ML20210T500

Startup Rept

ML20210T500
Person / Time
Site:	Calvert Cliffs, 05000000
Issue date:	05/12/1977
From:	BALTIMORE GAS & ELECTRIC CO.
To:
Shared Package
ML20210T454	List: ML20210T449 ML20210T457 ML20210T464 ML20210T470 ML20210T478 ML20210T484 ML20210T493 ML20210T500
References
FOIA-86-236 NUDOCS 8605300488
Download: ML20210T500 (112)

Text

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TABLI. OF C'DN'"I2i"E SifTIOh PAGF 1.3 INITIAL CORE LOAD P

2.0 POS"I CORE HOT F"JNC":'IONAL TESTS 12 2.1 CEDM/CEA Perfor:=ance Tests 13 2.2 Reactor Coola=t System Leak Test 15 2.3 Primary and Secondary Water Chemistry 16 2.k Pressurizer Effectiveness Test 17 25 Reactor Coolant Eystem Flow Tests 18 3.0 INITIAL APPROACH 73 CRITICALITY 21 L.O LOW POWER Ph7 SICS TIETS 3b k.1 Critical Boroc Concentratioc arad Soluble Boror.

Worth Measure =ents 35 k.2 Flow Reactivity Measurenent 37 k.3 Ecn-Overlapped CEA Group Worth Measurer.ents 33 k.k Overlapped Regulating CEA Creup Worth Measure =ects L7 l

k.5 Te=perature Coefficient of Peactivity Measurerents i9 4.6 Shielains I.freetivecess and Plant Radiation Level Measure =ents 51 4.7 Chemical and Padiochemical Tests 52 5.0 ESCA~ ATION M PC/-B TESTS SL 1

5.1 Turbine Generator Startup ar.: Atacspherie Du=p."

Turbine bypass Valve Test 55 5.2 Reactivity Coefficient Measurements 56 5.3 Plant Power Calibration

$7 5.k Shielding Effectiveness and Plant Padiation Levels SS 55 Chemistry and Padioche=istry Tests at Pcver 59 5.6 Process Computer Measured Variables Check 63 5.7 Dropped CEA Test 6L 5.8 Induced Xenon Oscillation Test 67 59 Core Power Distribution 63 8605300488 860521 PDR FOIA WILLIAM 86-236 PDR

TABLE OF CONT'E55, cor.

f SECTIQh PAGE 5.10 RCS Flow Measurement Test 79 5 11 Shutcovn from Outside Control Room 80 5.12 Loss of Offsite Power with Coastdow=

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5 13 Partial Loss of Flow Trip 7est 89 5.13. Trip of Maic Generater 93 6.0 U3rJSUAL LY:XTC 96 i

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Page 1 IhiRODLCTION The startup test program was or.ranized and ade'-istered by Balticore Gas and Electric Company (~MLE) personnel assisted by Combustioc Engineerina (CE). The startup test program consisted of se veral phases. Test results from each phase vere reviewed by a Tes Data Evaluation Grc:2p (TDEG) consisting of the bG&E Startup "l"est Coordinator, the CE Chief Test Engineer, and others as required. Test results fallinF Cr2tside of acceptance criteria received an additional review by the Flant Operations a=2d Earety Review Cn-ittee (POSRC) and were resclved prior to besinning the next test phase.

i The test phases were as follcrwe :

l l

(1) Initial fuel load (2) Post Core liot Punctional Tests (3) Initial Approach to Criticslity 1

(k) Low Power Physics Tests 1

(5) Ewalation to Power Tests Maximum licensed reactor core power le vel is 2=40 MWth.

The startup test program began on 8/10/76 with the leading of the first ' bel assembly into

?.he reactor vessel and was completed on 2/15/U.

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Page 2 1.0 INITIAL CORE LCAD

""he ir.itial loa.iing of the ecre cormenced en 8/10/76 and was completed on 8/22/76. because of prcblems with a similar tatch of fuel at another reacter, the fuel containins burne.ble poison reds was eff-loaded and returned to Vincisor, Connecticut for modificatien. T::e fuel loading reconszenced on 10/21/76 and was completed on 10/28/76.

Both fuel loe. dings followed the sequence shown in Table 1-1 and Fis are 1-1.

Figures 1-2 and 1-3 show the location by serial number of each fue2 assently and each 4

control element assembly (CEA).

l A plot of inverse neutron count race versus the number of fuel asse hiles j

loaded was maintained for esch of the separate detecter channels to ensure that the reactor vou d remain suberitical follcwing each loadi=g step.

The detector channels used were vide range log Channels A and O and teszpcrary i:ncore detectors A amd B.

The locations of the temporary detectors are s' ovn in Figure 1 k.

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""he reacter coolant system boron concentration was mai=tained at 1770 pp or higher during: both fuel handling operations with the borated water level at appror*wately 8 inches below the top of the reactor vessel cs tlet nozzles. A sta -tup neutron source was installed with the first assen'-ly to verify respomse of the neutron : monitoring equipment. Cor:=unicatices were r.aintained between the contrc2 rocus and refueling pool are9 duri=g l

all fuel loading operations.

During the first. core loadi=s the refueling pool and spent fuel pool remained empty umtil the lkCth assently was loaded. At that step both pools were floocied. Puel mevement was no proble= with the refueling pool flooded but additional lighting oc2y made the CEA an4 *uel assembly se-ial cumbers scre visible but not readable. At the ecespletion of fuel loading, core verificaticm was acecmplished with the refueling pool drained and the boratec water level at apprc:ximate y 8 inches below the top of the resetor vessel outlet ncuzles.

s Several problems with the fwel hawing equircent occurred durinc fuel load:

Refueling wachine Only a few proble=s were experienced with the re.^2elins cachine during fuel load. A hedst everload was experienced whe= atte=-ting to raise a fuel assembly from the urender tecause the upe= der was coving slightly past vertical. Jiew ecordina es of the upender positic: vere used to resolve the problem. The brakes malfunctienec due to a faulty rectifier box which was subsequently r=placed. An adjustment was performed on t.he lov to hign spee:i interlock when the hoist box veuld net chanse

• rom icv to high speed while lifting a fuel assembly. An air lesk developed or one of the spreacer actuaticc cylimders which was corrected bv replaci=g the cylinder. A"ter loading step 172 calle slack cccur ed early when trying to grapple a new fuel assently from the urender.

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f Page 12 2.0 POST CCFI HOT FU'.JCTION A1, TECS Prior tc initial criticality, f el and reactor internals were installed.

Several tests were ther perfermed before initial criticcity was achievec. A list of the preliminary tests are as follows:

A.

Performar.ce tests on Control Element Drive Mechanisms and Control Elennent Assembly positice irdientors.

b.

Reactor protective trip cirenit and manual scra= tests.

C.

Individua-rod drop time rJessurements at ne ROS flow and at full RCS flov.

D.

Final leak tests of RCS.

E.

RCC flow tests.

F.

Pri: mary and seccndsry water ehemistry tests.

G.

Pressurizer effectiveness test.

ii. Neutron response check of scace range monitors Neutron respocse check of source range mcmitors and reactor protective trap circuit and manual scra= tests were performed as ac initial ecodition to Initial Approach tc Criticality. All test results were within the acceptance criteria w.th the exception of rate of pressure increase durir.g the pressurizar effectiveness test and vithdrawal rate of 7 coctrol element assemblies.

'"hese problems are discussed in Sections 2.1 and 2.36 Additional tests perferned during post cere hot functiocals included:

A.

Check of turbine Eeneratcr as maximum RPM.

B.

RCS and steam generator instrument cslitration check.

C.

Incere detector recistance eseck.

N suMrummese Muduuuwmw oupuuuuuuuuuuumssummguussur AuguuPWWurf i

Page 13 2.1 CEDM/CEA Performance "ests Purpose The control element drive mechmuism/ cont.rol element assembly perfot mance tests were conducted to accomplish the following ob,l}ectives :

1.

To desmonstrate proper functioning of the CEA's and CE"M's unde =-

various reactor coolant sywtesa (RCS) temperature, pressure, andt flow conditions.

2.

To provide measured CEA withdrawal, insertion, ar d drop time data which will serve as comparison standards fbr futurre per-formance tests.

3 To perform a check of the position indication system and to establish proper functioning of the CEA operating and inter-j lock lights.

l k.

To verify that all regulating and sh:utdow CEA's have drop times for 905 insertion in accordance with Techetical j

Specifientica 31 3.k.

Test Results The CEDM/CZA performance tests were conducted at RCS teWures and pressures of 260'F/k90 psia and $32*F/2250 pain. The tests consisted of:

1 1.

A measurement of coil resistance.

2.

A check of CEDt/CEA withdrueral speedt.

j 3

Ti=ing of rod drop from full out to 905 insertian of all re-gulating and shutdown CEA's with full RCS flow (h reactor coolant pumps).

1 k.

Ten (10) additional drops of the fastest and sicwest Ns.

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5 Measurement of CEA withdrawal speed and rod drop time for all f

regulating and shutdown CEA's.

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6.

A check of all CEDM/CEA position indication, operating; and interlock lights.

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CEM withdrawal and insertion traces.-re analyzed and ad,*ust-ments were made to coil power programmers te ensure accept.able l

operation of the CEM's.

Discrepancies in rod positi an izxii-i cation were corrected by adjusting compu-ter setpoints and l

programmair.g. The results of the CEDM speed and drop tine tests are as follows:

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Page lk 1004/230 rsia l

CEDM Vithdrawal Speed (No Flow)

Fastest s'.cgle CEA - CDM 1 3k.1 f r /= n Slowes t single CEA - CEDM 11 30.37 in/cin Fastest deal CEA

- CDM A 5 22.1 ir./mi.n

- I Slowest dual CEA CEDM AS 20.05 ir/m:in Drop 7he to 90". Full Deser^ ion (No Flcar)

Slovest single CEA - CEDM 57 2.14 see Fastest sicgl.e CEA - CD M 55 1.95 se:

Slovest dual CEA CEDM k 1 95 see Fastest dual CEA CEDM 49 1.6k see 2604/4,90 nsia Dror Time to 90% Full Insertion ( 2 RCP's )

Slowest single CEA - CDM 56 2.24 Fastest single CEA - CEDM 3h 2.05 Slovest dual CIA

- CDM 43 2.03 i

Fastest dual CEA

- CDM41 1 90

$32*T/2250 pela

)

Drop Time to 93% Full Inspection (Full Flomr)

Slovest single CEA - CDM 57 2 37 Fastest single CEA - CEDM 35 2.17 Slovest dual G

- CDM 48 2.22 Fastest dual G CEDM h1 2.11 Drop Timme to 90% Full Insertion (No F_ew)

Slovest single CEA - CDM 57 2.14 Fastest single CEA - CDM 58 1 95 Slovest dual CEA

- CDM SC 1.86 Fastest dual CEA

- CDMLO 1.80 Conclusions CEA 1. 22, 31, 3h, 36, 37 and 61 all icdicated a withdrawal speed 4

greater than the acceptance criteria of 33, in/ min.

The G withdrawal incident is based on a reactivity additice ra which vculd require a CEA vorth of 1.03 x 10 ge of C.6 x 10-k

/see

/i:) at a withdrawal speed of 33 in/sec.

All CEA withdrawal rates are less ti:an 35 in/sec.

A CEA Motien Inmibit will prevent further CEA withdrawal upon excessive misaligna:ent of a CEA during group motien.

All-other test results were neceptable.

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7-Poge 15 2.3 Reactor Coolant System Leak Test l

Purpow A leak test of the Reactor Coolant System was performed to eteck for abnormal leakage from the primary system.

Tgst Results The leak test of the RCS was conducted at 2250 psia and cove red the following areas:

(1) Reactor coolant pus:p seal area l

(2) Reactor vessel head, seal (3) Steam Generator pri.sary manways (k) Control Element Drive Mechanisms and In-Core Instrument Penetrations (5) Stem leakage on all valves (6) Pressurizer heater penetrations

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The inspection of these areas showed no abnormal leakage ami the test vas considered satisfac*ary.

Cooelusions_

The RCS leak test aboved that the primary system was tight.mfter reactor vessel reassembly following fuel load and no abnored leakage was indicated, i

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Page 16

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2.3 Primary and Seeendary Water CE,-mistr v Purpose To establish, monitor, asd control primary and secondary water chemistry during plant heatup and post core hot functional tests.

Test Results All primary and secondary water ches* stry results durir.g post cere bot !bnetionals were either acceptable or when acceptance criteria vere.not met, ccrrective actice was taken to maintain acceptable conditions.

No unusual levels of soluble, corrosion pWucts or particulates were observed in Whe reacter coolant syster. during the t.e sting.

r Conclusiens

'Ibe data indicates that good chemistry control was maintai=ed during post core hot functional testing.

