ML20077H642

Final Significant Deficiency Rept Re Dropped Rod Methodology.Initially Reported on 791119.Westinghouse Performed plant-specific Analysis W/Approved Methodology. DNB Design Basis Met

ML20077H642
Person / Time
Site:	Summer
Issue date:	08/08/1983
From:	Dixon O SOUTH CAROLINA ELECTRIC & GAS CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
10CFR-050.55E, 10CFR-50.55E, NUDOCS 8308110173
Download: ML20077H642 (19)

Text

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'.e . . . l SOUTH CAROLINA ELECTRIC & GAS COMPANY POST OFFICE 764 COLUMetA. south CAmoWNA 29218 4 ' O. W. DIXoN. JR.

viCa m.S""' August 8, 1983 NUCLEAR OFenATioNS Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation l

U.S. Nuclear Regulatory Commission

! . Washington, D.C. 20555

Subject:

Virgil C. Summer Nuclear Station Docket No. 50/395 Operating License No. NPF-12 Dropped Rod Methodology License Condition 2.C.22 ,

Dear Mr. Denton:

The Virgil C. Summer Nuclear Station Operating-License Condition 2.C.22 requires that for operations above 90% of full power, the

! reactor shall be controlled manually or the D bank control rods shall be out greater than 215 steps until written approval is received from the NRC. By letter. dated March 31, 1983, from C.

'O. Thomas (NRC) to E. P.. Rahe (Westinghouse),.the NRC approved ~

y the Westinghouse generic Dropped Rod Methodology. ,

-Westinghouse has performed a plant specific. analysis for the.

~

-Virgil-C.' Summer Nuclear' Station,_ Cycle 1, utilizing the NRC approved methodology. The results of the analysis indicate that the DNB design basis is met. Based on this evalue . ion, South Carolina Electric.and Gas. Company (SCE&G) has concluded that the interim restrictions-for rod control are no longer required'for ,

Cycle-l. Attached are marked-up FSAR pages reflecting the new analysis. ,

1 By copy.of.this letter to Mr. James P. O'Reilly,-SCEEG considers this letter the final report on the significant deficiency

-concerning the dropped _ rod analysis first reported to Region II on November 19,'1979.

We request your_ expeditious concurrence in our position on this item in order that the-restriction on~ rod control operation can be removed.. If you have any questions, please let us know.

Ver truly yours, s

s 1

O. W.. ixon,. ,

NEC:OWD/fjc -

t cc: (See Page-#2.) 830809 ]A SM*iWl'aa?J_

s: .

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

~

i' Mr. Harold R. Denton August 8, 1983 Page 42 cc: V. C. Summer T. C. Nichols, Jr./O. W. Dixon, Jr.

E. H. Crews, Jr.

E. C. Roberts H. N. Cyrus J. P. O'Reilly Group / General Managers O. S. Bradham R. B. Clary C. A. Price

. A. R. Koon 1 C. L. Ligon (NSRC)

! G. J.-Braddick J. C. Miller J. L. Skolds '

J. B. Knotts, Jr.

NPCF File-(Lic./Eng.)

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qr -

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-4 . for reactivity insertion rates less than p 3 x 10 6K/sec,the rise in the reactor coolant temperature is sufficiently high so

- that the steam generator safety valve setpoint is reached prior to trip. Opening of these valves, which act as an additional heat ,

load of the reactor coolant system, sharply decreases the rate of.

. . rise of reactor coolant system average temperature. This decrease in rate of rise of the average coolant system temperature during the transient is accentuated by the lead-lag compensation causing the overtemperature AT trip setpoint to be reached later with re-

_ sulting lower minimum DNBRs. ,

For transients initiated from higher power levels (for exasple, see Figure 15.2-8) this effect, described in item 4 above, which results in the sharp peak in minimum DNBR at 4 3 x 10~ 6K/sec, does not occur since the steam generator safety valves are never actuated prior to trip.

Figures 15.2-8,15.2-9 and 15.2-10 illustrate minimum DNBR calculated ,

for min 4== and maximum reactivity feedback. ('j 15.2.2.3 Conclusions The high neutron flux and overtemperature AT trip channels provide ade-quate _ protection over the entire range of possible reactivity insertion ,

rates, i.e. , the min 4== value of DNBR is alvsys larger than 1.30.

