ML20024B239
| ML20024B239 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/01/1979 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | |
| References | |
| TASK-*, TASK-GB GPU-2314, IEB-79-05-02, IEB-79-5-2, NUDOCS 8307070471 | |
| Download: ML20024B239 (21) | |
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>.'r IE Bulletin No. 79-05 Date: April 1, 1979 Pcge 1 of 3 EVALUATION OF FEEDWATER TRAN$IENT A loss of offsite power occurred at Davis-Besse on November 29, 1977, which resulted in shrinkage of the primary coolant ';olume to the degree that pressurizer level indication was Icst. A recommendation to convey this information to certain hearing boards resulted in the attached discussion and evaluation of the event.
This discussion includes a review of a loss of feedwater safety analysis assuming forced flow, which predicts dispersed primary system voiding, but no loss of core cooling.
During the Three Mile. Island event, however, the forced flow appears to have been terminated during the transient.
Attachment:
Discussion and Evaluation of Davis-Besse Transients
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8307070471 790401 PDR ADOCK 05000289 G
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IE Bulletin No. 79-05 Date: Aprff 1, 1979 Enclosurs 2 Attachm nt I-Page 2 of 3 Z Cy?2T 7%CM M22'.CRANCUM 12i~IIC "COXn..a.w. NW IFICE!ATICN TO I.ICINSI::3 BCA2.35. - DA7 5-3IS3I UNI S 2 & 3 AND MI::* WD UNI 5 1. & 2",
DA'"Z3
.IA3CA3.T 8, 19 79, 71Cu. J. S. C?.7.S.7m"o J.7. S'"-' -. :..
3 I=spectic= and ' 'c ceme== Rape:: 3C-346/78-C6 dee.c. tented da:
- pressurize: level had sc== cffseale f or acyrexi=2:ely five-
. '.=i=utas duri=g the Nove=he:!- 29,1977 1=ss od off site pove: eve =:.
Thers are s=s i= dica:1==s :ha: che: 3&~>T pla=:s =ay he.ve p =b-
- le=s =ai=- d ; pressu=6== level i=dicati=s duri=g r.a sie=:s.
I= addi:ic=, u=da= car:a1= c di:1 =s such as icss cf f ardsa:er a: 100% pcver v'. h de ranc::= ::cla=: pu=ps -- ' g da pres--
s==1: e =ay veid c._sle:ely.
A special a=alysis has hem = per-f===ed c==:er=ing
- '-d < eve :.
This a=alysis is attached. as h eksure 1.*
3ecausa cf pressurize level =ai=:e===ce ;;ch-lams -de s' '=g =f da pressurizar =ay requirs further :evia'v.
. Also =cced during de eve =: vas he fae da: '"::1d vece eff-
. seals (less dan 520c7).
I= addizi==, 1:-was:==:ed. da: the
~..akeup f1=u===1:::i=; is li=1:=d := less das 16C gc a=d da: =akeup f1=v =27 be substantially S=eaterf $2= dis vaha.
- This i=fe =:c:1 = sh=uld be a:n # ed 6 1.1.sh: cf de require-
= ants cf GOC 13.
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DISC S3ICN AND I"7A*CA~~C-i The avu== a Davi.s lesse whid es=1:ed i= 1 css cf pre'ssurize: level i=di=a:i = has bet: reviaved by h72. and d e c = 1 si = vas : tat =ed
-ha: =s==revizved safety ques ' = exis:ad.
.The pressu-1:er, :=ge:her vi:h the :ta::c: cc:12 ; =akeup sys:e=, is '
desig=ed :: =a1=:21: d e 'prir.a:r systa= pressu s a=d va e: 1s rel vi:'-d-dair eperati:=al l'-d:s c=17 4uri=g==:=21, cpera:d.=g c =d1:1:=s.