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Page 17 2.k Pressurizer Effectiveness Test Purpose l

l The purpose of this test is to verify the effectiveness of the pressurizer in controlling reacter coolant system pres sure during plant transients at 225: psia and 532*F vith h reactor coolant pwsps operating.

'Iest Resu' ts The effectiveness of the pressurizer was determined by measuri=g the rate of pressure increase with the backup heaters on and both spray valves fully closed and by measuring the rate of press-ure decrease with the spray valves fully open without the tackup heaters. Two test runs were conducted and the results are as fellows:

R!3 PRESS 3 E OECREAEE PREST!1RE 3C* EASE (psi / min)

{ psi /nin) 1

/

100 5.1 2

102 Les:s than 9 3 Camelusiccs Loth runs indicated that the spray valves opened fully and reduced tt:e pressurizer pressure at a rate within the acceptan. e criteria.

a The test runs also indicated that pressurizer pressure would not ir. crease at a rate within the acceptance criteria of 1"T.5 + 3.5 psia / min.

21B spray line indicated a higher te:sperature than 21A with the reactor coolant system at a steady tenperature and pressure.

It was also noted that the RCS vc. ld slevly lose pressure without backup heaters when the proporticmal heaters were cperating.

"?ie s e indications suggested 213 spray valve was leeling vtics could cause a decrease in the rate cf pressure increase. After adf ur tments were made to both spray valves and alsc the auxiliary spray valve PCS pressure could be mainta.ined with propertional heaters only and 215 spray line temperature dropped to stout the same value as 21A.

It was concluded that the leakage problem was ccrrected. Plant cperating eccditions did not allev for a repeat cf the test af*er the adjust-mests were made.

i i

l

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Pase 15 l

t 2.5 Fea:ter Cc clant Cyste= F1)v Tests Pur;ose l

This test was cend. acted to determine reacter coolant syste flov rates and pressure dreps around the reactor coolant loccs for various reactor coolant pu=e (?CP) conbinations and to determine RCS flow coastdown cheracteristics.

Test Results Testing was conducted at RCS conditiens of 532*F/2250 rsia.

Reccrded traces of SCS flov versus time were made for each of the RCP cochinations listed in Table 2.5-1 and for the simultane:r.as trip of al.1 four RCP's.

Total core " low data was taken for all possible comhinations of three RCP's.

Total core flow following the trip of all four.*rP's is eer: pared with the predicted four pump flow ccastdown in Fir.:re 2.5-1.

The lowest three pump total core flow rate was measured with 22A, 215 and 223 RCP's running.

""h e fastest fiev coastdev= occurred when 22B PCP was tripped.

Conclusions The four pump flow coastdown was exactly as predirted and the total core flow was hG1, LOO SPM vhich is well atese the rir.irus acceptance criteria. Core flev vith less than four RCF's operating was measured fcr infor=stion purposes only since admiristrative re-

)

i strictions on plant operatic =s prevent, power operation with less than I

four RCP's.

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TABLE 2.5-1 g

REAC10R COOLANT PUMP Flow C0ATfDOWN DATA _

.t t

BCP's INITIALLY FIDW COASTDOWN PER*ENT OP MINIMUM i

RUNNING PUMPS TRIPPED DATA

'!UPAL FIAW (OPM) 10TAL FLOW 21A, 22A, 21B, 22B kolh00 108.50 21A, 22A, 21B, 22B 22B X

301909 81.60 i

21A, 22A, 21B, 22B 22A X

302225 81.68 21A, 22A, PIB, ??B 21B X

301555 81.50 21A, 22A, 21B, 22B 21A X

300580 81.?h 4

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Pase 21 1

3.0 INITIE APPROACH TO CPITICE'TY Initial criticality was achieved on l' /30/76 vith the Peactor Coolant System at.?260 psia and 532*F.

The icitial FCS borce concentration was 1770 ppe.

The initial approach to criticality wits initiated by withdrawing the shutdewn, part-lenatth, and regulating groups in a specified sequence until group 5 was at mid-core.

Plots of the 'nverse count rate versus the CEA withdrawal points were maintained during the withdrawal sequence. The Reactor Coolant Systen toren concentratice was decreased by dilution at a rate of 88 gpe to approximustely 1200 ppa.

The dilution was stopped for about k bours then continued at a rate of kk g;ma until criticality was acnieved. Group 5 was then used to nain*ain the reactor stable at 10-35 power.

Plots of the inverse count rate versus RCS boron concentration were sain*ained so that the precise time the reactor achieved criticality could be anticipated.

The CEA withdrawal sequence is given 1.n Table 3.1 ar# the initial criticality data is given in Table 3.2.

The Peactor Coolant System boron concentration versus dilution time is shown in Figure 3.10.,d DCS Inverse count rate multiplication versus CEA vithdrawal points an dilution time is shown in Figures 3.2 through 3 9 The response of the 'Jide Pange Leg Chm 9nels to the neutron flux / level was monitoreo during the approach to criticality. Tbc data indice.tes that more that one decade of overlap exists between the ecctinatier. of fission chambers and proportional coucters and fission cha=bers only.

Figure 3.1 shows the overlap data.

An inadvertant. reactor trip was experienced six (6) bours after startics:

the RCS dilution. The dilution was halted for four (t) hours while the rods were withdrawn then continued until the reactor was critical. There were no other probleus encountered while reaching initial criticality.

The measured and predicted critical bcroc coccentrations were in close agreement.

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I Page 22 TABLE 3.1 CEA VITICIRAVAL SEQUENCE CEA NWAL DOINT CEA GROUP INCHES VITEORAV'T 1

S utdown A 67.0 2

Elbutdown A Full out 3

slbutdown 3 67.0 k

Sibutdown B Pull out 5

m utdown c 67 0 6

a utdown c Pull out T

Ptart length CEA's Full out 8

Ruegulating 1 67.0 9

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TABLE 1.2 INITIAI. CRI""! CALM ".#"A paa h"JRED INI*!AL CONDITION RCS Temp. 'F 5 32 532 RCS Pressure (psis)

??65 2250 RCo's operating 21Ak3, 22A&B h

WFt".4 A (5 power) 1 5 x 10-3 10 h WICC B (Toover) 1.2 x 10-3 lo-h VICC C (5 power) 1 5 x 10-3 10-h WitLC D (5 power) 9 x 10-3 10-h I

CEA CROUPS (inches)

Shutdcyn A 1 35 135 Shutdown B 135 135 Shutdown C 135 135 Beg 1 135 135 Reg 2 1 35 135 Reg 3 I?5 135 Reg L 1?S 135 Reg 5 6.5 66.75 PLR 1 1?S 135 i

1:S 135 p=

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.,a' I >ds.F P-iY3 :C5 *~21'7' The Calvert Cliffs nit ? -. i t i a '. core e-r.sirtr cf C;~ faal u s liar ea h :ontaL=ing 17 fuel rm s/tt. nati-reison : cas ac: ' Contrus Elenent Assen:.ly Fuide tuteJ.

Fue. S3Settlier a~e diVdded ir.*.C 3 tyr*M Dy er.rie:.=ent, 'Iype A. 5 w.a '. Twelve.172 fuel rods in all Ty:a is sed several Cy;e C ft el mast-et ias are rerlaced wi..* turr.st le re i =en rc1:s.

Fenetivity :octrcl is ; re.ided ty =cl t1= torc-ir. the resete r ecois.t syste= (BCC) and by d5 CEA's.

C 3's sre Ins e--te d i nt e u.s wi-!. draw-frcr.

the core Ly mesns of C'. co. trol elane.t dr ive -: nee r.an i s==

( C E--s ).

.Nenty (iui C ud's are stta.:ncu te dual CIA's.

The C:' t" s ar-arrs.-rea it.to t en OLA rroup s ar.: aefined by f;=etior. pee-ferne:.

"rou 3 AE and C sre sr.utaest.;,roups ; tr.rourt 5 are rerulatir.F greurs; an: r trt le:.gt: rc4 3r:aps 1 an: 2 are rever shar inc. Fr oups.

L.::w Fower Fr.ysics Tests eccsist yrimarily of tre r.essurerent er resetivity vertr.s cf r-e:har.ise.s v:.ich vary the critical e ndition e' the enre.

To iceresse t' e secura:y snd speed of dats rollar*.icn, an snslo compu:ter n

wcies solves the ki:.etics ecuatien fer reactirdty wa.s sed.

."* ve rs I tech:.igees were ases ir. ccr.l,une. ion vitt. t$:e r-actiirity ecmp ster te-zzess ae Ch.A strt*1s.

A solutie toron eva; techr.ioue ccc.zister of a continaoas or st g dilatioc. or t.oratien of the EC~ si==0taneeus wit.h small ccr.;ensatins reactivity ecanges ir. CEA ; esitic:n.

~'he eactor was sept near critical iaring this evolution and

• ne resetivity remputer

rovisec a signal w=ich emld te trende

and correlated with OEA pesition as a fur.: tion of ti.ne.

Soluble toroc werth, temperatura, act flow coefficient of reactivity worth were neasu. red in additien te CEA r=-activity worth.

Other tests included primary syrteri er.er.ical sc c radicche-:lesl terts, snieldir effectiveness, ar.d pisnt radiatio. Level =essurene=ts.

I I

i 1l 1

iare ^'

a.1 C r i t i cal berc r. Con.? - t ra i r - sr 1 e '. 1 ' a 1-rcr. A r t r. '%ri

.re w.a t s T rl ose I:. a:dition te cant r ; re d c:r ve.e-.. se;ut '.e toren may te asac

.e c:ntrol react v;ty. r.

L' a cera.

-r.r a.eur.: cf reactivity cor.t re l c:ay be obtair.e ty

.,at ar.inir.c t he r**a t ! vity wo"*.h of tha selutie tcron nr.s measurire t he critical teren ccr.entration fer vsrieur cor. trol elezer.t ar.F. -t :y cer.fieura* i r.

Tast Iesait,

'.re reactivity w rt:_ e r t r.e relub'a V rrr. v*2s re==urec 'v diluting r

L:.e 3 enetor Cr.:,hn*.'yst e~ ar t i : t'a r*%: stirs eurtro; eier ent a settlies were at f.;l it. sert i on sr.2 sr.ut: wr. Creuo C CF1.'s were d:vn to 21' wi *.r.:rra-T r.e rod treur resetivity wnrths were measures with tr.e r setivity ee-ruter durir.c the i;1ution process arad the boron :encertration vss r.easare: at the ecdpoir.t of tha CEA ecmficurations.

Tr.a Pesetor Coolar.t f.vrte-was then berated tack t.o the cririnal coreen ratter. whila tha ect trel ele cr.t arser-tlies were telne witn:ravr. to a.1 ross cut.

i*y dividir.r tha chance ir! torce ecmcentration inte :*.a tcts chanr= in reactivity durier the dilution ar. Loration process, the solutic toror. worth was otta!r.ed. The reactor was held crit *:a1 at aptreximate;y 10-D rever during th-ec:*. ire process. At raricus contre; elar ent asse.tly cor.fisurstions,

t r.,a reactor ecolant system terer. concentratier. war alleved to st.atilize and a sam-;e of rea tor coe;ar.t was analyzed.

7'.e results are given in 7atie '.1-1.

Cemelusions i

Th.e measured critical toren eer.centratters and solutie boron verths for the variews contrcl elecer.- sssec5ly ec= figurations were within tr.e acceptanM criteria fer si; : easurad vslues.

4 I

ll e

l 1

d l

I 1

vt

f*~,

~

_ ' ' _. 'f.^

,L'

• l ~.*A., p r

, _. ; r..

't

_ r. #

a w.

l 1

TABLE 16.1-1

}

CRITICAL BORON CONCtJITRATIONS AND SOLUBLE BOR03 W)PTH j#

Ct.A IEITION 15)H0N CONCENTRATION. SOLUBLE BORON WORTH M1AMJHED CRITICAL

{

l (INCliff. WITl!DRA'dii)

(ppm)

(rpe5f 4..*')

ACCLITANCE CRITERIA i

I cp5 - ys"

[

1121 l

1o89 + 33 nra I

I l

1 l

ARO 10 %

1089 1.*5 prm g

)

I 1

(RegulattrF Groups 737 75.9 720 1 e? rpm l

rull In) 77 1 5 prm/% Ar-Grouts C - 21*

g

'~~

1 i

e AHJ IO'M

,?

77 1089 + 25 ppm

'1

.i 77 1 5 I ps@,1,-

i 1

1 1

l HC* at 2?50 ps! sad $ 3?"F.

J i

4 1

l 9

4 k

l l

J t

I j

P 1

=

e-c

-r

~

=

v Daze :-7 k.2 Flav hea-tivity Measuresent Purrose 4

The purpcse of this test is to measure tr.e e.'"ect of reacter ecolant system Ic:ss of "lov on reactivity.

Test Results The excess reactivity vms acj.sted to zero vi--

the control re:s.

21A reactor c clant pump (R;P) va.s then secured ar.: the r-activity ch ere notes on tr.e reactivity compu _er.