~

15.2.3 ROD CLUSTER CONTROL ASSEMBLY MISALIGIOENT s

I 15.2.3.1 Identification of Causes and Accident Description l

f Rod cluster control assembly (RCCA) misalignment accidents include: e

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t

.i s. .

b 1.  :. : =;;= ===  %;

l 0'!' 0F Pror e clr e/ RCcA, y, ///k g

2. k droge/,)(c ,fa,,f3 PQ

3. S/a //cas)e Nafbye/15.2-14 f((p, ~,5,k g; 7 J - . . . ..

--v---- , - , . - - - - , . , -------.--,,-,.,---,--.,,--e,,-r,_,na . , -n+ - . - - - , , - - , - - ,,_,,,,,__,,--------,--,--.,---,.-,--n,--

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

(- .

r 2. ped.as!sembly hank-

'. 3. Statica misaligned bly (see Tabla , _ _

. -2).

1 L - . .-

Each RCCA has a position indicator channel which displays position of

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the assembly. The displays of assembly positions are groupid"for the

m. operator'st convenience. Fully inserted assemblies are further indicated by a rod bottom signal, which actuates a local alarm and a control room annunciator. Group demand position is also indicated. The l assemblies are always moved in preselected banks and the' banks are always moved in the same preselected sequence.

A droppedACCA  :::- 'ly or _RCCA

__ 'ly b dank

- ' eee/.sdetected by:

. 7

.- 1. Sudden drop in the core power level as seen by the nuclear instru-mentation system;

2. Asymmetric power distribution as seen on excore neutron detectors or core exit thermocouples;

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3. Rod at bottaa signal;  ;

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4. Rod deviation alarm; .

5. Rod position indication.

Miss11gned assemblies are detected by:

1. Assynt.tric power distribution as seen nn excore neutron detectors )

or core exit thermocouples; 1

i

. 15.2-15' AMBleteH

., . ., , d

2. Rod deviation. alarm; .

3

3. Rod position indicators. -~

e The resolution of the rod position indicator channel is +5 percent of

. span (+7.2 inches). Deviation of any assembly from its roup by twice _

this distance (10 percent of span, or 14.4 inches) will not cause. power distributions worse than the design limits. The deviation alarm alerts the operator to rod deviation with respect to the group posh. tion in, .

execss of five percent of span. If the rod deviation alarm is not -

operable, the operator is required to take action as required by the Technical Specifications. ,' ~

If one or more rod position indicator channels should be out of's'ervice, detailed operating instructions shall be followed to assure the align-

~

ment of the nonindicated assemblies. The operator is also required to take action as required by the Technical Specifications. The operating instructions require selected pairs of core exit thermocouples to be 7

monitored in a prescribed time sequence and following significant motion (,

of the nonindicated assemblies. The operating instructions also call for the use of moveable incore neutron detectors to confirm core exit thermocouple indication of assembly misalignment.

15J2.3.2 Analysis of Effects and Consequences .

15.2.3.2.1 Nathod of Analysis e e h Ser-f -

1 -

Steady-se over distributions are analyzed for t ev, ant using the s' TURTLE [8] Code. pea. ting factors calcula y TURTLE are then used by ~the THINC Code to ca te tte DNB or the transient response to a dropped RCCA or RCCA bank th [7] Code is used. The code sie ulates the neutron kineti , rest tor t system, pr,essuriser, pres-surizer relief and ety valvesi pressurizer ray, st,eam generator, and steam g ator safety valvet . The code compute rtinent plant variab including temperatures, pressures, and power leve ,

. 15.2-16

- 16 2.%2.lMethod of Analysis.

Inser$ b , ~

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. 17.' One or more dropped RCCAs frcm the same group. g m

For evaluation of the dropped RCCA event, the transient system e response is caiculated using the LOFTRAN code. The code simulates the neutron kinetics, Reactor. Coolant System, pressurizer, pres-

{

surizer relief and safety valves, pressurizer spray, steam genera- l

. . tor, and. steam generator safety valves. The code computes perti-nent plant variables including temperatures, pressures, and. power level. [,

-~' J

. .__ L _. 9 Statepoints are calculated and nuclear acdels are used to obir.ain a

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hot channel factor consistent with the primary system conditions <

and reactor power. By incorporating the primary conditions from l 1

.. . the transient and the hot channel facter from the nuclear analysis, the DNB design basis is shown to be met using the THINC code. The transient' response, nuclear peaking factor analysis,-

and DNB design basis confirmation are perfcmed in accodiance with the methodology described in Reference 13. M~

M Statically Misaligned RCCA --

Steady state power distribution are analyzed using the computer -

coces as described in Table 4.1-2. The peaking factors are then used as input to the THINC code to calculate the DNSR.

