ML19332B260

Small Break W/Failed Power Operated Relief Valve. Related Correspondence

ML19332B260
Person / Time
Site:	Three Mile Island
Issue date:	02/11/1980
From:	BABCOCK & WILCOX CO.
To:
Shared Package
ML19332B231	List: ML19332B236 ML19332B237 ML19332B238 ML19332B241 ML19332B243 ML19332B245 ML19332B249 ML19332B260 ML19337A421 ML19337A422 ML19337A423 ML19337A424 ML19337A425 ML19337A426 ML19337A427 ML19337A428 ML19337A429 ML19337A430 ML19337A432 ML19337A434 ML19337A437 ML19337A439
References
86-1117679, 86-1117679-000, ISSUANCES-SP, NUDOCS 8009260380
Download: ML19332B260 (14)

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"hTED CORRESPONDM@Docket No 5 -289 Licensee's Exhibit No.

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BTW Document 86-1117679-000, "Small Break with Failed PORV," (February 11, 1980)

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_SMALL BREAK UITH FAILED PORY 5 h " 1 b k I Ih [ 9<> ". . ; -

1.. INTRODUCTION i

It has been established in reference 1, that very small cold leg breaks (<0.01) will repressuri:e to the PORV setpoint of 2465 psia if the auxiliary feedwater is delayed significantly. Since' there is a probability of the FORV sticking open aft 9.: being actuated, concerns have been raised regarding the impact of this consequential failure.

This report presents the results of an analysis of a 0.01 ft2 cold leg break with the subsequent fail'ure of the FORV to close.

2. SQS*ARY & CONCLt*SIONS -

As has been de=onstrated by the analyses presented in Section 6 of reference 1, )

small breaks in the pricary system will not cause a repressurization to the PORV setpoint unless all feedvater is lost to the steam generators. Under this j

, situatics, there exists a class of very small breaks, (.'.ess than 0.01 ft2 )

wherein the system uill repressurize to the FORV setpoint, An analysis is pre-sented herein for a 3.01 ft2 break, without feedwatet to the steam generator, which results in a rupressurizatica to approxi=ately the f0RV setpoint. t 20 minutes, the PORY wa.i actuated and was assumed to stick open.

As is demonstrated in Section 4, for the 177-EA lowered-loop plants, operator 1 action by 20 minutes to =anually actuate the two high pressure injection trains will keep the core covered. A qualitative analysis is also presented which '

demonstrates that reestablishment of auxiliary feedwater by 20 minutes, for both the 177-EA raised and lowered loop plants, will prevent core uncovery.

Therefore, a 0.01 ft2 break with no auxiliary feedwater can be mitigated safely with B&W's present operator guidelines. These operator guidelines require _ - . .

establishing feedwater to the steam generator as soon as possible, if ghe AFW is not available initially, and =anual init.iat. ion of the HPI upon loss of the cteam generator heat sink or saturated c.ond.i.tions in the primary system.

3. METHOD ON A"Al' ISIS Evaluations of very small breaks which ,- re.s. u.l.t. in repressurization phenomena cre presented in reference 1.

These-analyse.s

- - - . . .. de=onstrate that i.f auxiliary feedwater is delivered to the steam ge.ne.r.at. ors, the primary system would not repressuri e to the PORV setpoint. ...

Howe.ve.r, t.he analyses in i.eference 1 also f

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g3 -11176 79 -00 demonstrate that if feedvater is not delivered to the . steam generator within 20 minutes, there is a class of very small breaks, less than 0.01 ft2 , which will

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result in system repressuri:ation to the PORV setpoint. Since the PORV might stick open af ter being actuated, concerns have been raised regarding the impact of this consequential failure.

'An analysis of a 0.01 ft2 break.in the cold leg pump discharge piping, without auxiliary fecduater to the SG, was perfor=ed wherein the PORV was actuated and assu=ed to stick open. As has been demonstrated in reference 1, larger breaks will result in. auto =atic. actuation of the HPI system and will not repressurize.

While smaller breaks will:repressurize to the PORV setpoint earlier, less in-ventory would be lost out the break. Therefore, the 0.01 f t2 small break with the subsequent failure of the PORV is expected- to be the worst case for tran-sients of this type.

The analysis was performed using the B&W ECCS cvaluation model for the 177-FA lowered-loop plants.2 The analysis was perfor=ed using the same model and assumptions listed'in Section 6.2.1.3.5 of reference 1 with the :T1y changes being those =ade to reflect the PORV sticking open. Key assumpedens of the analysis are listed below.

1. The initial core power level is 102% of 2772 ESt' .

2. The core decay heat is based on 1.2 times the ANS standard. .

3. Operator action was taken at 20 minutes to manually actuate both HPI pumps.

4. The PORV was modeled as a leak path on the top of the pressurizer. The orifice area of .0373 ft2 was used, however, a C f 0.72 was utilized in D

order to reflect the proper relief characteristics of the PORY with the Nbody critical flow model.

