 29 C1 \\ 3, O l t. Aeministrative proceeures, Con _ 14 _20 _ ditions and Limitations Totals 7n Final Grade: 7.7 122 O ^ S

] m 7 H. REACTCR THECRY (14) H.

i. Two minutes af ter a reactor trip from full pcwer, a large number of neutrons are still present in the core.

List these neutrens and give their origin. (l.5) H. 2. Assume your reactor is on a 1 DPM startup rate at 5 X 10-9, group 6/7 is at 15% withdrawn. a) Whatshoulg?be the rod position when the acerator levels off at 10-Show all work, include any assumptions. (2.0) b) With maximum rod speed, how long should the operator move the rods to level off at 10-C? Explain how you arrived at your answer. (1.5) / c) If pcwer were allowed to increase, hcw far do ycu excect (1.0) the power level to rise? Include any assumptions. d) Will the same reactivity addition produce the same S.U.R. during core lifetime? Explain. (1.5) H. 3. Refer to figure 28. Control rod groups 5-7 reactivity worth at 5320F. a) Sketch the apprcximate reactivity worth of group 5-7 when the temperature is 5820F. (0,5) b) Expla'a why the curve will change. (2.0) H. 4 Refer to question K3. Assume tha individual responsible for taking the count rate reads 4.nd records the data immediately after the fuel element is inserted into the core. Would the resulting plot be more or less consenative? Explain ycur answer. (2.0) s H. 5. One of the limiting conditions for oceration at your station is the value of the moderator temperature coefficient. Is this value of concern early or at the end of core lifetime? Explain t'e reason (s) for your answer. \\. (2.0) r m" ( uG.-.- l *.' ra ,e-- wwww-g

1 5 ^ 1. FAD *CACT!VE MATER *ALS 5;NCLING. DIS CSAL AND -AZAP.DS (ia 0) I. 1. Pericdically, during ?:wer ::aration, it is necessary to purge tne reactor building. Describe the procedures and limitations that you, as shift supervisor, are responsible for during the purge.

nc'ude the folicwing:

l a) Administrative repuirements, includire individuals who I approve the release. ~ (1.0) b) What monitors are required to be in service? (1.0) c) Why is it necessary for the fire protection system to be Operational during the release? (0.7) d) Maximum react:r building pressure Orior :: release. -- ~ (0.3) I. 2. A smear is taken en a calibra:ed pressure gauge which was used in the waste disposal area. 1 Smear Area - 100 CM" Eackground - 15 CPM 5 minute count - 900 Scaler efficiency - 15: a) Determine if the gauge can be re:urned to the instrument sn00. l.3 b) Discuss how you wculd take a smear (s) for this gauge. (1.0) I. 3. Assume a reactor c0clant leak develops in the react:r building. a) What are licits on react:r coolant system leakage as a, function of locaticn or' identity of tne leakage? 2T (l.5) .,n J b) Assume the leakage collects in tht: reactor :uilding sump. Describe in secuence what ha::en:, Oc he water frca the (1.5) sump to its uldicate disposal. c) Frior to enterinc :ne reac r building :o search for the leak, what instailed conitoring system (s) can be used by p' the inspection team to evaluate One levels of radiation? (1.5) I. 4 'n opefa:Or' has dropped a set of keys en t:0 of a s;ent fuel element.in the storage ccol. He wants to fabricate a tool ti made of ten fcct long' one inch diameter sections of electrical \\ conduit joined by enuciings. In additicn, he has had a hook welded to a ipe cap to join with the end Of the c:nduit sections wnica will be used :: cick u: the key ring. Shc.id you alicw nim

assemole One ::ci and use it? Explain wnv cr whv r.c.

(2.0) Section : c:n:inued next : age i e ~.. A 4

2 2, ~ 3 ^ ^ s . 2_ I. 5. The maximum activity c::ntained in one cas decay tank is limited to Trv>. a) What is the limit and who is responsible to assure this limit is not exceeded? (1.0) b) What is the reascn for this limit? (1.0) k I 4 #? 7}' A 4.' < 0 */

2 o j ~ O J. SPECIFIC CPERATI"G CHARACT5RISTICS (14) J.