1 h . e*- a ,e wee.- mee.m.h=e ene- - ew w -.e .eum .e -q, ..e e eie .g.w . , , . .

• "

• h* weee 6mm s. e.g a um ,.veh ee m

9

.15.2.3.2.2 Razults (bcc b Se B ,,

dropped RCCA typically results in a reactivity insertion of -150 m. I yses have shown that with the core power distribution which e ses fo ,

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the drop of a single RCCA, the reactor may be return to full y r

er with the full power reactor coolant system temper ure (plus N_ measur t and control errors) without the DNBR going be .1.30.

This is v ified by the results in Table 15.2-2. .

r ~

Extensive an ses were performed to show that wit automatic rod con-trol the min

% DNBR occurs near the end of the ransient when the reactor coolant s tem has essentially returne to .its initial. steady-state equilibrium e ditions. Without aut ic rod control, the sys-ten will return to a y equilibrium cond ion at a reduced primary i

temperature as a result f the moderato reactivity feedback. As typical of a PWR uncontro ed respons , the return to power is mono-tonic and therefore power o rshoot is not a concern for this case.

E power overauuww af cus a esvy

- d ACCn Aussucuw usu vuly so ul6 isum l

the action of the automatic dc troller. For a given PWR system, the power overshoot is ess tially function of the rod controller characteristics. Large over oversho ts can result if the rod controller '

is designed to restor primary system c lant temperature or secondary system steam pressu e without regard for e core power level. The Westinghouse des uses a dual controller ch limits the power over-shoot. The reant feature of the Westingh use rod controller _is that it te l ates rod withdrawal well before t e primary coolant ever- l

~ age temper ure is restored to an equilibrium con ition. This not only I

-4*4=4z the power overshoot but also ensures extr margin to DNB.

San civity studies have confirmed that the maximum pow oversboot e a for the following conditions:

i i 15.2-17

j _

. Minimum moderator reactivity feedback corresponding to beginning j of core life conditions. ,

f '

2. ximum reactivity worth of the control bank.

Figure 15 -11 illustrates the transient for the following 1 iting conditions:

1. Initial P er: 102 percent of rated power -

2. Moderator Rea ivity Coefficient: Least Neg ive

3. Control Bank Reac vity Worth: 12 pcm/st p

4. Dropped RCCA Reactivi Worth: 250 p -

The initial reactor coolant sy tem t erature was assumed at its maxi-zum value and the initial reacto co ant system pressure at its ministan l

. 1 . i.;.  ;. 1;2. .;.e.2, .;..;.- f 11 p.... epese;ie . 02. 02.;.ie

. 15.1.2.2 for a discussion of in ia conditions. The selection of a value of -250 pcm for the mar rea ivity insertion is a conserva-tively large value for the rth of a s gle dropped rod. As a result l of a dropped rod, the nue r power will crease and the decrease will be sufficient to be det ted by the power r se negative neutron flux  ;

i rate trip circuitry a trip the plant before he reactor can return to ,

high power. The a ysis was performed only to etermine the limiting i

DNB conditions du ng the transient, thus it is i apendent of power

(.1 distribution. e DNBR was computed along the tran eut assuming con- g l stant design at channel factors to illustrate that 1 ting DNB con- l ditions oc r at the end of the transient. Therefore, y final equil-ibrium c ditions need to be analysed with the penalty as ociated for .

a dropp d RCCA condition as shown in Table 15.2-2. ,

i 15.2-18

s.

dropped RCCA group typically results in a reactivity insertion of 00 pcm which will be detected by the power range negative neutron f1 rate trip circuitry. The reactor is tripped within approxima y i