C:=1de _ :n--iz=:s, such as Icss cf effst:e p=ver a=d 1:ss =f feed-va:er, sene i=es ssul: i= p:d-
--r pesssure a=d veh=a cha=ge da:
a:e bey: d the ahili:7 cf this sys:e= := c:::::1.
he analyses cf a=d e=perie=ce vi:h such ::a=sie=:s shev, h vever, da: th ey c.a= b e sustai=ed vi deu: c =p :=is1=3 da saft:7 ef the ::ac:::.
he principel c==cer: caused by such :: -<1s=:s is' da: day nigh: causa veid' g 1:i che p f=ary c cla=: sys:e= tha: vce.1d lead to lhss of a':ility := ade-qua:ely. ::ci the reac::: c=re.
se saiety evalua:1 = cf :he less of of fs':e p:ver. ::a=sie=
shevs that, th= ugh. level i= dica:ie: is 1:st,
sc=e va e re=ai=s 1: :he pressurize: a=d the pressu:e d:es=== decrease belev abeu: 1600 psi.
h cid=4 fwr vwidiu, 1. 9 w a.u.,
the pressure =es de:: casa belev :he saturatio: pressure correspe= ding : the systs:
- rgera:ure.
1600 ps' 's :he satura:1c= pressure : r espeedi=g ::
605'7, whid is also the =axi=us allevable care cutle: :e=pera:::e.
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Yeiding 1: :he pr-l=ary syste= (excepti=g the pressu-izer) is p;ecluded i: :his casa, si=ce pressure d=es ne: decrease := sa:::a-t =.
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IE BulletinINo. 79-05 Date:
April 1, 1979
- , Attachment Page 3 of 3 N
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1"he safa:7 analysis f er =c = severs c=eid=v= : z=siah r, such as :he less of feedva:e eve =:, i:dia:as :ha: the vater volu=a c:uld dacrease
- s les's thau the syste= valu=e===1csive of :he pressuri:ar.
Duri=g such an eve==, the e=pt' ing of de. pressuri=== veuld be fc11 cued by y
a pressurs radueric= bel =v the sa:::atic= poi== and the fc=a:1= cf e ='l veids th:=ughou: =uch of the p d=a:7 systa=.
This.veuli==
result i= che icss ed c=:= c=cli=g because de veids veuld ha dispersed ever a large vol=== a=d forced $1=v wculd ; eve== :he= f = c:alesci=g s=fficiently := preve=
c= = caoling.
The high pressure c=c]a==
1=j ec=dc= p%s. are s-zned ac:c=2:1cally when the p d-':7 press =e decrezses bel =v 1500 psi.
Tharef=:e, any pressu = :=due:i== -hi=h is suffi=1r== := all== veldi=3 vill als res=1: i= varar i=j et:1= vhich vill :spidly res:cre the p d=ary va:a: :o nor-e lavels.
For :hase : mas =s, ve believe tha: he i=ah m :7 ef the p;e.tscri a a=d =c=21==o14=r skeup syste= := c==::1 s==e :::=sia=:s dcas===
l pr.; vide a' basis for recui:1=g =re czpaci:7 i= hese sys:a=s I
Ce=er.z.1.D esig: Critart== 13 of Appe d1= A = 10 w. 50 requires
- -M.:--
- a:i = ::===1:=: variablas ever dei; a= icipated' ra=ges fe:J"a=:icipa:ad.pera:ic-d ec:u n a==es".
Such.ec=ur:a= es are specifhally def1=ad := i=clude has of cifsi:a power.
Tae f ac:
- ha:.T c=1d gees cif scals at 520*7 is =c: c =ridered := be a dev-l.ar.1:n f:= t=1s requireme== b ecausa dis i=di=2:== is backed up by v.de Tz=ge e=pe =:::e i=di:::i = :h2= ex --ds := a 1:n -1: cf f0*?.
Ne.i:her d= ve c==sida the a.kaum fiev===1:::i=g := devia:r s1=:e tha a==.=::. cf :.akeup fiev i= e=:ssa cf 160 gp= d:es=== a: pea; := be a s *.~ '.. --..* a e - ~ ~. ' -
..u.a c 1.:s e. s...u.esa ec:::- ~ es. '
l
"'Se icss of presse:1:e: va a: level i=dicati== c= he c==sidered :=
devia:*. f:= GCC 12, b ecause dis leve' '-d'ca:1== pr vidas the =.in:1:21
er=s: cd da:ar 4-ing hz p r_== y c==12== i=w_=:::7 Ecuever, pr. si
cf a.lavel.i=di==:1:= dha: vculd c=ver all a=:icipa ed oc=u-z=:ss =27
=== be,.-sc:i=21.
As disc =ssad dove, he less cf feedva=e eve =: cz=
lead :s a===e=:a:7 condi:1= vhe=*'
== meani=gini level e=1sts, becaure cha e=: ire p:d- :f systa= c= : ' s a.s as= e..:e:
' :=:e.