1.. Lithium hydrcxide and hydruiFie are

.vo chemicals that are .,Jadded to the reactor coolant system. / a) Discuss when it may be necessary to add these chemicals. (1.5) b) '4 hat precautions must be observed when handling these chemicals? (1.0) J. 2. During the unit heatup, one of the steps requires a hard bubble degasificatien when the ccolant temperature reaches 3500F. a) Cutline how this step is performed. (1.0) b) 'What wculd be the ::nsecuences if this step was not performed? Assume :cwer operations with steady state and transient conditions. (1.5) J. ', 3. About two years ago, TMI Unit Cne went critical during a plant heatup, while the RC temperature was 5050F. a) What procedural steps were modified to prevent a repeat incident? - (1.5) b) What is the L.C.0. for minimum R.C. temperature for e r_ c criticality and what is the basis for this L.C.0.? a, (1.5) J. 4 When operating from 50 to ICC" rated :cwer with 4 RC ::ceps running, the staff is required to maintain cold leg differential temperatures to less than 50E. a) Mcw does the operating staff ccmply with this limitation? Include controls the staff must manipulate. (1.5) m N b) Why is this conditicn recuired? (1.0) J. 5. Assume a reacter startup is in prcgress and within 5" of rod positicn for criticality. At the same time, an I.C.S. representative is " fine tuning" the control system.

Suddenly, all turbine bypass valves go full valves open. Assume (i) one main feed pump in service and (ii) R.C. temperature decreases at 30cF/ min, a) What protective device (s) shculd respond to trip the reactor if no action is taken?

.,,0..,, (.0. 3 ) b) What imediate acticn is required by the cperating staff? (0.7) Section J continuec on next cage qy 4.. y>. .t l'

s ~.. -. 3 1 _~- J. 5. c) Assume no acticn by tne cceratinc staff. '4 hat systems will respond if :ne turbine by-pass valves cannot be snut. Attempt to shcw if the plant systems w1ii cF will not terminate this transient without opera:ce action. (2.5) ti =. a. g' y

m j K. FUEL HANDLING & CCRE PARAMETERS (14) K. l. Listed below are several fuel hancling devices and maneuvers. List and explain the ::urocse of the interlock (s) during eacn maneuver. a) Control red handling mechanism while inserting and withdrawing control reds. (1.0) b) Rotaticn of the fuel transfer baskets and the main fuel (1.0) or spent fuel storage handling bridge. -g c) Fuel grapple mechanism and hoists when inserting or removing fuel in the care. (1.0) K. 2. Assume a malfunction en one of the safety reds in group 2 -esults in a decpped red. The reactor is at rated pcwer and the reacter demand station is in manual. a) What immediate acticn is required by the operator? (1.0) b) Why is the abcve action necessary? G' (1.0) c) Assume trouble shooting cannot begin f.or several hcurs, would the axial pcwer imbalance and/or the quadrant pcwer. - - tilt beccme the limiting condition for operation? Explain ' 'i your answer. (1.5) K. 3. Initial' fuel laading at most PWRs develops as a " slab". The source is located in the center and the fuel is leaded outward (see figure K-3). Temporary chambers (A&B) are located in the vessel and subcritical multiplication plots are made for each section of the slab loading.

3. - -

a) Sketch the expected plot of subcritical multiplication for each chamber. (1.5) b) Explain any difference in the plots. (2.0) K. 4. During the in'tial pcwer escallation and immediately following a control rod interchange, the rate of power increase is limited to 3"/hr. between 75 to 90". pcwer; also, a 5 hour hold is required at 90". power. a) What are t',e nor al maneuvering rates between 75 to 90". pcwer? Provide transient and centrolled conditions. (0,5) b) Why is it necessary to perform a control rod interchange and why are power increases reduced to 3"./hr? (2. 0 ) K. 5. At: ached is figure lA, Core Excess Reactivity, for your facility. Explain the shape of the curve during the first ICO EFp0. (1.5) $ $ I.& ~

2 j m 2 L. ACMINISTRATIVE PROCEDURES, CONDITIONS AND LIMITATICNS (14) L.