2.5 s conds following the drop of a RCCA. The core is not adver ly ,

affecte during this period, since power is decreasing rapidly The most se era misalignment situations with respect to D at sig-nificant pow levels arise from cases in which bank D i fully inserted ly withdrawn; a 12 foot misalignment rror. Sultiple with one R'CCA -

.- independent ala , including a bank insertion limi alarm, alert the

\

operator well befo the postulated conditions ar approached. The .

bank can be inserted to its insertion limit vit any one assembly fully -

withdrawn without the R f alling below 1.30 ,

The insertion limits in th Technical Spe ifications may vary frcu time to time depending on a numbs of limit g criteria. It is preferable, therefore, to analyze the misa igned CCA case at full power for a

,, position of the control bank as e ly inserted as the crit. aria on minimum anK and power peaking f or will allow. The full power in-

!\ sertion limits on control bank nu then be chosen to be above that position and will usually be ictated y other criteria. Detailed results will vary from cy e to cycle d ending on fuel arrangements.

For Case I shown in T le 15.2-2 with bank inserted to its full power insertion limit and ne RCCA fully withdrawn, DNBR does not fall below 1.30. This case s analyzed at 102 percent o full power with~the in-creased radial aking factor associated with th misaligned RCCA.

DNB calcula ons have not been performed specifica11 for assemblies

~

missing om other banks, however, power shape calcula ons have been done required for the RCCA ejection analysis. Inspec on of the

. f- powe shapes shows that the DNB and peak kW/ft situation i less severe t the group D casa discussed above assuming insertion 1 ts on the her groups equivalent to a group D full-in insertion limit.

l i 15.2-19 i l

l

.s

. l

/S. 2. 3. 2. 2 Results InsJ B 4 I /. J One or more Dropped RCCAs Singie or multiple dhopped RCCAs within the same group re:: ult in a ,

negative reactiyity insertion wnich may be detected by the power a range n*gative neutron flux rate trip circuitry. If detected, the!

reactor is tripped within approximately 2.5 seconds following the drop M the RCCAs. The core is not adversely affected during this

- period, since power is decreasing rapidly. Following reactor '

trip, normal shutdown procedures are followed. The op'rator e .nay manually retrieve the RCCA by following apprcved operating proce-durt:s.

. e .

For those dropped RCCAs which do not resuit in a reactor trip, power may be reestablianed either by reactivity feedback or con-trol oank withdrawal. Following a dropped :cd event in manual rod control, the plant will establish a new equilibrium condition.

Th: ::e"ibri= ;;r ::::,without control system interaction is

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r%e pbuer kcread )

N*"* e-

• a e em -a.-w -

m e me-e -m ee ese-Weteh+<--6enum = -==-er Omm e*= a e gum use 6 e es e--asee gemOeeee..we = w = m . . , , , , , , , , , , ,, ,.

-- ege a w w em. - a .gie e -w ,se m,g,em.g.,e, .. mogi, ,e

& he * - mes--eip ee** - N- **M6M e + + gmumb N

• MN-m e - h e emoeeun- e O- e w.es =-h--- m.ee ne e.w ow- m , , , , , ,, , , , , , , , , , m,__.,,,,

eem me h, wh e. em aemme e e en e ,, , ,a es- e .- -

t

b. . .

. monotonic, thus removing power overshoot as a concern, and '

establishing the automatic rod control mode of operation as the -

limiting case.-

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.. . 3.,

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'For a dropped RCCA~4 vent in the automatic rod control mode, the

,. Rod Control System detects the drop in power and initiates control bank withdrawal. Power overshoot may occur due to this action by I the automatic rod controller after which the control system will

. insert the control bank to restore nominal power.- Figure 15.2-11

-~

. shows a typical transient response to a dropped RCCA-(or RCCAs) in automatic control' - Uncertainties in the initial condition are included in the DNB evaluation as described in Reference 13. In all cases, the minimum DNBR . remains above the limit value y

/,30.

2g. Dropped RCCA Bank A dropped RCCA oank typically results in a reactivity insertion greater than 500 pcm which will be detected by the power range negative neutron flux rate trip circuitry. The reactor is tripped within approximately 2.5 seconds followihg the drop of a RCCA Bank. The core is not adversely affected during this period, since power is decreasing rapidly. Following reactor P trip, normal shutdown procedures are followed to further cool down the plant. Any action required of the operator to maintain the plant in a. stabilized condition will be in a time frame in excess of ten minutes following the incident.

