ML19322C299

Forwards Rept on ECCS-small Break LOCA Analysis in Response to

ML19322C299
Person / Time
Site:	Crane, Bellefonte
Issue date:	01/23/1979
From:	Mcfarland J BABCOCK & WILCOX CO.
To:	Tira Patterson TENNESSEE VALLEY AUTHORITY
References
TASK-TF, TASK-TMR NUDOCS 8001160822
Download: ML19322C299 (3)

Text

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Babcock & Nilcox 1

Power Generatien Grevo P.O. 8ex 1260. Lyneneurg. va. 2:!n Te!ephone:(804) 334 5111, January 23, 1979 Letter :To. D-3132 File Ref: UkM-2/12314 Ref,Ltr:

K-5020/L-27-73

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t Tennessee. alley Authority i

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l kCO C------a Avenue fqf Inoxville,

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Attenticn:* Mr. D. R. Patterson 5

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'ISa,, 15/10, o Chief Mechanical Engineer R C o.cnes 3ellefente. Iiuc1' ear Plant Units 1 i 2

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Contract :To. 71C62-5h11k-2 B&W

Reference:

. iSS-15 1.

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Subje *-

• -C' 3:eak LCCA Analysis Gentle:en:

IThe attached repert fis in' respecse to ycur reference letter.

Please let us kncv. if further discussion is required.

.Very truly yours, James McFarland Senior Project Ma:.::ger

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14 O e ca-b By 1

Robert E. Lightle Asscelate Project Mar.ager Attachr.ent

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W. 3:ent '7ade J..L. Atchisen E.,L. LcEan M[knbD'R~YnLo i

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The Babceck & Wife:x c:rreany / I:taerisne:: 1857 3

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• 9 s o Respont c to' TVA Letter K-5020, Escrgency Core Cooling Syste= --

Se 11 Break LOCA Annivsis ::4.M-2-14 (A.1). Acril 27.1973 Via TVA Letter K-5020, TVA transmit:ed to 3&W a report encicled. " Decay Heat Rcrioval During a Very S=all. LOCA for a B&W 205-Fuel-Assc=bly PPR." by C..

7 7Midhclsen,,.dacad January, 1073.; This report. presents a st=plified; hand-

'halculation;reviewofther=allbreak. transient)andpotedtialconsequ'encEs for very s=a11' breaks not'explic'itly examined within the s=all break :opical

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for the 205 EA plant, 3AU-10074A, Rev. 1.

Within this paper, the following concerns were expressed for the very s=nll breaks:

1..How is decay heat rc=oved?

2.

Will syste= repressuri:stion occur? If so, could a s= aller case be a vors: break?

3.

If the opergeor isolates the break, will systes repressuri:stien occur?

If so, will the pressure relief valves be subjected to slug or two-phase Elcu?

Responses to these concerns are developed in the subsequent paragraphs.

,3efore discussing these cencerns,.$ general over0iew of the s=all break tran-

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ksient;in'a 35W 205 plan: needs to be briefly discussed.

S=all LOCAs'can be yieved as a slev.::ansien: during which :hetRCS can be described l asia. sealed.

amano=eteri 3ecause of the internals ven: valves, nol extensive _stea= bubble;

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iw111 form within the reactor vessel while any significant liquid inventory (resains in the loop. Many e= peri =en:.' have been run which shev that so long

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as a fluid, with qualities less :han 7C or so, covers the cere, no adverse core temperature excursion vill occur at decay heat pcver levels.

Thus, er.y problems with s=all breaks will only occur af ter the KCS loops have depleted their inventory.

Decay heat re=cval frc= the core region is no probles as stated' above. Ecv-ever, decay heat re= oval frc= the sys:e= as a whole needs to be exa=ined further. There afa two ways of re=oving decay hes: frc= the sys:e=; via the break and/or via the stes= generator.

Both of :hese ice =s are discussed is detail in the TVA letter.

For :he very s=211 LCCAs of interest in this dis-cussion, it was shcun that the break alone is"not capable of re=oving all :he decay heat and hea: re= oval via :he' steam genera:ce is necessary.

While :he T7A-predic:ed break size : hat this occurs at was :ot checked quan:atively, the ac:t 1 break si:e : hat it occurs at is.incensequential. Such a break Psize}doesaristwherethesteamgeneratorsarenecessary.

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g The role of thef steam generator as a heat re= oval source is basically as described inLeheiletter. 'In'icially, na: ural circhlatica vill be =ain:ained in the syste= and the necessary hest re=cval is easily acce=plished.

