Text

Summary of Natural Circulation Alternatives for Long-Term Core Cooling at TMI-2
	ML19220C685
Person / Time
Site:	Crane
Issue date:	04/10/1979
From:	Karrasch B; BABCOCK & WILCOX CO.
To:	;
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ML19220C682	List: ML19220C681; ML19220C685;
References
	ACRS-SM-0079, ACRS-SM-79, NUDOCS 7905140028
	Download: ML19220C685 (33)
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L ff Dabcock &'.'li'cox April 10, 1979 A

SUMMARY

OF NATURAL CIRCULATIO:! ALTER :ATI'/ES FOR LONG-TER I CORE COOLING AT TMI-2 prepared by B. A. Karrasch 9

261 d'Dchnwwr 79051400f8

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TABLE OF CONTE.';TS INTRCD CCTIC'i SUFC!ARY DISCUSSION STEADY STATE RESCLTS TRANSIE':T ANALYSES OTSC TEST PROG?JJ!

ACCEPTN?CE CRITERIA DURING OFE?dTION RECO.'C ENDATION APPENDICES REQUIRED I:iSTRU::E:.- ATION REFERENCES b

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INTRODUCTION:

The TMI-2 long-tera cooling mode proposed by 35U utill:cs RCS natural circ-ulation core cooling with the steam generator secondary side in a solid flowing water condition for heat removal.

This ultimate decay heat recoval code is a key feature of the following proposed sequence of events to achieve a stable, cold safe shutdown condition at TMI-2.

Phase I:

Reduce RCS temperature to approximately 230 F by steaming the A OTSG through the turbine bypass system with one RC pump running and RC pressure controlled to a value greater than the pu=p liPSH,using the pressurizer in a normal code.

Phase II:

With the A OTSG stea=ing, the 3 OTSG (and closed secondary system yet to be installed) will be slowly filled solid with water and a transition will be made to remove RCS decay heat with the 3 OTSG solid. The A OTSG will be isolated and the RCS temperature will be reduced to approxicately 100 F with the 3 OTSC.

Reactor coolant flow and pressure conditions will remain the same as Phase I.

Phase III:

The A OTSG (and closed secondary system to be installed) will be filled solid with water and a transition made to remove RCS decay heat with both A and 3 steam generators flowing solid with 100 F facdwater.

Reactor coolant flow and pressure conditions will remain the same as Phase 1.

. nn m'7 77 20J

i Alternate to Phases 11 and III: The B OTSG will remain isolated and the A OTSG will continue to remove RCS decay heat during the transition from a steaming secondary at 230 F to a solid water secondary at 100 F.

This scheme would not utiliae the B OTSC; however, the transitian operation is more difficult with respect to steam line water hammer and maintenance of a stable RCS temperature and pressure.

With the reactor coolant system at approxicately 100 F Phase IV:

using normal RC pressure control and secondary side heat removal with a solid system (between 3000 and 5000 gpa at 100 F), the reactor ecolant pump will be tripped and natural circulation core cooling will commence.

Acceptance criteria for core cooling will be established and long-term cooling of the core will be caintained with natural circulation.

Phase V:

With natural circulation for core cooling and a solid secondary system for OTSC heat re= oval, RC pressure can be reduced to a minimum value required to caintain the RCS in a sub-cooled condition.

Cur plan is to fill the primary system solid, including the pressurizer, and maintain pressure control with a enkcup pump designed for such an application. To maintain a stable sub-cooled margin, we envision a long-term RCS pressure between 20 and 50 psia.

NPGD and During the past week, analysis and testing has been underway at the Alliance Research Center to define and understand the various alternatives available for core heat removal with natural circulation.

The analyses were 99 264 J1:ctrcJ tcward obtaining data to define an optinun long-term cooling mode.

rhe impretant censiderations include:

se, ital 1.

Core natural circulation coolic.g. for various re n' t,P configurations,

with one or two OTSC's 'n service.

2.

OTSG natural circulation cooling performance with various secondary side water flowrates to the unir. through the main or auxiliary feedwater nozzles.

E.pected transient performance during the transitian f rom forced to natural 3.

x circulation including specific acceptance criteria for the operator to determine if adequate core cooling is achieved.

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The pref erred mode for natural circula tion core cooling is to use both OTSG's solid on the sceendary side, with a flavrate of 3000 gpm entering the OTSG through the main feedwater no::les and exiting the unit through the stcan outlet nos:1cs.

This code vill provide a =aximum core flowrate

(> 800,000 lb /hr), a ninimun core.1T (< 30 F), and a minimca reactor coolant average ti.mperature (< 120 F for a 100 F OTSG feedwater teaperature).

The secondary sit'e OTSG cooling is a stable, forced convection mode, which transfers alt the primary system energy above a tube elevation of 30 feet, thereby proiidir' high column of cold water for enhancing the primary side natural circulation. This rode provides a driving head similar to that obtained with the OTSG stea=ing with a secondary side level at 30 feet.

The solid secondary side code of operation has a distinct advantage over a steaming mode in that a much lower reactor coolant system temperature can be achieved. The solid configuration will result in an RCS te=perature very close to the OTSC facdwater temperature (approxinately 100 F); the stenmiig mode of operation can only obtain RCS conditions equivalent to the saturation temperature at the lowest achievable senam pressure (approximately 230 F).

In addition, the use of the main nozzles for CTSG feedwater addition has been shoun to yield a predictable and unift.m primary system heat re= oval suitabic for natural circulation.

The use of the auxiliary no::les for OTSG feedwater addition, with w: ter exiting the main no: les, should also remove the primary heat a t an elevated point in the unit.

However, the flow distribution and uniformity of cooling is uncertain and the feedwater flowrates are limited by syste= design and GTSG tube crossflow velocity concerns.

In addition, major secondary plant modification would be required to implement reverse flow through the OTSC main feedwater no:21cs.

Testing i.

o perforacd on the 19-tube stcan generator at the Alliance Research Center confirms that feedwater addition through the main nozzles wi th water exiting the stcan outlet nozzles is the preferabic code for natural circulation.

The advantage of using both secas generators instead of only one is an increase in the core natural circulation core fluwrate of 10 to 20 percent o

and a decrease in the core outlet tecperature of about 5 F.

Extensive analysis has been independently perfor=cd at !PCD to confirm that the difference between using one or two OTSC's is not significant frca a natural circulation standpoint; one OTSG in service will provide adequate core cooling.

We believe, however, that the uncertainty in local core conditions and cooling requirements, the need for heat exchanger redundancy in the long-term cooling mode, and the case o' transition and operation with a solid water secondary for two OTSC's versus one, makes operation with two lcops a superior mode.

The effect of a greater core resistance on the natural circulation cooling capability has been evaluated and deemed acceptable.

A core resistance of 60 times the normal value has been assumed in the calculations, and the reported results are acceptable for either one or two steam generators in operation. The difference between a normal core resistance and a core resistance 60 times normal (indicating a significant blockage) is a factor of two in core flow and AT.

This favorabic :=sult is due to the effsetting effects of systen resistance, flouratg and te=perature dif f erence to sustain a s table natural circulation condition.

