ML18046A746

Provides Status on SEP Containment Analysis for Facility. Containment Design Temp Is Exceeded in Some Cases.W/O Stated Encls

ML18046A746
Person / Time
Site:	Palisades
Issue date:	02/17/1981
From:	Vreeland D LAWRENCE LIVERMORE NATIONAL LABORATORY
To:	Butler W Office of Nuclear Reactor Regulation
Shared Package
ML18046A745	List: ML18046A744 ML18046A746 ML19351A119
References
CON-FIN-A-0241, CON-FIN-A-241 TF81-029, TF81-29, NUDOCS 8106250285
Download: ML18046A746 (69)

Text

I TFBl-029 W. Butler Containment Systems Branch Divisio~ of Systems Safety Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.

20555 NUCLEAR SYSTEMS SAFETY P~OGRA~

February 17, 1981

Subject:

~EP Containment Analysis and Evaluation for the Palisades ~~we~

Plant, Docket No. *50-255.

LLNL Containment Analysis Support _fo~r SEP, NRC F!H #A-0241

Dear.Mr. Butler:

This letter is prepared in partial satisfaction of the LLNL Contaihment Analysis Support for SEP, NRC FIN# A-0241.

The items cor.;pleted are covered in Task 1, i*<hich is the Palisades SEP facility.

These are subtasks a, b, c, d, e, f, and g.

Subtask d, which is the secondary pipe brec.k ana1ysis was also completed.

The results of this work ar~ pro~ded as Enclosure 1. provides a report which describes the analysis of a post-accident containment p~essure and temperature response to a p~imary an~ secondary system pipe break.

The analysis was perforriled by Energy, Inc., Idaho Falls.

Mass and energy release rates utilized in the analysis were calculated using both *RELAP4,MOD6 and MOD7.

Calculation of the post-accident containment pressure: and temperature was done using CONTEMPT-LT/028.

Several cases 1*1ere run for both the primary and secondary system analyses.

For the primary system pipe breaks, the calculated transient reflects a

. post-accident containment pressure from 66 to 69.7 psia and a temperature from

'273°F to 325DF depending on the particular assumptions made.

The containment design pressure and temperature for Palisades are 55 psig and

. 283°F.

There.is, therefore, a 0.0% to 5% margin between the calculated post-accident containment pressure for primary system breaks as identified in Enclosure i and the design pressure.

However, the containment design temperature is exceeded in some cases.

• The containment response to a secondary system pipe break is also gi~en in Enclosu1-e 1.

This analysis was a main steam line break \\':hich considered the blwodown of beth steam generators with several difficult assumpticns.

The calculated transient reflects a post-accident containment pressure from 97 to 107 psia and a ~emperature from 42QOf to 465°F.

Both the pressure a~d te~perature exceed design by a substantial margin.

W. Butler February 17, 1981 TFBl-029 Page2 If you have any questions concerning this matter or require additional

~nformation, pl~as~ let me know.

DGV:l gd s102.17/1933u cc: Cw/ encl)

W. Russel i. ~ --

S. Br m*:n ( 2)

C. Tinkler (3)

(~1/0 encl)

G.: Cummings B. Bowman Sincerely, David G. Vreeland Principal Investigator I.

~:~- Chang fr May 1:, 1.981 Contain~ent Systems Branch 1Division of Operating Reactors Office of Nuc l f:4r-Reactor Regulation U.S. Nuclear* Regulatory Commission Washington, O.C.

20555

Dear Mr.Li:

FIN A0241 Containment Analysis Support for the Systematic

• Evaluation Program.

Main Steam Line Break Analysis, MSLB, for Slowdown of One Steam Generator

1.

Palisades Plant~-Automatic Initiation of Auxiliary Feedwater System at Palisades Plant, Docket 50-255 License DPR-20, January 21, 1980 letter from R. W.

• Huston of Consumers Power Co. to Dennis L. Zieman of NRR, NRC. -

2.

Palisades Plant--Proposed Technical Specifications Change Related to Contain~ent Spray Initiation Time, Docket 50-255 License DPR-20, Nove~ber 24, 1980 letter from 0. P. Hoffman of Consumers Power Co. to D.