6a i=:::due:1:= := A pe=di= A (las; paragrap$)
I: sh=.1d be :::ed -22::
P
- sces=1:es cha: fulf'7 '
e== cf s== cf the cri: aria =cy== = alwa7s be l
appr:7-12:a.- Tcis i=::cducti== also s:::ss tha: decartures f..
- he-Cri:eria =ust he idc=:1fied and justified.
The discussic= cf GCC 13 i= the D vis 3ess e 75A.T. 11.:s the vaca: l a el i=s tr.=.e=:2: i =, but does==
=e::1c= de possibility of icss of va:c: level indica:1: :
duri=g ::a=:ie=::
Thi.s appare== c=1ssic= 1= the safe:7 a=alysis vill be subjected to fu::her.reviev.
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l IE Bulletin No. 79-05 Date: April 1, 1979
,Page 1 of 3
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LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Bulletin Subject Date Issued Issued To No.
78-05 Malfunctioning of 4/14/78 All Power Reactor i
Circuit Breaker Facilities with an Auxiliary Contact Operating License Mechanism - General (OL) or Construction Electric Model CR105X Permit (CP) 78-06 Defective Cutler-5/31/78 All Power Reactor Hamer, Type M Relays Facilities with an With DC Coils OL or CP 78-07 Protection afforded 6/12/78 All Power Reactor by Air-Line Respirators Facilities with an and Supplied-Air Hoods OL, all class E and F Research Reactors with an OL, all Fuel Cycle Facilities with an OL, and all Priority I Material Licensees
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78-08 Radiation Levels from 6/12/78 All Power, Test and Fuel Element Transfer Research Reactor Tubes Facilities with an OL having Fuel Element Transfer Tubes 78-09 BWR Drywell Leakage 6/14/78 All BWR Power Paths Associated with Reactor Facilities Inadequate Drywell with an OL (for action)
Closures or CP (for infomation) 78-10 Bergen-Paterson 6/27/78 Ali BWR Power Reacter Hydraulic Shock Facilities with Suppressor Accumulator an OL or CP Spring coils
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r IE Bulletin No. 79-05 Cate:
April 1, 1979 Page 2 of 3 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS (CONTINUED)
Bulletin Subject Date Issued Issued To No.
78-11 Examination of Mark I 7/24/78 BWR Power Reactor Containment Torus
' Facilities with an OL Welds for action:
Peach.
Bottan 2 and 3, Quad Cities 1 and
- 2. Hatch 1, Monti-cello and Vermont Yankee. All other BWR Power Reactor Facilities with an OL for information 78-12 Atypical Weld Material 9/29/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-12A Atypical Weld Material 11/24/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-123 Atypical Weld Fhterial 3/19/79 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-13 Failures In Source Heads 10/27/78 All General and of Kay-Ray, Inc., Gauges Specific Licensees Models 7050, 70508, 7051, with the subject 70518, 7050, 70608, 7061 Kay-Ray, Inc.
and 7061B Gauges 78-14 Deterioration of Buna-N 12/19/78 All GE BWR Faci-Components In ASCO lities with an OL Scienoids (for action), and all other Power Reactor Facilities with an OL or CP (for information)
IE Bulletin No. 79-05 Date: April 1,1979 Page 3 of 3
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LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS (CONTINUED)
Bulletin Subject Date Issued Issued to No.
79-01 Environinental Qualifica-2/8/79_
All Power Reactor tion of Class IE Equipment Facilities with an OL, except the 11 Systematic Evaluation Program Plants (for action),andall other Power Reactor Facilities with an OL or CP (for in-formation) 79-02 Pipe Support Base Plate 3/8/79 All Power Reactor Facilities with Design Using Concrete an OL or CP Expansion Anchor Bolts 79-03 Longitudinal Weld Defects 3/12/79 All Power Reactor Facilities with in ASME SA-312 Type 304 Stainless Steel Pipe an OL or CP
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Spools Manufactured by Youngstown Welding and Engineering Company 79-04 Incorrect Weights for 3/30/79 All Power Reactor Facilities with an Swing Check Valves OL or CP Manufactured by Velan Engineering Corporation 6
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e ENCLOSURE 2 LIST OF LICENSEES AND CONSTRUCTION PERMIT HOLDERS RECEIVit4G IE BULLETIN 79-05 FOR INFORMATION Baltimore Gas and Electric Ccapany Docket Nos. 50-317 ATTN:
Mr. A. E. Lundvall, Jr.
