ML20114E412
"Draft Meeting" is not in the list (Request, Draft Request, Supplement, Acceptance Review, Meeting, Withholding Request, Withholding Request Acceptance, RAI, Draft RAI, Draft Response to RAI, ...) of allowed values for the "Project stage" property.
| ML20114E412 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 07/08/1966 |
| From: | Gaske M Advisory Committee on Reactor Safeguards |
| To: | Thompson T Advisory Committee on Reactor Safeguards |
| Shared Package | |
| ML093631134 | List:
|
| References | |
| ACRS-GENERAL, NUDOCS 9210120161 | |
| Download: ML20114E412 (13) | |
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HDERANDUM To Dr. T. J. Thoapson i
Indian Point 2 Subcommaittee Chairman From M. C. Caske, ACR2 Staff Earvin C.
Gaske' i
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Subject INDIA!1 Po1NT 2 SU3CONNITTEE WETIIIG. JUNE 13, 1966 i
Attached is a draf t of the Subcommittee minutes for the June 23, 1966..o l
Indian Point 2 Subcotenittee meeting which was held in Washington, D. C.
Copies 'of the minutes are being distributed to the other ACRS members who attended the meeting, in the enrent they wish to coassent, and the remainder of ACRS for information.
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MINUTES OF I
INDIAN POINT 2 SUBC(39GTTER MEETING Washin8 ton, D. C.
June 23, 1966 l
The purpose of the Indian Point 2 Subcommaittee inseting held en June in Washington, D. C, was to continue the Subcossaittee review of Consolidated 23, 1966, Edison Company's application for a construction pcruttt for the Indian Point 2 facility. Present at this meeting were the follovirs:
_ACRS LeBoeuf. Lamb & Leiby
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T. J. Thompson P. A. Gifford E. B. Themas S. H. Hanauer A. A. Upton H. W. Hewson N. J. Palladino Consolidated Edison Consultants M. C. Casko, ACRS Staff G. Brown Renulatory_ Staff C. R. McCullough--
W C. L. Allen
_ estinahouse Electric Corp.
R. 3. Boyd L. 1. Kopp R. N. Andrawe D. R. Mueller
- 8. S. Beckjord P. E. Norlan A. R. Collier z
F. Schauer H. J. Cordle R. L. Waterfield R. J. French J..Proctorp.C9nsultante.a.
G. A. Harstead W. Laster Consolidated Edison Company J. S. Hoore R. C. Nichoin W. J. Cahill, Jr.
L. Porse H. W. Dierman J. G. Russell J. J. Grob, Jr.
H. L. Russo J. A. Prestele T. Stern C. F. Soutar United Enaineers & Constructors 8..B. Barnee J. H. Hamearth R. O. Ishoff D. Rhodes D. Room-1 ACM
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gxecutive Session Dr. Thompson stated that the possibility ot a reactor core melti a molten mass of material has previously been considered for faa and forming The probina of recriticality was the primary item of concern
- eactors.
m been placed lalow some fast reactors to separate the molten fissionabl
, and dividera have al should a core melt down occur.
e materi-Preliminary Safety Analysis Report.Dr. Hansuer cal.ed attention to th ent tot the Ind;.cn Point 2 reactor pressure vessel will have the capabilitThis' par breaching of the containment by tho molten core through the use of w t 4
y of preventing cavity. lie believed that a er in the there esy not be au unacce,ptsbie acfety problem Lavolved withif Consoli che Indian Point 2 facility.
core melt down at Third Supplement (Question 9 reads:The question was raised whetner Quastion 9 in the reliability of the emergency core cooling system.) is present because ofP on the part of the applicant or at the request of the Regulatory Staffa desire It was reported that, on Juno 14, 1966, the DRL Staff contacted th the applicants regarding Dresden 3 and Indian Point 2 relatire e ACRS Dffice e
y of a core melt down accideot without ve to (1) the course the reliability of the emergency cooling systems. core cooling systems in operation and (2) representatives of the applicant for each reactor and informed them that neith t..a ACRS nor DRL was requesting that written information b er to the above two items.
Commonwealth Edison representativese provided in regard that no written Laformation would be submitted regarding Dr indicated to DEL Third Supplement which they were in the procese of prepa esden 3.