Crutchfield of NRC

• Attached is the MSLS ~nalysis for the blowdown of one steam generator.

• This would*be the worst case analyzed provided a fix was imposed to prevent both steam generators from blowing dawn.

Iri this c~~e the single failure would be loss-of-offsite power with a diesel aenerator failure.

The available containment heat removal system would t~en be re~uced to

• hm spray pumps and one air f c.n coo 1 er. The mass and energy re 1 ec.se* data used in this analysis were taken from ref. 1.

The assumptions that went into arriving at these mass and energy release rates were found to be conservative and in agreement with the SRP provided a failure of a MSIV is not considered.

The assumptions used in the containment response calculation are based on the proposed technital specifications discussed in ref. 2 The results of our analysis show that the calculated peak pressure is 58.5 psia reached at 67 seconds.

This 1.2 psi below design.

The calculated peak temperature is 413DF reached at 37 seconds~ Therefore, based on this analysis, a fix which would prevent the blo~down of both steam generators would limit the calculated peak pressure to 1.2 psia be 1 m.; design.

Yours truly, b~~d1.~J David Vr:::e1c:n.:

Principle 1n\\*estig2:~o*;..

At tc.chrn;;nt

SEP Containment Analysis and Evaltiation fa~ the Palisades Nuclear Power Plant.

Contents 1.0 Introd~ction an& Backgrotind

. 2.0 Containment Functional Design 2.1 Review of Palisades Containment Design Analysis 2.2 Primary System Pipe Break 2.3 *secondary System Pipe Break

• 2.4 *Reanalysis of* Pali*sades Containment Design 3.0 Primary System Pipe Break

. 3.1 Initial and Boundary Conditions 3.2 Slowdown Phase 3.3* Reflood Phase 3.4 Post-Reflood and Containment Response Calculation 3.5 Containment Response Resul~s 4.0 Secondary System Pipe Break 4.1* Assumptions 4.2 Containment Response Results 2

2 3

. 3

• 4 4*

5 5

6 1

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_, uu 1.0 Introduction and.Background On J~nuary 1, 1980 the Office of Nuclear Reactor ~egulation (NRR~.

initiated a two-year prograi11.with Lawrence ~ivermore National Laboratory

~LLNL) titled Containment Analysis.Support for the Systematic Evaluation Program (SEP).. !This-program is directed toward resolution of SEP Safety Topic VI-2.D; Mass and Energy Release for Poisible Pipe Break Inside Containment, and Safety Topic VI-3, Containment Pressure and Heat Removal Capability.

The containment structure encloses the reactor system and is the final barrier

• against the release of radioactive ffssion products in the event of an accfdent.

The containment structure must, therefore, be capable of withstanding, without loss of function, the pressure and temperature conditions resulting from postulated LOCA and steam line break accidents.

Furthermore, equip~ent having a post-accident safety function must be environmentally qualified for the reswlting adverse pressure an~*temperature

~~....-:="'C" conditions.* To accomplish th~ obj~ctives or~n~p~ogram, first,.th~ existing docket*information was reviewe~ and evaluated and then additional analyses

~s

. were performed Krequired.

The purpose of this report is to docu~ent original anal.jses performed by the LLNL on the containment *functional design capability of the Palisades Nuclear Power Plant

. an~ ev~luate~existi~~-~naly~~;:io~

conformance with current NRC criteria *. :*._

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'2.0 Containment Functional De~ign Palisades is a Combustion Engineering Ph~ licensed to operate at 2200 t.,,..

'll'i I.

• The primary coolant system is a two loop system consisting of h:o steam generators with two cold leg loops per steam generator.

The containment systems include the containmeQt structure and associated systems.

These systems include containment heat remcival systems, containment isolation syste:11s and a combustible gas control system. f

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• l The containment ts a steel-lined,

,:i... _~ r: *{\\:n: t:1 pre-stres ~ed\\-'-po~s.t-tens ion concrete structure with a net free volume of 1,640,000 cubic feet.* The containment structure houses the nuclear.steam_ supply system~. including the reactor, steam 5enerators, reactor coolant pumps and pressurizer, as well as certain components of-* ttie* engineered safety fe.atures systems.