50-318 Vice President - Supply P. O. Box 1475 Baltimore, Maryland 21203 Boston Edison Company M/C Nuclear Docket No. 50-293 ATTN:
Mr. G. Carl Andognini, Manager Nuclear Operations Department 800 Boylston Street Boston, Massachusetts 02199 Connecticut Yankee Atomic Power Ccmpany Docket No. 50-213 ATTN:
Mr. W. G. Counsil Vice President - Nuclear Engineering and Operations P. O. Box 270 Hartford, Connecticut 06101 Consolidated Edison Company of Docket Nos. 50-03 New York, Inc.
50-247 ATTN:
Mr. W. J. Cahill, Jr.
Vice President 4 Irving Place New York, New York 10003 Duquesne Light Company Docket No. 50-334 ATTN:
Mr. C. N. Dunn Vice President Operations Division 435 Sixth Avenue-Pittsburgh, Pennsylvania 15219 Jersey Central Power and Light Company Docket No. 50-219 ATTH:
Mr. Ivan R. Finfrock, Jr.
Vice President Madison Avenue at Punch Bowl Road Morristown, New Jersey 07960
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Maine Yankee Atomic Power Company Docket No. 50-309 ATTN: Mr. Robert H. Groce Licensing Engineer 20 Turnpike Road Westborough, Massachusetts 01581 Niagara Mohawk Power Corporation Docket No. 50-220 ATTN: Mr. R. R. Schneider Vice President Electric Operations 300 Erie Boulevard West Syracuse, New York 13202 Northeast Nuclear Energy Company Docket Nos. 50-336 ATTN: Mr. W. G. Counsil 50-245 Vice President - Nuclear 50-423 Engineering and Operations P. O. Box 270 Hartford, Connecticut 06101 Philadelphia Electric Company Dock.et Nos. 50-277 ATTN: Mr. S. L. Daltroff 50-278 Vice President Electric Production 2301 Market Street i
Philadelphia, Pennsylvania 19101 Power Authority of the State of New York Docket No. 50-286 Indian Point 3 Nuclear Power Plant ATTN: Mr. J. P. Bayne Resident Manager P. O. Box 215 Buchanan, New York 10511 Power Authority of the State of New York Docket No. 50-333 James A. FitaPatrick Nuclear Power Plant ATTN: Mr. J. D. Leonard, Jr.
Resident Manager P. O. Box 41 Lycoming, New York 13093
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3 Public Service Electric and Gas Company Docket No. 50-272
' ATTN: lir. F. W. Schneider Vice President - Production 80 Park Place Newark, New Jersey 07101 Rochester Gas and Electric Company Docket No. 50-244 ATTN: Mr. Leon D. White, Jr.
Vice President Electric and Steam Production 89 East Avenue Rochester, New York 14649 Vermont Yankee Nuclear Power Corporation Docket No. 50-271 ATTN: Mr. Robert H. Grace Licensing Engineer 20 Turnpike Road Westborough, Massachusetts 01581 Yankee Atcmic Electric Company Docket No. 50-29 ATTN: Mr. Robert H. Groce Licensing Engineer 20 Turnpike Road Westborough, Massachusetts 01581 Duquesne Light Company Docket No. 50-412 ATTH:
Mr. E. J. Woolever Vice President 435 Sixth Avenue Pittsburgh, Pennsylvania 15219 Jersey Central Power & Light Company Docket No. 50-363 ATTN: Mr. I. R. Finfrock, Jr.