CIA P P va.s then started and 22A RCP was secarea.

All fc ar (b. reactor ecolant pu=ps were runninir anc flow cccaitior.s were stat.e at tne time esch rump was secured. The res.Alts a.re shown in Tacle L.2-1.

TABLE L.2-1

• 353 OF F.OW FEACTIV"fY O FEC"

/

IN IT I A*

C'de L ITIC% C PCT Ct.CL? EL CFA13E I P.1ACTIV!"T (Cects )

I

/

21A, 21c, 22A, 22B rurming 22A

+.36 21A, 21b, 22A 22b rur. ning PlA

+.35 o

gnelusioes PartiLJ Icss of reactor coolar.. flow does not cause acy appreciable chant,e in reactivity.

w

rwe 3d

=.3 hein-Overiar ped CEA Orour Wort. ?'a ssure-rent e F.trr.os e Reactivity con

• rol is usuallY rrevide-: by reacter coolar_t systea

< aluble boron curins ncrr.al creratien. Aoditiensi control is previ:ed by mevatle cor. trol ele =ent ar se=t lies (CEA) arrar.res in syr::netri cal groups in the cere.

C hr-nur:er of Cd's in each croup is snown in Table 4. 3-1.

"'he C=J. freup wort h menruremects ve7 perforned over their full rsr..;e of mos er.'

ir a n.en-cverlappec r.ede.

Tes t Hesults The CLA reactivity worth.s s e cessur-s by insertina or withdrawins toe CM grcup while diluti.g er torati.rg th,- reactor cociant system to bcAd the reactor critical. The incre -=ntal worth of the group could tten be obtained frcan the reactivity computer traces.

because of the close similarity tetve-n Uni. I and Unit 2 cores, the Unit 2 group verths were assced to te in close agreect'ct with the corresponoing ' nit i group verths so that tre, ri.nre of tre acceptance limits were en ch less. The cririr.al intention was to sessure the worths of the reguAsaicF Fro-Ts cnly and assume thee shutiovn group verths were close tc the measure: Uni 1 shuf.down croup werths. Einee C11. Group 1 was sli.-atly lower than tbc acceptance limitt, shutdevn Group B and C worths were also maasur-d.

'"be part lengti CEA groups i

I vere not measured since they are not,. ed durin caorr:al operation.

Toe measured CM group vorth ar* shewn in Table k.3-2 and Figures k.3-1 tcrough A.3-6.

Cccelusions CEA Group 1 vert ~. va zensured to be slightly low-r than expected while r

all other mess.ared gcup wort:s vere well' within e.ceeptacea limits.

l 1

1 i

l l

i

e 78$'.5L.3-1

,R:QCTIVF fEdOL F:i;~"TO:1.0F CEA r.RO*F3 i

c2A CROUP ENBER g5GEHOFC?.A's CONTROL _FI;C*IG;i l

~

A 16 rarc'y n

8 cafety l

C 16 Safety I

1 8

Power Irrgulatter.

2 6

Power Begulating 3

P Tover Perelatina, h

h I'over hrgulntinc i

5 9

Power Regulating j

i PLR 1

.h Arial Povar Chaping l

..s j

PUI 2 h

Axial Fover Shaping

}

i 4

4 i

e 4

i l

q

i

/

l CALVERT CLIFFS UNIT 2

INTEGRAL GROUP WORTH l

BOL, 1st OYCLE, NOMINAL 532

• F, 2250 PSIA l

9 I

t

. t0 i

. 99 l

I

. 50 I

. 45 cc M

=

. 40 e-M O3

. 35

)

o i

i O

CE

.30 y

cc CE QW

.35 l

Z_

j 20 l

.I5 10

.05 0

le 20 30 40 50 E f, 10 80 80 les lie 82e 130 140

\\

INCHES OF WITHOR ALIL 1

1 GROUP l

l16URE t.

3-l I

i

'~4+--

._,,,c..__4..4-.

- - + _ _ -.. -., _ _ _ _. _.,., _ _., _, _ _ _, _.. _ _, _ _. _

TAbi,E L.3-?

CI:A fil,TJ.GHAL REACTfVTTY WORTl!C

(:leminal 532 F, ??50 psta)

'Ft.A GitOUI' NU'4hEH i)F CEA's MtEURED WoliTil (s_1 )

PREDICTED WORTil (% A ' )

I

^

A 16 r:qt !!nneurert III 8

.959

,966 g 01.9 b

C 16 1,P60 1,1n1 3,0*f.

i.

_.)

1 h

,891

, 9 <.1

+

6!.'s P

8 806

.77A +

41.,

-__.-_L__..

3 8

,908 91? 1 01.9 i

L 4

.369

.353 +,o U 5

9

.558

.55? 1 0b?

= _

(1) Hensured from 19.$ inches withdrawn to fully with<1rawn.

i l

n

.g 1

CALVERT CLTFFS JN1T 2

INTEGRAL GROUC nOFTH BCt. 1st CYCLE NCMINAL 532*F 2250 C5: A I 30 l

g i

• 1.2311 I.20

~

l i

......I 1

1 1c i

i 1 QQ

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

l 4

I

.._.1..__.._...._..

g.9o 1

i 4

hf

{

4

.8o

...__y.-..--._...._.-.....

x E

l 1

C3 I

l 3 *Fo O

l

-.t

..i i

O CC I

.6o f.

-_..... a _. __. ------.4_

a cr i

a a

.p...

o I

LJ l

5o

~z

.t.

i.

4o t

I

.......i.

i.

e

.3o

. _...i......:.

.20 l

I

_. ~ - -.. -.

..j.....

.j.

I I

.j t

i

.: o

.i j

1 4

l

..j..

.l.

t l

j 0

c to

o ao as sc 5:

se sc i::

itc i;c n:

$4:

'NCrE CT. : T - O R A '*

• L GROJP C F 8 3U;E 4.3 - 1 h

'_'.'.y*

f*,

.. - ~

_~

~

,, w

~, _ '

G t=

y ', s.,

, _ g.

. ~,,

l

.,,_.--..y-----

.,.-4

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

CALVERT CLIFFS UNIT 2

INTEGRAL GROUP WORTH 1

BCL, 1st CYC E,

NCMINAL 532*F. 2250 P51' l

J l

13" i

I e

i 2:

i 9

1.1.

~ ~ -

I l

3 i

i t ::

?

1 l

I

}

i f

I

/

g gg C

}

7

-tf93 9

44 e

i

.8:

L-

- ~ ~ ~ ~ -

---

l-x g

O

'I 73

...- ~

o O

e 6

.g:

..s-.

CZ I

m C

j d 5:

~ - -

-~~

---t-----

z

,a; 1

l

,3g j

4. oo ;

e..

.e e

0 1C 20 3:

4:

1 i;

et 5:

Y:

!!s

??:

?4C F

b M WP g

$ I GROUP i f I GUFE 4,3 - 2 i

s 1

1 1

MI

i CALVERT CLIFFS UN1T 2

INTEGRAL GROUP WORTH BOL, lst CYCLE, N O M I N At.

532

• F.

2250 PSIA 30

....._ k

{_ :.:- _

. - - i...-.:.-.-.

i i

I i

20 5

}_

---

S----.----

i

}

h l

I 10 6

4

-[.---

1 i

30

.__..._q__...-

2

/

I

^

j f

90 c

h

-1

)

/

i l

80 (6066

}

. I'

... - j --..

i

.ra j

_i j.

-e l

l l

l_

l f..

_ _1 i

t i

/

i 90 t

---

2--

i

/

h

+~

80

.i i

i

.30 I

t

.i i

. _..}

...d'. -. -

i...-.

?

~

l

_.}

{

{

[

' i' j

j 2

i

___I

... j -

l i

i t

to 3

- i l

i r

t _

is 20 3c 40 SG E0 te so go ict tic 920 130 14:

INCMES Cr wtTHDRALAL GRCUP 2 FIGUCE 4.3-?

i I

l I

l i

CALVERT C L I F'F S UhIT 2

INTEGRAL GPOUP WORTH BCL. 1st CYCLE, NCMINAL 532 F

225C P5'-

i 3:

3 20 i,3:

3 ga _

'O A

g e;g.*

a::

/

et

.50

=

/

cr o

3 7e b

O a:

.oc cz X

o a. 5.n, Z

I e

.4C

.3C l

.2c l

I

,i c 1,

t Ic 2:

4:

6:

7:

e:

s:

i:e i:

irc 13 :

is:

1 NCdE5 ;F Wi 'H 40.aL i

GRUUP 3 l

F GU;E 4.3

• i 1

i i

l

L m

... x.

.,. n

~

,f CALVE ~T CL: FFS U t. I T 2

INTEGRAL 3ROUP WORTH BCm. Is: CYCLE. NOMINAL 532 F.

2250 PSIA 10

,I I

i l

3 i

1 i

.o i

{

t i

s

..s t

a

.so x

e.

cc O

t 3

2 t

t 1.3699

.5 C

o cc

..e

~

cz cc o

t

_a z

.i3

o.

0 -

---

2 ie 3:

5:

s:

r::

e 2:

l

s. c _ : s _ r..

a..r. =....

t GROUP 4 g GU t. 3-3 p

".$e

-

• Q g,%. s&*

.Q.

.er w & ** l.

-_,t*~.~^

,. s ?, h,'$ '_ ',R R.

E. -

^

4 n

- ~

i i

.e-,.,e..

.>...-#.r

,....w.e.

CALVERT CLIFFS UNIT 2

INTEGRAL GROUP WORTH I

BOL. 1st CYCLE NCVINAL 532 3,

2253 PSIA 1.30 1.2C l.IC 1.00 g.90

<3 M

1 x.50 e-.

~~-

- - ~ ~ - - - - - - - - - - - - - - -

~-

O rr l

3

,70 Q

O a:

.60 acz e

et

{

o j

{

g f.,., 9 ).

d.50

~

g j

.t

,t Z

t

.s0 F-------

l j

.J0 l

t

.i

,t I

I

.20 1-------

-f,---------.

-k. -- - -- - - -

---

+----J---

.10 i

1

\\

0 i

)

c to

o sc 4:

Sc so 7:

s:

sc ti:

Irr sac iso INCHES Cr olTHDCAWAL GRCUP 5 FIGURE 4.3-6 W

w -

r Pare ~

Ovaris; rea.%.: - i "rr -

'.*e

'-f.

ura ent-Pu po e Fas:L.,r pow-r ;e..

ay t e

!.'r.t rn1 '_ a d by sequenti=_; ir.s-rti n or vi t r.c.ruval e r

.-ai.. -e ir.c

..trd e:- ent ss s e= bl ie s. Pareant of overlst is ce'ect<

i n -

r.

a re. =.tiv e;v constar. insertian rate of pcsitive - r ne. r.:ve r-s t i. i ty ever tse full rsr.. e cf ZEA :.ove-ment.

T echt.ie s !

-i rie - ' < r., a l '. -w CD croup i:.:ertie n '. imits i

ss a rar.eti;r. -r -

ter - var va..

v.rieur. ir.s-rt i er neeurs at s

.. S"' ne n ive r* set ivity 4erc ; over

.:.a t-t r

rr^.ifn..aly e

incert ed dt.-

t-C._*, ' :.

it. terr._ ret. ivity verth fer tne re. '. st i r..-

7. c.r at e in a - c tar.a : ed mode was messred. The maair s1':we; ir.s a rt i c.r Lt caro rver ir arTrexi.=atelv 50" witz.drav41

_n IEA : ur.

-ne

.euarene nt was

==_:e with the re s: t e r c oc '.ar.. c; st e:- st

? *F sn.

2.^ 50 s i a si r. = th e T u-ve devele;e: free th<

--as are ant i-ad i r.

.aking e stirnt-d critics; coc:iticn e s' e Dit! cs ; r.

r to a r-seter sttrtur.

Test :esalt:

The over.arred intar s1 re activi.y certh :" the re 21 stint-CEA grJa;; va =casarad y sier-ly withd avin-tha CEA. rour r it incre-me=ts usine the =A.usl seaantial =oda vr.Ile bera.ine tha reactor cocl a:.

syctes to no;d pes-r consta.=t.

T-s reacti--ity traec of this CEA c:c vecent was re aced t.: obtain t.c ra;ationsni; between CTA grtr.:7 positions una reacti rity wr arth a. t.ese y sitions. Tne total ever-larre CT.A sorth frr reEulsting frc arn 1 -hrour.h ! van 3.5;M,

• which was well witni= the acceptance li.1.s of 3.!"

+.22 W.'.

Figa.re L.L-; show: tre interral vor n for Groups 3. I ana 5 from full eat to stcut W with:=rawn en ^2reur 3.

Cor.1 sier.s The 0 r rlstped CEA vertia '- - the ra a:ntir.r rroute i:: v i. r. f r. t r.e j

sceart a.ce criteris e.d th - curve < rive; from thi-aasure=er.t is

• nde q u=..s for use a.s

.r.

cpe: att ent.1 t e:1.

i 1

1 l

l

~ *.