Con Ed repre-e ir emphasized that neither thsy nor the ACRS was requesting such written inf submission.
DEL ormation.
Prior to the June 23, 1966 which might ha discussed at Subcommittee meeting, con Ed requested a list of items Con Ed, and a copy of this list is attached.the Subcommittee meeting. Such a list was prov tea adght want Dr. Thompson stated that tha Commit-be perforned regarding the problem of the consequences of a c earch own accident.
in regard to item 3 of Attachment 1 (reactivity transi that sufficiantly comprehensive and did not indicate how c serious accidenta if reaccivity coefficients slightly different from those a e to were used in the analy' ses.
of characteristics of a number of boiling water reactors and pressu ssumed reactors (list forwarded to Ceumsittee members on 6/23/66).
five Westinghouse ranciors in succession have the same general reacti it He pointed out that tsristics and that San Onofre is the first v y charac-in this series.
l He indicated he believed chat, if the ACRS approves the proposed operation of
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San Onofre, the Committee may have established a precedent for the eli s. ole s
al}oVAbh resetivity characteristics of subsequent Ueatinghouse reactors.
Ha also pointed out that the Committee had reccatly acted in a favorable manner regarding che Brookwood application and that t's Brookwood reactor has essential-i Ly the same reactivity characteristics as those of Indian Point 2.
Dr. Thompson stated he believed that the Indian Point 2 reactor could be made as safe r:1stiva to reactivity treusients as any i operation. He said that Con Ed might have to rutreat somewhat and that, perhaps, the ACRS should inform Con Ed that they cast place a burnable poison in the coro during the first part of the life of the initial core.
He stated tnat Con Ed could place a boron stainless steel or otter suitable poison in rapty positions in the fuel essemblies and thereby make tha moderator temperature coefficient significantly less positiva during the initial operation of the reactor.
ReRulatory Staff DEL reported that a.5% reactivity insertion during full power operation would bring the hottest fuel pellets in the core tc a just molten condition.
A 1%
insertion from full power would result in 11% of the core reaching a molten con-diciou.
If 11% of the core were te nelt, it might jeopardize the integrity of the primary system.
The DRL Staff reported that their consultants for structural matters, Newmark j
and Hall, had most of their questions regarding the containment structure answered in Supplenent No. 3.
DRL understood that they still had one minor point i
which was unresolved, but DRL had not yet secured the final report from their consultants.
Mr. Boyd stated that the DRL Staff presently believes the proposed Indian Point 2 plant is an acceptabla one.
They consider Indian Point 2 to be a "suburhann re actc. .
formation to convincecaDr. Hansuer inquired whether the Staff was satisfied it had enough in-itself that there is not a great extension of the contain-ment through use of the isolation valve seal water system and that failure of the system will not cause greater leakage than the system is intended to eliminate.
The DRL Staff indicated they were satisfied with the now seal water system which has been proposed by Westinghouse.
An intermediate size pipe break in the primary system is considsred to be the worst type of pipe rupture.
In event of a large pipe break, the system will rapidly blow down,and the low head satety injection system pumps can inject water into the reactor pressure vessel.
safety injection pumps can handle the situation.If the pipe break is small, the high head Mr. Boyd stated that Coo, Ed had elected to provide the information contained in Question 9 of the Third Supplement on their own initiative.
The DEL Staff indi-cated that they did not believe chare was a significant difference between Indian Point 2 and Dresden 3 regarding the consequences of a core melt down accident
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and the ultimate fate of the molton fuel. Mr. Boyd said he recognised the incon-siste cy of the Staff's present position that the core will amit for fission pro-n duct release and metal-water reaction considerations but that the core will not melt and cause a problem as a result of the molten mass of material which would be present.
Mr. Proctor reported that he and another person at the Naval Ordnanco Laboratory had both developed models regarding the effects of a pressure vnesol rupture.
J Both of these models gave approximately the smee results as the model used by Westinghouse. If a longitudinal rupture occurn4 the pressure vessel would move outward approximately six inches and contact the concrete shielding located around it.
Simultaneous r:.pture of the pressure vessel head bolts would result in the head passing through the top of the containment. Westinghouse still believes that such a rupture of the pressure vessel head bolts is incredible.
if ths head were to fly off in such a manner, the entire core would probably be lifted.