Vice President 260 Cherry Hill Road Parsippany, New Jersey 07054 Long Island Lighting Company Docket Nos. 50-322 ATTN: Mr. Andrew W. Wofford 50-516 Vice President 50-517 175 East Old Country Road Hicksville, New York 11801 9
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Niagara Mohawk Power Corporation Docket No. 50-410 ATTN: Mr. G. X. Rhode Vice President System Project Management 300 Erie Boulevard, West Syracuse, New York 13202 Pennsylvania Power & Light Company
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Docket Nos. 50-387 ATTN:
Mr. Norman W. Curtis 50-388 Vice President Engineering and Construction (N-4) 2 North Ninth Street Allentown, Pennsylvania 18101 Philadelphia Electric Company Docket Nos. 50-352 ATTN:
Mr. V. S. Boyer 50-353 Vice President Engineering and Research 2301 Market Street Philadelphia, Pennsylvania 19101 Public Service Electric & Gas Company Docket Nos. 50-354 ATTN:
Mr. T. J. Martin 50-355 Vice President 50-311
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Enoineering and Construction 80 Park P' lace Newark, thw Jersey 07101 Public Service Company of New Hampshire Docket Ncs. 50-443 ATTN: Mr. W. C. Tallman 50-444 President 1000 Elm Street Manchester, New Hampshire 03105 Rochester Gas & Electric Corporation Docket No. 50-485 ATTN:
Mr. J. E. Arthur Chief Engineer 89 East Avenue Rochester, New York 14649 Metropolitan Edison Company Docket Nos. 50-289 ATTN:
Mr. J. G. Herbein 50-320 Vice President - Generation P. O. Box 542 Reading, Pennsylvania 19640
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UNITED STATES NUCLEAR REGULATORY COMMISSION 0FFICE OF INSPECTION AND ENFORCEMENT f
WASHINGTON, DC 20555 APRIL 5, 1979 IE Bulletin 79-05A NUCLEAR INCIDENT AT THREE MILE ISLAND - SUPPLEMENT Description of Circumstances:
Preliminary information received by the NRC since issuance of IE i
Bulletin 79-05 on April 1,1979 has identified six potential human, design and mechanical failures which resulted in the core damage and radiation releases at the Three Mile Island Unit 2 nuclear plant.
The information and actions in this supplement clarify and extend the original Bulletin and transmit a preliminary chronology of the TMI accident through the first 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> (Enclosure 1).
1.
At the time of the initiating event, loss of feedwater, both of the auxiliary feedwater trcins were valved out of service.
2.
The pressurizer electromatic relief valve, which opened during the initial pressure surge, failed to close when the pressure decreased below the actuation level.
3.
Following rapid depressurization of the pressurizer, the pressurizer level indication may have led to erroneous inferences of high level in the reactor ccolant system.
The pressurizer level. indication apparently led the operators to prematurely tenninate high pressure injection flow, even though substantial yoids existed in the reactor coolant systen.
4.
Because the containment does not isolate on high pressure injection
.(HPI) initiation, the highly radioactive water from the relief valve discharge was pumped out of the containment by the automatic initiation of a transfer pump.
This water entered the radioactive waste treatment system in.the auxiliary building where some of it overflowed to the floor.
Outgassing from this water and discharge through the auxiliary building ventilation system and filters was the principal source of the offsite release of radioactive noble gases.
5.
Subsequently, the high pressure injection system was intermittently operated attempting to control primary coolant inventory losses through the electromatic relief valve, apparently based on pressurizer level indication.
Due to the presence of steam and/or noncondensible voids elsewhere in the reactor coolant system, this led to a further reduction in primary coolant inventory.
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IE Bulletin 79-05A April 5, 1979 Page 2 of 5
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kipping of reactor coolant pumps during the course of the transient, to protect against pump damage due to pump vibration, led to fuel damage since voids in the reactor coolant system prevented natural circulation.
Actions To Be Taken by Licensees:
For all Babcock and Wilcox pressurized water reactor facilities with an operating license (the actions specified below replace those specified in IE Bulletin 79-05):
1.
(This item clarifies and expands upon item 1. of IE Bulletin 79-05.)
In addition to the review of circumstances described in Enclosure 1 of IE Bulletin 79-05, review the enclosed preliminary chronology of the TMI-2 3/28/79 accident.
This review should be directed toward understanding the sequence of events to ensure against such an accident at your facility (ies).
2.
(This item clarifies and expandi; upon item 2. of IE Bulletin 79-05.)
Review any transients similar to the Davis Besse event (Enclosure 2 7
of._IE Bulletin 79-05) and any others which contain similar elenents from the enclosed chronology (Enclosure 1) which have. occurred at your facility (ies).
If any significant deviations from expected performance are identified in your review, provide details and an analysis of the safety significance together with a description of any corrective actions taken.
Reference may be made to previous information provided to the NRC, if appropriate, in responding to l
this item.
3.
(This item clarifies item 3. of IE Bulletin 79-05.)
Review the actions required by your operatirg procedur.es for coping with transients and accidents, with particular attention to:
a.