$ g. Statically Misaligned RCCA The most severe misalignment situations with respect to DNBR at significant Power levels arise from cases in which one RCCA is fully inserted, or where bank D is fully inserted with one RCCA fully withdrawn. Multiple independent alams, including a bank insertion limit alan., aien tne coerator l

l l

l. - . - . ,. . - - _ .

. well before the postulated conditions are approached. Tha bank

. can be inserted to its insertion limlt with any one assembly fully withdrawn without the DNBR falling below the limit value. l The' insertion limitr in'the technical specifications may vary from time to time depending on a number of limiting criteria. It is preferable, therefore, to analyze the misaligned RCCA case at full power for a position of the control bank as' deeply th'serted ~as the criteria on minimum DNBR and pikte= peaking factor will illow. The' ~

full power insertion limits on control bank D must then be chosen to be above that position and will usually be dictated ny other cri teria. Detailed results will vary from cycle. to cyc.1e depend- ~ ~ ~ ~ ~ ~

ing on fuel arrangements. -

For this RCCA misalignment, with bank D inserted to its full power insertion limit and one RCCA fully withdrawn, DNGR does not fall below the limit value. This case is analy:ted assuming One initial reactor power, pressure, and RCS temperatures are atlheir nottinal values including uncertainties (as given in Table 15.1-2), bu't

~

with the increased radial peaking factor associated with the-misaligned RCCA.

f DNB calculations have not been perfomed specifically for. RCCAs missing from other banks; however, power shape calculations have -

beer, done as. required for the RCCA ejection analysis. Inspection of the power shapes shows that the DNS and peak kw/ft' situation is -

less severe than the bank D case discussed above assuming inser-tion limits on the other banks equivalent to a bank D full-in insertion limit.

For RCCA misalignments with one RCCA fully inserted, the DNBR does not fall below the limit value. This case is analyzed assuming

the initial reactor power, pressure, and RCS temperatures are at their nominal values, including uncertainties (as given in Table l

15.1-2) outkith the increased radial ceasing facter associatec

)ith the misaligned QCCA.

l ,. .

DN8 does not occur for the RCCA cisnifgneemt incident and thus the -

ability of the p:-imary coolant to nemove heat' from the fuel rod'is

! not reduced. The peak. fuel temperature corresponds to a linear I

heat generation. rate based on the radial peaking factor penalty a associated w'ith tiHMiilsaligned RCCA and the design axial power distribution. The resulting linear heat generation is well below ~

{

that which would cause fuel melting.

Following the: identification of a RCCA group misalignment condi-tion by the operator, the operator is required to take action.as . .

fequired by the plant technical specifications and operating . _, . .

instructions.

l 1

1

)

S

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

15.2.3.3 Conclusions q L g h {9 3 is shown that in all cases of dropped single RC , th'a DNBR remains j greate han 1.30 at power and, consequently, dropped single RCCA l

' ~

does not cau ore damage.

- l Tor all cases of dropped a he reactor is tripped by the power

. . . . . . . \

range negative neutron f rate and consequentifdropped banks do not cause core dama .

l

. For all e s of any bank inserted to its rod inserti imits.with any

~

sing 1 CCA in that bank full: withdrawn, the DNBR remains er than O. --

l 15.2.4 UNCONTROLLED BORON DILUTION 15.2.4.1 Identification of Causes, and Accident Description Reactivity can be added to the core by feeding primary grade water into

, the reactor coolant system (RCS) via the reactor makeup portion of the chemical and volume control system (CVCS). Boron dilution is a manual operation under strict administrative controls with procedures calling for a limit on the rate and duration of dilution. A boric acid blend system is provided to permit the operator to match the baron concen-tration of reactor coolant makeup water during normal charging to that in the RCS. The CVCS is designed to limit, even under various postu-lated failure modes, the potential rate of dilution to a value which, af ter indication through alarms and instrumentation, provides the oper-ator sufficient time to correct the situation in a safe and orderly manner. '

l The opening of the primary water makeup control valve provides makeup to the RCS which can dilute the reactor coolant. Inadvertent dilution f rom this source can be readily terminated by closing the control 4

, 15.2-20 l

1

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

15'.2.3.3 Conc 1usions

{yC i

For cases of dropped RCCAs or dropped Danks, for which the reactor is tripped by the power range negative neutron flux rate trip, there-is no reduction in ,

the margin to core thermal limits, and consequently the DNB design basis is

~

1 met. It is shown for all cases which do not result in reactor tripihat the l DNM remains greater than the limit value-and, therefore, the DNB -des +gn is l

set.

.- . : ._ u For all cases of any RCCA fully inserted, or bank D inserted to its rod insertion limits with any single RCCA in tnat banx fully withdrawn (static l

.nisal'ignment), ne DNBR remains greater than the limit value.

so4 -

i 1

=

0 e

s.

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. I

= . .

10. Geets, J. M. , " MARVEL - A Digital Computer Code for Transient Analysis of a Multiloop PWR System," WCAP-7909, June,1972. l

11. Mangan, M. A. , " Overpressure Protection for Westinghouse Pressur-ized Water Reactors," WCAP-7769, October, 1971.

, 12. Geets, J. M. and Salvatori, R. , "Long Term Transient Analysi.s _

Program for PWR's (BLKOUT Code)," WCAP-7898, June,1972.

u. noci A , r, et a /. , " D J foof 7%dA% C for Ny& A 5 ,na P/a.r& '%e*'h -

/oa n ,9-n (A,ex, *edij ) a!/ ocRP- /oz ye -/s

(/ von - /%priefary), 7ase , /t/2- -- .

G e

e *

(

4 15.2-66

_9 TABLE 15.2-2 '

, t .-

r MINIMUM CALCULATED DNBR FOR CASES OF ROD CLUSTER CONTROL ASSEMBLY a SISALIGNMENT AND DROPPED ROD CLUSTER CONTROL ASSEMBLY.

o (1) a Radial' Power Min -

Peaking Factor (FAH) D R- .

J i ..

~

Cases Anal ed  ;

n . .- -- . - < - .

i

~ - Bank D at inse tion limit, .  :. - .

E-2 fully withd wu(Md . . . . .

C1xtar Cor. trol embly , , ,

2 - -Misalignment)--- - - - - 1.6 ._ >1:3 ______ i

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

.d Dropped Rod Clustar Cet rol Asembly C-9 1,51 > 1. 3

_. Dropped Rod Cluster Control Assembly F-10 1.59 > 1. 3

. .. Dropped Rod Cluster Control -

Assembly E-11 1.56 > 1. 3_

2  :

Dropped Rod Cluster ~ ntrol _

Assembly B-10 ,

1.59 > 1. 3

' Dropped Rod C star Control Assembly R- 1. 5 > 1. 3

.' (1) kaaes include 15% uncerta:.nty allowance in F .

eye 5 e i 15.2-76 s

l -

l

3

- T* 110

/

5 l l l l 1 1 I i 1/

• sE g =

100 -

• 90 -

,I 1

x .I

,-T - -

E g

u u.

80 hg * [ REACTOR TRIP ON .-

LOW PRESSURE

~ , , ]

70 ,

sc / .

mL

! =

0 pA ,

~

20 - .

5#- \\ 1 u0 \  ! -- l 7 24c0

\ s

/ = - --

l 5E \ _

'y i.2200 N \

ae \

0 a 2000 s -

20 \

E 1800 0.8 / \

\

!, LEGEND:

=

~

0.6 / .

WI AUTOMATIC CONTROL f

EE 0. 4 gg === WITH T AUTOMATIC CONTROL

= c

" 0.2 -

1

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0.0 (I I I I \ l l I O 30 60 90 120 150 180 10 2140 270 300 Oe/efe lh '"'

SOUTH CAROLI ELECTRIC & GAS

. VIRGIL C.SUMME NUCLEAR STAT 0!

Transient Response to Dropp l

Rod Cluster Control Assemb$

Frinure 15 J-ll

e s e a

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. een 11 .

5  :

I t.

3 to -

y see z

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we -

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u  :

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m l 1.2 . - - = = = . '

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un  :

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em ucosmos. .

Inse, Fgare I

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T ran s -en : P.esponse to Dropped Rod Cluste Con croI Asse-t ty i

rigu re 15. 2-11 1

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._ -_ __ ._ _._ _ ._ _ _ - - . _ _ _ _ _ - - _ . . . _ - - -.-