Expccted transient pnrformance during the transition f rom forced pri=ary system flow to natural circulation is predictable and stable.

Fron an initial condition with the RC pump running and pri=ary and secondary temperature approximately 100 F, a stable natural circulation condition will be achieved within a half hour following the pump trip.

The cold leg temperature will decrease slightly (due to the pump power loss) and remain stable at about 99 2b 100 F.

The core outlet temperature will increase by about 20-30 F within 10 minutes and be observed on the hot leg RTD in Icss than 20 minutes.

During the first hour after the pump trip, the reactor vessel heatup with no primarv system ficw would only be 100 F.

Acceptance criteria during the first hour of natural circulation will be provided to the operator and primary systen pressure will be maintained to assure that the reactor core outlet temperature remains 100 F sub-coole.1 at all times.

When the operator observes the increase in hot leg tc=perature indication, a stable natural circulation condition will be confirmed.

e e 99 2bU 9

DISCUSSIO::

A.

Steady State Analysis The results of the steady stato natural circulation analyses performed to date are presented in Table 1.

Four different reactor configurations were evaluated to determine the sensitivity of various conditions and assumptions on the natural circulation core flowrate and core te=per-ature drop. The configurations studied include:

1.

Two loop operation with both steas generators steaming at 230 F (20 psia) at a 30-foot secondary side level (95% on operate range).

This configuratica is similar to that which has been tested on the Oconee L* nits and forms the basis tc. a considerable amount of analysis at NPCD. These cases have been used to provide a bench-mark on the OTSG beat transfer characteristics for development of a driving hea. and for confirming RCS loop a? characteristics.

Figure 1 illustrates the sensitivity of the core natural circulation flowrate with loop al (the driving head gain) and loop pressure drop (the driving head less). The flowrate will seek a stable natural circulatien condition based upon the loop aP and the resultant core aT.

The key to obtaining a maximum flowrate is to hot eold) at remove the primary system heat (i.e., change t

' T as high an elevatica as possible in the steam generator.

Our testing and analysis confirms that the primary heat is all trans-ferred above the liquid / steam interfaca (i.e., the icvel) on the secondary side of the OTSG. The calculational results presented conservatively assu=c that the primary system tc=perature change occurs as a s:cp change at the height of the OTSG operate range icvel. h9

.b

The ef fect of increased core SP ans also been evaluated to determine the core flowrate and temperatur drop sensitivity.

The following types of analyscs have been pertsraed at NFCD to conclude that the THI-2 core resistance in its current configuration could be as high as 60 times the nominal value, indicating a high degree of core blockage:

a.

Core aP calculations based upon a postulated core configuration.

b.

A comparison of RCS flow meter readings, with one pump running, before and af ter the TMI-2 incident.

c.

A conservative estimate of core flourate and pressura drop in the current TMI-2 core configuration using the actual decay heat level and the difference between the cold leg teaperature and the core outlet te=perature as determined by the core outlet thermocouples [i.e., core flow = 9 decay and AP = (core flow) ].

AT These analyses have provided a range of core AP values which have been included in the evaluations described in Table 1.

The effect of increased core AP on the natural circulation flowrate is illustrated on Figure 2.

The analyses have shown that the natural circulation flowrates are adequate with the core in its current configuration.

2.

Single loop operation with OTSG A secaming at 230 F (20 psia) at a 30-fcot secondary side level-0TSG B isolated. This configuration has been evaluated to provide a comparison of two loop versus singic loop operation.

The single loop calculations confirm that the net core flow will be 10 to 20% less in tMs configuration than with both steam generators in service.

The resultant core AT will 99 270 W

increar.c about 5 F (depending upon the decay heat icvel) and is still acceptable for core cooling. These analyset were perfor:acd to confic.i an accep tabic condition should an emergency situation require an imacc*iate transit ion to natural circulatica prior to the planned sequence to a solid steam generator secondary side.

3.

Single Icop operation with OTSG A in a solid secordary side mode with water addition through the cain f eedwater nocales.

Steam generator heat transfer analyses and testing have confirmed that a 3000 gps feedwater flourate to the main feedwater nocales will provide a primary to secondary heat transfer characteristic similar to that achieved with the OTSG s: caring with a 30-foot water level. The majority of the heat rc= oval occurs above the 30-foot level in the GTSG with a 3000 gpa f1 curate.

If the flowrate is increased to 5000 gpm, the driving head for natural circulation is further improved to about 35 feet.

The calculational results confirm that adequate na tural circulation flow and core aT can be obtained with a single steam generator operating in a solid condition.

Additional analyses were performed in this configuration to determine the ef fect of reduced core decay heat levels.

These cases were run at 2 and 3 El to provide a comparison with values of core flow and nT at 5 :.W.

As can be seen from Table 1, natural circulation core flowrate and core IT are both reducce for lower decay heac values, and core cooling remains acceptable.

4.

Two loop operation with both steam generators in a solid secondary mode with f eedwater addition through main noz:les at 3000 gpm.

This is the pref erred code for long-term cooling at TMI-2 and the results are very similar to the two OTSC's s teaming case.

Again, the solid flewing water secondary sys tem at 3000 gpm induces g 271 d ',

a high heat transfer interface in the OTSC's and acceptable core na tural circula tion coolin;; is achieved.

The steady state natural circulation analysis has reculted in the following conclusions with regard to long-term cooling at TMI-2:

1.

Adequate core cooling uith natural circulation can be achicved with either one or tuo steam generators in service.

2.

An increased core resistance due to blockage decreases the natural circulation f1 curate and increases the core iT.

However, it has been shown that acceptable core flow and AT can be maintained uith significant increases in the core resistance due to blockage.

3.

A'-se te natural circulation flowrate can be achieved with the sr.can geacrac,r(s) in a steaming or solid code if the eff ective heat transfer height is =aintained at 30 feet or greater using a high level for stearing (3C feet) or a high flourate for solid (3000 gpa).

4.

Adequate natural circulation coolina can bc =aintained at reduced core decay heat levels.

. 7 99 2'/c

c B.

Transient Analvsis The results of the forced flow transition to natural circulation cooling are presented in Figures 3 and 4 Transient analyses were performed at core decay - :: levels o f 2 and 3 =cgawa tts to better understand the time dependent responses of core flowrate and to per-ature change following the loss of forced cooling.

The bases for the analyses are as follows:

4 Core Power - 2 and 3 Megawatts Reactor Coolant Pump Trip at Time 0 OTSG A Solid with 100 F Feedwater into the Main Nozzles at 3000 gpa OTSG 3 Isolated Core Resistance Factor - 60 The transient responses of core flow and temperature confirm that a s=ooth transition to natural circulatica ic achievabic.

Following the loss of forced flew, the reactor vessel heatup slowly induces a temperature gradient between the reactor vessel and upper OTSG and n atural circulation occurs with no operator action. The core flowrate reaches a minitus about 1 minute into the transient and reaches a stable condition between 10 and 20 minutes. The core outlet temperature begins to increase and reaches a maximum vclue 4 to 5 minutes into the transient and a stable condition at about 10 minutes.

