ML19224B811
| ML19224B811 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/30/1978 |
| From: | Sternbert D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19224B810 | List: |
| References | |
| FOIA-79-98 PNO-780330, NUDOCS 7906260478 | |
| Download: ML19224B811 (1) | |
Text
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jed PoELIMINARY NOTIFICATION - NOT FOR PU3LIC DIscLO50.'tE y)
Ho: PN-NRC:! -40 Date: March 30,1978 PRELIMINARY NOTIFICATION OF EVDG OR UNU5UAL OCCURRENCE This oreliminary noti fication constitutes EARLY notice of events of P05SIBLE safety o_r oubile interest sieni ficance.
ine lnformation ce?seateo 1s is initially received witnout ver1fication or evaluation 7Aq t s casically all that is known at tne time of this noti fica on.
IT secuta sE 5PECIFICALLY horio f.-AT THIS 60TIFICATICDAY Cta ka INFCP.*ATION THAT LAiiR MAY SE CETER: DINED TO BE i?@CCU.4 ATE ilR INCCNS ISTEffF.
Facili ty:
Metro.colitan Edison Ccmpany (DN 50-220)
Three Mile Island Nuclear Station, Unit 2 Middletown, Per.nsylvan!a
Subject:
REACTOR 1 RIP AND SU3 SEQUENT BLO'AD0',a 'dITH E5 ACTUATION A reactor trip accurrad.at 2:30 a.c...P2rch 20, 1973, durin h 3 sys.t m E
- sting following initial cri ticali ty.cn Parch 1.3 Tae-cause of the trip was datannin+4 to ba the loss of a vitar bus produced a/ the trip of alt:'r-nl:e pcwer required by the test in progress acccmpanied by a lor.s of the ner.ai bus supply inverter due to a blown fuse.
The vital bus provided p:,.ar to the runnic ; pra ncnitor '.thich ir.dicated a loss of loo,' ?!cw beca.:se one RCP ws cut of service due tc a failed anti-reverse rotation devic 2 and the sicr.al frc,the other running pump was lost.
The PPS carrectly tri::-:d the reactbr.
kather cen?oner.t en tha vital bus tac tha rc er apsly f c -.e crimary syste3_EQCtrcratic__reli3_O/alveyhicjl_U1_ils_ogrL_c_rj_a_lIdl.
c4 ccwer.
IliTs resuitec in a c&pressurin: ton or tne primary.
e, a prr. sura
'oT5~iGsi, all SI pu: os started.
Part of the SI system includes the pcst accidant Sadlern Hydroxide (iiaCH) being injected into the iiCS.
Primary pH increased to about 7.5 before power uas restored to the bus and the event was terminatad.
Contact:
D. M. Sternberg 255 A. B. Davis 252 E. C. McCiM 210 Preparea by EKt.
Section Chief ist.
Branch Chief an Ci s tri bu ti on_:
E. L..Torcan, Executive Of fi.er for Op-raticas suppcet, HQ H. D. Tho nburg, Director, Division of Reactor Uper3 F. ions Inspection, HQ G. Klingler, PN Coordinator Transmittad to HQ
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PRELIMINA?,Y NOTIFICATICH - OT FOR PU',LIC DISCLOSURE IE.I For a 83
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- APR 0 41979 SSINS 8150 MIM02uiCQi FOR:
E. L. Jordan, Assistant Ifirector for Technical Programs, Division of Reactor Operations, II:'dQ FEO.u -
R. C. Lewis, Acting Chief, Reactor Operations and Nucl:ar Suppert 3 acch, Region II S T ECT:
P0m7IA PR03'IMS ICBTIFIED AT BA3C0CK A'iD WILCCX FACILITIES IN FIGARD TO TE THREE MILE ISLAND CCCUP:.INCI Eased coon prelicinary informatics fr 2 Reactor Inspectors dispa: chad to the Cr-stal River and Oconee Facilitie: to inns ti:;1*.e 7,eneric coac- : a of
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the Three Mile Island incident, the f ollowing potential problens hav-be-a identified.
Scre ci these natter n a'i be ' nc.m and evaluated, however, the; were identified in our initial re-iew.