The containment is designed for an internal:pressure of 55 psig and a temperature of 283°F.

2.1 Review of Palisades* Containment Design Analysis There are two separate calculations which make up the containment design

. r analysis.

First is the mass and energy release analysis for postulated LOCA's.

This consists of a blov.*down, reflood and post-reflood phases.* The results are mass and energy release rates into the containment.

For PW~'s there are two possible break types which must be analyzed, a prima~y system pipe break and a secondary syste~ pipe break.

A break on the ptfmary side

~-

generally results in the most.severe pressure~~-~-ponse in the containment while a break on the secondary side results in the most severe temperature conditions in the containment.

The second calculation which is per.formed in the containment design ana1ysis is the containment response calculation~ This results in the contair.ment temperature and pressure response to the mass and

_energy release from the postulated breaks.

The acceptance criteria used to evaluate Palisades'* Conta~nment Design

. Analysis was based on the Standard Review Plan (SRP).

In order for the containment design analysis to be found acceptable both the mass and energy release and containment response calculation must meet the acceptance criteria specified in the SRP.

2.2 Primary System Pipe Break The SRP specifies several acceptance *criteria c.pplied to the mass arid release analysis for prjmary system pipe bre~ks. these are bred~

Tl/ ?c..h'>a.d~s

• Fs AR

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location.

In the Gabe of G1nRa the most severe mass and-energy release rate calculated for containment design was done assuming a.double-ended cold leg discharge break with no accounting.for the r~flo~d phase or energy in the

~econdary system.

Since this does not meet the acceptance criteria sp~ci~ied in the SRP or*previous}y accepted methods by the t~RC staff, this analysis is unsuitable for containment design c~lculation. Since the mass and energy release rate analysis is found unacceptable, so is the containment.response calculation based on the mass and energy release rates.

2:3 Secondary System Pipe Break The most r~cent se~ondary system pipe break analysis.that was r~viewed was submitted by Consumers Power Co. to the U.S. NRC on January 21, 19S0.1 In this analysis* a main steam 1 ine break (MSLB) analysis v:as performed.

In this analysis the blm1dov:n of one ste9.m *generator with feed\\'i'ater isol"aj:ion and

--~

• ~

1 oss-of-cff site pm*: er was considered.

Ho*r1ever~~e analysis did not_.:address the possibi.lity of a single failure of one of the main steam isolation valve$

v1hich could lead to the blm*;t:;:;1*m of both stean generators. Therefore, the analysis was considered incomplete and.unacceptable.

A.more thorough discussion of the MSLB analysis is given in Section 4.0, Secondary System Pi~e Breaks.

2.4 Reanalysis of Palisades' Containment Design As mentioned earlier in Section 2.1, Review of Palisades' ContaiDment

.Analysis, there are two separate calculations which make up the containment design analysis, the mass and*energy releas~ rate and the containment 1

Palisades Plant - Autmoatic Initiation of Auxiliary Feedwater Ssyte~ at Palisades Pl~nt~ Docket 50-255 - License DPR-20, January 21, 1980 letter from Roger\\.!. Huston of Consuiilers Po\\:er Co. to Dennis L. ZiE;-;;an of i~RR, NRC.

response.

The mass and energy release rate calculation can be the result of either a primary or secondary pipe break.

The primary pipe break generally results i~ th~ limiting condition for calculating the peak pressure inside the containment.

The secondary pipe break analysis generally is the most limiting I

case for temperat:ure conditions inside* the containment.

Both of these an~lyses were performed and are discussed below.

3.0 Primary System Pipe Break For a primary system pipe break there are three phases in calculating mass and energy release rates. These are the blowdown, reflood, and post-refJood phases.

In each of these phases the calculation was done in accordanc~ with

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1 th: SR~ \\'l're;e 11 pcss1 ble un_der the c-en-st-r-a+rrt cf the computer codes used.~In general, the analysis was done in a manner that conservatively establishes containment design pressure; i.e.! maximizes the post-accident containment pressure.

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(J 3.1 Inftia1 and Boundar*v Conditions The initial and boundary conditions for this analysis were defined to satisfy the requirements of the Standard Review Plan.

The single failure I.

assvmpt ion for these analyses was a loss of one di es el generator.