There is about a four minute time delay in the response of the hot leg temper-ature measurencat due to the approximately 10C0 f eet in the reactor vessel upper plenum and hot leg piping. The cold leg tc perature drops slowly to closely match the OTSC f eedwater temperature due to the loss of the approximately 5 =cgawatts of pumping power.

The transient natural circulation analyscs have resulted in the following conclusions with regard to long-tera cooling at T!!I-2 : _

99 2,/a

S 1.

A smooth tran:;ition f rom forced flow cooling to natural circulation can be achieved by tripping the reactor coolant pump and observing core outlet tempe ra tu re.

There is no reason to slowly reduce the RC pump speed for a more gradual transition to natural circulation.

2.

The reactor coolant system flow and temperature will reach an equilibriua value within the first 1/2 hour of the transient; the response of the hot leg temperature =easurement occurs within 5 minutes after the core outlet temperature changes.

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C.

OTSC Tcat P re r.m The ability to achieve and maintain a stable natural circulation flowrate is dependent upon the clevation dif ference between the heated core outlet temperature and the transition to cold Icg temperature in the OTSC tubes.

This transition print in the OTSG is in turn dependent upon the heat transfer characteris tics of the unit.

If the primary to secondary h:at transfer can be obtained at a high clevation in the OTSO, the driving head frca the density difference will be improved and natural circulation flowrate will increase.

The primary to secondary OTSC heat transfer =cchanism.uhile in a steaming mode is boiling at or abent the level of the secondary side water.

Extensive analysis and testing of the OTSC in the " pot boiling" mode has confirmed that the primary system temperature transition occurs above the level of the " boiling pot."

All calculations performed with the OTSC in a steaming moda :onservatively assume that the primary system cold leg temperature is available for driving the natural circulation flow at the 30 foot level.

Use of the steam generator as a water to water counter flow heat exchanger is a more dcairable condition to obtain during a long-ters decay heat cooling mode.

The primary temperatures caa be maintained much nearer the temperature of the incoming feedwater to the OTSO.

In order to determine OTSC characteristics in a solid mode, a test program was conducted at the Alliance Research Center on a 19-tube, full-length, steam acnerator.

A natural circulation flowrate of 700,000 lb/hr was simulated on the primary side and forced secondary side cooling was injected inco the main feedwater noncles; flow exited the unit through the steam outlet no::les.

Feedwater flevrates were varied from a scaled value of 100 gpm up to 5000 gpm. I 99 2'/a

The rer, ult: of this test program are presented on Figure 5, a plot of feedwater flowt ate versus the OTSC heat transfer clevation.

Ilca t transf er cicvation is defined as that icvel above which all primary system heat is transferred to the secondary syctcu fluid.

That is, the height at which one can assume the primary cold leg tcmperature is available for driving natural circulation. The figure sEcus that a heat tremfer cicvation of 30 f eet can be obtained if the fecdwater flowrate is at 3000 cpm or higher.

A 30 foot elevation head in the OTSG primary is adequate to achieve natural circulation as demonstrated by the calculations in the previous section.

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9

D.

Acccotance Criteria Dorin" opera t ion The succccc or f ailure of natural circulation as a core cooling mode depends upon the value of the core I.T that can be maintained.

The key obj ective during plant operation in this mode in to maintain a primary cold leg temperature as low as possible and observe the resultant het leg tc=perature.

lhe acceptance criteria for success of the natural circulation code 1: to maintain the het leg temperature below the sat-uration tc perature which would cause bulk boiling.

Figure 6 illustrates the propcsed ?OD criteria for natural circulation:

to naintain a 100 F sub-cooled cargin to bulk boiling using the plant instrumentation in its current degraded state.

The large errors which have been imposed on the pressure and temperature instrucentation make it imperative to keep the RC pressure as high as possible at the time of pump trip. This will allow a large hot leg temperature increase to occur before boiling and assure a reasonabic ti=c period to achieve a stable natural circulation.

If the RC pressure is raintained at 500 psia, the hot leg temperature can reach 340 F (from its initial condition of 110 F) before action must be taken.

An analysis of the reacto vessel was perforced to determine the potential heatup rate with zero flow into the vessel. This analysis provides a bounding bulk fluid heatup rate to indicate the amount of time available to the operator to take action before a boiling condition could occur.

The high probability of achieving a stable natural circulation condition irdicates th.it such a reactor vessel battup could never occur.

The analysis is provided to show that the operator has at least one hour to confirn natural circulation before any action must be taken. 9 hh

/I

REC 0r F:Q ATIO:. :

!!atural circulation has been shova to bc un acceptable ceans of heat removal for long-term coolir; at TMI-2 with the core in its current config-uration. Use of cither one or two stcan generators is fensible if the proper secondary side heat transfer characterictics are established and maintained to recove the primary energy near the top of the OTSC.

In addition, the expceted transition process from forced cooling to natural circulation will previde a continuous and stable core cooling condition which can be =onitored and controlled by the plant operator.

B i'J, therefore recotmends that a planned transition to natural circulation core cooling be implemented at TMI-2 as soon as the degassing process is completed.

Both stesa generators should be utill:cd in a solid flowing water condition uith approximately 100 F f eedwater at 3000 to 5000 gpn entering through the main feedwater no :les.

The sequence of events for this transition, as described in the Introduction of this report should be as follows:

1..

Reduce RCS temperature to 230 F with a OTSG steaming.

2.

Slowly fill OTSG B solid with water and begin rc=oving primary systen energy with the B OTSG by gradually increasing feedwater flow until a stable condition is reached at 200-230 F.

When a stable condition has been established, isolate OTSG A.

3.

Reduce the RCS temperature to approximately 100 F by increasing the feedwater flowrate to OTSC 3.

Fill OTSG A colid uith water and prepare for operation. O 99 q/O L

4.

Slowly begin feeding OTSC A with 100 F feedwater and establish a stable condition with' both steam generators removing decay heat with a 3C00 gpm feedwater flowrata.

5.

Establish RC3 natural circulation as follows:

a.

Throttle feedwater flow to both steam generators to establish approximately 25 E aT between fecdwater tc perature and OTSG secondary cutlet tc=perature.

b.

When a stable condition has been establsihed, trip the running reactor coolant pump and increase feedwater flowrate to both OTSC's to 5000 gpm within 3 minutes. Biaintain at 5000 gpa.

c.

Ibintain RC pressure at the initial condition value and observe both A and B hot leg temperatures.

d.

Compare the hot leg temperatures to the acceptance criteria on Figure 6.

If the temperature exceeds the limiting value, start a reactor coolant punp.

c.

When stable natural circulation ccaditions have been achieved, reduce RC pressure to the proposed long-term cooling value between 20 and 50 psia.

The above sequence of events will establish a stable and safe natural circulation condition for long-term cooling at T1!1-2.

All starting or stopping of reactor coolant pumps should be avoided until the pump is tripped to induce natural circulati n.

In addition, 36N recomrends that the sequence of cvents be imple: cated in a planned and controiled manner, i.e.,

we should not wait for a complete failure of all four RCP's before cctablishing natural circulation. We should, however, have an alternate decay heat removal system installed and cady for operation prior to the transition to natural circulation.