I Mr. Norian of the DEL Staff reported that it appeared that the only pressure vessel failure that the reactor containment could not withstand is simultaneous f&ilure of the prassure vessel head bolte.
Consolidated Edison Representatives Mr. Beckjord stated that Westinghause had concluded that the ahielding structure -
around the reactor pressure vesset would not fait violently so as to cause missiles in the event of a longitudinst rupture of the pressure vessel. West-inghouse has assumed that a longitudinal crack would proceed down the wall of the pressure vessel but would stop'at the bottom hemispherical region which is located outside the area which might be made brittle by irradiation.
Mr. Beckjord read a surmaary of Westinghouse's cocclusions regarding rupture of the pressura veesel below the flange. The statement was se follows:
" Control rod drive mechanism shield, reactor vessel shield, lif ting rig, control rod drive mechanisme, stude, that part of vessel and flange above breaks, internals, core barrel and core -are acceler -
ated to a velocity of approximately 85 ft/see at an elevation of approximately 25 It above starting point.. With no collision with crane, the above masses would rise se additional 112 ft approst-mately to a point about 21 f t below the top of. containsusat. Tt a collision with crane occurs, the initial mass together~with crane would rise between 30-39 f t based on inelastic impact. leaving a distence of 16-25 f t vertical rise to the point where the crant girder projection would touch the containment."~
Westicanouse has etudied reactivity inserteon accidents involving.the' release of large quantities of sensible heat to the reactor coolant. These studies indicate that the pressure vessel might f&il due to the-quasi static pressure risa in the vessel. The quasi static pressure is the pressure which persists in the pressure vessel after the initial shock wave pressure risu.
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i If pressure wire continuously raised in the pressure vessel, the concrete shielding surrounding the vessel would be contacted by the vessel and the shielding would eventually fail. The concrete shielding is capable of with-standing an internal pressure of approximately 750 to 800 psi. The design yield strengs (650*F) of tne pressurn vessal head bolts is 9300 psi. Mr. Baekjord stated that the prinsry system piping and steam generator tubing would be expected to fail before the pressure vessel.
He believes that hhe previous Westinghouse estimate of 500 - 550 f t rise of the pressure vessel head, in the event of simul-taneous rupture af the pressure ves.wel head bolts, is conservative and that the vessel head might not rise that high.
Mr. Beckjord said be would classify the possibility of pressure vessel failures as folk est 1.
Simultaneous pressure vessel head bolt failure - incredible 2.
Circumferential pressure vessel f ailure above the naanles; but Eelow the flange - not a finite probability 3.
Failure of the pressure vessel below the aczabis - extremely remote Con Ed does not intend to place a microphone in the coniainment to assist in the detection of leaks within the containment. Failure of four adjacent head bolts would prevent reactor pressure frosa being maintained. Westinghouse has recos:xnended to Con Ed that one half of the head bolts be inspected th' ugh use of ultrasonic testing and dye penetrant iesting at each refueling. They also have reco::rnanded that all of the head bolts bc visually inspected at each refueling.
The reactor core would be expected to rise if the pressure vessel hard lifted from the top of the pressure vessel.
Mr. Beckjord stated that the l'roblem is alleviated somewhat because most of the water in the pressura vessel is located above tha core. It van his opinion that the core would probably rise approxi-mately 20 to 25 feet but would not leave the pretsure vessel.
Dr. Gifford stated he believes the degree of concern relativa to off-site emargency procedures should have some relation to the number of reactors pres-eat at a sita. Con Ed har had some discussions with the Nov York State Police regarding arrangements for amargency off-site evacuation. Con Ed has not, how-ever, tested emergency procedures for major accidsnts because of possible adverse public relations affects. No arrangements have been nada with a nearby vaather station for weather information to be provided in the event of an accident.
Dr. Gifford suggested that such arrangements might be mada and periodically tested.
Dr. Doan c.tated that Con Ed seems to have developed tempes2!!s plans forohandling releassa of radioactivity at Part 20 limits rather than for the more significant releases which might result from possible accidents.
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d Dr. Hanauer pointed out that, if a group of intelligent people can nacidevise j
emergency plans for use.during accident conditions, the reactor supervisor can j
not be expected to devise such plans on the spur of the moment at the time of an accident.
Dr. Hanauer inquired how well con Ed would know what quantity of sadioaepivity was being relemed from the reactor stack in the event of a major. accident.