1. If any safety limit is exceeded, the shif t supervisor is required to notify the station / unit sucerintendent and have the reacter in hot standby conditions within ene hour.

a) What parameters and limits are considered in the safety limits for your station? Wr 4 n -/ " -

.us (1.5) b) What instrumentation in the reactor protecting system should prevent reaching a safety limit? <ar p er_-

(1.5) 3,y (0.5) c) What are the plant conditions for hot standby? L.

2. Title 10 CFR Part 55, Operators Licenses, defines an operator as any individual who manipulates a control of a facility, a) Define centrol as it applies to this definition and give two examples.

(1.0) b) Is it necessary to be a licensed operator to manipulate a control? Explain. (0.5) c) Assume an operator was injured during a. motorcycle accicent. He is disabled for 5 months. When can he return to normal (1.0) a. o ! ';. - o duties? u n' e.- L.

3. When a shift is relieved, a number of items must be completed.

a) What shift records are maintained? (1.5) b) What conditions are required in order for an entire shift to be relieved? (1.0) c) List the shift organization required by the technical specifications during (i) fuel leading and (ii) plant, v.- " ",v (1.5) ~ < - startup. ,c L.

4. Assume you are the shift supervisor in charge of your shift.

For each condition belcw provide (i) i=ediate and subsequent action and (ii) the reason for your action, including any administrative limits. a) The unit is shutdcwn for OTSG tube inspection. You are informed that the sodium hydroxide tank was inadvertantly lined up to the decay heac pumps. About half of the tank volume is left. (1.5) b) The plant is at rated pcwer during the winter. The auxiliary operaccr infor,s ycu that the vent (s) on the 3WST tank seem to have frozen due to ccndensaticn and ice for ation. (1.0) Section L continued :n next : age 7.3 NO 3

3 g m 2 .~ =. L. 4. c) Curing a test of the containment isolation valves, cne of the valves begins to leak at :ne packing gland. The test engineer suggests he can "back seat" the valve by holding in the open relay with an insulated scree, driver at the m.ctor control center. (1.5) '- ^ '.a, A.ltnU s 5

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_______.,3 hble 1 }C TE**A!iECUS 10 C?.9 20 Ap;endix 3 j 7-Table 1 li Tab'.* !! e Er.er y Col 1 C 1 II e Col 1 Cal II e FEV ;er ALr 'dat er l Air Vater v terial Hair-u re disinter:2:i:n se/c1 ue/:1 ue/-S m:/:1 o -6 -S A-41 1.83 h 13 2x10 9 Ex10. 4 -5 Co-60 53 y 25 3x10-7 lx10-3 1xlo 5xta I-131 8.1 d 0 36 9x10' 6x10-5 1x10-10 3x10-T r.r-85 l 10.8 y 0.Ch 1x10-5 3x10-7 -k Ni-o5 25 h 0 59 9x10 4x10-3 3x10 d -7 lx10 4 Pu-239 2.kx10 y 0.cc8 2x10-2 1x10" l 6x10* l 5x10 -5 l -11 1 -6 -9 i Sr-90 23 y lx10 l 1x10 3x10 3x10 -o Xe-135 92 h 0.25 hx10 1x10-' Any single radienuclide with T > 2 hr 3x10 l 9x10-5 -10 -6 -9 1x10 3x10 g which does not decay by a cr sp:ntaneous I fissionj Neutrens per c d-Average flux to deliver Neutren EnerivDTV) ecuivalent to 1 re 100 ::re in 10 uurs 6 Thermal