The in it i a 1

~-

power was specified to be 102% of safeguar~s design rating or 2690.76 MWt.

A steady~state ~ass and energy' distribution was provided in the primary and secondary coolant systems consistent with the conservative core power.

The break flows were ~alculated using a discharge coefficient of 1.0, with the Henry-Fauske correlation for subcooled and the.Moody correlation for saturated fluid.

The safety injection flOi~S.,.,*ere minimum, corresponding to *the diesei ge~erator failure.

The mass and energy release ana~ysis was performed with RC:tJ.P4 i*;OD5.

Steam quenching by the-.safety injection 1*:ate:- cccurred cue *tc ~

r f..

the hoo:ogeneous equ{librium*(HEM) assur.:ptions of the RELAP_4Y.DD6 code.

All of the safety injection water temperatures were defined* to be 90°F.

Screiil was assuined to occur witb a low pr.essurizer pressure of.1750 psi a.

~ 1.0-second delay time was used in the model for conservatism; however, th~

I moderator reactl.vify feedback caused core shutdown before the control rods were effective. -The main coolant* pump po*8er v;as tripped off at the time of the break..

Steam generator isolation was initiated one second after the break and the valves were assumed to compi2tely close in five seconds. A 15-psia

.constant ~ontainment backpressure was assumed to maximize mass and energy release throughout the blowdown.

The end of blowdown was defined a~ the time the pri~ary syste~ pres~ure reached the contain~ent design pressure of 5~ psig.

a-_) -w:i.~ ~evv./,./!'J A.~vi<'.,.JI;",,{_ for

~ode..

The RELAP4 input deck was obtained from rrnc...._.Additiona1 information required for the analysis was obtained from the Palisades FSAR, and telephon~

conversations with C. Tinkler of NRC*and D. Vandewalle of Consum~is Power Company.

A thorough discussion of the modef-~anefte found in the Metffodology Report for the Palisades Nuclear Power Plant.

3.2 Slowdown Phase The blowdown analysis results are summarized in Table 3.1 and Figures 3.1

_through 3.4.

Table 3.1 itemizes the energy sources for the duration of the blowdown which ended at 20.4 seconds after the break.

The total energy released during blowdown was approximately 253.4 million Btu.

Figures 3.1 through 3.4 provide break flow and enthalpy out the break.

The accumulator flows start after 16 seconds c:nd do not reach maximum flow rates by the end-cf-blowdown.* The pu~ps coast down at different rates.

The

• pu~p nearest the break reaches zero rpm before two seGonds because of reverse flow through the pump.

The pump~ were not allowed to reverse, providing a cc;.ser'.'~tht:ly high resistan:e 1*1hich al 10-~::*s so;e flo\\>1 throuch the stec.."71 ~

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• g~nerator side of the break.

The other pump in the broken_.. 1~op coasts dm.;n to zero rpm.at a_bout 11 seconds.* The pump~ in the unbrok~n loop continue/ to have a posi"tive rotation throughol,lt the_ blowdO\\"m~ although it decreases to 500 rp:n 1n about 10 seconds.

Although the.scram occurred at about eight seconds, l

moderator reactivity feedback had already reduced the power to less than 7-1/2% of the initial power.

The mass and energy release rates and energy sources were qualitatively compared to the CESSAR results for a. double-ended suction leg slot b~eak with the sc.me area.

The similarity of the results sugg~sts the RELAP4 calculated blo1:1dm*m results are reasonable.

3.3 Reflood Phase LL The reflood analysis for the double-ended pump-sfection break was assumed to immediately follow the LOCA blowdown analysis:

The analysis __ 1:ias performed

-~

using REL~.?4 MOD7.

Within the limitations o*(R.~t=AP.4 MOD7, the analys1s was 1-~

performed in accordance 1*Jith the requirements of Section 6. 2.B of the Standard Review Plan (SRP).

• Initial conditions for the start of the reflood analysis v:ere based-on the end-of-blowdown {EOB) results.

EOB was defined to occur when the primary

_system pressure fell below the Palisades containment design pressure of 55 p"sig which occurred at 20.4 seconds after the start of blm<Jdm~*n.