Recognition of the possibility of forming voids in the primary coolant systen large enough to comprcmise the core cooling i
capability, especially natural circulation capability.
b.
Operator action required to prevent the fornation of such voids.
l c.
Operator action required to enhance core cooling in the event such voids are forined.
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'IE Bulletin 79-05A ApriT 5, 1979 Page 3 of 5 4.
(This item clarifies and expands upon item 4. of IE Bulletin 79-05.)
Review the actions directed by the operating procedures and training instructions to ensure that:
a.
Operators do not override automatic actions of engineered safety features.
b.
Operating procedures currently, or a're revised to, specify that if the high pressure injection (HPI) system has been automatically actuated because of low pressure condition, it must remain in operation until either:
(1) Both low pressure injection (LPI) pumps are in operation and flowing at a rate in excess of 1000 gpm each and the situation has been stable for 20 minutes, or (2) The HPI system has been in operation for 20 minutes, and all hot and cold leg temperatures are at least 1
1 50 degrees below the saturation temperature for the existing RCS pressure.
If 50 degree subcooling cannot be maintained after HPI cutoff, the HPI shall be reactivated.
c.
Operating procedures currently, or are revised to, specify that in the event of HPI initiation, with reactor coolant pumps (RCP) operating, at least one RCP per loop shall remain operating.
d.
Operators are provided additional infonnation and instructions to not rely upon pressurizer level indication alone, but to also examine pressurizer pressure and other plant parameter indications in evaluating plant conditions, e.g., water inventory in the reactor primary system.
5.
(This item revises item 5. of IE Bulletin 79-05.)
Verify that emergency feedwater valves are in the open position in accordance with item 8 below.
Also, review all safety-related valve positions and positioning requirements to assure that valves are positioned (open or closed) in a manner to ensure the proper operation of engineered safety features.
Also review related procedures, such as those for maintenance and testing, to ensure that such valves are returned to their correct positions folicwing necessary manipulations.
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IE Bu11ctin 79-05A April 5, 1979 Page 4 of 5
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l 6.
Review the containment isolation initiation design and procedures, and prepare and implement all changes necessary to cause containment isolation of all lines whose isolation does not degrade core cooling capability upon automatic initiation of safety injection.
7.
For manual valves or manually-operated motor-driven valves which could defeat or compromise the ficw of auxiliary feedwater to the steam generators, prepare and implement procedures which:
a.
require that such valves be locked in their correct position; or b.
require other similar positive position controls.
8.
Prepare and implement immediate'ly procedures which assure that two independent steam generator auxiliary feedwater flow paths, each with 100% flow capacity, are operable at any time when heat removal from the primary system is through the steam generators. When two inde-pendent 100% capacity flow paths are not available, the capacity shall be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the plant shall be placed in a cooling mode which does not rely on steam generators for cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
When at least one 100% capacity flow path is not available, the reactor shall be made suberitical within one hour and the facility placed in a shutdown cooling mode which does not rely on steam generators for cooling within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or at the maximum safe shutdcwn rate.
9.
(This item revises item 6 of IE Bulletin 79-05.)
Review your operating modes and precedures for all systems designed to transfer potentially radioactive gases and liquids out of the primary containment to assure that undesired pumping of radioactive liquids and gases will not occur inadvertently.
In particular, ensure that such an occurrence would not be caused by the resetting of engineered safety features instrumentation.
List all such systems and indicate:
a.
Whether interlocks exist to prevent transfer when high radiation indication exists, and b.
Whether such systems are isolated by the containment isolation signal.
O
IE Bulletin 79-05A April 5, 1979 Page 5 of 5 10.
Review and modify as necessary your maintenance and test procedures to ensure that they require:
a.
Verification, by inspection, of the operability of redundant safety-related systens prior to the removal of any safety-related system from service.
b.
Verification of the operability of all safety-related systems when they are returned to service following maintenance or testing.
c.
A means of notifying _ involved _ reactor operating personnel whenever a safety-related system is removed from and returned to service.
11.
All operating and maintenance personnel should be made aware of the extreme seriousness and consequences of the simultaneous blocking of both auxiliary feedwater trains at the Three Mile Island Unit 2 plant and other actions taken during the early phases of the accident.
12.
Review your prompt reporting procedures for NRC notification to assure very early notification of serious events.