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APPENDIX 3 LIST OF REFEPI:CES S

This Appendix will be included in the final report and include a ccaplete list of all calculations and related test data and backup caterial for the information contained in this report.

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SUM!ARY OF NATURAL CIRCUI.ATION ALTERt:ATLVES Reactor Configuration Core Flow Core AT Tcold hot Two OTSG's Steaming at 230 F With 0.8x10 - 1.2x10' lb/hr 15-25 F 230"F 245-255 F a 30' Secondary Level 60 Tines Normal Core Resistance 0.8 - 1.2x10 lb/hr 15-25 F 230 F 245-255 F 10 Times 1:ormal Core Resistance 1.1 - 1.6x10 lb/hr 11-19 F 230 F 241-23.9 F 6

Normal Core Resistance 1.5 - 2.3x10 lb/hr 8-13 F 230 F 238-243 F One OTSG Steaming at 230 F With 6

o o

o a 30' Secondary Level 0.7 - 1.1x10 lb/hr 20-30 F 230 F 250-260 F (60 Tires Norr.al Core Resistance)

T'eo OTSG's Solid With 100 F Feedwater at 3000 npm 0.8 - 1.2x10 lb/hr 15-25 F 105 F 120-125 F (60 Tires Nornal Core Resiscance) 6 tr.e OTSG Solid With 100 F 0.7 - 1.1x10 lb/hr 20-30 F 105 F 125-135 F Feedwater at 3000 gpm (69 Times Nornal Core Resistance) 6 3 FN Decay IIcat 0.7 - 1.1x10 lb/hr 10-20 F 103 F 113-123 F 0

2 L' Decay lleat 0.6 - 1.0x10 lb/hr 8-15 F 101 F 109-116 F w

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N. ra afrun 4/6/79 Tril CORE A55E59 ENT 0 EJECTIVE:

Setermine the esents within the TMI primary. sjstem insofar as tcey m')ht ir.cice te cere ccafigura tion.

"iThJ2:

C;re ce sity ch>.r;?s were inferred ' c.n reduced radiation transmissica to i.mter ediate and source rin;e detectors.

F.2actcr cc:lart s,st'? pres sure, -eac ter inlet and c.tlet t2 rera t.re s, ar.1 sica c ?nc-2 t;r pres s_re 2,d levels we-e obtained f em reccrder p! ts.

Operator acticns were cetaired fr:m computer events ronitor.

System behavior was theorized by a prccess of hypothesis and test for data c ;n s i.- t e r.cy.

'!r.en pess10le, sirple calculations were perferced rela tive to hca t up, boilef f,

and volume changes to cor. firm reascnableness of process uncers tanding.

Time infarred fece graphs and/or recorder charts.

Esti. mated +3 r.inute a:c racy for event specification.

RESULTS:

The sequence of inferred events one hour af ter turbine trip is attacnad.

This sequence states as fact many items that are ccnclusions and hypotheses that seem to fit the available data, and it should be used accordingly.

1lhere facts are unable to be tested, a (?) is shcwn.

Also atteched is a diagram sunrariz.ng ccre density behavior in the ;eriod frca 70 to 220 minutes as deduced frca ion cha.?.ber traces.

In these traces, cf.anges in transmissivity are interpretted as fluid /ccre density changes.

Changes in slope are considered important as indicatica of phase changes or ele.atica change.

Also included in the diagran are plant and peccess informatica censidered relevant and calculational notes.

CONCLUSIC_'5:

1.

Sisnificant core damace occurs in the one hour fericd from 6 a.n. to 7 a.m. -

lon chamber signals suggest significant voiding at this tire.

Significant superheat occurs.

Relatively little inflew to Icwer plenum is noted during early porti3n of period when significant heating of voided core occurs.

Spike in son cr smters at 146 etoutes suggests struc ture change.

Radia tion noted in c on;a inment. More tr.an enough time and hea t cr ists to cause c.e t al reac tions i f ccoling res tric ted.

99 k _c.,L 7. 2

Ty CORE ASSESSMENT Ft;e 2

0.'Cl t'510'is ( r,n t. ) :

?.

Grcater dar. ace on E-leo side of core then A-lec.

Significant ficw noted cn A-lee throuchout period t>etween 6 a.m. a c 7 a.m.

as sigr.aled by preferential heatup of outlet RTD and more rapid te ;erature decrease on A-leg irlet RTD during latter part ef pcried.

i.

Cc re a ureater th.n cc bined aP of ou o and stean cenerator.

Wher.e cr pres.ure is redaced folic.ing 7 a.m., cold h;l ar.d 3eal sat.:r fic-frca pu p and pump discharge tc.sird outlet ter;crature detector in the pu p suc!icn.

This conclusico ccald b2 errci.ecas if the A-3 r s-atc r ca tie t t3p were cooler thar. ti.x t arc c f 10..dr pier.um and HF I wa te r.

E id i.i O921 iy,

magni tude c f ter r arature surges indicate A-le) in highly v aporous die te.

7 Ep*strse flo

occurs in A-leg starting about 2 p.m. and lasting until E3 p. m.

During this period, the plant is at low (500 psi) pressure and bicwing dean with flood tanks providing cold water to icwer plenum.

t.

Addi ticnal darage curir.c ceriod following 7 a.m. cessible, but not__cy_etain.

Outlet tenperatures at superbeat or saturation and cold water f rca I and/or core ficed tar.ks in inlet plenum suggests core ficw between 7 a.m.

and 2 p.n.

Icn detectors suggest scce density change, but nothing approach-ing the response noted at 6 a.m.

From 7 a.m..(204. minutes) en, HPI, core ficod, A-venerator and relief valve suggest ample heat remeval capability.

1 On the other hand, high4 P and reverse ficw suggests Icw flow given lack of pump.

5.

F.etal/Wa ter Reacticn Rate might be bounded by Eydraulic Data from 130 to 176 minutes, the pricary sys*.em prcsstre rises folicaing valve This risc night well includ co.. -ibutica frcm H associa tcd wi th closure.

retal/wa ter reac ticn.

Duringthispressurerise,theinc$redensity decreased by a factor of 1.5 to 2.5.

Wi th availatle pressures ar.d tciperatures,

it should be possible to estimate H fraction in steam and thus, reac tion f rac tion.

it might also be possibib to bcund the encrgy added in additica to decay heat which caustd pressure rise, thus scoping exothtraic reacticn.

.: (y.

OS 9 9 L 7.;

4/b//9 IM! SLO,E!. E Cr EVENTS Of!ER THE F 1 RST E7J" F O'. I OW ;NS TU R3 I N E T RIP 73 min.

- E-lcop fle.< steps, E-Gene a ter pressure begins to dtcrease, pri:aary f ressure cuasi-stable a t 110C psi, with relief valve '.i! 2m a.d c::c. c f l ov, I e t r.t a - s a t ur e t i or., T inlet subcc.eled, b?I of f (?), A-Gene a tcr les ci slowly dre;;ing, B -Gt ocra t:r level increasing.

"0 min.

- Ir.crea:.e in 3-Gene ater level ec:urs.

9 0 - l J C -" n.