At that time, the core power level had dropped to 159.41 MWt or approximately 6% of the initial pm*~er. The accumulator flows had been initiated on 1 ow cold 1 eg pre~sure trips of 262.5 psia which occurred at about 16 seconds into the blowdown and had reached a total of 5900 lbm/sec at the start of reflood.

The reactor coolant pumps had coasted down and the rotors were locked.

For the reflood analysis, the primary system was initialized at the cont2in~ent design pressure, 69.7 psi a.

-fh~ prira2ry system junction flo~s

~~

were zeio except for-the accumulator and lower ~nu~ inl~;-~nd outlet junctions.

Heat conductor temperature and primary system state conditions were established based on the EOB ~onditions. *Core power continued to

~ecrease according to the ANS decay heat curve.

A natural. c1rculation heat transfe'r model v:as used in the steam generator secondary to maximize the. energy transfer qi.tes to the break.

The* primary coolant pump. rotors were assumed locked to conservatively provide resistance to flow.

A closed valve was modeled in the intact cold leg of the b~oken loop

.. to conservatively increase the flow through the ste~m generator.

For numerical stability of the RELAP4 computer code, the Emergency Core Cooling System (ECCS) flow was modeled as being injected directly into *the downcomer at a temperature of 300°F.

Plant specific information was predominantly derived from a RELAP4 Reflood input listing for the Palisades powe~ plant which was obtained*from the

-:i-r--~

Nuclear Regulatory Commission_(NRC), and from-~Palisades Fina1.S?-tety analysis Report (FSAR).

Several sensitivity calculations were performed to evaluate vJ~ious input model arid code options.

The results of the sensitivity studies are documented in the methodology report. The Palisades reflood transient results are 8

presen~ed in Table 3.2 and Figures 3.5 through 3.~ Table 3.2 is a summary of th~ en2rgy balance at the beginning and end of reflood.

Figures 3.5 through 3.8 pro~ide break flow and enthalpy out the break.

The accumulator flow is initiated at 5900 lbm/sec and quickly rises to 6400 lbm/sec.

The flm-J remains constant until 40 seconds, c..nd then is ra,t:ped dvi>m to 0 lbm/sec at 50 secon..ds \\*ihen the accumulator is empty.

The HPI flow comes on at 0.6 seconds and remains at about 650 gal/min for the duration of the trc:nsien't.

The LPI flow comes on at 7.6 seconds and vai-ies.in mc..gnitude between 400 and 600 lbm/sec for the duratton cf the transient, depending on the pri~ary system pressure.

The primary system pressure starts at 59.,7. psia~_increases to 160 psi a at 20 seconds> and then slowly decreases to 100 psia.

T~~ pressure increase can be attri.buted to steam binding in the primar.y system.

As the ECCS water

~nters the tore, it boils away faster than the generated ste.:.m tan escape l

through the -break"~ After 20 seconds, the core is quenched and the steaiil generation rate reaches a new pseudo-steady-state with the break flow.

Normally, the end of reflood is defined as the time when the core recovers to within two feet from the top of the core.

In the case of. Palisades, the

.maximum mixture level is less than seven feet at 50+ seconds into the transient which.is still four feet below the top of the 11-foot core.

Ho1'l'ever, the core-stored energy was essentially removed at 30 seconds into the transient.

Tha reflo6d calculation was extended to 100 seconds to determine when and if the steam generator side break flow would beg_in a rapid decay *expected

• ~

• ~

after the accumulators err:pt i ed at 50 s econds-:~-:Tuce the. re.pi d fl ow-decay did not oc~ur, the reflood calculation was continued beyond the time when the containment calculation predicts the peak pressure and temperature at 84 seconds after break or 64 seconds after start of ref lood.

Because the safety injection water was assumed to be at 300°F, the extended duration of the

_reflood analysis is corisidered to provide a conservatively high energy transfer rate to the secondary.

3.4 Post-Reflood and Containment Response Calculation The tontainment model used was based on a CONTEMPT deck received from the NRC.

The mass and energy flows to the containment were replaced and the

• remaining data carefully checked against the FSAR and other sources. The analysis was performed using CONTEMPT-LT/028.

The heat structures used ar~ listed in Table I ' 'j '.* *, '

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• ~ ;.\\11-7 L::.