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a 7 ), l 4_ 9 U. 5. ATCMI' ENIRGY CC'O'.!SSION m "?ATION RE?CRT Applicant License ap liac :or: RO M SRO / / Reactor Nuo kjj b ki b d. Location L. Currently a licensed operator here? Yes / / No @ VRITTEN EXAMINATION: Operster (Section 55 21): Administered by Date 11/ f Waived / / Graded by 1). Grade h3 AS) :, 32 %, Cf,A", DF1, EX :, 73 7 :, cM:. Category Grades: 's Senior (Section 55.22): Administered by Date Vaived / / Graded by Grada Category Grades: H I, I %, J I, I

, L OPERATING TEST: (Section 55.23)

Date Waived / / u ul. Hot @ Cold / / Ad=inistered by / EVALUATION: Senior Operator - - PASSD / / FAILID / / Operator - - - - - PASS D @ FAILED / / CO.TtENTS : RECCS ENDATION: / M pprove for h ) (Operator) License / / Do not approve f or (Senior) (Operator) License DATC: I l, EXAMINER vu ISSUE (ST::TCR) (0?ZZAICR) LICINSI (( [Mh,, MY: / Chief, Operator Licensing 3 ranch [ L/ NOTE: 'he operator and oral exa=ination check sheet =ust be ce=pleted for each applicant. If he has failed the operating and oral test at the senior level, the check sheet should be cc=pleted to indicate his evaluation at the operator level. c; a q --s%

k m \\ N: OPERATING AND ORAL EXAMINATION SD.ARY RF. PORT

• PACE SRO RO NO.

i 1. Operating Demonstration 7[///((((((((((//) ~ 1.1 Pre-startup and Instrument Checks l 3l l -1.2 Console Operation I/(([/(((([/(((((([/ a. Manipulations l$ i l b. Understanding l l f I 2. Faei1iey Equipment ((((//((((/(((((((( a. Major l l $ l b. Auxiliary l $ l c. Engineered Safeguards Systems l 3 l //[/////[//((/(((([/A 2. Iaseruncnea aon a. Nuclear l $ l l b. Process l 4 Reactor Protection. l $ l l _(((((((((((((((([/1 5. Procedures a. Normal l $ l b. Abnormal 3 c. Emergency l N l l 6. Reactivity Effects (Except Console Operation) f 7. Ad.ninis trative Requirements lS l l 8. Emergency Plan l3 l l 9. Radiation Protection l bl l 10. Responsibilities l $ l If an SRO applicant is unsatisfactory at the SRO level, complete the RO coltann to evaluate his adequacy at the RO level.

• For cach unsa t j s fac tory ("ll") Jist the page number (s) of the operation-oral exarnination notes on which the unsatisfactory responses are explained.

70-479 2. ' - ' ~

J 3 4 5 c _- - zr '< t A 'fA t t' [. Etcm E,c.,b e 8%=G[ e 0hwa EvLwe 5. 06~e v u l ct 40b 91 QL 4 LLLOL ct 0 a &r.r, t I. (ckli( ta b O ff U ?_kQ.b , hh. hscud do ( 0* @,k DPL M Q c[cma O I c~ c m C.'p, p 4 dv2w 20 R c-Cdc,J.1. eta _ 2 tv% {SPM0+ [ Ao IO -T T.R. 3 & % ~,7 g h h ~.4 T 1% \\cJ f( d \\q' DL p,r1 (JM QibPM, 2C. [ k \\un Kr.-w \\o L un k kk{ p2 T1 \\.w.0 O f -D.k 0N \\ Y $^ C*

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• ~ ~7 roCa% f % mLPwe/

Fo< If(G L b,nr.7 . w AJ1 8.S re mwti w ~ Tod ) k[ Akh dd 1 4 +.aa Dog ~ 2 LL gn to hcTrs~e{e~q. I 73-175 G*d g '9 e a G -e y c, W2" r 1 ",5 6(oO (f 73) Cwav7 C To eepn m L -fo r i D o pf u, t i

m i.b LM .Y- <t N -. 07/06/74 - - ~ ,pj5 Revision 3 - i . r. .h ~ ,'jtSTI?.ATED CRITICAL ROD l'ar!T:oN .Je [ Table 3 - Ref. Cond. 5 C'r, 0 TI', No Xe, No centrol rods