- Cold fluid en 5-Gr r. era tor c auses ce:r c s s ar i;a tie-of :-irary syste' with der.si ty char;es ir, c;re regica.

.3 nsity de:rcesis nearly li -arly for S rinutes as ;rcsssu e f e:rcises.

Magni tude i s 1/3 of subse:ues t der:s i ty ct.c: ge.

Te :eratu.e 3t saturaticn i n Loth i nl e t a nd cu t'. e".

i eg s.

E-C2' 2ra tar pre < 2t re begins te 'all co :nc ident wi th 'ia*.hing, suggc : ting retiuced ?.ea remeval due to reduced 'luid density.

100 r:.tp.

- A-loop flew steps.

A cuench cccurs from fall tark of cold water from heaJ or B-lcop or ;ossibly from co : iarrel c5c:k valse leai:;e, al:5: ugh source is uncertain.

C ench irferred from nearly Ites s.ise density increase.

S-Ser.c ra.:.r ;r 2 s sure drop s c;s.

A-Ger.eratcr levei begi r.s to r se ar.d pres sure drop contin.es, sucgesting centinued icw deasity in cr.rary side cf A-Genera tcr.

100 - 116 nin. - Pressere continues decrease, water new in ccre rehea ts.

Rou:h calculation estimates 12 minutes re;uired ta reheat two tires ccre volur.e uo to sa turation (f actor of 2 is alle.cance for inflos).

Inflow indicated by inlet temperature decrease.

Core cutlet at or near saturation.

(

116 min.

- Voids begin to form in ccre.

Vcids (rather than water density change) carked by slope charge in scurce range detecter.

Outlet begins to shcw superheat.

Pressure centinues to fall.

A-Generator level stabilizes, indicating some helt rejection.

Rate of pressure decrease in A-Generator decreases.

116 - 124 min. - Upper half of ccre voids, superhrated steca in cutict.

Hot spot probably steam cooled or voided at 121 minutes.

Signifi-cant steam ficw in A-leg and S-leg cera inlet at 124 r.inutes.

Significar.t steam flew in A-Icg cetlet, less in B-leg.

124 - 132 min. - P.est of core voids a t 123 to 132 rir.utes.

It 132 m'netes, gate valve on pressurizer relief valve closed by opcrctor.

Pressure bcgins to rise.

Inlet terrera ture indicates little,17 ary, flat frca /.-Ir.g to inlet, but soTo lea feco S-leg to inlet.

A-1cg superheit ra te of a t leas t IC[f/ min.

.t-132 - 150 min. - Core apperrs voided with sort dens ity reduc tic, as pressure rises. Cotic t plenu, cor.tains super ncited s team.

Scca flew from B-leg to le..ar plenar, little or na fle# f ron A-leg.

At 146 minutre., the neu tron dele:to-s indicate one or rcre changes in core structure possible.

e/9

/r/7V o

El'SFCuihCL CT EVENis AFItR THE Tage 2 I P. T HOL'E F OL L u I h', T"T 3 :N: TRIP 50 ein.

- Indication of steam cn bcth gene ators since li ttle or no heat reseval.

Lcwer flew rate in S-leg finally raises cutlet terterature detecter te sucecheat indication 30 rinutes later t h a r. A - I c ;.

Can ta i r.me n t raciatier s;ike reparted, Adiabatic calcula tion (cnarac tcris tic cf d y cut) ><culd ;.crmi t metal /

sater reaction in 6 ein. at het spct

-d iC-! 5 r i nu tes f er fuel red of average heat.

Sir.ple calcu::::en woule perr.it ccre voli. e and dc.ncc.mcr volume at core height tc be toiled out at 152 minutes.

50 - l'6 c.in. - Fressure continoes to ircrease with e>.re ential ac;ea c".:e from 600 psi a t 130 minutes to 2100 at 17 6 r.. i r. t t :.

Da tl e t tarpera turc s in bath 1(;s indica te superhea tca s t;i.

Ir:ce sec ficw at inlet indica ed by dec sasing teraccatures, with crea ter ficw f ecm the A-leg.

Lic:le or no cooling incicated at eitaer genera *cr.

ip;arent ex;occntill rise sucgests aat: atalytic H, generation.

Ccre densi ty indica tcd by icr cFc-bers decreased lbss than expected from change in pressure, suggesting H 2

presence.

76 - 180 min. - Qucr.ch of core and colla;se of voids indica ted.

Pressare discharge fecm relief valve.

Lcwer pienum water forccc into cold legs, with g-eater ficw to A than B ( A ccatlins less water) and scre ceoling by the A-Genera tor.

Main steam isclation vah e and turbine bypass valve cn B-Generator closed a t 120 minutes.

Outlet teapcrature indicates superheat.

Quench suspected to be frcm water or steam frcm A-leg.

l.

30 - 204 min.

. Quench water / steam reheats and, in 7 to 11 r.inutes, voids again begin forming in the core, although not as extensively I

as before (about one-half).

Little ficw (pessible tackficw)

I in th0 A-leg as indicated by hot water at inlet ter.perature detector. More flo.I cn B-leg and tcward inlet plenum.

Outlet te.rpera tures at sa turation.

Relief bicck valve oper.ed at 204 min.

34 450 min. - HPI turned en.

Pressure drops to 1200 and is maintained between 1200 and 1700 psi until 5.5 hcurs, when tre relief b'ock valve was closed.

Whcn valve closed, pressure rose to 2100 psi and remained there through pericd.

Thrcughout peried, cutlet temperatures were near saturation and inlet teeperature generally decreased as the result of HPI flcu.

Periodic surges of Icwer plenum water move into cold legs wher.ever system pressure decreases, r. ore so in the A-leg than B-leg, indicating greater vapor volume in A-1(1 Ccre voids or significant der.sity decrease noted at 223 minutes, associated with pressure surge and backfica from Icaer plene-tc inlet legs., Sicu ccrrection or collapse noted in this cese, 99 zy

TMI Sf0UENCE Cr EtEf.TS ATTER THE Page 3 F R5T HOUR FOLLC'JItC TUP.E.:.NE TRIP 7.5 - 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months - kelief bicck valvc open'd at 7.5 homrs and pressure drecs to 500 psi in 1.5 hcurs.

At S.5 hears, the core ficod ac tuatn.

Outlet tc ;eratu es show super 5e2:ed steam.

c m ra tors cease to re ave heat fc r 2 heurs s tarting et 2.0 bc.ers.

Earing this 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period, inle t ie.pera ture decreases and outlet ter;e-a ture is at st;cchea t ccr,di tion.

Smcil d,ar;e in co. c density occurs a t 9.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

10 - 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months - Reverse flew in A-leg becins at 10.5 hears.

E -1 c p c.ay ba. e tried to reverse a t 14 r oars.

L :ca plai ed lesel change in 3-Sercrator at 11.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months tha t ap. ears to be f rer: charginc, alt 5cugh logs and data sheets ccr.tain no reference to such action.

Frimary side contractica palls lower ple %. water into E-le.

Even t'.ough MTI ficw is increased at 12.5 heur.s, re. rse fice c : tinues in A leg.

Relief valve aga'n ciesec et 13.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and pressure ri:e begins.