Con Ed repressatatives did not know the range of sensitivity ~of the presently insts11ed Indian Point 1 stack monitor nor the range of sensitivity of monitors to _ bei i.
installed for Indian Point 2.- Dr. Hanauer cautioned that the need for stack j
monitors with an adequate range' for possible ladidanPoint 2 accident conditions was being indicated to con Ed at an early. stage.
N 47 psig containment design pressuro La - greater than the. pressure _in the
_i water in ene cooling coils of the air recirculation system located inside the
!l containment.
h pressure in the coils _is~approximacaly.20 to 25 peig.
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Dr;;. Thompson was concerned that, during ' accident conditions there might be a leakage of radioactivity to the river water circulating through the coils.
Dr. Hanauer pointed out that, in the event of a loss-of-coolant accident,- three sets of pumps are required to function to remove heat through the residual ~ beat j
exchangers.
N service water system supplies coolant water from the river to the reactor plant component cooling heat exchangers. N-component cooling heat -
exchangers in turn provide cooling water to the residual heat exchangets located inside the containment structure.
Thec a residual heat exchangers then remove heat from the containment internal recirt.ulation system.
l-h isolation. valve seal water system has been redesigned to ine rporate the use of valves.which seat at both the inlet and outlet portions of the t alves.
can be injected into the. area between the two seats to prevent the leakage-cf Water L
gases through the valves when they are in a closed condition.
. being used, _ in liey of the seal vetor ystema described in the ' application, in These valves als order to lessen the possibility-that failure of the ~ system might result in a
. leakage path from the containment.
h applicant saaintains that,the new design j
.is merely a change in detail and does not intend to submit an amendeont describ-ing the new system.
Con Ed reported that there has been experience _in the gas industry with the typ of valves being proposed 'for _use in the contaitement iso-t lation valve seal water system.
Following a loss-of-coolant accident and injection of the contents of the refuel-ing water storage tank into the containment, the water level._in the containment structure will be approximately 3 feet above the floor level.
h attached sketch L
was drawn by Westinghouse to show the level that the water would reach relative to l
to the cora.
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Westinghouse supplied the following table of items which would be in operation if two of the three emergency diesels araidparablet INDIAN FOINT FIANT SAFETY INJECTION OPERATION EMERGENCY POWER i
I (Injection Phase)
(Internal RecirculationL l'
Compouant Horse Power Component Horse Power l
4 fans 1400 2 fans 700 i
safety injection pump 350 1 residual heat pump 250 a
1 spray puisp 350 2 service water 3 service water pumps 800
- .pumpsa
.m a r 1200-1 racirculation pursp -
350
'l component cooling-2 pump 200 i:
l Total' 3150 2450 j
Approximately one hour af ter the occurrasca of a loss-of-coolant accident, exter-i nal pumping of water through the safety injection system and the containment, spray system is ended, and a shif t ia made to the internal recirculation operational' mode.. Approxim6tely 10 minutes are required to start and stop pumps, close valves, trip containment' fame off the lina, etc., during the shif t.
.Dr. Hanauer inquired whether there would be adequate cooling through all of the charcoal filters if-two of the air recire:ulation system fan cooters are turned off af ter the filters 1,come_ laden with fission products. There is one charcoal filter is series with
- A t fan.. The discharge free each fan passes into a coasson discharge duct. West-Lyouse maintained that, if two of the fans were turned off, there would _ be-hufficienou back flow through the other fans and their associated charcoal filters to prevent the charcoal filters from buradgg. The refueling water storage tank has a capacity of 320,000 gallons. It.will take between one and two weeks to mix the boric acid solution required to fill the tank.
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I Con Ed will establish a criteria as to how much water must be in the refueling water storage tank to permit reactor operation to proceed. Westinghouse reported that, at a temperature just above freezing, the borou concentration in the refuel-ing water storage tank would be approximately a f actor of two below the satura- -
tion concentration.
Dr. Newson inquired as to the reason the reactor pressure vessel is not located deeper in the cavity at the botto 4 of the containment in order to obtain better assurance that the core would be flooded if water la injected into the containment.