Once a steam bubbic of sufficient size necesscry to fill the U-bend at the top of l

the hot legs is for=ed, nscursi circulatien vill cease. -The inter =ic: ant

{ natural circulation discussed in, the letter vill act occur.due 'to' the.slev l

Inasurs' fLehe s=allibicak.: snsient. Once natural circula:1cn ec ses, :he I

'systes vill repressuri:c se=ewhat until the SC pri=sry side liquid level drops I

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a below the SC secondary side level and condensatica heat transfer is established.

During this period between the natuial circula: ion and condensation hea:

re-moval modes, the letter expecsses concerns tha: the I1guid inventory within

the sys

c= will be~ denleted at c race in excess of the races for the breaks aecause at. the par:ial repressurization c:. the sys:e=.

rtA analy:ed,.,.~. n.

is concerned that this ultica:cly vill resul: in =cre core uncovery than : a:

shewn in the s=all break topical rep e t ' " -10 0 7.'. A. 'This is not the case.

I During the natural circulation phase, it is obvious that the s= aller :he break,

.the slower the loss of syste= inventory and the lenger the peried of na: ural circulation..After natural circula: ion ceases, sys:e= pressure vill be cen-trolled by a " volume balance." That is, the syste= pressure vill balance at a poin: where the.volu=a of fluid discharged through the break equals the volu=e of stea= being created in the core. Since the cold leg fluid enthalpv re.ains unchan;ed during a s=211 break ::ansient, the volu=e relief cut the break increases vi:h increasing syste= pressure and b:cak size.

The cel-

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u=e of s:ca= being generated in the core decreases with increasing pressura.

As the break decreases in si:e, the RC syste= vill repecssurice :o a higher value; thus the volu=e relief out the b:cak necessary to catch the volu=a of stec= being crea:cd decreases. Taerefore, the syste= inventory will be lost at a,slove.r ra:~e as break siae decreases. Once the SC becc=es available for cendensati:n heat re= val, the pri=ary syste= pressure vill depressuri:e to app

'-' ely the SC secondary side pressure.

Since the seconds:7 side of the SG vill respe'nd in a si=ila': =a:iner :o tha: of :he 0.05 f:2 break analyzed in the tcpical, :he pri=ary side pressure response, folleving :he,

advent of condensati:n hea: re== val, vill be si=ila :o :ha of the 0.05 f:-

b:cak. Thus, for the smaller breaks, the sys:e= inventory vill always be greater than that for the 0.05 ft2 break and the core' vill always re=ain covered and vill not undergo a te=pera:ure excursion.

In the paper concerns are raised rela:ive to isolation of the break af:e natural circulation is los:.

'I'ha scenario presented in the le::er is rease=-

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able.

Should :he break be isola:ed a: that t'i=e, sys:e= repressuri stion :c.

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the pressuri:er safe:y valve setpoint is p:cbable. Two-phase or liquid fl:v

,through the safety valves vill also probably occ"r." Cace the sys:e= de:le:es 7E '._

sufficien: inven:ory co-establish :endensation heat transfer ac ess :he the sys:e= vill depressuri:e a:;d no further loss of inven:::y vill occur.p Tae N

O" core vill re=ain covered f:: :his secenario and no te=perature excursi:n f

.She'uld the pressurizar safety valves beco=e da aged because of :he occurs.

two-phase flew ouc :he valves, the response of the sys:e= veuld then be l-- g si=ilar to that presen:ed 'in the ?,5jg for the pressurizer safety valve scuck open accident and no core uncovery occurs. W g

As far as the appropria:eness of the operator using pressurizer level indi:a-

. tion to trip :he F?! pu=ps, 3 G agrees tha: the level indica:ica is'not a reliable indication of :he s:s:e of the RCS. Ecuever, use of :he pressurizer level indication, 'along vich sys:e= :e=perature and pressure =easuremener to ensure that the systc= is still in a subs:ancially subcooled s:ste, vill p:c-vide sufficient guidance for operator action.

In su== ry, while the T/A ps;:er raises valid conec:ns :nd gives a de: ailed exacinntica of :he small becak ::ansient. :he s=all bec:k cpical repor:

provides sufficienc :nalyscs to ensure :he abili:y of chc 3 C 205 pt n: EC3 syste= to cent:ci s=all bec:k in :he RCS.

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The RCS and main cteam system pressure shall not exceed 110%

L-of the design pressure.

4.

Resultant doses shall not exceed 10 CFR Part 100 limits.