Increasing pressure pushes HPI wa ter to.eard A-leg gen 2ra tcr anc pushes icwer plenum water into inlet of 8-leg.

Pressure ir.cre2se 3rd its effects extend from 13.5 to 14.5 hcurs.

During the irisode, cooling occurs in A-Generator.

Some densi ty cc.;nge noted in core as pressure increated, then a general dcrsity decrease cccurs.

Suspect hot fluid from upper plenum pushed thrcugh core.

At 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months A-loop Primary pump turt.ed on and significant events end.

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OfflCir.L USE 0[:LY 4osos ATTACIFEl.'T I STATUS OF THE CCRE Dr. Ivan Catton, ACRS Censultant The core assessments by Kaufman of EG&G, by D. N. Roy of B&W and by the NRC Staff differ on specific details of core cooling during the accident. Rey are, however, in reasonable agreement on the extent of damage. Se core was partially or totally uncovered on at least two possibly three cecasions for periods ranging frca several minutes to several hours.

The major damage occurred during the initial uncovery period which occurred at about 126 mintues into the incident and lasted for ap-proximately 50 minutes. D e core quenched at 176 minutes when the pressurizer block valve was closed and the pressurizer c.ptied into the core.

The amount of zirconium oxidation was estimated by approximating the amount of hydrogen in the bubble, in solution in the RCS, in the con--

tainment and used in the containment explosion. H e estimate leads to 30 to 45% of the circonium being oxidized. %e rapid quench fol-lowing the initial dryout is believed to have caused a great deal of cracking of the oxidized clad.

OFFIClf L USE ONLY

, Calculations based on xenon released indicate that sizeable fractions of the core saw temperatures as high as 1600 C for several hours.

The high temperature fuel and steam are believed to lead to many fines in the system.

The bottcm 1/4 to 1/3 of the core is thcught to be reasonably intact.

The top 2/3 - 3/4 of the core out to 1/4 radius is believed to be in a heterogenous state. Were are many hypothesis as to what this part of the core looks like. Some p stulate intact guide tubes and grid spacers covered with cracked pellets and zirconium oxide. We question of natural circulation cooling of this region has arisen.

Temperature measurements indicate that the flow through the central region is less than through peripheral regions.

Natural circulation cooling of the core should p se no problem pro-viding the water is maintained in a subcooled state. Were will nost likely be local boiling in the central regions.

Removal of the core frcm the vessel will be a difficult task.

Some special equi;nent will be needed to vacuum the debris from the highly damaged areas.

99 301

TilREl: M l l.E I S L'.N D CORE INTECRITY ASSESSMI:NT INDUSTRY.\\DVISCRY GF. CUP Agreed to ar.d understapd by W.

Bixby D. Ditmore R. Muench C. Solbrig e

99 302

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_9,__m.._n p Y_ _C.O..N.C.1.'.I S_I O.N S The bot t o:.i 1/4 - 1/3 ot t he core -i 3 likely to hate ret.t:ned its basic s

rod' d i n..-n 3 i on s sp. icing., etc., but lirobably has physical i nt eg r i t y ',

e.g.,

censiderable clad 3urt'se oxidation.

The top 2/3 - 5,'l ot the core cola.ac i s o.- t likely to be in a het eregeneou3

at e wi th a peripheral anna t u ; of t'elat icely int act nut highly oxidi:ed, balleened and pe rfora t ed c laa t'ut t rod with a cent ral core re ion which has experienced

.c ry su' : t :.nt ia l da :ag., e.g., r ui'st an t ial e lad bal loon ng, elad pe rfo ra t ion a

red bewing, excessive c a,i exidat ion (15 - 45", total core clad oxidation with leen tly ur to !"0M, ed si;ni ficant e lad ar.J ex ide f ragment a t ion.

In spite of a high deg: ce of fuel damage in the upper central core region it is likely tnat the general core structure retains some physical integrity; 2.e., cont ral guide tubes which rapport the grids largely intact, remaining

,rartial segments of distorted and oxidi:cd cladding remaining intact, very possibly cost of the grid cells are still occupied by cladding and the upper ends of the heavily danagcd rods are probably hanging from the top grid spacer.

The highly damaged top central ccre region cost likely has considerabic debris (clad, clad oxide and fuel pellet fragments) pile-up at the grid spacers.

The result ing effective core flow area blockage due to the fuel.. condition based on flow resistanee elaculations, is substantial and may be as large as So to 90'.; with local fuel assecbly blockage i n exc ess o f 90'..

The degree of flow blockage inferred to exist, while substantial, should not preclude adequate care cooling under natural circulation conditions.

!!cwever, local coolant boiling in the most highly blocked areas can be expected.

We believe that the core is stable enough that it will retain coolability during natural ci rculat ion operation.

99 303 e

The ea t i re core an.! c::peci.:11y t he guide t +e-which are st ill intact

.re suspected to have received s nough oxidat iou so t iia r oxygen incursion into the :ircoloy c:u: sed t he rc :ainine clad to be ext remely brit t le.

!!cnc e, the fuel

.ss ei;.b l ies n...y no t retain physical integrity during removal from t he core, if removal i< done in the usual manner.

'G RCDL'C l 107.

The ahave arsessm.' :t of the Three 5 tile Island core integrity is based upon a combinat ion of quant itative and qualitative c') servat ions and assumptions dcrited frc:- bath the deduced accid at performance and subsequent steady-

. tate performance in single loop-one reactor coolant pump operation.

The evidence s largely circumstantial und thus these conclusions rely heavily on deduction rom numerous indirect observations.

CRE COOLING DURING THh ACCIDi!NT At Icast two different scenarios for core cooling durin; the accident have been put forth.

These two scenarios are presented in:

a.

The T3tl Core Assessment by N.

Kaufman of EGGG of 4/6/69, and b.,the Otl Core Condition Assessment by D. N. Roy of %W of 4/3/79.

While these reports differ on the specific details of core cooling during the accident, they are in agreen ent that the core was partially or totally uncovered on at Icast two, an possibly three, occasions for periods of several minutes to several hours.

Both also agree that the first period of significant uncovery occurred at approximately 5:45 A.St. on 5farch 28, 1979 (approximately 100 minutes following the initiating transient). This initial uncovcry apparently lasted for approximately on' hour anc' was terminated when the pressuri:er block valve was closed and pressuii:cr liquid flowed into the core gutnehing the fuel rods.

Beyond this ini t ial uncovery and recovery there were appa rently two subsequent periods where part ial uncovery is beiieved to have occurred, 9

core DA31u:l; him1NG nil: ACC i til NT liceause ot' the N igni fi cant rediletion in heat transfer coeffieient at tt ndent

3 to' core uncertry, rt ha t ion 4 o f *.co-t o-three orders o f r..*gn i t ude in conveetare Ma t trans fer s veira i cot, the :.e l ro.1, und..ubt ed l y ex;n rienced very. i gn i fican t overheating a..d....... g during tl.e ii.itial uncotery period.

The. sequence vf event s during. thi s core dan:n e ;mriod i, 1ikely to have been manifested by init ial elad b il lovning. then in rfarat ion.