For Babcock and Wilcox pressurized water reactor facilities with an operating license, respond to Items 1, 2, 3, 4.a and 5 by April 11, 1979.
Since these items are substantially the same as those specified in IE Bulletin 79-05, the required date for response has not been changed.
Respond to Items 4.b through 4.d, and 6 through 12 by April 16, 1979.
Reports should be submitted to the Director of the appropriate NRC Regional Office and a copy should be forwarded to the NRC Office of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, DC 20555.
l l
For all other reactors with an operating license or construction permit, this Bulletin is for information purposes and no written response is required.
Approved by GAO, B 180225 (R0072); clearance expires 7-31-80.
Approval I
was given under a blanket clearance specifically for identified generic l
problems.
Enclosures :
1.
Preliminary Chronology of TMI-2 3/38/79 Accident Until Core Cooling Restored.
2.
List of IE Bulletins issued in last 12 months.
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Enc 1bsura 1 to.
IE Bulletin 79-05A April 5, 1979 PRELIMINARY CHRONOLOGY OF TMI-2 3/28/79 ACCIDENT UNTIL CORE COOLING RESTORED TIME (Approximte)
EVENT about 4 #4 Loss of Condensate Pump (t = 0)
Loss of Feedwater Turbine Trip t = 3-6 sec.
Electromatic relief valve opens (2255 psi) to relieve pressure in RCS t = 9-12 sec.
Reactor trip on high RCS pressure (2355 psi) t = 12-15 sec.
RCS pressure decays to 2205 psi >
(relief valve should have closed) t = 15 sec.
RCS hot leg temperature. peaks at 611 degrees F, 2147 psi (450 psi over saturation) t = 30 sec.
All three auxiliary feedwater pumps running at pressure (Pumps 2A and 28 started at turbine trip).
No flow was injected since discharge valves were closed.
t = 1 min.
Pressuri:er leve1' indication begins to rise rapidly t = 1 min.
Steam Generators A and B secondary level very low - drying out over next couple of minutes.
t = 2 min.
ECCS initiation (HPI) at 1600 psi t = 4
.11 mi n.
Pressurizer level off scale - high - one HPI pump manually tripped at about 4 min.
30 sec.
Second pump tripped at about 10 min. 30 sec.
t = 6 min.
RCS flashes as pressure bottcms out at 1350 psig (Hot leg temperature of 584 degrees F) t = 7 min., 30 sec.
Reactor building sump pump came on.
(
2-TIME EVENT t = 8 min.
Auxiliary feedwater flow is initiated by opening closed valves t = 8 ' min. 18 sec.
Steam Generator B pressure reached minimum t = 8 min. 21 sec.
Steam Generator A pressure starts to recover t = 11 min.
Pressurizer level indication comes back on scale and decreases t = 11-12 min.
Makeup Pump (ECCS HPI flow) restarted by operators t = 15 mi n.
RC Drain / Quench Tank rupture disk blows at 190 psig (setpoint 200 psig) due to continued discharge of electromatic relief valve t = 20 - 60 min.
System parameters stabilized in saturated condition at about 1015 psig and>about 550 degrees F.
t = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />,15 min.
Operator trips RC pumps in Loop B t = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 40 min.
Operator trips RC pumps in Loop A t = 1-3/4 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> CORE BEGINS HEAT UP TRANSIENT - Hot leg temperature begins to rise to 620 degrees F (off scale within 14 minutes) and cold leg temperature drops to 150 degrees F.
(HPIwater) t = 2.3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> Electromatic relief valve isolated by operator after S.G.-B isolated to prevent leakage t = 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> RCS pressure increases to 2150 psi and electromatic relief valve opened t = 3.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> RC drain tank pressure spike of 5 psig t = 3.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> RC drain tank pressure spike of 11 psi -
RCS pressure 1750; centainment pressure increases frem 1 to 3 psig t = 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Peak containment pressure of 4.5 psig t = 5 - 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> RCS pressure increased frca 1250 psi to to 2100 psi
,-._e.-
. TIME EVENT
(
t = 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Operator opens electromatic relief valve to depressurize RCS to attempt initiation of RHR at 400 psi t = 8 - 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> RCS pressure decreases to about 500 psi Core Flood Tanks partially discharge t = 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> 23 psig containment pressure spike, containment sprays initiated and stopped after 500 gal. of NaOH injected (about 2 minutes of operation) t = 13.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Electromatic relief valve closed to repressurize RCS, collapse voids, and start RC pump t = 13.5 - 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> RCS pressure increased from 650 psi to 2300 psi t = 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> RC pump in Loop A started, hot leg temperature decreases to 560 degrees F, and, cold leg temperature increases to 400 degrees F.
indicating flow through steam generator Thereafter S/G "A" steaming to condenser Condenser vacuum re-established RCS cooled to about 280 degrees F.,
(
1000 psi
\\
Now (4/4)
High radiation in containment All core thermocouples less than 460 degrees F.
Using pressurizer vent valve with small makeup flow Slow cooldown i
RB pressure negative t
l
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e
c IE Bulletin No.79-05A Date: April 5,1979 Page 1 of 3 3
LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Bulletin Subject Data Issued Issued To No.
78-05 Ma'1 functioning of 4/14/78 All Power Reactor Circuit Breaker Facilities with an Auxiliary Contact Operating License Mechanism - General (OL) or Construction Electr,1c Model CR105X Permit'(CP) 78-06 Defective Cutler-5/31/78 All Power Reactor Hammer, Type M Relays Facilities with an With DC Coils OL or CP 78-07 Protection afforded 6/12/78 All Power Reactor by Air-Line Respirators Facilities with an and Supplied-Air Hoods OL, all class E and F Research Reactors with an OL, all Fuel Cycle Facilities with an OL, and all Priority I Material Licensees 78-08 Radiation Levels from 6/12/78 All Power, Test and Fuel Element Transfer Research Reactor Tubes Facilities with an OL having Fuel Element Transfer Tubes 78-09 BWR Drywell Leakage 6/14/78 All BWR Power Paths Associated with Reactor Facilities Inadequate Drywell with an OL (for action)
Closures or CP (for information) 78-10 Bergen-Paterson 6/27/78 All BWR Power Reactor Hydraulic Shock Facilities with Suppressor Accumulator an OL or CP Spring Coils e
r_.,
m.
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IE Bulletin No.79-05A Date: April 5,1979
('
i Page 2 of 3 LISTING OF IE BULLETINS ISSUEDINLASTTWELVEMONTHS(CONTINUED)
Bulletin Subject Date Issued Issued To No.
78-11 Examination of Mark I 7/24/78 BWR Power Reactor Containment Torus Facilities with an OL Welds for action:
Peach ifettom 2 and 3, Quad Cities 1 and
- 2. Hatch 1. Monti-cello and Vermont Yankee. All other BWR Power Reactor Facilities with an OL for infomation 78-12 Atypical Weld Material 9/29/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-12A Atypical Weld Material 11/24/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-12B Atypical Weld Material 3/19/79 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-13 Failures In Source Heads 10/27/78 All General and of Kay-Ray, Inc., Gauges Specific Licensees Models 7050, 70508, 7051, with the subject 70518, 7060, 70608, 7061 Kay-Ray, Inc.
and 7061B Gauges 78-14 Deterioration of Buna-N 12/19/78 All GE BWR Faci-lities with an OL Components In ASCO
- (for action), and all Solenoids other Power Reactor Facilities with an OL orCP(forinformation)
(
IE Bulletin No.79-05A Date: April 5, 1979 Page 3 of 3 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS (CONTINUED)
Bdlletin Subject Date Issued Issued to No.
79-01 Environmental Qualifica-2/8/79 All Power Reactor tion of Class IE Equipment Facilities with an OL, except the 11 Systematic Evaluation Program Plants (for action), and all other Power Reactor Facilities with an OL or CP (for in-format 1on) 79-02 Pipe Support Base Plate 3/8/79 All Power Reactor Design Using Concrete Facilities with Expansion Anchor Bolts an OL or CP 79-03 Longitudinal Weld Defects 3/12/79 All Power Reactor in ASME SA-312 Type Facilities with 304 Stainless Steel Pipe an OL or CP
~
Spools Manufactured by Youngstown Welding and Engineering Company 79-04 Incorrect Weights for 3/30/79 All Power Reactor Swing Check Valves Facilities with an Manufactured by Velan OL or CP Engineering Corporation 79-05 Nuclear Incident at 4/1/79 All Babcock and Three Mile Island Wilcox Power Reactor Facilities with an OL, Except Three Mile Island 1 and 2 (For Action),
and All Other Power Reactor Facilities With an OL or CP (For Information)
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