Mr. French of Westinghouse reported that burnable-poison could be placed in the core without seriously affecting the power distribution. He stated that, if one weight percent boron stainless steel were inserted in the core, only 8% of the original boron in the stainions steel would be present at the and of the first. core cycle. Westinghouse indicated that, if found necessary, a burnable poison could be added to the fuel assemblies near the time that. reactor opern-tion is to begin.
Dr. llanauer cancioned that, in view of the fact that rod pro-graming is very important in the prevention of reactivity accidents, this matter will be looked at' carefully at the operating license stage.
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Westinghouse presented the following table:
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I; LINES OF FRwwuCN FOR CORE MELT TRRDUCH
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l Reactor coolant systee integrity
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a.
Designed to quality control standards 1
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b.
Pressure vessel above NDI j
c.
Ductile pipe 2.
j Safety injectiou systes reliability a.
Redundancy j,
b.
Five puxape (3 high head, 2 low head) not including recirculation j
Pumpe c.
Eight injection points
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d.
Three diesels,-3 instrument and control ch=anals Roactor vassal and reactor coolant pipes supported for double end-i e.
ed pipe severance-lf f.
Failure analysis i
3 Recirculation 6
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3.
Safety Injection Perforzaance i
Approx. 207, of installed deluge flow will prevent melt thW a.
b.
3 diesels - only two required:
For the core melt through accident, Westinghouse has postulated that approxi-mately 607, of the core is in a molten condition at the bottom of the reactor pressure vessel and that the bottons of the vessel is located ;in' water.
Several inches of insulation will be present around the pressure vessel, but this in-sulation is of a reflective type.
It consists of alum 4=== foil placed between i
stainless steel sheets. Mr. Stern stated that, if steem is' generated'in the l
insulation, water can ea===deata freely through the issuistion, and any. steam l
formed would be readily removed. Asmusstag 601, of the sore is metten and pre-sent in the bottom of the pressure vessel, and that the after heat generation-l-
rate is 2% of full power, Westinghouse believes ~ that the.aore would met asalt lg through the bottom of the pressure vessel. - Westinghouse, however, had me ex. :
per4= ament evidenes to support their oostantion that there womid be eniedequate I
hast flow throud the presesre veneel.
It une potated est that,. if 60%\\ef the core were actually molten, the Anel pe11ste reanimies la a solid form would probably fall to the bottom of the'veneel. The questies'une esteed as te the affect the crust, whien muld form on the' outer surface of the molten itsel,.
would have on best trane,a characteristica. Westtagbouse bee'salaulated ths.t.
. if the core were molten and dropped into water, there weeld be a 15 peiirise in the presours taside the en=*minamme.
It vae suggested to Westfacheese that they might wish to consider the malt through ' accident a general' problem and publish a report on it that would not be related.to a particular reactor ACF'
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i Weseinghouse presented the following table which represents the sequence.of eventa l
if all of the core is allowed to salts SEQUENCE OF EVENTS WITH NO ACTIVE QUENCRING OF CORE Di REACFOR VESSEL s
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blowdoun 0 - 12 seco.
i j-1st some melts 145 esca.
i bottom of pressure vessel dry 575 sees.
l-j_ j barrel melts 'ott.m y; eenu.d wy 2300 secs.
c.
I melt through pressure vessel 2300 sees.
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Executive. Session l
Dr. Thompson believed that con Ed might be asked to cesiders i
1.
Pcovision of meena of insuring an adequate==i===== floodtag depth
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to cover the core for any benake in the primary system.
I 2.
Placing a burnable poison in the oore so that the questity of boron i
in the coolant would be limited to the===$=mg of 1500 ppa during j
reactor operation.-
3.
Placing a large U tube between_ the Indian Point 1 and Indian Point 2 i
contai====*= to provida for bl* of pressure into the other _ con-l taia===e in the event of excessive pressure in eitherfa=*=i====*.
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4.
Providing for cooling water to be added to the seeendary side of the f
i eteen generators in order to *== steen and remove heat on the j
l primary side in the event of a lose-of-coolant aseideas.
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5.
The addition of a holddous device to prevent the eere fasa rising in j
the event'of an accident.
l in Dr. Theepsea decided to present the above/taformet di.,msssene with the app 11-sent and te_ present theos as _his oun fadividual ideas. -The asemer of handling e
eesh of the tem items la the list presented to Ces Ed for oral discussies et
- the suboeumittee asettag'was agreed upon.