5. ,The PORV was opened at 20 ninutes. This is consistent with the operator guidelines for a LOCA with no feedwater to the steam generators. However, if the operator'had not acted within this time frame, approximately a 2 minute delay in operator action would have resulted in the PORV betag actuated automatically. - -

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4. RESULTS Figures 1 through 7 show the system' response during the transient and Table 1 presents a sequence of events for this accident. The resultant system pressure response of a 0.01 ft2 celd leg break with no A N is shown in Figure 1. This particular response is due to (1) the loss of the SG heat sink; (2) no automatic HPI actuation prior to the loss of the steam generator heat sink; and (3) the opening of the PORV and actuation of the HPI at 20 minutes. As seen in Figure 1, the pressure initially decreases follouing the break opening. During this de-pressurization period, the reactor trips, the pumps trip, the pressurizer c=pties, -

and the steam generator secondary inventory boils off. With the loss of the SG heat sink, the primary system starts to repressurize before the ESFAS signal is reached. Therefore, the HPI is not auto =atically actuated. The system pepres-

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surizes to 2350 psia by 20 minutes at which t1=e the PORV was assumed,to open.

This is only 115 psi below the PORV setpoint which would have been reached ap-proximately 2 =inutes later. However, the operator is instructed to manually open the PORV if the system repressurizes and the SG heat sink is lost. Thus, the opening at 20 minutes is not totally arbitrary. During the system repres-suri=ation the pressucizer level increases (Figure 2) and when the PORV is opened the pressurizer rapidly fills with two phase mixture. At the time of the PORV opening, the two HPI pu=ps are t:anually actuated, and due to the addition of the cold =akeup water and the additional leak path area, the RCS depressurizes.

The inner vessel mixture height is shown on Figure 3. As can be seen, operator action by 20 minutes to manually actuate the HPI prevents core uncovery and a minimum two-phase mixture level of 4.5 feet above the top of the core is main-tained. Long term cooling is established at 25 minutes as the injected'HPI l fluid exceeds the core boil-off. Thus, the acceptance criteria of 10 CFR 50.46 are sa.tisfied. . .

1 While the analysis performed herein addressed the effect of operator action to manually actuate the HPI by 20 minutes, the effect of operator action to manually restore the auxiliary feedwater within 20 minutes can be qualitatively assessed.

As has been'shown in Section 6.2.1.3.5 of reference 1, actuation of the aux'iliary feedwater system at 20 minutes for a 0.01 ft2 break results in a rapid system depressurizatica and the subsequent actuation of the HPI. For the case analysed Berein, the depressuri:stion effect of the auxiliary feedwater would be faster . l than that shown in reference 1 due to the effect of the loss of inventory through the PORV. Thus, the HPI would be actuated earlier and long term cooling

Du -11176 79 -00 would be established' faster than that saoun in reference 1. Therefore, no core uncovery is expected if the operator only actuates the auxiliary feedwater sys-tem within 20 minutes and, contrary to the guidelines, does not manually actuate the HPI.

1 S. APPLICABILIIT TO DAVIS-BESSE 1 I

The analysis presented herein required that no auxiliary feedvater is delivered to the SG during the small break transient.

This situation is considered highly

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unlikely for Davis-Besse 1 because the auxiliary feedvater system is safet/ grade.

However, it is expected that the Davis-Besse 1 plant can' safely mitigate the accident if auxiliary feedwater is restored within 20' minutes.

The analysis presented herein was performed assuming operator action at~20 minutes to canually actuate the HPI. Due to the icw shutoff head of the HPI pump at Davis-Besse '., this operator action would not provide adequate protec-tion for this event ett Davis-Besse 1. "However, as previously discussed *, restora- l tion of the auxiliary feedvater system by 20 minutes is Ecpected to prevent core uncovery for the low tred-loop plants. Because of che raised-loop arrangement of the Davis-Besse u' tit, more of the loop water is available to drain into the

• vessel. , Thus, operar.or action by 20 minutes to restore auxiliary feedwater ,

at Davis-Besse would provide more =argin relative to core uncover than would ex-ist for the icwcred-loop plants and compliance with 10 CFR 50.46 would be easily  !

assured.

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86-1117679-00 Table 1. Secuence of Events Event Time, s

1. 0.01 ft2 cold leg break occurs o,o

2. Reactor trip, loss of feedvster, and RC pump trip 54.5

3. Main feedwater coastdo a ends 60.0

4. SG secondary boils dry 270,o

5. PORY opened 1200.0

6. EPI is =anually initiated 1200.0

7. Long term cooling established 1510.0 e

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REFERENCES I Letter J.H. Taylor (B5W) to S. A. Varga (NRC), " Evaluation of Transient

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Behavior and S=all Reactor Coolant Cystes Breaks in the 177-Fuel Assembly Plant," by 7,1979.

2 B.M. Dunn, et.'al, "B&W's ECCS Evaluation Model," BAW-10104, Rev. 3, Babcock

& Wilcox, May 1975.

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