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t.:-:t J ~

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.:J 3.3.

All the structures are represented in rectangular geometry.

The thermal conductivity and the volumetric heat capacity wer~ checked for the four materials used:

steel, concrete, insulation, and* air (gap).

I The heat capacity was found to be about t\\*10 orders of.-m~gnitude low for insulation and was changed.

Tag ai11 i /Uchida boundary conditions were used for all heat ~tructure surfaces except the base s 1 ab, which was as surn::d to be covered with v1ater.

The Tagami peak t-:me used was 20 seconds, 0.5 second before the end of blowdown.

The basic assumption was that off-site power was lost and that one diesel

• generator failed to start. The cooler and spray pump start times are based on the generator loading sequence.

It was assumed that one fan cooler was operating and that it started at 23 seconds ~fter the break. *The heat re.11oval rc.te 'n'as variable, rang_ing from 97.5 MBtu/hr at a containment temperature at 3S0°F to 3.0 MBtu/hr.at l04°F.

The one operating die~el generator \\>ias_.:O!-so assumed capable~-bf powering _two spray pumps_.

Both pumps together were capable of 1.34 Mlbm/hr (2700 gpm) with spray efficiency of 90%.

The containment spray st~rted at 84

~econds and used water from the Refueling Wat~r Storege Tank until 30 minutes when the tank was empty, at which time the water source switched to the

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The heat removal rate for the shell and tube heat exchanger in

. the containment spray system is computed in CONTEMPT.

The entered parameters were:

the product of the heat exchange surface area and the overall heat transfer coefficient was 2.28 MBtu/hr/F; the.cool.ant inlet temperature was 114°F; and the coolant flow rate was 2.0 Mlbm/hr.

~

General initial conditions are given in Table 3.r.' Initial conditions for the primary sy0-em refer to the end of blowdm*:n.

No \\.;ater 1-;as introduced to tit~ '1.,..,,,_ \\

C.,; the Elry.. *e1l as an initial step input.

The evaporafi on-condensation model in c c-r-t1t 1"n,..,~ /\\ t' the ~

wc.s bypassed unt i1 the en! ef b 1 O\\*:do\\\\*n.

The frc;ct i en of 1*:a 11 or-

-ro-

cooling coil condensate ~ransferred from the

.the pool.was set at 0.92. The heat and mass r~... ~~ ~ ;.~-*.::J I: 1" *~ :*.l t..*~ :""!

l

~

i~.~ s\\f z.i t~~ ~~

super~ea"ted-Efi.=y.;:?11 ~a-tmosphere to tn-.~A i 11 !"\\~

transfer.multipliers were set at 1.0, and the temperature flash op~ion was u~ed. -

Tvrn methods of treating the post-reflood period were used for this I

analysis and.thfee different assumptions made about the mass and energy release during re~lood, resulting in six cases.

The blowdown mass 2nd energy release was the same for all cases.

Peak flow was about 77,000 lbm/sec at 525 Btu/lbm, and the blowdown ended at 20.4 seconds.

The reflood data~

~-1-.*1.-l te- -rrcs.r~W. ~

1 asted 100 seconds~

120.4 seconds after the break. The core was not covered, or even two-thirds covered at this time, but it was substantially cooled. Therefore, the end of the RELAP4 reflood run was defined to be-the end of

~vc-r, *~~)._

reflood...a-s the accumulator flow had been ramped down to zero between 60 and 70 seconds after the break.

During the post-reflood period; decay heat, heat from the secondary

--~

system, ::rid heat. fr*om the hea~ structures in""tri@iJrimary system are released to the* containment.

The decay heat is released over the duration of the run based upon the ANS standard decay heat curve plus 20% and an ultimate reactor pm'.'er of. 2638: t*n*Jt plus 2% for instrument error (excluding pump heat).

The heat from the secondary system (61 M3tu) and the primary heat structures (53 i*iBtu) was all released by one hour after the break.

A linear ramp to zero was used.

The amount of heat released to the containment by the secondary was determined by obtaining the stored energy in the water.in steam generators and in the SG tubes at the end of the RELAP4 reflood calculation.