C1:.d oxidation rate would have

'. creased sig:ficantly

-hc e la.! t s:yera t u*e increased.

The conclusion t '.4 s i g n i f ie:.n t fuel perforation occurred Juri:'

the init ial uncovery is support ed b;- the steep ir.crt ase in fission ;roduct activity in t he reactor tuilding the period from 3: 43 to 6:45 A.M.

on March 28, 19 -'9.

While the subsequent two periods of uncovery are 1ikely to have resulted in additional fuel damage, the most major change of physical conditien of the fuel is likely to have occurred during the initial period of uncovery when the fuel was initially unfailed..\\ttachment #1 letter P. N. Marrict: to D. C. Ditmore

" Mechanical Candition of Three Mile Island Core," 4/13/~9, presents the results of a rough evaluation of the possibic core damage scenario.

It is estimated that the peak cladding ter.perature increased from less than 10000F to the range of 2000 to 2600 F in the first 10 minutes of the' initial uncovery pericd.

It is estimated that most fuel rods experienced cladding s

temperatures in this range within the next thirty (30) to fifty (50) minutes.

The exception may have been that fuel elad in lower power bundles may have been tclow 2000 F, i.e., peripheral core bundles may be less damaged.

fuel cladding would b. expected to experience perforation in tbc range of 1400 to 0

1500 f for the internal pressurization of the TMI-II fuel rods and the measured reactor system pressure during this period.

Prior to perforation individual fuel rods woult! have everienced considerahic clad ballooning possibly up to 2X the initial diameter. Clad ballooning and perforation is not likely to have been co-planar across the core althcun,h preferent ial ballooning within the hottest core regions axially would he expcet ed. " Ballooning i s typically 99 305 semewhat randem due t o stat ist ical variat ionr in t he claihling.

It wa a-

..l.

suge.ested that co-planner baliuoning..ght have been po.sible consiJering t he ro s t u l a t t J s l r..

rate of initial uacovery, lhn.ever, experienee and a simple calculatien indicate th.it elad overheating wauld not be expected until t!.e fluid level d:.ipi ed sc l1 1.elow t he r.;idplane, t hus grea t ly rcJue ing t he likeli!wod e t 1 cali:td co-plai.arity.

The : ire claddi g i3 believed to F. ave e.sperienec).. g r. i fi. an t oxidation Juring the rise n temp ratarc with the nighest rate 3 e veri.nced above 1500"r clad tenperature, 1! i s postu;at ion is consisten* with calculations of ar.e u n t of evo;ved hydrogen which imply s.sidation of from 15 to W. of the fuel claddi'i see attachmcat

'1, " Assessment cf Cool ed H; Jr ' gen," by J. W. Thi esing, '/12 /79).

It is likely that local eladding oxidatica of essentially 100* of the wall could have occurred at perforation locations during the first uncovery, and possibly at other locations in the subsequent periods of uncovery.

After approximately one (1) hcur during the initial t covery period. The core was recovered with liquid. This occurrence is likely to have quenched the hot clad and clad-exide causing high thermal stresses and resulting in significant fragmentation of the most highly exidi:cd cladding.

The two subsequent uncovery and recovery incidents would be expected to have oxidi:cd additional eladding and resulted in additional fragmentation on q2enching.

This fuel rod and cladding damage sequence is likely to have led to a condition of substantial buiIdup of : ire clad, clad oxide and fuel pellet fragments (debris) at the grid spacers, which would present a very significa..: flow blockage, particularly when combined with the local effcces of rod ballooning and bowing.

The greatest damage would most likely exist in the hotter central (radially) region of the core, and the upper portions of the core which were likely uneovercJ for the longest periods.

g s

/

99 300

Txperience w i t h core heat-up an,1ys i s duri ng a l o :: uf coolant accident

ht, e"re conpanents; i.e., cent rol ro! i. iide t We<, will a gge :. :,,

.t

...v s

'. sve espe ri enced t empera t u res ect:pa rable in magnitude to the peak fuel clad,

ad thus those: irceniun component s will he highly oxidi:ed anJ brit t ic.

15hW \\TIONS SUMD.IflAI N lilE ACC li'!..NT A numhe.r of observat ions of plant performance a..-ing ". ingle loop --

evidence One reactor coolant purp operat ien provides addi t ion..! e ircumstantial

'hich lends same indirect support ta this core Jamage scenario.

.S_ elf Pawcred

'.t ut ron Det ector (SPND Damace)

Nith the except ion of one detector at level 5 (there are ccven levcis of 5P.ND's starting in the lower core region) all detectors above, te secor.J 1evel are not corking. The reasons for their malfun< tion is unknown.

However, since essentially all SP.;D's above the low'r 25', of core volume are not functioning, it is likely that this portion of the core (the top 2/3 - 3/* of the core volume) r2 ached higher temperatures and thus experienced.rore damage than the lower 1/4 - 2/3 of the core volume. This tray be a consequence of lower temperatures during the accident due to shorter periods of uncivery and lower heat s

generation rates.

In-Core Thermaccuoles Fif ty-one (51) of the original fifty-two (52) core exit thermocouples appa ently survived the accident, of these, forty-nine (49) remain intact and are now recording temperatures at the core exit which appear to be accurate. This fact is significant because the thermocouple Icads pass up through the core regien in the center thic.ble of 52 of the fuel assemblics (locations shown in Attachment #3 ) to a detector (junction) approximately 4 - 6" above the core exit.

T:.ese thermocouples t ravel up t he same tube a s the SP.ND's.

The thermocouple The cont inued availa$ ility of these thermoccupies 99 307 wire melts at 2500"F.

sugge t s that the center thimbica and other non-fueled tubes; i.e.,

control rod guide tubes, have remained intact so that the core has ret a ined some st ructural a-

6-

n spit e of an e.s [ cet el hic.h degree of fuel tegr ity la

!.c app r r.;, ions i.e., d i-t o rt ion e l.nl ; erfora t ion and os id.it i sa.vid f ra gment a t ion.

f :.ir,c,

Travel t ag I na re t': N (~1 l i') Insertion

' he insertion of t he single core Tli' was at t empted.

T he probe : stuck i

.t a level of 3.5 feet inte the core from the bot tom.

No further travel was This is addit ional qualitative ceidence of p aibly and the probe re.a.iins. stuck.

to 1/3 of the core volume.

the degree of core damage l'eing less in the bottom 1/4 this tube, while possibly highly distorted, retains enough It indientes that inte3,rity to present fur:hcr movement of the TIP.

Core Flow Resistance / Flow Blockage B6W and PNL have performed ir.dependcat assessments of core flow flow area blockage.

B5W resistance which ! cad to a conclusion of cignificant has estimated the increased core flow resistance could be as high as approximate-ly a factor of 60X. This estimate was apparent!y confirmed in two ways; 1) comparison of reactor coolant system flow meter readings, with one pump operation, before and af ter the accident, and 2) a post-accident single-loop operation core heat balance using the estimated decay heat and measured core coolant temperature change to determine flow. These assessments pr:,vided a range of valve.

core resistances up to a factor of 60X increase over the pre-accident A factor of 60X increase in core flow resistance is appro inatc1; equivalent to an effective core flow area blockage of-90* Jk/a2 ~ 60;

^2 -- t[l/ 60; (k/ah At (1-d 1/00)100 ~ 87*.