"m.

,pa a.v pf.. " ' * -

'?-

) ', -Q g gir' N 4">-4p J 4.

g

- h

__ 4

r

_..an w,

4

CALVERT C LIFFS U NIT 2

INTEGRAL OVE R L A P GROUP WORTHS FROM ZERO POWER PD;L TO FULL OUT

BOL, 1st CYCLE, 532' F, 2250 PSI A i6 t

l a

i i

l l

[

/

e r

.e l.

i i

i n

1 I2 l

-)

l 6

.f I

l l

i IO l

f I

I I

s i

i<

l

<3 i

,l

,i t

e j

I t

+

r i

e i

+

1 1

l l

e I

l l

l 1

a i

i i

i i

t I

4 i

i i

i l

2

/

.t l

I i

i

.I i

I 1

Ii ri i

j GRour a o ao se ;

taa ;

0 40 22 120 GROUP 4 0

40 80 120 GRour S 1

INCHES elTHDRAWAL l

4-1 FIGURE 4.4 -l

.3 Af,,

',s.

e, -

]

?.

' h,-

i.

i*

i I

^

z,4

.=;

i

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

t Pare i.9

"'er ernt re Jo 'ri cie-t cf ;- n-t t ei - -

"%: _ resect s l'ar cre

'. h e ::4erste- * - +ra.ura Oxfficien. cf r* activity is closely reistes t; the isc the a1 te peratt.ra c oe f ficien* (ITC ) v'.er i

the reacter is lar.on free a. : the rever lav'I is telev tha r.: int of sclin, nucle ar nea.. f ar.:e it 1-not rcssible to chsnce tha moc rrator te=perature withe t chand..R the temparature of all core cocper.er.ts, t he isothe =al te=reratur+ coefficient it =eature:

at various rod cr.fi, ratic s and reactor encian. te=reratures so j

that ar. esti= ate of tse ree-rator tacrerst'are ceefficient M O.'

at

)

full power can :,e ext. arola ec frc= t he zero povar maasurer.ent of the isother=al enperature scefficient.

Test :-as its 1

The isotr.-resl f.e:perature coefficie.t was =essured at the dasired I

Cf.A positier. by peeressine tr.en increr.inr the reactor cools.nt syst.em te=rerat.:rc sr.: not::: the resetivity charre cr. the raaetivity c o=; ut er. The st a f. en s eral nestar an ecoldown r.ms was ave: ased to ott1Lin the iset *. -rmal tec* erat =re :oefficient.

"'he CO measured at all recs c'at..14 T rc teren, ero pcVer, and the "."C measured st. 737 77 beron, Orcup 7 with:iras.cn 20", reculatine groups full in,,

cro F. war, were riotted arsinst Soron coreentratinc.

By usi s an astima e cf.f.e c:.-.ieal beren ec.:entratien at 70% and 100%

pover, the isotmer.a1 tenr+ sture coef fi ci-nt at 70% and 10C0 power couLd be ext rapciated. ITC :sta is shown in Tab'e L.5-1.

Conclusi er.s The mescu er.' VCuas of tha ; retr.er=al tenperature coefficient are vit..ir. tha sece:* ar.ee tritaria an:: t r.a **0% snd 1:30 rever - derater tec;-cr stu.-a eceffi iente catrarc' ne i from thasa ITC ::w=asure er.ts are ve 1 sithir. it.c I &.i t e --ceif:e: oy Teennical free fication 3.1L.1.1.

'* ~

.,e e.

m o_

,e M

.'d #4.bh~...

g.-

y, n-g.

N

I g '{

g

,g, if.

.. 1

l

/

Page 50 TAB *_.E 2. 5-1 IS3THERVX 70' PEA?JE CO_N_O!EN" DATA cEA caccP resr-:os rr EX""EAPOLA""ED.E (Ixons wITur=wa)

(no 'tF/*P)

(nc 2f.sf/*r)

Ac c e p*. ar. -e W asured Crit eria 70" Power 100% Power All Rods Out

.2h

.25 +.1

.6

.7

.3 f

Op C 20" Withdramrn

.78

.81 1 2 j

?

5

( Re ar. Om. Pull In) i

+

/

1 a

RCS at 532*F and 2250 psia.

Il' i

I

}

l 4

i 1

l I

s.

g

. - m..

.g S[

w

)

AW '

'1

,/

- a...

Sh i e l;ir.

r.f fe c t iv-r..z s -

2r. : i. sr - %J i 'it - - "..* v e ; vaa:urerar- -

Pu g3 T:.is test vss perf: rne2 tc t re.. :.a a coer reher.sive survey c' : sdist ier.

le.el arou:.a the ; 1 ar.t in. er..-r s; and Unit 2 :entain=rr.t ir. T utieuiar.

Tnis sarvey vac tr.a t,as. s for se ermining later tuild ; of ralicartivity

)

in specific pir inc ana cer=4 cnar.t

• and te evaluate efficiency.mf n.: ;ainr.

Desic. basis dose rates fer tr.e saveral zones a.* listed in : TAR Cee.icn i

11.'...

Resalis 1

All except - of tr.e sa v y points locateu outsida U-I eent air.= ant i r.d i c at e r. dc.* rat a s le ss t hsr.

1 r.res /h r. The e ssurestle sese rate =

were attrit tacle to y.ipes cer.ta; sine exhusted ier. exchanaer resin er icr. eachar. car ver.

cases durir.c the tra.sfer of resin.

1 Only s few -asurestle er.za dose rates at : r.c =~as arentle reatron de se rates were c t t air.e i vit r. n Unit 2 c ont ai n=,-nt as part cf the Fadiatien Survey. All of the observed dcse rates were ottair.ed st the 10' elevatior..

Durinc a ro. tine ras: ole.cical su.--vey of scr.e unsr.ielded areas, some ex;<ctea aose rater vere otservec. These were locatec.*ust within the reactcr vessel cavity ha*.cn and tear a hot leg ; er atrs*. ion threugh itA=

prima. y sr.ield.

Con:1 siens i

Cociariscn cf the data cctained caring Uni-1 lcv pcVer rhvsics tests incica.e fewer aress where neasurasble cosa ratee were observes and the ma.xi=2:r.

ger.eral ares cose ra es are ernsistant with criteria presertec in FEAR rectic: 11.2.1.

.(_lx;

^

,&Q Q, f

'kf

&A

I Tar * %.*

l l

"' 4:*. i O * *.a-* i e 4 '. 5ar*F

,[

' rte *.i c s } and Puricr-The r% ; ose. f i t.a cha-i C sr ~ rsdi :Or --i :4 '.

• • ts sra to cetern r.a

sseli:.* corr sien :a.a. fi e s t : r. p rc..:uc t activ.e

-ve s at

hi; :ur.

ra ;

urar.i.:= cor..rt : ut ier. tc ar.ivity- ; eve:s

.c asrly dat. et icr. of falle4 fuel.

Test Iesalts baselir.e Cor sicr. ft ly:

,ue tc crene:.csi ; rc Ga vit. the n-line fi.t +r rys.em, =o 50-100 liter filter-: s ar.; ;. vas a la te te cry.r for tna ccrresice stucy for

.PI"!.

C,clutie Fe, a ana Or were sli fo.nc tc ta let s.ha:. the =inirsam setect able cer.centrat.or..

Fissic. and I. :tiva-ice Tr-2 art S i!..-Ur ft dv oceaase cf ::.e limi t e: :_rntier. cf tne Pr" only thr(e RCE sa=rles were snaly ze: f c r.7 emi.t e rs. One res lts are t elev

.'in uci/ml):

v I '-

rlU Csl3~

CoS5 N 31 I133 L at a-

.1/30/76

1. 5:L-c

.u:E-C 7.-2L-7 3.cLE-T 12/1/7;

.331-6 i.c3E-f L.72 -7 12/ 3 /7*.

1.51E-6 2.0LE-5 b.22 E-6 5 22 E-7

-~he generally low - eve;s of a:.ivity, the unkncv. centribution of recycled toric acid f o= U-; ar.d tne lir.ited :at e. availst.e re.e cor.elusier.s from 2

this dsta cesningless. The ap:. earar.re cf crud a:tivation products (Cc55, i

Mn5f-) and va cr activstien pr :ucts

,F t; in lov levels it set t.neArectec.

i t:. i L= bui1=-Up:

2ne li.hium =ailcup 2 rirr LTr..as -xr-etea to te s C.1 f rz/:ay.

~ne

.itniu= res 1.s fcr E 2r showed th s o te trua.

A lithiur. nycroxi a ad ition was =ade cc. 11/3D/'i t o ad,' ast -.e lithium ecc. centration c 1-2 ppc tut bect se cf the diluticc cf the 503 the eencer.tratien was secreased t o <. ppm.

Durification -vster L.F. :

o cru samp_es fr:s filter ir et a== cu.let vere ecllectec when e ud levels were

-! 3.01 pr=.

Tne EF results i1.e.,

t r.l et 4 0. ';; and ou-let 4 0.v1), there fere, so :.ot in icste a. rue purificatien systec 27.

Due to no operatic. cf tne c.

ex:ha.rers, analysis cf Ni, Fe, and Cr on icn excha.:.rer influent and eff;uer.t were net perfer=ed.

Insolutle I,i, Fe, asa Cr vera arialyzes across tne CVOC purificatf or. filter but all resul s were.ess than the r. n.=u:. detee.atl-concer.tratic s.

Tne l

purification filter ;F (tne ice. exensnrers not havir.c teen in service) for

) cuitters was aeter :ined via cross #f a.d scan analyses, bat due tc the auw nt, act:rity 1-ve '.s no use

• ul case vns obt ained.

v

,-a e3 a

Vol :re Contr 1 Tar.r.:

Due to tne fact tnat fission rasen were at cr be;c-t r.e i r m;, ni=2 detecta' ale activity, no builc ; of fissicn.ases were ot r erve: ir.

the

'..'T.

Sa=;1es of reactor cociant ecllected car nr

• P'S chevec ne fission gas.

_aca.ase of reco=menda.ier.s ma:e te 1:=it oxygen ngress I

intc tr.e RCC sz a result of tr.e Unit I core diffarential press ze grotlec, hyars::r.- vss metere: ir.to Ur.it 2 ECC r.aFa-up vster a.d a nycrocer. over;ressure was establishec in the VC7 d.rinr LPPT.

Oxyge=

levels ir. Unit 2 F.0C r.ever exceeded tr.e CE re-cor=er acc s7e:ification aurir..; tr.e entire icw power pr.ysics tect p rc.tr a=.

.'or.: 1.s i cns Che=ical anc rsdic ene=iesl data was reviewe- * * -~

out LPPI ar.d all results were eit:.er seceptat.* cr when apprepriste, corrective acticr. vas ir.st itute. to aer.ieve acceptatie eencitier.r. No meanir.rfal sata was obtai= ed or the perfor:s:.ce cf the CVCf ic. exchangers and filters cue to r.o operatier. of tr.* ior, excha..gers ar.d icv levels cf suspecded solics and rasice.etivity.

I '

11 1

Par e $L f.

r C. i O!.

.'0 PC' ~r.E Ti.CTE hv.-sesiatic= te power tests were cond : tea t o dete--.ir.e sc-tuilt p.aa.. enaracteristics aurinr-ste ac state and.rar.ciant orarstier.s frc=

4erc tc IWL power anu tc c-.,or strate stat the f ar. is. eara le of l

withrtar. disc the accidents a-d tre..sient s analyzec i= the FfAS.

Terts were i erfor=ec at majcr plate sus cf 20%, 50>, i%, and 100', r ever. Tests perforcea in the escalat inn

.o ros+r prew: ram includec:

A.

.'ar:.ine generatcr stest ; and stmospherie cu=r/t.ct ine tyrass vsive test (O to 10%).

h.

.'esetivity ecefficier. me usure.ent ( 50%, 1'>05 ).

i i

l C.

-lar.t pver calibratier.

%, cs, 50%, 80%, 103%).

1 D.

Flemt raciation level messurez.ents (205, 5,5, 60%,~200%).

' 1

\\

L.

Cnecistry anc radiochemistry tests (20f., 50%, Sc%, 200% ).

/

F.

Process computer measurec variables check (20%, 50%, 80%,100%).

G.

.torpec CFA test ( 50- ).

j 11. Incace-Xenoc oscil'.atic:n test (5CS).

o 1.

Core pcver distribution (D to 100%).

J.

SCE flev =easurement test (20%, 50%, 80%, ICt%).

E.

Cnutdcr.n frce. cutside the control rocxi (5C%).

L.

..oss cf cifsite pever vi h coa.stdovt (10%).

M.

Fartial loss of flow test (50%).

N.

Wir of usin Eeceratcr (1005).

1 1

e r

,s..

y n.4, y-,s.