Specifically, the follovi:3 aeth21 or. potential problecs hav: been 2dentified:
1.
East transients frc= high power cause fluctuations in pressurizer level that require cperatar artic to correct.
Ir sone instances these at.iots result in primary systra cooldown.
2.
Pressurirer lezijl___i;; ru.nentatation is not safe ty__ related bey :d General Design Criterica 13 of A,opandix A to 10C;350, !_h-re fo re,
sressuri er level is not necessa ru..
rt.
y_availa.,a to att tyte an ;c
- or for post accident recove y.
In addition, this ipstrumentatica da g
- .ptJave saaled reference legs which cean. the reference le n a h l
- .:i durin: tr2nsients resulting in erranecus level indicition_..
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features differ frca Westinghcuse P'a design which considers pre 32nri:3e level safety related and provides reactor trip and safeg'iards functicas and further designed with sealed reference legs to prevent flashing.
3.
The integrated control syster: (ICS) which controls primary and seccndary systems, including feedwater, steac generator.: vel, turbine, steaa bypass and rod control are_not sa_fety r_ elated and ara __ powered tr;:2
_ single vital electrical bus. In addition.pressurizar level and e e sure g;Lntrols a re_not safety related and powered. tron;nge_pr!.euugg Problems associated with this design include the following:
a.
During transients the turbine is soneti=es automatically tr.tas-ferred to nanual at fixed load while the ICS is reducin; reactac power.
This creates a ciscatch transiaat between turbia-load and reactor power.
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b.
The failure of the ICS power supply causes all controls to decand 50*. which places transient on plant if power level is at c.her than 50%.
- 4. ' In regard to IEB-79-05, both Crystal River and Oconee Plant procedures
$ require the operator to secure reactor coclant pumps when the pressuriner e=pties during an accident.
This procedure negates the assertion in IE3 79-05, Encicsure 2, thet void for ation en e ptying the pressurirer are dispensed by force flow.
In addition, the evaluation 1 ovided in IE3 79-05 also addresses transients as a result of a loss o. cff site pcwer.
In all cases, the loss cf site pcuer is assteed to result in a icss of the reactor ecolant pumps.
This fact. would certainly negate the f orced flow assertion.
5.
The accident anal / sis for a loss of feedwater flow froc high power indicates in sone cases that the pressurtzer cay go solid resulting in passing water thrcu;h the po,. r cperated relief valve and the safety val;es on the pressurirer. Two concerns as a result of this possi'oility har been identified.
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a.
The safety valves have been designed to pass water, a conc:rn exists as to whe.her the power operated relief is also desi;;ned to pass water?
b.
' "here is - 1"% que :: tor operated isolation valve devn strea=
cf the power operated relief.
The concern for this valve is
- whether the valve is G signed to close against full designed flow L.through ahe power operated relief ?
6.
Should a c:all loss of coolant exist, the current system design and cperatin; procadures allow for automatic release of reactor coolant external to the containment. These releases occur sia tihe sump pucps and reactor coolant drain system which pu=p automatica'lly to tachs outside the coatainment. This release would centinue until automatic safeguards actuatice occurs isolating containment.
In addition, the release of coolant could also result in the loss of water inventory assu=ed to be in the containment surp for FPSH requirements for decay heat purps during recirculation phase.
7.
Neither facility has hydroses reco=biners and rely strictly upon venting of containment for hydrogen control following an accidect.
Tae review of these itecs indicates general applicability of all itees at both f acilities exce.ot itees 1 and 3.b which do not appear to be problems at the Oconee Units.
Ia order for a tho_o ch evaluation of generic aspects of the Three Mile Island incident to ther Babcock and Wilcox facilities, the above itccs 2
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APR Q,t e,,
ma E. L. Jordan should be investigated at other B & W facilities. We are presently continuing cur investigation into these ite=s at Crystal River and Oconee.
We vill advise whe= = ore detailed or additional infor=ation is available.
/
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R. C. Lewis, Acting Chief Reactor Operations and Nuclear Support Branch cc:
- 3. H. Grier, RI J. G. Keppler, RIII K. V.
Seyfrit, RIV R. E. Ingelken, RV ee
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