PNI. has employed the CC"Pd code to perfom an.,1yses c,f a 26 channel si:aulation of the B 1-II core to reproduce the measured core exit temperature distribution during singic loop operatien.

From their work they estimate 60 - S0*. p,ross core blockage, with local regions with blockage in excess of 9 5 *.. 1be local high blockar,e is implied by t'.e wide va-iat ion in core exit

~

99 303

.~.

caulant t e:ipe ra t u re: Juring single loop c[ station u t t a cle". n ;

'3J.

.. i t h.:

r i h ra l core region. o t he en t ral core region.

.;.ria t ica o f ~ 100 i-from the pt i

s simple car, hent balance using t he avera: e core exit thermoco.ple readi.igs n the p. r:t hery of t he core also le ids to the conclusion of an et':ect ive core flo-2rea bleek.;ge of ~ o 5;'c e i fi ca l l y, the measure core coolant 3t in sore the core peripheral region was - S.3"F (see attaerment '3) on the average on 1/10/70 (this i s t yp i ca l of p e rfo r:.u n e e o f t he pa s t week or 80),t.it!' 8:nc.le leep -

anc purp o;' era;ien.

N1thoat flow blockage the core ecciant at should be ~ l F eore

.:. d i r.g ; e ;oop - ene pu.ap ope rat i on.

':i s :s ccafirred by simpic ea! ulatione Is~

attachuent = r), and by measurement, since the measured cold leg-to-hot Icg at daring this ccndition is di F.

The r:,tio of unbIcaed 4 t ts hlceked A teere eore is reficctive cf the ratio of blocked to unblocked core coolant i' o w

.c.,

1 W unblocked

.11 or a flow area blockage et. ~ S 9,..

r blocked d.a The trend of increasing temperature from the outer periphery to the centrni gion suggests again the pos sibility that the centr.1 core region is the core:

most highly damaged.

Further, the peripheral exit t. cpers ares are close enough to the core inlet temperature that they suggest that the -inimum occurs flow blockage /in the peripheral region.

It should be noted that the one exception to this central to peripheral trend is a core temperature reading in the center core region which is much below all temperatures, including the inlet.

Another observation of significance relative to flow blockage resistance is that _.

tran for of pt ;.ing frca the S/G Al to A 2 pump shifted the Iceation of the highest exit coolant temperature to another quadrant in the central core region. This suggests variation in inlet lower plenum flow and pressure distribution attendant to single pump operation. This is consistent with analyics in the Oconee reae;or per formed by T. %t t of Technoldgy for I:nerg'y Corpc. rat ion, in Knoxville, Tennessee.

o/q 90

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1.etdo n l' low Block.y.e The letdown Iine has given some evistenee of finw h i n. t....,.

i,.. v,........ ; n nw

8

..- a l t froa Ic31stanet, folio. ing t he acei.!ent.

thi; flow pluckage c.uld carryover t f.-eme d eb r i s f rc.; tV et re region.

<imila; f l o.. rc:istance har keen observsd, ceurdin; to I.. N.i! ! T*ll - ! shift u;ierei or, when s y st em c rud has been shocked loc e i n D'l -l. As a conseque+ ice t he obse rvat ion is not conclu ise, "at is qualitatively eensi,. tent with fuel f r;.gr. at a t ion.

I f the filter veeame closged over an e;ght hour per iod, a it. r t h., a cc ident, it could be frem

rud.

If it Sceame el,gged in a ten ainute peri,, it is pro: ably frca large particlc4 (such as a fuel).

The actual occur

.".ec should be checked.

Also the ahiiity of the sus;'ect ed ficw rates in the core to levitat e f el part icles should be cheekcJ.

NATURAL CirCill.ATION CCM COCI.ING A.iSESS"ENT The The pec'.ious considerations indicate the core config'tration is :, table.

calculatiens presented here show that natural circulation enn cool the core in its present supposed state.

A number of natural circulation analyses have been performad with varying,

degrees of flow blockage frca 0#. to :s 900 blockage by P.GW and L*;EL.

Typical results obtained are presented in Table 1.

These results indicate that adequate core cooling can be achieved with substantial core bicekage; i.e., up to 93*, without gross core boiling.

Since che core condition in TM1-II is estimated tu be no worse than # 90*. effective core flow bleckage, a stable condition with no hulk core boiling is expects.d.

Some local boiling in the most restricted flow areas would be expected. Ifowever, the intended operation with high subcooling s ill preclude any 1:ulk boiling.

RECOWEND \\TICNS FOR FliTURE WORK The results from calculations being performed at various locations will be very important in verifying the reactor core medel of the core prerented here.

For example, Carl Oben:hain of IN!:1, calculat ed t he guide tube temperatures and showed that t hese t empe: at ui es closely fol los.ed t he fuel cladd ing tempera t ures.

Thi s indicat es that the guide tui.es are probably brit t !c in egions where the 99 3i0 cladding is compl et el y des t royed.

Furt hee calculat ions could give a bet t ty

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.I.Tr nl.r.ef:!'.ll.1: \\ T..B.'.l. u.t'll April 10, itC9 I;eac t o r i ond i t ions :

t in -

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t e o r t-exi*

( t.t

- 3..? F )

core

~te 9eore 3Yi;

~

6

ornal ope: -len - 4 p.np s ore flow 137.9x10 lb.hr Sinp,le loop - one punp ccre flow 1/S - 1/4 x full core flow * (next page)

Uab_ locked lare !!:at T..: l a n c e Sinc l e Lorn 'S intl e Pump a) 1/8 of full core flow C

Qcore

Ncore p ateore P/I1) 6

t. orc - Q 'F0 (5%".O (3.413x10 ITr ' E.1 o

,v1.1 F C

3 (137.9x106 lb/hr ( i9 E.1, C,gy.e 8

IboF b) 1/4 of full core flow

- 20F d t Core a*

B_ locked Core (Current C.sndition) Heat Balance - Single Looo/Sincle Pump C

e re p ;tcore) blocked (W

C (W

eore P d t eore) unblocked

=

eorchlocked (a core) unblocked l blocked W

=

Neorcunb;ocked eore) blocked A unblocked at (a teore)lalocked ~S.50F t

(3 eore)unt>1ocked-I-24

[sblocked 1-2 F Range

.11

.24' A unblocked 8.N 89 -

76'. flow blockage 99 iJ S9 - 76*. core block.y,e ir. peripheral regioh liccau.e of hi gh expec t a t ion t ha t no r:aa l siny,le pu.up operat ion core flow is clo.ci to 1/S t' n o rma l four pump operat ion core flow, t he expect edje ul t

.5

.trI.\\Ci rir.T e1 0>nt inued is closer to 3%.

\\nalyn s by 1. Stott an.! Br.N inJieut e - 1/S of t he no: m:i1 four pump flow through he core n:id. r one rump op. rat ica, :nost of t he five 1 y-p tssing :!.e core and flowir.; in reverse mode in non-operat ional loops.

G G

O 0

.

ML19220C685

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