ML19220C054
| ML19220C054 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/09/1977 |
| From: | Ross D Office of Nuclear Reactor Regulation |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7904280132 | |
| Download: ML19220C054 (2) | |
Text
,
~s Dis tribution set Fi NRR Rag. r u.e RS3 File Watt chron SC
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Docket No. 50-320 ME50RANDUM FOR:
D. B. Vassallo, Assistnnt Director for LWRs, DPM FROM:
D. F. Ross, Jr., Assistant Director for Reactor Safety, DSS SUILIEC1:
THREE E ISLAND UNIT NO. 2 STEAMLINE SP2AI ANALYSES The staff position en steamline break,nnnlyses is enclosed. The Reactor Systems Branch has compared the analyses presented by the applicant to the requirements identified in Issue 1 of NUREG-0138.
The staff finds that the analyses do not deronstrate that adequate safety-grade systems and components are provided to mitigate the consequences of postulated brenka in the secondat7 systes. Mini -
requirements for annlyses and guidance aa to when credit can be taken for non-safety-grada components are provided.
While the Rs. actor Systems Branch has enkan the lead in preparing this evaluation, it must be noted that determination of acceptability must be made by all of the respan.=ible branches. For exa=ple, the Analysis Branch must approve the codes and -mdeling, the various Engineering branches must approve the adequacy of systems and components, and the Plant Systems branches nust concur from their various disciplina stand-points.
Origir.a1 sf4T.-2 6 D. 7.Ron Detwood F. Ross, Jr., Assistant Director for Reactor Safety Division of Sysr*== Safety
Enclosure:
Staff Position ec:
(See attached list)
Contact:
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Three Mile Island Unit No. 2 Stea::1 Line 3reak Analyses Docket No. 50-320 S. Hanauer R. Heinemein D. Ross J. P. Knight R. Tedesco G. Lainas V. Benaroya T. Ippolito R. Bosnak Z. Ros: toc:7 H. Silver S. Varga T. Novak S. Israel J. Watt 86 180
i STAF7 POSITION Three Mile Island. Unit No. 2 The applicant =ust de=enstrate through analysis that adequate systa=s are provided to sitigate the consequences of stea-line and feedwater line breaks.
As stated in Appendix A, the applican =ust address these accidents in ter=s of whether the break is postulated to result from a seis=ic event or not.
When iniziated by a seismic event, the breaks are l'~d ted to occurring in non-qualified piping but the systa=s and conponents utilized to =itigate the consequences after considera:ica of single - failure =ust all be safety grade.
Their sufficiency =ust be de=cns: rated assu=in:
the worst single failure. When not initiated by a seismic event, the postulated pipe break can be assu=ed to occur anywhere in the secondary systas but credit can be taken for the action of non-safety-grade co=ponents as a backup when considering single failures in category I co=pon en ts.
As a sini=us, the staff requires that analysis be provided for the following u: cases wi:h additional tables and discussion :o suppor: the contention at they represent the nos: severe cases relative to core or contain=ent, break location, size, onsite vs offsite power, single failure, etc., as required in accident analysis.
The four cases are identified as follows:
1.
Stas-line break berween steam generator and =ain stea= isolation valve (seis=ic piping).
86 181
. 2.
Stea=line break after nain steam isolation valve (seismic event breaking non-seis=ic piping).
3.
Feedvater line break between check valve and steam generator (seismic piping).
4.
Feedwater line~ break before check valve (seismic event breaking non-seis:ic piping).
Table 1 provides a guide showing which co=penents in the present syste:
can be taken credit for in the analysia of each case.
In order to take credit for a backup, a single active failure in a si=ilar safety-grade co=ponent cust be considered.
As =ay be noted, there are no active safety-grade isolation valves in the feedwater lines.
The MSI7's are safety grade; however, their closure eine nust be accounted for if credit is to be taken for these valves.
The staff's evaluation of previous analyses is presented in Appendix A.
86 l82
. TA3LE 1 GUIDE TO CREDIT 1CR CCSTONF_;TS IN SECCNDARY SYSTEF. 3REAK XLM.YSIS Seis=ic Non-Seis=ic Seismic Non-Seis=ic Sten-line Steamlinc Feedwater Feedwater CASE 1 CASE 2 CASE 3 CASE 4 31SIV*
Yes Yes Yes Yes Turbine Stop 3ackup No 3ackup No Turbine Control Ne No No No Check Valves
- Yes Yes Yes Yes Seartup Backup No 3ackup No 6" Block 3ackup No 3ackun No TJ Cont cl 3ackup No Backup No 20" Block Backup No Backup No Feedpu=p C
C C
C 3 cost Pu=c Syste:
C C
C C
- Safety-grade cc=ponents which should be censidered for single active failures *. hen applicable.
Backup-Cc=cenent =ay be utiliced as a backup for a safecy-grade :c=penent.
C-Conseria:ive ceasedcun characteristics nay be utiliced.
/
O7 80
'1 0 J
APPENDIX A EVALUATION OF STEAM LINE 3RF.C ANALYSES Three Mile Islcnd U.it No. 2 Introduction The applicant has sub=1::ed reanalyses to decenstrate that the cafety related equip =ent =1:igates ' he consequences of postulated breaks in t
the secondary systen.
The NRC released NUREG-0138 in Nove=ber 1976.
Issue 1 of that docu=ent includes a s:aff position treating the use of acn-safety grade equip =ent in nitigating the consequences of postulated s:aan line break acciden:s.