Assu~ing that this was based on 32°F and th.at the entire steam generator would be at 212°F. after one hour, the amount of heat avai 1 able to be rel eased was computed to be 61 MBtu.

This is* conservative because the contain~ent pressure will not decrease to atmospheric pressu~~ in one hour, and so the secondary syste~ will be hotte~ than 212°F.

For the primary heat structures, the energy s_t6red in all of the* *heat structures used. in the ref1ood model, except the core (fue~ rods) and the 'steam generator tubes, was used in the model.

It

. was conservatively ass.urned that all this metal would be at 212°F. after one hour, \\dth the tli-ffereni=:e (53 MBtu) being released to the containment.

During the post-refl_ood p_e~iod, two different methods were used which differ orily i.n the manner in which the. energy from the secondary.system and

. primary heat structures is. rel eased after refl ood.

/

In Method A,~J1e pr-i-riiaqr_

• c~f.:g;-1t ;;,**wu-mjP+ i~'? uct'iu;aMI. the energy is released directly to

• l<f>t.."tA..:n*~~f\\r Withcu.' ~~' * * 1 -T.

. fr1mM"fj

. t!~.cl6.A7 s.y~"ti!.,...,

t:he er;,Xal 1.

The a.11ount of mass accompanying this_ energy rel ease is requfr.ed, and it is obtained by assuming that the heat is used in converting :water at.

saturation to steam. *A typical value for the heat of vaporization at the

• pres~ures experienced in the containment for the fifst hour is 925-Bt~/lbm, and this has been used to calculate the mass release rate.

This m~thod is

--~

~"=t:r...:_~..J¥ conservative~ s i nee after some* time t~water in the priJiiary system

~

will cool ~a below the boiling ~oint and most of the decay heat will go into

. heating the water to saturation and leaving very little to generate. ste~ *.

-;;1:...Ziu-cti~

• The systems (HPIS and LPIS) that inject water into the_ primary are not modeled in. Method A since the mass and energy flow from the primary.i~

• already calculated as described above.

I

~**.... h d B t°h 68UTE'I __ \\

...1. "'"h d

n 1*1e1... o

,..w..;;.e P"".lrnary cor.:;:nrrtn".enc-45 aet:i're iF1.t ** ~ffee11&1. e ecay

\\.._

heat and heat from the secondary system and from the primary heat structures*

+t,,o~ 4

~d,,,- eool"'-"'r ~;1stc.rri is passed to the prl!Tlary compartment/\\. The model "in CONTEMPT then determines how much steam is produced, hO\\'I' much heat* goes into increasing water temp~rature~ and so ~n. This is much more realistic than Meth6d A since it allows the steam production to decrease with time.

It is still tonservattve

• since.the hea.t. input has been calculated to be conse*rvatively high..in Method S, injection into the primary ts explicitly ~ode1ed. The LPIS

{5000 gp~) and the HPIS {450 gprn) both take water from the refueling water storage tank until it empties at 30.minutes.

After that, only the HPIS c,-.,,..-tt~;l"ln\\0.~ ~ "'r continues, ta~ing water from. the cii=y11*c11 poo-i-.

There is no heat exchanger on Cc;.-\\~;......,~~ s...,,:..r either of these systems.

The di-j*11e1l ps~ is gradually cooled~~ since the water tha~ ~s recirculated through the containment spray system does pass through a heat exchange~.*

Three as~urnptions were 4onsidered for the mass and energy release to the containment during reflood.

In the first assumption, only the steam flow from the SG side of the break was used from the RELA~4 results.

To be (-e~... ~~iall..

• ~-F-C j) conservative, it was then assumed that all of this dry stea.'Tl was superheo.'ted to 1300 Btu/lbm (about 5006F), and the energy release rate was obtained by multiplying the steam flow rate by 1300 Btu/lbm.

The actual effluent enthalpy is about 1200 Btu/lbm for the first 20 seconds of reflood and gradually decreases to about 600 Btu/1 bm after that.

Thus, assu;ning that* ori)Y dry

• --*~

superheated steam is released is*~

conS:e.Lvative.

~--*

.*-.*~

In the second assumption, the mass and energy flow rates from the SG side of the brea.k were used, but the energy flow v.*as augmented to account for superheat i ng.