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i 11detM Edh..hmtives j
Dr. Tbepeen referemoed the tea items la the list provided Con Ed for oral dio-l-
i cussion at the Subcommittee meeting. He stated that cea Ed shoald be prepared to disones these items at the full Ceamittee anotias as-felleest a
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1.
Be preparud to discuss.
2.
Be prepared to discuss.
3.
Be prepared to discuss briefly and indicate the general means of re-treat if the magnitude of the positive void coefficient is consider-ed too large.
4.
Be prepared to discuss.
5.
Be prepared to discuss briefly.
6.
Be prepared to discuss, especially the operatin6 experience with the type of valves proposed for use.
7.
Be prnpared to discuss, particularly possibility that internal pressure of containment may exceed the pressure of the water in the cooling coils of the air recirculation system.
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This item appeared to be reasonably well do===ted in the Third Sup-plement.
9.
tais item also appeared to be sufficiently well documented.
f 10.
Emergency procedures would not be discussed at the full Comeittee meet.
in8, but Con Ed should keep emergency procedures in inind for the operat.
ing license stage.
Dr. Hanauer stated that a question which should be explored is what would happen if the core melted through the pressure vessel.. Westinghouse indicated they be-lieved that, when the molten fuel melted through and hit the water located below, there would be rapid chilling and formation of solidified uranium oxide.
Dr.-
l Thompson a phasized the need for adequate information to be provided to the full Committee regarding the core melt through accident.
Westighnuse reported they are considering the insta11stice of an opea stainless steel tank at the bottom of the contain==r.
The tank would be located below the pressurs vessel and would be raised to permit circulation of water under it.
The bottom of the tank would be lined with fire brick. Ta the event the oors melt-ed throusi, the pressure vessel, approvimmealy a one-foot thleh mesa of would be deposited on top of the fire brick. Heat would be transferred by boil water l
on top of the molten D02 and by the passage of some heat through the fire brisk to p
water located below the tank. It was enggested that the bottsus of the tank might have to be slanted to prevent the occurrames of steam blanketing below the tank.
l At the end of the meeting, con Ed revised their commitaset to inspect one-half of the pressure vessei head bolts by ultrasonic and dye penetrant testing at seeh re-fueling. They agreed, instead, to inspect all of the bolts in this sinneer at each refueling.
i Atto (1) Subjects for Oral Discosstoa...
(2) Charr ACRS e e e Mcc/evb
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sonJacrs von ogA. DIsConslos l
AT INDIAN pouff 2 SUSCOWGTTEg METING, = JUNE 23, 1966-i-
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h possible course of a core moltdown accident without core cooling i
systems in operation and poesihte addittomal means of coping with this type of acaident, if necessary.
(Discuss, in particular, the ultimate fate of the molten fuel and any effect it any have on con-l tata==== integrity.)
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2.
Reliability of-the ame 3ency core cooling system.
(Discuss the l-differsatial pressure which any esist across'oore componente during i
primary systen blow down accidents and the ability of the componente i
to withstand the forces favolved. Discuss.the offact that disarrange-ment of core cosponents may have on coolias of the core by the emer-i gency coolant system. also, discuss the amount of pressure vessel movement which could occur without -impairment of the proper function-ing of the core emergency coolant systes.)-
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Reactivity transients.-
4.
Current plane regarding desismias against the-longitudinal rupture of j
i the reactor pressure vessel.
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. The capability for simitaneous operation of the emergency core cool-l ing systems and the containment spray eyetems.
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The isolation valva seal water system regarding:its vulnerability to i
malfumations which might violate contaie====e integrity.
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Ealiability of the service water system during accident conditions.
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8.
The degree of inspection of the centsiammet liner welds following
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fabrication.
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9.
i Requirassats, above those stated La Section III of the As* Boiler _ and Pressure Vessel Code, which will be.in effset-for the Indian Point 2 2-reactor pressure vessel.
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A brief description of all. arrasigements, procedures, and special equip-j asse provided for use la emergencies which might-affect persons locat-i ed off-site. _ Discuss the degree to d ich emergency plans have been
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tested and the results of such tests.
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AIT. I to Indian Point 2 Subcate Minutes of Has held June 23, 1966 lf L
j ACRS j
MCG/eyb
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