^

CAUTION: USE IllE 'i3."F FUi:L CURVE AND THE 532*F ROD CURVE [ / J, k 3 EFPD ' 1.a. Core Ags.,(EFPD) / b. Execan reactivity due to fuel (Fig.1) + / 7, [,- %AK/ 2.a. Rod worth of Cp 1-7 at present position - 3. 5f%ax/K x (Fig. 2A/B) b. Rod worth of Cp 8 '(Fig. 4A)

• 3((%aK/K s

c. Total rod worth Gp 1-8 (2. + 2.b) - 3, 9 3 C%5K/1 3.a. Present Boron Concentration / 9 Z a p:3 b. Baron Worth (Fig. 5A) / L[, $ %6K/1 .L 4.a. Te perature coefficient of reactivity (Fig. 7A)g,/2//A %AK/K b. Present Tav bl

• F Reactivity due to temperature. change c.

- (4.a) (Tav - 532*F) + O %aK/} 5.a. Reactivity due to Zenon Y %aK/) 6.a. Power at time of last shutdown // %FP b. Time since last shu tdowrr I/2d DAYS Reactivity due to Samarius buildup (Fig. 9) d c. % K/K 3 7.a. p (n e t ) =p (f uel)+p (rods )+p (b o ron)+o (t emp)+p (Xe) +o (Sm) , p (ne t)= (1.b)+(2. c)+(3. b)+(4.c)+(5.a)+(6.c) - ), 2 3 ( %sK/K 8.a Estimated @m 4 Critical Position (ECP) 9/ % withdrawn on Cp [ NOTE: In actuality these rods will not be wit.hdrawn until af ter deboration. b. Rod worth of Gp 1-7 at ECF - 2,3 / $ %3K/K (Fig. 23) c. Rod worth of Gp 8 at ECP .3 [ $ %dK/K (Fig. 4A) d. Total rod worth at ECF ~ (3.b;+ 8.c) - ), '7

aK/K I

rX6 20.0 w- ~ ~ ~ ~ ~ ' ~ ' ~ " * * *

., ),./ .........- - -....-- ~. Q 4 07/06/74 _ l' Revision 3 2 4._.,,. q . /<t ."h .-)? .../S.e-apdue to rod vitt.drawal ),23i %AK/K n + g' (8.d ) - (2.c) d .g ap (boron) = p (net)+p (rod withdrawal) 9.a O %AK/K (7.a) + (8. e) = f: 10.a Initial Boron Concentration (Cr ) (From 3.a) ppmB ~ y

b. Reactivity due to Initial Boron Concentration

' %.a : - %AK/K s 1.- (from 3.b) - sj i l

c. Change la reactivity due to deboration j f.p.

~ Y,I.. _ j,h? (10.b) - (9.a) %aK/K ..-c j '. c : j, (Fig. SA) ~ ~ /329 p d,~. Final Boron Concentration for 10.c ,. _....,; s .g, ,,,~ 7 I/-(- %AK/i D^. t' % g,= - s; A,g,e ( ;M' ,.'s.., , Reactivity of Gp 1-7 at final position (8.b).M I f.-*.. 'W ".'j' *C ll.a ...., g,,, [ %AK/: -lg g. ,,.g, c. ..,., y.sr +t.7,.. g,..,.p g..gegi.g.5.._ ' f, 7 / ~ 'b Add 0.5%AK/K to ll.a

2. [/ 5

%AK/ i... -6, c Subtract 0.5%AK/K to 11.a --.}-3flM.4bldt::y !.a&S.yfx, r;tgi@)fyyEQy;%W D.kQ'f=ggg Q g[.*i. g ...g. ..y. d Rod Positions of ll.b N,5 .. on Gp Rod Positious of ll.c [2 _%W.D. on Gp 3,n* - !'. 4:

• 8 *...r 4-e.

-~ e ~-.. 'll-,m'L'. ep CALCUfATEDBY DATE-TIME APPROVED BY DATE-TIME RETUFl? TO 'TCLEAR E':GI';EER /IN'7 21.0}}