At 70 psi a, saturation is about 310°F, and the specific enthalpy of steam is 1185 Stu/lb.'TI.

At this pressure, the specific enthalpy at 500°F is only 1282 Btu/lbm.

Whi]e the SG tubes are a.little below 500°F at the start of refl ood, their temperature is on the order of 3S0°F at the end of reflood.

Therefore, adding 100 Btu/lbm for each pound of stea~ flow is I

I

\\

\\

~-

'~

y conservative.

The steam flow rate used to calculate this added

'-' v energy was th~ same as that used in the first assu;.;ption.

The additional energy was about 8°/o of that computed by RELAP4 at the beginning of refl-ood arid about 6% at the end of reflood *

]n the third assu;nption, the mass and energy released from the SG. side of the bi*eak by REU\\!'4 1*.*ere used direct.1y._ The 1 i~uid phase fal 1 s to the pool c.s released.

Naturally,. this case results in lower peak temperatures and pressure~. than the superheated-steam-only case, but i~ is more realistic and it is conservative.

For all three assurnpti ans, the.release of water from the pump side of the.

break is i gnoreiJ.- *

~~

... ~l"ri'\\e...,.:\\ SM,....~ -

~ell fle-&l.

The This release is all liquid phase and goes directly to the

-?c!vVvP amount of water in the*~ has a negligible effect on the u>1icur.-~-t:

temperature and pressure h,is~ory of th~ vapor r~gion. The RELAP4 modt.*l had to use ~CCS water at 300~F in order to avoid instabilities, whereas the ECCS e,P c-fJ L "&>'

t 'I

• water ).f1'"'" actual..:H:y w:ff1"" be about l00°F.

Since the water coming out the pump side of the break will have had no contact with the core and little with any of the metal enclosing the primary system, it should not be significa-nt1y*.

warmer than when it 1 eft" the accumulators.

In view of the 1arae difference betwee~ t~ctual and the model ECCS water temperatures, neglecting the 1 iqui d flow from the pump side of the break is P!-~~-~~h:~ more realisti~.:~~~ncluding it~

3.5 Containment Results The results of the CONTEMPT runs are shown in figure j_g through 3.14.

Figures 3.9 and 3.10 show the results for the case where only dry, superheated steam flow from the SG si d.e *af the break was cons i cered~ and the energy re1ease rate during reflood* was obtained by muftiplying the steaiii flow rate by*

. 1300 Btu/lbm (which is approximately the specific enthalpy at S00°F and 70 psia).* Figure 3.9 shows the results for Method A and Figure 3.10 shows the results for Method B.

The two cases ate identical to 120.4 seconds since the difference is in how the mass.and energy releases are handled after i:-eflood.

The figures show thatthe contain::i2nt atmosphere.reached aliiiost 70 psia and 3256F at 84 s~conds just befcire the containment spray began.

Since only dry

. stea!n \\*:c.s rel eased 'to the containment d:.iri ng ref lcod, the cent ai n:i"lent spray '

has an i:ii:nediate and *dramatic -effect on the containment vapor region temperature and pressure.

,Th_~ peak pres.sure equals the containment design

't' pressure of 69. 7 psi a. and tho pe.:k terpere.t.ldre ex,i;~l:i~1e-de.si9~..._..,,..e, du: e

_,.r In vie\\°' of the -e+tt. c..1_ly conservative assumption of *releasing only dry, SUP'\\'=rheated steam during reflood, this is not considered significant.

It is inconceivable that the ~uperheated steam could flow from the steam generator to the break with the saturated water and not mix to form a homogeneous, two-phase flow.

Since the Method A assumptions are not suitable for a long-term model, the run shown in Figure 3.9 was terminated at two hours, while the Method B r'uri in Figure 3.10 was. continued to ten days.

The results of the tvm meth*ods are quite close at two hours.

The dip in the atmosphere pressure and temperature at 30 minutes (1800 sec) in Method B is due to the shutdown of the* LPIS at the time v1hen the RWST runs dry.

This does not show up in Method A*since the primary system is not modeled._

The change in-~~e at 30 minutes i.. riithe Method A result is due to the fact that the source of 't'later for the

-. