ML20024B570

Discusses Results of Small Break Analysis Performed to Check Adequacy of Present Auxiliary Feedwater Level Control on 205 FA Plants

ML20024B570
Person / Time
Site:	Crane
Issue date:	03/30/1978
From:	Nirodh Shah BABCOCK & WILCOX CO.
To:	Roy D BABCOCK & WILCOX CO.
References
TASK-*, TASK-GB GPU-0613, GPU-613, NUDOCS 8307090243
Download: ML20024B570 (12)

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.5 status of 205 TA Small Break thCA Analysis March 30. 1978

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The purpose of this neso is to discuss the results of the small break analysis performed tc check the adequacy of the present auxilliary feedvater (AN) level

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control on the 205 TA plaats.

ne analysis shows that the present level control will not keep the core cool-

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able for certain small breaks. Potential fixes are presented and discussed.

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but have not been implemented or analyzed at this time.

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Introductios We have, at present, an NRC approved small' break 14CA analyses for the 205 FA

[4 ',Q plants (BAW-1007&A Rev 1). De approved analyses were done using an AW level f:

Subsequent to the analysis, the level control has under-sona a design' change where the TOGC system now controls the auxilliary N at a d.1-j.

control of 40 feet.

The old level (40 f t) s,-

4 foot level in the secondary side of a steam generator.

~ control is non-safety grade whereas the new one is safety grade thus protecting

..d A new LOCA E.'

the steam lines from the steam generator " overfill" condition.

analysis is now necessary to verify that the new level control will sitigate N:

the consequences of a small break in the RC primary piping wholly in compliance g

p-with the requirements of 10 CFR $0, Appendix K.

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i Characteristic Importance of' a small Break 14CA 2

Small breaks are defined as ' ruptures of the RCS with leak areas o'f 0.5 f t or A small break LOCA compared to a large one, is a slowly depressurising

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less.

transient in which the ECCS injection..which is inversely proportional to i-the RCS pressure, will not match the core water boiloff for a prolonged period.

It is important to make sure that the core is adequately covered by a two phase mixture during the transient to avoid a fuel-cladding temperature excursion.

The four important phases of a small break LOCA can be described as follows:

(1) forced circulation coastdown, (2) quiescence period blowdown, (3) heil-ing pot period, and (4) long ters cooling.

i Reactor trip occurs soon af ter the break and concurrently,'with the assumption of the loss of the offsite power, the main pumps trip thus beginning a coast-i down of forced circulation in the RC systes. The quiescent blevdown period.

During this period the RC loop flow may begins once the pump speed is lost.

exist for a while due to natural circulation caused by the steam generator heat removal' process. During this quiescent period the core boiloff is generally

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offset by the hot leg water draining directly into the core and the SG pri-F mary water draining into the reactor vessel downconer. The available HP1 water also assists in asintaining the vessel inventory. The duration of the C

O loop drainage can be prolonged if the secondary side pressure of the steam M.

3 generator remains below the primary pressure and the A W level control is high enough to condensa the primary system steam and thereby partially offset pd-','

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the systes beiloff.

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The boilina pot period begins once the available loop water is depleted.

$,5 During the boiling pot phase, the reactor vessel water inventory continues to deplete due to the core decay heat, primary metal heat transfer and flash-Q],

i ing..Make-up water to the reactor vessel consists of the ECCS injection and p.g 1.ona tera coolina is y

the condensation caused by the cold injected water.

j established when the aske-up water exceeds the boiloff in the reactor vessel.

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After long term cooling has occurred, the liquid inventory in the vessel ia.

creases gradually and eventually covers the core, if it had uncovered earlier, iQ J

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by a two phase misture.

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,q Method of Analysis _

The analysis uses a modified CRAFT 2 code and a hand calculational method to b([\\

La T The CRAFT 2 ~ '-"

7' develop the history of the reactor coolant systen hydrodynamics.

code was modified to include a new steam generator heat transfer model (froth r

to froth heat transfer and steen condensation model). The hand calculational This.

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aethod is applicable during the " boiling pot" phase of the transient.

method is applied for economic reasons to minimize the CRAFT 2 code calculations.

[F The CRAFT 2 model uses 25 nodes to simulate the reactor coolant system, two Q*,.

nodes for the secondary system, and one node for the reactor building.

A schematic diagram of the model is shown in Figure 1. Control volumes (nodes)"--

E, E

in and around the vessel are all connected by a pair of flow paths to allow the occurrence of counter-current flow. The break is located in the cold leg..

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piping either at the lowest point in the pipe at the pump suction, or at a g

point opposed to the high-pressure injection nozzle at the pump discharge, or in the core flood line joining the CF nozzle. The Wilson, Grenda and Patterson

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I average bubble rise model is used for all nodes.

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I Within the secondary side of the steam generator and the core node a time Initi-variant multiplier is applied to the calculated bubble rise velocity.

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ally a high multiplier was used to match the steady state mixture level in 3

the steam generator in the seconda y loop. After seras, the multiplier was

,I ramped down to a value of 2.38 in a:cordance with the ECCS evaluation endel

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I requirements.

l The following assumptions are made for conditions and system responses during 3

j the accident:

l The reactor is' operating at 102% of the steady-state power level of 3800 1.

MWt.

The leak occurs instantaneously, and a discharge coefficient of 1.0 is 2.

used for the entire analysis. Bernoulli's equation was used for the sub-f

/

cooled portion of the trsasient while Moody's correlation was used in the two phase portion.

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No offsite power is available.

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The reactor trips on low pressure at 1965 psia, 5.

The safety rods begin entering the core after a 0.65 second delay from

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f the time the reacter trip signal is reached.

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The reactor coolant pumps trip and coast down coincident with reactor.

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One complete train of the emergency safsguards system fails to operate, leaving two CFTs and only,one high pressure injection and one low pressure

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injection system available for pumped injection to mitigate the conse-1

4f7, quence of a cold les break. For the CFT line break, only one CFT and one

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9 high pressure injection system is assumed available for providing ECC 0

'J va.tx fluid to the vessel. Credit is taken f:r the cross-connection of the EFI

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The auxilliary feedwater system is assumed to be available during the W

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transient with a set level of 4 feet (cold). The signal error band is.

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included in the actual computer code simulation.

'l ESFAS signal error band is considered inthe analysis to signal the actua-9.

G.{::3-y tion of the MPI.

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O TvDes of Break Analyzed ~....... ' ' ---

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Two breaks are analysed in the present study. (1) 0.05 ft W

z at pump discharge. The suction break was selected since the -

hb and (2) 0.05 fe break location is the lowest in the cold leg and minimizes the time to drais

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'2 the available RCS water. The discharge break'was selected since the HFI flow h>/

injected in the break, node is quickly lost through the break. Thus,'for the s[g_x:r discharge break, the core receives less make-up water which prolong the onset i

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of long term cooling.

Results of Small Break Analysis L,h

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0.05 ft2 s,tge,e yo,,s,eeto, (3.g....

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The break is assumed to occur at the bottom of node 9 of Figure 1.

The q

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4 analysis takes credit for two CFTs, one HPI and one 1.PI pump.

d,. f The core pressure and fluid inventory is shown in Figures 2.and 3. re-g '

spectively. The reactor trip occurs around 11 seconds after the break.

+

During the subcooled blowdown, the RC pressure hangs around 1700 psia g..,

thus preventing the ESFAS signal on RC pressure. Finally, the ESTAS is O

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actuated on the containment building high pressure signal at 110 seconds

^

and the HPI starts at 145 seconds. The loss of natural circulation occurs p,

around 125 seconds. The boilies pot phase of blowdown begin by 800 see-p/

onds and the onset of long term cooling occurs at 1500 seconds. The F'-

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core is always covered by a solid water, hence no cladding temperature

I

excursion will occur.

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The break is assumed to occur at the bottors of node 10 of Figure 1.

The analysis takes credit for two CFTs, one BFI and one LPI pump. The break U.

j saalysis is similar to the suction break except for the location change.

R'.

Essentially, is of the HFI is not available as core make up water due to sb.

its loss out the break. This loss is not an assumption.*but rather it is C.; '.'

calculated by the computer simulation.

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Figures 4 and 5 show the core pressure and the core fluid inventory re-fa "

spectively. The reactor trip occurs around 11 seconds af ter the break.

Q' During subcooled blevdown phase, the'RC pressure first hange around 1700 2..

pais and repressurines af ter 40 seconds. The ESFAS signal is reached at h'A N

110 seconds due to high containment building pressure signal. The quiss-

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~.J cent blowdown phase begins by 130 seconds at which time the natural cir-Z.. ~

.I culation is also lost. The EFI flos-injection starts at 145 seconds and.

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by 200 seconds high quality sixture is exiting through the leak path and

.r the system assia begins to depressurize. Af ter 900 seconds the systen G

experiences a steam condensation in the steam generator thereby increas-F - '.

ing the system depressurization. The available loop water is drained by.

1200 seconds. At that time, the syntaa pressure is below the secondary r--

SC pressure and system is well into the " boiling pot" phase of depressur-1 isation. The core begins to uncover at 1600 seconds and continues to de s.m so until about 2600 seconds when long zera cooling is established. The f,,

('-

core is covered to a level of 6.5 ft.

0.' 4. -

Summary of Results

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P. J The results of 0.05 ft2 brasis shows that the consequences of the break at fM

.yump suction can be safely altigated by an HPI pump alone whereas changing the position of the break to the bottoa ef the discharge node shows that the

-t' core is uncovered to a level of 6.5 ft and the core remains uncovered for a b-"

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long time. Under the present calculational method, such a core uncovery any F !,

result in a very high cladding temperature (greater than 2200F).

p..:

Recommendation e-The present results for a discharge break are not acceptable in comparisos s.

3 to the BAU-10074A small break analysis ressults or by 10 CFR 50.46. The

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" fixes" which can improve the results are discussed below.

c.

1.

Analysis (model) research:

Y Check the calculational method, nodal diagram, etc., for an unrealistic E

behavior and ways to counteract.

2.

Zaloudek-Moody leak discharge model=

i The present analysis uses the Bernau111 equation for the subcooled liquid t.7 Ioak discharge. Use of the Zaloudek equation is estimated to reduce the a

subcooled leak rate by about 40 and delay the core uncovery by about L

75 seconds.

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10% uncertainty on DH af ter 1000 seconds:

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The decay heat (DH) curve used in the present analysis has a 20% uncer-

$.} '3 tainty margin over the ANS DH data as specified by the NRC regulation.

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A literature search shows that only 10: uncertainty is realistic after L' M Y

1000 seconds into the transient. ANS also reccamends 101 at this time.

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It is estimated that the use of the reduced uncertainty will almost cover T.

-i the cote by a mixture. This fisi could require a rule change, it will hg, -

definitely require a model change.

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4.

SC level sets RJ e b~..

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y-Use of a 40 ft aux feed level control in the past has resulted is as ac-ceptable small break analysis. A level control higher than the proposed Ni'7

- 3 3

4 f t level should provide faster systes depressurization.as well as estra *

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available water supply to the core by condensing steam on the primary h

Y 'p side of the stems generator. Also, the high set level depressurizes the t-y secondary side thereby providing additional temperature difference for hy';:

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heat transfer from primary to secondary side of a steam generator.

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Operator.aetfan:

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Tio*^M Tuo types'of operator action can be taken to improve the systes depres--

q.

s surization.

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Depressurize the secondary side, e'.g., by opening the atmosphere mod -

(l +; )

.1 ulating dump valves.-

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Depressurize the primary side, e.g., by a,ctivating the pressurizer '

.q spray system (only small benefit is expected since the spray unter comes off the EP1 line).

g d

'6.

1.arger HPI espacity V l.1 Q.'

a

. A 20% increase in the HPI capacity will significantly shorten the oos'et

• of the long ters cooling as well as reduce or prevent the core uncovery..

l4' -

t-7.

ESFAS set points b-.

It is desirable to have timely ESTAS actuation on the primary signal in-

/,

stead of the backup system (containmer t building pressure) used in the present study.

i 8.

Dump to sumps L

)

A larger leak area is provided by timely opening of the hot leg dump to sump line. The additional leak area increases the primary systes de-pressurization. Opent::t this valve would place all accidents in the lov pressure range. This would mean that (cr any, break size or location no 4-core problem would occur.

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Schedule EFI-U I to get 2 HFI's:

By delaying the UI ESTAS set point, the other inactive HF1 (where a

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diesel failure is postulated) may be diesel loaded instead of the LFI thus provid"ng two HFIs for nitigating the transient.

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10. Ce to 2200y cladding temperature criterias d).-

3 Lirx Implementation of this criteria will allow the core uncovgry to the ex-

[*i;,

tent that the cladding temperature should not exceed 2200 F limit and so-

-4 other cladding da.sge criteria set by 10 CFR 50.46. Without other sys-

@f tem improvements ECCS does not believe this change alone will suffice.

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Steam to steam heat transfer la SC:

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In the present analysis, no steam to steam heat transfer exist between py '!.

the primary and the secondary side of a steam generator. Only a small benefit is expected by implementing this mode of heat transfer.

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