..e,,

m.-

eme,....-.--,n,_e,.

I

/

Par * * ',

1

. o r :. r.e C --r.

r a t s r. t artu: =_- : Atre :

rie Da- /T r: ' --

• nar n

• /_41_d *=.**t

_1__rc :- _ -

.r.e J rycse of n_F test v a. - te:

(li Verify the y rcrer c; erat ien cf.urbir.e centrole, cer.erater controls,

ana su; port rysters.

(. 3 orice, tr.e t.:Tir.

e er, erat er t e tout 1% power.

(a, Verify tr.e proper cyeration of -.e turcine tyrass /atnospherie dump valve co:.trc. sy r.t ecs.

hst ResC ts Reactor power was increasec by cilut ion frc= 10-ki p er to apprcxinately 10, power.

'.he tu.rt ir.e tyrsss valves were oTer.ed as the reacter ocver increasec to =ain.ain reactor coolan system te=perature e.t 532 F.

At approximately 3. p.over, the a,tr.ospherie du=p valves were used te hold reacter coolant syste= temperature eOnstant while mincr = sir.tena=ce was performec. Tnis initial ope 7ation cf the turbine bypass and atmespheric du=p valves was satis factory.

Initial cperation and startup of the turbine e,enerator was conducted in accorcance with Of LIA and included ne followinc.:

(1) Laergency te>rin oil pu=p autcr:stic start test.

(2) Tri; test cf all reneat stops, e overnor, and interceptor va ves.

l (3) Cneck cf turtine sha*t eccentricity a=plitude.

(43 :*ain turbine lute oil systen operaticnal check.

1 I

( ';, Aatc=atic staat test cf electrc-hydraulie centrol syste= pu..).

( t.1

". e s t. cf overspees prcte. ion :c=trclier.

j (T; Generator voltage rerZater te..

Cc:.01:sien::

. tartar 2:.c tri; testint of the turbine renerater var ec=pleted satisfacterily a r.c it vu p ara 11elec an 1:ned te _inimu= lea:1 in accordar.ce with OI L3A t:.en Icaa-c.: to ' J po er ir. ac corcar.re with OP-2.

Initial cpera. ion of tha tartir.e tyra s anc at=ospherie dum; ralves was satisfactory.

l l

e

/

1 l Fora 54 li i

I 3.2 Feaativity Coefficient Measureenent s Purrose

'l A variat,le Tavg test was perfert:ed at 5C% and 100% power to deter:mine the iscinermal temperature and pcVer coefficients at these power levels.

Test Fes dts The 50- and 100, variable Tavg tests were performed with Creup 5 at 103 inenes witndravn.

The isotter-.a1 te==perature coefficient was measured by raising tse generater output 10 MVe and hcidinr 4 T pcver constant by inserting CLA 5-1 while Teold dropped approximately IC'F.

Generator output was tnen Icvereo to its original level and CEA 5-1 was used to hold A T rover cor.stant while Teold returned to its original value. The power coefficient was measured by de:reasi=g the generator output by about 34 MWe and inserting CIA 5-1 to hold Tavg constact. Generator output was then

}

raised to its original value a=d CEA 5-1 was withdrawn to hold Tavg constant I

until d T returned to its original value.

These cycles were repeated at l

least three times for each coefficie=t acasurement and an averamte was taken j

i of the resulting data.

The measured and predicted values for the temperature and power coefficient at each power level are shown in Table 5 2-1.

l 1

/

Conclusices Tne measured values for the Isethermal tearperature coeffielent and Power 1

Coefficient agreed with the predicted values and were well within the

]

acceptance criteria.

l "JLBLE 'i.2-1 5

i REACTIVITY CCE~FFICD1'T >'EASIREMENTS Noctinal Peactor Isothermal Temperatura Power Ccefficient Power O.s Coefficie=t(x1G-* e /"F)

(xlO" F /$ power) l Measured Predie ed Measured Predicted 50

.11

.11 1 1

-1.07

-1.e4 13 100

.28

.21 1 1 937 99 1 3

/

F age 57 5.,, Plant Pobe r f al. cratien Par; or.-

ine purpose cf tr.e test was to:

(1) Deter =ir.e cere therr_al never ty means of a secondary heat halance.

(2) Verify t.at plant co=puter ccre thermal rever calculations were 1.n agreement vita hand calculatiens.

(3) Detercice tre tnermal output of the reactor at low power using a prir.ary system calorimetric.

(L) Perform ;relir:inary calitration of the nuclear power and si T power safety ci::.ancels.

Test Results A hand calculation of the plant secondary therzal output was performed at the 205 power plateau and compared with the computer ccre therinal outputt 1

calculation. Tie computer indicated a power cf 592.7 Mrith which agreed s

closely with the hand calculation of 591.2.Wt.h.

At approximately 5%

reactor power, the Reactor Protective System (RPS) Power Channels were calibrated using the J T power calculation. This calibration was repeated at a reactor power of 8%.

At 10% reactor power all safety channel pcwer instruments were within +.5% of Res A T power. Calibration of the pcwer range ir.strument channels was also perfoezed at the 50%. 60% and 100%

plateaus.

Conclusions The bana calculation of the plant seecndary thermal output was in close agreement with the plant eenputer value. Calitratiec cf the power ranse safety and control channels was aceceplished at each major test plates =.

l

N <- P

'. l.

Cr.ieldir.g Lifertiver.ess and F ant P s: t iti er. > vel t i

Parrose he test was conducted to eva'.uate the adequacy of plant radiatioc shielding, to determine radiation levels at varying power levels in the plant, and to deterv.ine radioactivity build :zp in specific components and systems.

Test Fesults i

Radimiten surveys were conducted prior to initial criticality and at the 205, 50%, 80%, and 100% power leve'_s.

A few areas inside Unit 2 containment in-dicated radiation levels above IR/hr. These were in the area below the steam generator cold les primary piping.

• tis area is below the secondary sh eld and is not nor= ally accessible duri=s operation. Neutron dose rates inside Unit 2 container...; ranged fronc 5 to 500 nR/hr. A temporary neutron shield was installed above the annulurs between the primary shield and the reactor vessel flange prior to initial. criticality. Additional shielding was also installed around the equipment access hatch, the emergency personnel hatch, I

and the reactor cavity necess hole at the bottom of the primary shield.

here were no areas in the con-tainment that 1.udicated unexpected high levels of radiaiton. A neutres dose rate of.2 MR/h.r was measured outside the Unit 2 containment in the area abowe the equipment access hatch. No measurable dose levels were observed in any of the rensining uncontrolled access areas surveyed.

Radiation levels sneasured in tSe auxiliary building did not exceed expected levels. Chemical and volume control, system piping and components were the

==fn contributors to the measu: red levels.

Cocclusions he data indicates that the neutron shield installed in Unit 2 is as effective 1

as the shield installed in Unit 1.

The area around the equipment batch out-side Unit 2 containnent exhibited jtzst measurable' neutron dose rates of about

.2 mR/hr. Bese levels are ecznsidered acceptable since the total dose con-tinues to be well below the F52R criteria of.5 mR/hr in the uncontrolled

~

access areas. All areas surve:yed exhibited 60se rates that are consistent with criteria presented in FSAR, se = tion 11.2.1.

~

Page %

.s. k Jr.em i s t ry__hsa kaatcena=irtre

.-ets st Pcwe r i%rrct e j

Tne pur;,ose for con:2:tir.g en _:stry and radici:hamistry terts at power is to correlate ectrosicc sate ar.: setivity cuild:up witn power level, to maintmic pr;, mary an: secondary netistry in a:-rordance wit: Calvert Cliffs radiation s:.2 chemist.ry proce: ras ar.

te verify that moisture carryover from eacn s ea= esenerstor is 1-:s tnan. ~ 'L.

fest S.sults Prima:y anc secor.cary chemistry tests we-- ecccueted at the 2C%, 50%, 80%,

and 1 L. pcser p;atens and pri=sry coolant an:alysis was p*-rformad at the j

senedwea trips frce these plateaus. The tests performed included corrosion j

and fission product ar.alysis

• _dthia= tuild-up and volume control tank gas l

analysis.

"'able 5.5-1 shows t=,e er.emist:y tests performed at the various power levels.

i The routine chemistry tests oc stes= renerater condensate, feedwater and main steam systems snowec tha. One=istry was maintained within procedural li=its or corrective action was taken as necessary to achieve acceptable

(

ecnaiticca.

'Ihe =oisture carryever tests were inconclusive because steam 1

renera.or imp rities never exce-ded 1000 times the minimum detectable concertratices vr.ich are necesssry to perfom carryover testing.

Corrosion and fissio.2 product t alldup tests ve:e conducted for informational purposes only. Corrosion produ. t tests consis.ad of an anm7ysis of corrosica:

product concer.trations, a go mm sean, an.: a microscopie examination of a filter sample from tne reacter roolant system.

The specific concentrations of Irce, Chrceium, a== Kickel r-mained below LC ppb during EDT.

Ca==a scans of filter samples revealed no u= expected types or activity levels of corrosicr. procucts. Microscopir examination of these samples showed only a few icent-fiable i=p rities.

Pri=sry coolant sa=ples wer-analyzed for fissice. procuet tuildur durin.r sll test Fower '.evels and a-tivity levels re=aicses te!.ow 10-2 41 fee.

Fissier. precuet activity level was predominately

'cdines teesase no C'.*CC icn ex::.anrer was inse-vice when the samples were

.axen.

Laily itt.ii = sa=ples fro = the reactor cerolant system filter inlet were snalyz.ca to =cr. iter :.ithian tai' dur.

The averare rate of traildur was about

. h pp=/ cay vt.ich is sliently lower than

.he pradicted rate.

Zamples were crawn froc tha velee contre; tank vapor space to assure hydrore.

crer pressure anc to determine fission Fu.s levels. The fission asses identified 2r. the

'.*C sre snown on latle 5.5-2.

Fem: tor coolant systen sa=ples were takan ans an,aly ec tefere and after asch c f the seheruled trips. No sirnificant

..snge in tr_

analysis was note: fc levi =r the f05, 505, aos c 5 trirs. Liouid s a=ple: were crawn 1/2 hcur ter re anc a'ter tha 100% trip.

'"h e s a=p le s siffere cnly slict.tly in fissi-. pro.!ucts but the after trip sa ples idtntified crros en prxuets (Cr-51, Mo

.a., F -59, sna M:-5k) not previously identified.

A " burst of ccrresten product: ofter. acemmpanies a trip an: isentification c1 these isctc;-s was expected.

Cenel.s: ns

""he ene=istry ar.d raciochemistn tests ce.auete : at pcwer indicated that primary ano se or.sary ene=istry met ac cept ance :riteria and the che=istry I i.

i 1_ ~ J Z M_T -

~

l

s Pase r m control systems coJd te crerased ar desif:LeJ ir. monitori=.c anc ttaintsir.ir.

primary ar.a =ccor.ca.ry syste=s s-ithir. specified limits whe= require:1.

/

/

a 6

I ir

l TABLE 5 5-1 CHMISTRY AND RADIOCHmTSTRY TFSTS AT POWER i

l i

TEST POWr8 LEVEL (1) l 20 50 BC 100 TRAF.IEYP i

Corrosion Product Testing X

X X

X 20f, S0T, 80f,100T Trip i

Fission Product Buildup X

X X

X POT, 'iOT, 80%,100% Trip Lithium Buildup X

X X

X VCT Gas Testing X

X X

X I

I 1

1 l

4 s

1 l

l i

J i

i I

l 3

4 9

l I

~

- --. - - - - - - - - - - - - - = - - - - -, - -..,

.g p,,.y

_y.,.

/

o

y e;

TAiLE 5. !-2 vc r ssIon CASES POWER LEVEL (")

M ACTIVITY LEVEL (uci/ce) 20 h -95 1 36 x 104 Ze-135 5 56 x 10-h e-87 3 35 x 10-k A~hl T.93 x 10-2 4

tr-8B 1 90 x 10 50 Ze-133 5 27 x 10 h tr-85 8.05 x 10 k Ze-135 19h x 10-3 K_~BT 1.16 x 10-3 Ar-h1 3.33 x 10-1 Kr-88 1.20 x 10-3 t'

100 Ze-133 2.23 x 10-3 i

Kr-85 1 5L x 10-3 4

Ie-135 L.58 x 10-3 N1 3.18 x 10-1 i

Kr-88 2.hk x 10-3 4

i 3

t..

1 J

h i'

J i

4'

?.

'.. i,..."

-+

-.,.,~,,.

K....{...

-:.: n-

~

s ve f\\

a

'. t.

Irsee:c s. ater 'nasur-i vari s: ' e: ?..ce e

! 41 C se Tne purpose cf t _is test was to com are p rocess ccep'ater remaings with meter reaci=.gs fcr selected para =eters.

Test resuit Readir.Es frem meter incientions were ecmpared with their ectrespondir.a compu.er pcints a. reactor pcuer levels cf 200. 50%, 30%, ar.d 100%.

"'he compa-isor. incluces reactor pcwer, reacter ceclant system tecu ratur=s, presse es, levels, and flows; OLA pcsitiens, teronometer, reactor prctective systez ir.citatio= s stem. cenerator pressures, stear. generator icvels, and feeswater T;cws. Most cf the :ceputer points =Areed closely with their correspor.cir.g meter incications.

Conclas ir>ne The process computer will record val.;es in elese agreament with the zeter indication of the particular plar.t arameter.

"'he cca:puter readouts can be used to mos=itor parar.eters tr.at are critical to plact operation.

~

NC ^

~~ Z '

M-__

-jmu'.GP-Ed !

~

i Para 6%

%( " rcrpe s _ CrA T. s t harroce e

i Tae purpcse of this test is to verify the ability of the execre detectors to sense a dropped CIA anc to : monitor plant response to a step load chanse at power.

• est Results "l'vo rod drop tests were conduetca. CEA 5-58, a low reactivity worth sinrle i

red on the core perip6ery, was used."or the first drop. CEA 5-1, the histest smorth CLA, was usec fcr the second drop.

Both rods were dropped with the reactor at SCS power by opening each 210 volt circuit breaker at the coil power programmer. Ordy one charnel cf excore r=aelear instrumentation imdicat-d a rc,d drc1p alarm vnen CEA 5-53 was drorped.

""hree channels indi-cated a roc drop when CEA 5-1 was dropped. Remeter power decreased apprc:xi-notely one percent when CLA 5-54 was dropped acei riant conditions varied only slightly. Beactor power decreased approximately 95 following the drop or CEA F1 and plant conditices stabilized within two minutes following the droF-hrbir.e load was Icwered by reduction of the valve position limiter. This mhled turbite first stace pressure to rapidly stabilize followins the CIA drty.

Pla=t conditions folleving the drop of CEA 5-1 are shown in Figures 5.7-1 assi 3 7-2.

i Conclusiens I

Both rou drop tests produced a rod drop indication. CEA 5-1 produced a rod drop indication on three chm =nels of exeore nuclear instrumentation. Only one enar.nel of nuclear instr anectatice indicated a rod drop alarz vten CEA 5-58 was dropped. In the oriEir.al design this alara was used to initiate a turbine ruchack and cause an autentie withdrawal prohibit. Since the design no longer inel ades a turbine rar.back and since t? e reactor is not operated j

im autocatie, it is not necersary to have redundar.t alarm actuation for a g

dropped CEA. The plant respended satisfactorily to the step loa! change.

t

~

1 I

i t

a s

i

$1

.f '

m o

I i l

l 1

MM" MM'TFF Mi'$ ' ' )

CALVERT CLIFFS UNIT 2

CEA DRCP TEST ROS PRESSURE 224o i

l i

l t.

_ I

4. -

g 2 tic t

n 1

,i- -

j j

n 4

i

- 22eo r

r-

'j' e

cL w

}

I w fise l

i

-~

8

/

cr 3

i s

m W fios -

j j

j at I

t I

CL t

f

.i i

i i

s i z e -.

.ie a

is 2s se 4e se se to es se 1%e sie TIME (SECCNDS)

I STEAM GENERATOR PRESSURE see

.........;.........-. f o

- ese -

m v

........ 4 a

ese.

wcr 3

.........~.-- -.

m w o2e cm i

^

L I

see

.io in rs se ao so so 7:

e:

so iso iso TIME (SECCNDS)

FI GURE 5.7-1

~

4 a

CALVERT CL2FFS U%IT 2

CEA DRCP TEST NUCLEAR PCNER 3g 48-c-

.... 1 48 aW I

I

3...

....'.3 w

.l.

r o

Ca.

I 42

.l t

/

48

. ~ ~..... - -. -

1 i

38 18 0

10 28 30 se St EC 70 SC 50 1:C s9C TIME (SECON05)

I RCS TEMPERATURES i

570 e

1

'

....;.e.:...

.:... t.....

j; 3

3 588

^

w

' '-- ^ :

a:

'~S' j

ISI E#I; 3

I cr E

^ g;1.;

'.41.

.. _:: ;;1.

w

l:.

a, 34.

,."..l.-..

a:ST CSL3 i

w

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e.

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.e e

-lC 0

1C 20 3D Af a3 gg yg gg gg ygg g;p TIMC (SECONDS)

. I 9

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h, r

I I

FIGURE 5./-2

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t

,, _ [ yg eyy,e 4 ;s 9

..g a-W-

^

~ *. py ;.

p.

l p:zr-iT

.c.c ee... :.. :.

t-il h :en ~. ~ n t L25.1::..

6ne pari ose c f tnis test vs.: to evalaste the abilitr of tn* reactor core to self-a cap axial Xen.on oscillations initiated by C'I.A ecrement.

Tent Result:

A Xeno*n transient was indure: in ne ore during the 50% ;cver test plateau.

Cz.A Oro2p

's was siluted te apprezimately 6c inches withdruvn for six (6)

s. oars tr.en toratea to all rceds cut.

This permitted a buildup of Xenon ir.

the upper regior. of t:- cora.

'.~nen t=e rods were terated to all rods out,

a Xenor. oscillat.cn eccmenced.

~~nis es;iliatien was monitored by the plant eceputer ans by.he reacter prcie:tiva systam calculated ralue of the Axial Onape Incex (A01,.

The eseillatien e:curres along tne axial dimension of tLe core ar.d the nagr.ituse of each peak cecreased with time.

The ma.zimum licea-heat rate measured by the II.0A progra= sr.d corrected for =ncertainties was 7.02 kv/ft. The mini =:am DLG vas 11. '"6. bcth values ver-well within the acceptance criteria. The St.afe Ar.nealina Facter (CAF) for each power rac.re safety channel was also measarec.

Th-CAF corrects the detector cirnal t.:3 account for tne sistance frt:x: the detector to the remetor core and corree.s for the signal produced ir. tne upMr :etector by neutrons fror-the lover part of the core and the sirnal producad ir. the IcVer det-ctor by neutrons "rcut the upper part of the core. The SAF is determined by plottinc the INCA A I versus tne ACI (EXT)

• rem the reactor protective system.

'"he slope of th.d s plot is the shape annealing factor.

~he ratio of tte difference in power generated in the lower and spper halves of the core to total core power is the Axial Shape Index. A plot cf the ASI and Feak linear heat rate were main".ained througnout the Xenon transient.

Cor. :l as ior.s Tne axial Xenon oscillatien was self-ca=penine to a riable pcver distritu-ion ana tne DNbR and peak linear heat rate durire the oscillatier. re=ained vi him the acceptance li=its.

As tr asured scape annealing factors vere incorpersted into reactor ;rctective system setpoints.

I l

\\

j l

l l

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

~.

i

/

/

I'

}

Par ~ th l

I s. ')

ore ic.sr..ntri: tion Puri ce.;

Core power cistribution merasure.cnts were perfor ec du-ins: EPT to verify that fuel assent.ly pcver fractions, axial pcver distribu.icn, peak linear meat rates, anu DNbR's were within acceptetle limits.

l' Tne aereptance criteria fcr the core rever distribution are listed below:

'l (1) F el ssse=tiv rever fraction: Puel asse=bly power fractic= is defined as tr.e ratio of the avera. e lir. car heat' rate (LIIR) in a fu,al assembly to tr.e everage 'liR over tee entire core.' The measured value shall be witr.in + 5, of the predictec value.

(2) A2ial rewer distribution:

"*ne =casured core average axial power

{

l distribution shall te cor: pared with the predicted distribu ion for general agree: men..

a

(.$ ) Peak Linear heat Fate It shall always te less thar. Ik.8 kv/ft. as i

s irdicated by INCA program.

I l

(4) DEbR: It shall always be greater.han 1.3.

\\

Test Re ults Core power distributiens are calculated by analysis of the incore detector readings. Their locations are shown in Figure 5 9-1.

This analysis is t

performed by two plant computer prcgrams.

"he first prog wn, autonatically l

converts t.be voltage signal frca the detector to the correct neutron flux level. Tae !ccore analysis precari (DCA) converts,neutree flux levels and various other' reactor parameters, and om, de=and calculates several incore data.

Tne I!CA proEram assumes eighth core symmetry. Puel anse:sbly power

~

Traction at tce 20*., 53%, 80% a d 105 test pcver levels calculated by the DCA progran 's ccacpared with predicted values in Figures 5.9-2 through 5 9-5 In every case the measured power frsetion is well within the acceptance l

limits.

Axia'. power distributions were calculated by the DCA prorram at each 6-st power level.

Figures 5.94 throur.h 5.9-9 show that t!y= ccgrison tets-en predicted and measured dis.ributions at each power level is in close i

agreement.

T e aepartare from rueleate Leiline ratio (D*;ER) was calculated

)

using t:e '4-3 correlation and i= ne case 'did tne DNEP. drop te'ov the acceptacee limit of 1.3.

The measure:

.EB never exceeded the acceptance limit of 18..o av/ft.- All reasurenen c va

• prformed at steady state Xenon JI I conditicns with all rear,s out.

The resvi:,s ire shown in Table 5 9-1.

}

I i

Cenelasiens The peak LER si: net exceea.1k.5 kw/ft., the =inimum D.TbF van rreater than i

1.3 and tne axial pcuer distribution and fuel assembly power fractions vere vithin acceptance limits.

i

'I I

.i

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l

/

Mt T

l

)

l TABLE 5.9 1 CORE POWER DISTRIBtfff0N SIM(ARY Measured Acceptance Limits Reactor Power 14 vel 205 50%

A01 100%

1 6

4 Minimum DNBR Not 12.9k T.1 5.0L Oronter than 1.1 Measured 1

Maximum IJfR (Kv/ft) 2.28 5.0L 9 90 10.?7 tess than ih.8 I

Largest Difference Between Measured and Predicted Fuel Assembly 1.0%

?.bf 32%

3.05 tess than 55 Power Traction Axist povar distrthiittnn Peak 1.?06 1.275 1.273 1.260 Predicted

[;}

j Po%

50%

40%

100%

i t

1.275 1.255 1.255 1.282

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?

CALVERT CLIFFS UNIT 2

'BOL.

IST CYCLE POWER FRACTION 1

DATE/ TIME 12/10/76 2049 POWER 599.4 WW th tNCA BURN-UP 9.5 MWD / MTu INCA x xyy R0D POSITION ARO 100% X f"E

, y+

! XX.X XENON ECuluBRiuM FUEL TYPE INLET TEMP. (Tc) 533.78F 0.694 1.064

' ~'

~ - - -

PRIMARY PRES. 22 4 AdP.Sl A 0.3 1.2 C

C BORON 0NETER 898 PPW PEAK LHR ?.28 kW/FT 0 6ti 1034 1.005 1.030 j

40.5

+i.1 B

MIN. ONRR NO T. C AL CUL ATE D...

g g

g 3

REACTOR POWER,?0,,*/o-0.827 0 9 8.2 1.047 1.0-67 1.093 40.8

+ 0. 2

- 1. 6

2. 3 2.1 1

t 8

A 8

A 0.563 1.030 1.077 1.0 92 1.113 1.124

.. 't. 7

. 0.1 0.0 l.3

. 0. 2 i.3 i

0.' 34 6

c.

A 8

A B

A

+ 0. 5 1.004 1.027 1.'060 1.0 99 1.082 1.12 9 1.l29 C

el 3

+ 2. 3 0.2

+ 0. 3 0.0

0. 9

1. 4 0.832 g,

g a

g a

g a

+.0 0.963 1.040

-1.089 1.105 1.121 1.127 1.128 f.131 5

. 2. 9_

+ 1.0 40.5 l.0

-0.9

- 1. 5 0.1 f.3 8

A B,_

y A

B "A

B

'l A

I i

FIGURE 5.9-?

i

CALVERT CLIFFS UNIT 2 w

i 80L.

IST CYCLE F

POWER FRACTION i

DATE/ TIME 12/21/ 76 - 1522 s

POWER.1305;8 MW1h_.

[

p INCA BURN-UP 84 i Mw0/MTU 5

INCA X XXX ROD POSITION ARO PREDICT-lNCA 1004 X g

t xx x 2

XENON _ EQUIL.!3RIUM -

-l--~

fuel TYPE PRID4CI

{

U INLFT TEMP, (Tc) 540 7 i g-v689 1 056 p

PRIMARY PRES. 2253))SI A

. i,9

.i 7

[

BORONGMETER 795 PPM C

C PEAR LHR 5 04 kW/p; 0 603 1026 0 991 1.033 17

-19 I

401 705 MIN. DNBR NOT CALCULATED

({-

C C

B A

REACTOR POWER 50 /o 0 818 09/7 1045 1.0 76 1804 I

f

-l.1 07

-1.$

- 0.9

- 1. 2 i

C B

A B

A E

E-0.950 1.023 1.018 1.10 0 1121 1140 i-

+ 0.6 0.4

+0 6 0.6

.I 6 03 0 622 C+

A B

A B

A l

g 1.086 1.020 1.081 1.112 1099 l 151 1.14 9

~

i

-02

+24

+00

+ 1.1

+ 10

+07

+00 0.808 l

C+

B l

A B

A B

A i

b 0 945 1.033 1.093 1.114 1.133 I 145 1149

! 151 i t

2 C

N 1

i I

- 0. 2

+ 0. 2

+ 1.4

-01

+ l.6

+ 0.0

+ 1. 6

+ 0.3 B

A B

A B

A B

A

=

i e-l

{

FIGURE 5.9-3

---

._.h 1

CALVERT CLIFFS UNIT 2 BOL.

IST CYCLE POWER FRACTION DATf/ TIME I/5/ 77 I40?

P0WER _20 4 7 9 uw th INCA BURN-UP 325 8 U AD/ Wid. _

s

~

ROD POSill0N ARO INCA 1 xxx PREDICT-INCA XENON 10UlllBkluy 10% X PREDIC1 XX FUEL TYPE

' INLLI 1EMP. (ic ) ~54 3.l o f "'~

0 676 1.047

-~ ~

PhiMARY PRES. 2253.5 PSI A. _,

,p,

,7. 3 BORONOMETFR 7 3 7 PI'W C

C PEAK LHR B.05_3W/FT 0.590 1.014 0.9 97 1036 Il F.?

0.6 0.0 MIN. DNBR.T.10 c

e g

3 REACTOR POWER 80 */o 0.803 0966 1043 1.086 1.l l 5

. l. 4 08

.? 0 l I,

.l.b l

C B

A B

A 0.934 1.013 1.083 1.109 1.146 1158

0. 5 0.3

.03 09 e 1. 5 0.0

~

C+

A B

A B

A 3

l.068 1.014 1.084 1127 I l19 1.174 1170 C

0.782 C+

8 A

B A

B A

1. 8 0 929 1.025 l l00 1 126 1155 1165 l173 1 171 C

+ 1.1

+ 0. 8

+l.0

0. 4

1. 6 0.4 2.2

.0B 8

A 8

A B

A B

A flGURf 5,9-4

CALVERT CYCLE UNIT 2 BOL.

IST CYCLE POWER Fi!AClION l

f w

DATE/T IME J2j/ 77 1321 POWER 2546 I Wwth INCA BURN-UP 628 C WWD/ MTU ROD P051110N AR0 CA rxxx p,

100% X X E NON [0 UI L.I.B l? t 0 M._ _ _..._.._

PR[pl(1 ---

XX X 4

+

INLET TEMP (Tc) $44.6' F FUEL TYPF

-'m e

I O

PRIMARY PRES. 22 53.1 PSIA f

03

.i s BORONOMETER 705 PPy C

C l

0.579 1005 0496 1 038 PEAK LHR 10. 2 7 K W/fT.

-l 2

-l 7

- 0. 2

-02 Mir DNBR _5.04 C

C 8

A REACTOR POWLR 100 *L 0797 0 960 1041 i 0 9.'.

I IP I

~

)

1

' 10

-05

-19

- 2.0

-20

'f C

R A

B A

0926 1.009 1089 l 114 1.16 0 1.16 9 l

+05 03

+0i 14

+0 9 0.4 0 $97 j

g 3

g 3

g

~

l 061 1 021 1086 1137 1130 1 191 I 115 l

C

+0A

+ 2.9 03

+04

-03

+ 1.1

+ 0. 7 0 765 C

B A

B A

B A

+2 4 0.919 1.024 1106 1.132 1.170 i 173 1.178 i 72

+16

+07

+07 07

+1l

+03

+ 2.6

+13 8

A B

A B

A B

A I

FIGURE 5.9-5 l

o l

l t

i l

{

~

i

-- 7

=

i i

e 6

i a

l I

l 6

s

~

I i

l z

L3 I

i e

1 l

g W

i o

e K

~---- h

- ~ ~

d spy i

l w

w x

an u

~

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v Z

o e

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,i

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2 3

I I

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w v" O

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w g

13 i

r

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3 p_

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Q l

l v =>

au Q

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e a

o.

"o l

o,

'a u

cx l

3 2

w m.

i g

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_c o

i

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=

a a

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s' m

i I

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I i

l i

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l

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=

\\

o o

o a

o

'( cd /d ) 11012 7d D N 13 V 3 d T viry I

w,.._.-_.

g gm_

l L.

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CALVERT CLIFFS UNIT 2

BOL 1st CYCLE l

50% AXIAL POWER DISTRIBUTION VS. CORE HEIGHT is 1

i' 3 OduOO R.,

i i

)

l erasuais C09

-/

\\

l F

' % PRIDitif0 l

e E

O l

"0 I

.1 l

~

a N

I i

< 0.5 f

I

(

0.3

(

i

! l i

0, s

- -..... 1 - - -....

\\

1 134.7 123.0 109.4 95 7 81 0 44.4 54 7 41 0 77 1 13 7 0

DISTANCE ABOVE CORE BOTTOM (INCHES) d flGUR E 5.9-7 Ae

. m.,,,,...,,.~,-.3.- e.

CALVERT CLIFFS UNIT 2

BOL lit CYCLE 80'/. AXIAL POWER DISTRIBUTION VS. CORE HEIGHT I 5

~ '

- ~ ~

~ ~ - ~ ~ - ' "

- - ' ' ' - - ' - ~ ~ ~ ~ ' '

' - - - - -~~~ '

~~~ ~ --

' - ~ ' - '

g'3

_..0 0 Il O O O O ()

EI

~

/

g g,y a.

2 as 0.9

- ~ ~ ' * '

l'-'-

~~

~

~'

1 l

l w

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Q g,,,

t, L

\\

\\ stAIU8tB i

I

0.5

~

0.3 0.1 nii in o iou ei.

no su se r oo ni ni o

J DISTANCE ABOVE CORE BOTTOM (INCHES)

FIGURE 5.9 8

- 3 1

1

(

4 CALVERI CLIFFS UNIT 2

{

BOL lit CYCLE 100'/. AXI AL POWER DISTRIBUTION VS. CORE HEIGHT I3 o dT ~5 a ti I

~

~

l t)

/

N

.e

)

i I

j D

1 me l

j O

~ - - '~ ~ - - -

'- ~ - ' - - '

--~ - - - - - - - - ' -

o.9

-~ -

\\ PetglCl[$

f o

l i

()

l 2

i 1

y0.7 w

\\

l S-OfA!UllB E 05 i

4 N

1 I

l

,j i

j J

t

{

ol-nu ou io u

,s.:

n.o su so n.o su or o

i f.

DIST ANCE ABOVE CORE BOTTOM (INCHES)

J j FIGURE 5 9 9 il il L

...___-._.,.m.,.

4

/

Pape

~'9

'.1L F '.' F i v "c a s a ree.er.

Zest P.rrose Tr.is test was perfor=ed tc mess.re tr.e tc al r~setor ecolar.. flow uning te=perat ;.re rise meress tr.e reseter ecre with four ra=_etor roolant pt.:=ps rur.r. inE.

)

T st F.ec Its Total reactor cociar.: flow was calci.im* ed by messurinr reaeter coola=t temperat re rise across the ccre ans equatine tnis therr.a1 cutput to the plant se:or.d arv tr.ernal c.: put.

Ter. (10) sets of tem cratu.re data ve re recorses over a ;eriod of cne 1.c.ar a. the 20%, 50%. SC5 and 100% power z

plateaus. The results are as fcilows:

l Rt.AC~OR M.TR (1' M*.

F.0"' ( MS r-- )

20 5.32 l

50 L.16

{

00 L.06

\\

100 L.10 I

Ccnelus Ier.s j

i The total calculated flow is well atcre tne acceptance criteria of 3.7 a 10) gpm reactor coolant syster flow with four reacter coolant ramps I

operating.

j l

i i

1 r

i

..y

l s w..

3 5.11 Onut.:own fram J t si.:e Tor.t re '. um Purnese The purpose of tr.is test is to demer. strate that the reactor can te trirpec and the plant ma'etained in tne not stutdovr. condition frem outside tha control room.

I Test f<es fy 3

The reactor was manually tripped free. 50% power by opening the reactor trip circuit treaxers in the ea:'e a reaatn. room.

Two operating crews were useo to perform tne test. The emerrency cr'ev left the contrel room after verifying the trip and procee.1ec tc their ecnergency stations to establish local control of the RCr and steam renerators. The cornal crew remainec l

in the control room to resume norma'. control of the plant, had a need arisan to terminate the test.

Oteam cer.erater levels were maintained by local control of the auxiliary feecvater system. FCS temperature was stabilized and controlled at about 510*F.

31s was lower than originally intended l

because an atmospberic cump valve did mot fully close on automatic signal.

RCS pressure also stabilized at a va.lue lower thsn intended but no problems were encountered maintair.ing this pressure.

Plant parameters following the l

trip are shown in Figures 5.11-1 throus.h 5 11-3.

Canclusiens The plant was shut down and maintair.ed in hot shutcown from outside the control reon successfully. RCS pressure and ter.perature were stabilized lower tnan specified while all otner test results were within the acceptance J

criteria.

I j

i l

i d

1 l

g gy' Wit, M,

- M A.< M U E C

,. ~...

i CALVERT CLIFFS UNIT 2

S/O FRCM.CUTSIDE CONTRCL RCOM NUCLEAR PCWER se i

7--.-.-.-

i i

i l

i i

se-I i

i i

6 i

i I.

t.

[

se I

. t.

i t

ag 1

l r

t

,,i j

a

.. _....i.

t i

g kJ l

E t*

l l

r' t

a-i

. {.

?.

+

e j

l l

~

i I

[

j 1

<g e

l t

]

g

.s e 3

.i e o

to to ao ao se ea 7o ee se tos sia T4WE (SECONDS)

RCS T OCLO i

?

r--- -- * - - - - r ses 3

.........l'...

I I

..l 4

.t....

g t

I-i see

..l..

,1 8

.j

... )

I l-l u.

535

\\.

.... z.

...... l'

. u.

.u-l i

-l-w t

m $3e

...... l.. : _..

.....l....;...-.l.

3 g

... u..

m l

.=.;

j-n..ag _;. _. :...

't t

ocw sr$

i; l

0-2

. !...:: n: _. :._...

.. u.:. j.......j. :. :._.-...

1 I

i i

1 \\

l' j

-l

...... { :..... :.... : uiu.;aa..

1 i

1 g

i g

sis

:)

i

. :=

..t.,_

g.

ses

.se o

is se ao ao so se to se en ice its ilWE (SEC0hDS)

F I GURE 5.11 -1

.y,,.

x

..e -... w.3

e.,

CALVERT CLIFFS UNIT 2

S/D FROM OUTSIDE CONTROL ROOM I

l RCS PRESSURE 2 2 3 e

.g.

l..:

g.

. Ql........

e n~.

a.

n : :-

t

'su

..:.:.i.

r I --

g q.-u y:--

n:' n:.:..

\\ ;:.;

=

nn: -

:: : T l

1 22ee n

... _[;;,.;,_,[

n.

.l..l.:.

r...

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S/D FRCM OUTSIDE CONTRCL RCOM PRESSURIZER LEVEL 2as.

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• cce cf 'Ffoite Fever. i th Cent :cwn Purtc:e The gurpose of performing the loss of offsite power trip test was to accomplisn the follovira objectives:

(1)

• to evaluate system reliability during a tots 1 loss of AC power.

(2) To desaonstrate the correct operation cf the turbine generator assisted reactor coolant pump coastdown circuitry.

(3) To obtain reactor coolant system natural circulation data folicwing a total loss of flow.

i Test Results The 50C.rN ring bus was split so that all Unit 2 plant los *s would be isolated from the grid and depecd on internal electrical sources noch as batteries, j

diesel generators, and turbine generator assisted RCP coastdown.

1 The reactgr was manually tripped from bO$ poser using the trip pushbuttons f

on the reactor protective system panels.

""he reactor coolant pumps remained j

connected to the main generator for 20 seconds. During a turbine renerator assistea coastdown the reactor coolant pumps are povered by the main 5

generator until the cor.nection is terminated by less than 605 generator out-put voltage or at 20 seconds following the trip whichever occurs first.

both main feedvater control valves ramped closed in 12 seconds and steen j

generator pressure was maintained below 1000 psia. The reactor coolant system was maintainec below 2500 psia and the stema dump and bypass control systems returned the plant to hot shutdown.

I 1

Some control room indication was lost for approximately 8 seconds after the j

trip while diesel generators 12 snd 21 started and loaded. Natural circulation into the h"S was verified 5 minutes after the trip. Evidesce that cooler water was reaching the hot leg was indicated by a drop in Thot. This showed

)

that an initial therma' driving force for natural circulation was established.

l Thot then decreased 16*F in 40 minutes indicating that sufficient natural circ.11stien was occurring. In addition, the temperature difference between the RCS bot and cold legs decreased to ik'F in 10 minutes. plant parameters i

following the trip are shown in Figures 5 11-1 through $.ll L.

i Cone 1asions d

The reactor was placed in a hot shutdown ecedition following a loss of offsite power. A turbine generator assisted coastdevn of the reaetor coolant pumps 4

1 occurrec as desirned. Dsergency diesel gererators and batteries functioned pro;erly by proviaing electrical power to vital instru=entation and equipment.

Natural circulation of the reactor coolant syster. van indicated 5 minutes after the trip and all plant systee.s and parameters respondeo in accordance with the 4

acceptance criteria.

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1 LCSS OF CFFSITE POWER TRIP 1

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Page 89 5.13 Partial Lors of F;cv Trie Tes:

_Purrc :e This test was performed to n: essure pla=t response to a partial less of rea.: tor coolant flev while at pcver.

Tes. Ecsult s The trip was initiated by securing 22A resetor coolant pump with the plant at =ccinal o05 power and with four reactor coolant pumps raing. Recorder tra.:es of various plant param:eters were obtained for the icitial portioc of the transient. A reacter trip on RC3 lov flow occurred approximately one second after the reacter coolant pt.mp was secured. Locer 22 f!cv coasted to a level slightly less than 25% of tctal flow in a period of arproximately 30 seconds. Total flow drc; ped to 77% in the same period.

Loop 21A temperature increased L%efore startin6 a decline.

"he ple.nt was returned to a bot shirtdown condition without excessive transients ic plant parameters.

Figures 5.13-1 through 5.13-k show plant parameters for tvc :ninutes following the reactor trip.

Conclusions A reactor trip was initiates oc a signal of low reactor cociant system "Icw.

Reactor coolant system coastdown occurred as expected and t2:: pla=t respenced in accordance with design.

CALYERT CLIFFS UNIT 2

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I 80% PARTIAL LOSS OF FLOW TRIP RCS LOOP TEMPERATURES

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F GLGE 5.13 -4 i

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o Pare 23 S.14 Trir of Main Generator Pu-cce The Furpose of this test vss to evaluate systes reliabil!ty during a loss of generator load from ncminal 1005 power and to rather powr defect worth, measurement data.

Test Pesults The 500rN ring bus was split so that tripping treaker $52-63 caused Unit 2 main generator to experience a loss of load.

After the plar.t stahilized, a reactor startup ccatmenced i=nnediately so that Xenon reactivity worth could be followed. A embination of reactor coolant system borations, dilutions, ana CEA movements were u. sed to hold the reactor critical at a pov=r level below the pcint nf adding nuclear heat.

By using the reactivity values of' solutie boron worth and CEA positic: that were ceasured during Icv power physics testing, the reactivity worth of the Xenon was folltved. The pcve-defect at 100% was then sneasured by extrarclation of the transient Xenon worth curve tack to the time of the trip.

This extrapolated value equalled

-1.05> 4t' vnich closely agees with the power coefficient censured durins the 10Cf pcver variatie Tave test.

The reactor coolant system was returned to its nominal values without ar.

excessively long cooldown or pressure transient. A turbine trip occurred directly from the loss of load.

Six main steam safety valves lifted following the turbine trip. This was partly caused by the failure of the atmospheric dump and turhine bypass valves to operate in " quick cren" ar.d atmospheric du:np valves to operate in automatic. A defective k-1 relay in the reactor regulating system caused the improper operation of these valves.

All other results were within the acceptance criteria.

Pla=t parsmeters

'following the trip are shown in Ficures 5.11.-l and 5.lk-2.

Conclusions A turbine trip was initiated directly by a loss of lead of the main cenerascr.

The plant was returned to a hot shutdown condition in a satisfact:ry ear.ner within 10 minutes followi=r, a trip from 100% power. Ienen reactivity worte.

ard power defect worth data werecollected.

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CALYERT CLIFFS UNIT 2

100% PCnER GENERATCR TRIP

'N S T E c. M GENERATCR PRESS'JPE

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!n orcer to avoid tre possibility of this protier. in the core.

Calvert C:.iffs Unit 2 core, all fuel assentlies were eff-loced ar1 the assoitlies containir.c t rr.able poison pir.s were returne: te Winaser, Cor.meeticut for me-ificatice.

3.e elapsed tima for ft.el eff-load, =cdification, and reicsd represented a 10 week delay ir. the start ; test program. The off-1 cad sequence is civen in Tatie 6.1.

Uo otr_er e.ajcr dele.ys were experienced during the startup ter. pres ram.

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Tar-M 7A,-- (,, i nrEl Acre SLY tr oc rr.*JE:: E l

STEP 50.

PJEL A. CE'C-1Y ;0.

CEA *;3.

FPOM CDP I.CCA'- CN 26 2A327 C1 279 27 2B258 2711 28 2BC58 2v11 29 2Bo60 2V15 30 2A?57 DS 2v16 31 3C33 2717 32 2AOD7 2n7 33 2B231 2n6 j

3b 2A32k CK 2n5 35 2b325 2T13 36 2AC54 E0 2T11 37 2Bc6S 2T9 38 2AO.5 BD 2T7 39 2B022 2T6 LO 2A0ok 2T5 41 2BO'6 2S5 b2 2AD'3 E3 206 h3 2B013 2S7 kb

?*4012 2S9 j

h5 23036 2S11 46 2A067 2S13 k7 230 7 2S15 I

h8 2AOC" '

Ek 2S16 h9 2B021 2SI-2A05S f 50 D9 2RIT u.

9 g

e a

g g

,.b

.-. - -. -. ~.. - - - -

Page 99 TAS T 0 1 RTEL ASSEW1Y OG4f D EE*UE%0E L

STEP NO.

FUEL ARSTMBLY NO.

CEA NO.

FIM_ CORE __2,0 CAT _OX 51 2B020 2R16 52 2AOk1 CJ 2R15 2R13 53 23038 Sk 2A013 C5 2R11 2R9 55 23035 56 f

2A002 BC 2RT 57 2B057 2R6

/

58 2A021 AH 2RS 255 59,

2B032 60 2037 2N6 61 230L6 257 62 2A032 EK 259 2N11 63 2B009 6k 2A03k ca 2N13 65 2B028 2N15 66 2A035 2N16 67 2B00k 2N17 68 2A017 El 2L17 2L16 69

,2B0h3 70 2A053 CH 2L15 71 220kT 2L13 72 2A001 OL 2L11 2L9 73 23055 Th 2A030 E3 2L7 75 2B053 2L6

$=

z

e.s w -

/

Pare 1 %

TAP /I 6,21

. FUEL ASE5' GLY CPGAD RE0*.T:CE TTTEP KO.

FUF1 ASED'31,Y M).

CEA No.

FP'M COE %.0CATIOtl

.a_..

76 2A02 8 M

2L$

1 l

7T 2.E339

1. 5 2J6 78 2A352

~

i 79 23056 2J!

M nM9 31 N9 el 23079 2J11 32 2A050 C9 2J13 83 2B049 2J15 l

Sh 2AoS9 RJ16 i

l 83 2303h 2J17 1

66 2A003 D8 2c17

)

87 ET:063 2016 l

1 68 2A020 CG 2G15 l

l 09 23Q45 2013 90 2ASk5 C3 Poll E

s

?

91 28037 e

2G9 w

92 2A0h3 5A 207 93 20017 206

{

9k 2Aoid Ar 2c5 95 128078 2P5 i

96 24003 E2 2F(t L

h 97 2R027 2li j

l 98 2A066 2P)

I i

99 isokk Pr17 f

100 2A056

-i-2F13 r

4$5

[

$l 3

E Ag, * /r.

b' c'

h,:. i e -:. C k

y:

. 3

+

= - -

Page 101 TABLE 6.1 FUEL ASEE' GLY OF.". DAD SE2*uTICE STEP NO.

IUEL ASTIE E Y 30.

CEA N3.

FP.OM CORE LOCATIDN 101 23359 2n3 102 a005 E5 2n6 103

Boko 2nT 10k 2AC16 2n7 105 Boll 2n6 106 2AC25 CF 2nS 107 2Bc23 2n3 108 2Ao39 n7 2E11 109 2BOSO 2E9 no 2Ac19 B9 2r7 n1 2Bo6h 2E6 11 2 2A3k2 2E5 113 2B01h 2D5 I

uh 2AC2h B1 236 4

115 2BC2h 2DT 116 2Ac29 BG 2D9 117 3069 2Dn 118 2A051 C7 2D13 119 2B061 2D15 120 2A010 D2 2D16 121 2B012 2D17 122 20105 D7 2C17 123 20005 2B17 12L 2B018 2C16 125 2A031 CE 2C15 l

i I

$ ts#

a Me e

h*

-w.'l

^

i l

Para 10P TABLE 6.1 FUEL ASSEGLY OF.F_.OAD SCUE';cE STEP NO.

l'JEL ASSCGLY NO.

CIA NO.

FPOV CORE LOCA'"IO*I 126 2B051 2Cl3 127 2A062 3L 2C11 128 2B073 2C9 129 2A065 3S 2C7 130 23015 2C6 131 2C103 AD 2C5 132 2 col 6 2B5 133 2c205 1r 2B9 13k 2c201 2B7 135 2colo 30 236 136 2c035 2A8 137 230k8 (Neutrt:m Source) 2Bil 138 2C21h C6 2B13 139 2CO31 2AlO 1ho 2C216 2B15 1kl 2 Coo 7 I:1 2B16 Ik2 2CO21 2A12 1k3 2CO2h 2Alb ikk 2C108 CA 2Db ik5 2C00k 2Ch 25067f 2Ek Ik6 IkT 2A0kk A9 2Fh 3

4 ik8 2B3k2 '

2Gk ik9 2AoS5 A1 2J4 g

2B030;l 2Lk

l l

l i

Pne 1C 3 l

TI,3LE 6.1 FUEL ASSE' GLY OFF: CAD SEQUE";0E

)

STEP NO.

FUEL ASSDGLY NO.

m NO.

FEM CORE LOCATIO.,

j 151 2A009 A3 2Nls 152 2B016 2Rb 153 2A038 ac 2sh 15h 2B080

.7th 155 2clo6 DC 2Vh 156 2c017 2VA 157 2c111 AS 2r3 f

158 2c012 m3 159 2B065 FS3 160 2A0k8 AT 273 161 2B07k 2 33

~

162 2A061 In$

EIS r

163 2 BOT 2 2J3 16h 2A063 A4 2G3 165 2B006

.2F3 166 2c10h AS 2E3 1

167 2c027 233 3

168 2c202 A2 2 12 169 2c211 2sg 170 2c01h Al 272 i

171 2c006 2E2 172 2305k 2L2 173 2C20h A3 212 17h 2c207 2R2 175 2c026 Ak 2S2 P

• b

} _

'r b

P p

--g-,

~,,m,,

n

-,-~-w-*-

w-

,nr.

,-m

-e,,-

o*+

r

~ow a

__.->,aw w

--r vv esw-wewe-eo v-e,

l i

/

Fue 2 c-1 TAB ~ E 6.1~

_FU_EL ASSE! GLY OFF.Sc SE Ur :cs 1

-]EP EO.

TUEL ASSEBLT fi3.

CIA *go.

FROM CDRE LOCA"' 0 i 176 ecoo9 2T2 177 2003d ppi i

178 20022 en 179 2co19 2K1 100 20037 2El 181 2cial IE 2D18 l

162 2C029 pgig 183 23008 /

2E18 1Bk 2A026 Ac pyyg 185 2n029 2c18 186 2AOOS Ay pyyg 367 2Bok1 pgg 188 2A02d H2 23111 189 23026 png 190 2A036 RE 2s16 191 2Lo66 ria 192 2cic7 pr py13 193 20021 2 VIE 19h 2c109 m

2ng 195 2 coo 3 2v19 IM 23001 l 2319

.o 197 2AGkT EE M9 196 2DC70

• 5 19 159 2A033 p."

229 200 2B062 I mg I

l e

d

. ~.

i

/

I Page 165 TA3LE 6.1 RJEL ASSE' GLY OF5"OAD Sl~;NE!!CE STEP NO.

MIEL ASSE' GLY NO.

m NO.

FROM CDP.E LOCA'" ION 201 2A06k B5 2G19 202 2B002 2F19 203 2c110 Bk 2n9 20h 20002 2n19 205 2c206 cs 2J20 206 2c210 2G20 1

207 2c033 c8 2F20

/

208 20028 2E20 209 2B010 2L20 210 2C212 D3 2N2O 211 2C208 2R2O 212 2C018 Db 2S2O 213 20015 2T20 214 2C036 2P21 215 2coko 2x21 216 2c023 2':21 217 2C039 2F21 i

l I

l l

.