In applying this posi:icn to the valving design of the Three Mile Island Uni No. 2 secondary systa=, the staff noted the dependence on non-safety grade cc=ponents and s rste=s, not as a backup but as the first and so=e-tines only line of defense.
This situa icn was discussed with the applicant in Dece=ber 1976 and the staff's evaluation is presented herein.
Su==ary The NRC positica, applying to breaks in the secondary systa= not caused by seis=ic events, per=its tahing credit for non-safety grade equipmen:
as a backup for single active failures in category I equip =ent.
Such credit cannot be taken when evaluating breaks in non-qualified piping resulting from a seismic event.
The applican: =ust de=custrate for breaks in non-qualified piping ini:iated by an earthquake, that only category I systats and cc=ponents will ni:igate the ccusequences given b0 lod any single active failure in :he category I cc=ponents.
.- The applicant has presented analyses based on sitigating the consequences of postulated secondary system breaks using non-safety grade equip =ent solely.
These analyses are unacceptable because (a) credi: is taken for non-safety grade syste=s and cc=ponents as the first line of defense; (b) single failures are censidered only in non-safety grade equip =en:;
and (c) the applicant has not addressed breaks initiated by a seis=ic evant.
Descriotion of Current Sys t gn The steanline systes is illustra:ed scheratically in figure 1.
The stes= lines are seis=ically qualified from the steam generator to and including the sain steam isolation valves (MSIV) located just curside of containment. Nonqualified steamlines e:ctend frcm the MSIV to the turbine stop valves which e=pty directly into the stea= chest.
Stess is =etered from the steam chest through four control valves to the high pressure turbina.
The sain steam isolation valves provide safety-grade ce=ponen:s for isolation of each stea=line.
The turbine stop valves p cvide a ncn-safety grade backup.
6 5
.. The feedwater systes is illustrated sche =atically in figure 2.
The feed-water lines are seismically qualified only fros the check valves located just outside of contain=ent to the steas generators.
Upstrea= of the check valves the line branches into 20" and 6" pipes for a parallel section.
The 20" section contains the feedwater control valve and a bicek valve.
The 6" section con:ains a feedvater startup valve and a block valve.
Up-streas of the parallel section is a feedwater heater, turbine driven feedwa:er pu=ps fed by the booster pu=ps and feedwater heater syste=s.
Discussion of Current Design and Analysis A break in the secondary systes is sensed by the feedwater latching sys:e=
which initiates closure of the =ain steas isolation valves, feedwater startup and control valves, and the block valves in series with the feedwater startup and control valves.
The turbine stop valves are closed on a turbine trip initiated by the reactor protection systes.
The latching systes senses two pressurer in each of the four steas lines.
Lev steam pressure in one steas generator relative to the other stern generator isitiates isolation of both steas generators through closure of the valves noted above.
The latching systen is being upgraded ro a safety grade systas except as discussed later.
Of the valves utilized for isolation, only the =ain steam isolation valves are safe:7 grade.
Their closure eine of 116 seconds will per:1: both steas generators to bicw down prior :o isola:f.on of the stea= line.
86 186
m The turbine stop valves are not safety grade. Nor= ally these valves are considered as an acceptable non-safety grade backup for the =ain stea:
1 solation valves when both have si=ilar closing times.
In this case, with a 0.5 second closure ti=e for the turbine s:co valves, the consequences of the accident change significantly depending on whether stea= line isolation is accomplished with one set of valves or the o ther.
The feedwater control and startup valves are not seis=ic Category I and are located in a non-se#e-#c area in non-seisnic piping.
Due to their location, even if the valves the selves were safety grade, the piping and wiring would not be seis=ic Category I. The closure ti=es are 12 seconds and 6.6 seconds for the startup and control valves, respectively.
T-ediate isolation of feedwater is critical for accidents depending on the MSI7 for stea=line isolation.
The block valves are not safety grade.
Due to their locatice near the startup and control valves, they have si 112: proble=s in being i= proved to safety grade.
These valves each have a 3C-second closure time and their motor operators receive power fro = non-vi:a1 buses.
I: is not currently known whether these valves could provide rapid enough isolation if they could be i= proved to safety-grade quality.
86 iB7
. The staff has evaluated the analyses of secondary s7sta= breaks (Appendin 153 of the FSAR and :he response te question 21.49) in ter=s of the require =ents identified in Issue 1 of NURIG-0138.
The analyses presented by the applicant are s"--'ri:ec in Table 1 and nay be described as treating two events.
The stea=line break =ay be considered to have occurred on either side of the MSI7 and is analyzed first considering the failure of a turbine stop valve := isolate the confaulted steas generator and secondly with the failure of a con::ci valve to isolate the feedwater line.
A feedwa:er line break be: ween check valve and steam generater is presented considering the failure of a feedwa er control valve.
Because of the lack cf relevant safety-grade equip =ent in the steamline and feedwater line, caly non-safety-grade equip =en were used Oc nitigate the consequences and also censidered in the failure =cdes and effects analysis.
This absence of safety-grade equip =ent is unacceptable because 1: violates the staff posi:ica stated in NUREG-0133, especially if even:s initiated b7 a safe shutdcwn earthquake ar2 considered.
The applicant =ust also de=cnstrate that stable long-term cooling can be achieved.
The short tern of previously submittad analyses failed to de=enstrate this systes capacili:7
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