15.6.1.3.2 Methods of Analysis

,The ; analysis.of the effect of the; inadvertent $ opening :of the.

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Qpressurizer safety valves was performed using theismall break h model documented in ESW topical report BAW-10074A, Revision 1, "Multinode Analysis of small Breaks.for B&W's 205-Fuel-Assembly Nuclear Plants With Internals Vent Valves." The small break model used in that topical was demonstrated to ccmp'ly with the requirements of Appendix K to 10 CFR Part 50 in Appendix A of BAW-10104, "ESW's ICCS Evaluation Model." Flow thr, ugh the open safety valve was, calculated by the Moody critical flow correlation..The leak _ area. is limited by thelorifice area!of the safety valve,iwhich is 0.03 fta.

In order to obtain rated flow i

through the valve at.the valve rated pressure, it was necessary to use a discharge coefficient of 0.75.

Other assumptions used for the analysis are as described in 3AW-10074A, Revision 1.

15.6.1.3.3 Results of' Ana'lvsis' Following the inadvertent opening of a;: pressurizer safety or =

relief valve, the design pressure of the RCS is 'never exceeded.

The^RCS depressurizes to approximately the secondary side b_

pressure of the steam generator.

The secondary side pressure is maintained at apprcximately 1250 psia by the safety valves en tte steam generator.

Since the design pressure of the main steam system is 1250 psia, the system is maintained at below 110% of its design pressure.

Figure 15.6.1-1 presents the inner vessel liquid volume (lower plenum,. core, and upper plenum) for this accident.

Since the y core remains covered by-liquid throughout the transient, pool.

inucleate-boiling.willLbe maintained and the cladding temperatures will remain within a few degrees of the water temperature.

No, (metal-water oxidation is incurred, and no abnormal core geometries lor fuel rod ~ damage is ~ caused by the low cladding temperatures.

By 3200 seconds, the injection rate'from the ene HPI? exceeds the combined leak rate and core steaming rate and will pro ride long-term cooling capability.

Therefore, compliance with the acceptance criteria of 10 CFR Section 50.46 is ensured.

The discharge coefficient of 0.75 was chosen to match valve perf ormance with the Moody correlation.

If 0.75 is assumed, the rated flow through the valve at the valve's rated pressure is calculated by the' Moody correlation.

Although this matches the valve performance, it should be noted that the discharge O

15.6-2

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coefficient assumed for this accident is not important.

The

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system's response following the opening of a pressurizer safety valve is similar to that calculated for small breaks. 'DSWJ topical report;SAW-10074A, Revision 1, discusses the. consequences of small" breaks and demonstrates that thetcore remainsfcovereds

,throughout the transient f or breaks with?, leak : areas Tless than: 0.1 fte.

Since the leak area for an open safety valve 'is only 0.03

~f t *,' the core will remain covered thoughout the transient, and no fuel rod damage will occur regardless of the discharge coefficient assumed.

15.6,.1.4 Barrier Performa_nce The inadvertent opening of a pressurizer safety valve does not result in fuel damage or excessive pressure in the Reactor.

gCoolant. system or main steam system sinceipool nueleate. boiling,

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C maintains f the cl' adding temperatdrgiwithin 'a 'f ew' degrees of th'c

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water temperature 'and peak pressu' es never exceed the Code design r

limits.

15.6.1.5 Radiolecical consecuences Since no fuel clad damage occurs, there will be no increase of, radioactivity in the reactor ecolant or in the steam.

Because the deses are a function of the amount of steam released, the potential radiological consequences are bounded by the consequences of the loss of ensite and offsite ac power to the station transient (section 15. 2. 6).

15.6.I Failure of small Lines carrving Primarv ecolant Outside Containment 15.6.2.1 Identification of causes A break in fluid-bearing lines that penetrate the containment could result in the release of radioactivity to the environment.

There are no instrument lines: cennected to the RCS that penetrate the containment.

However, other piping lines from t,he RCS to tne Makeup and Purification System and the Decay Heae Removal System do penetrate the containment.

Leakage through fluid penetrations not serving accident consequence limiting systems is minimized by a double barrier design so that no single credible failure or malfunction of an active component will result in loss of isolation or intolerable leakage.

The installed double barriers take the form of closed piping, both inside and outside the containment, and various types of isolation valves.

The most severe pipe rupture relative to radicactivity release during normal plant operations occurs in the Makeup and Purification System.

This would be a rupture of the letdown line just outside the containment but upstream of the letdcwn control valves.

A rupture at this point would result in a less of i

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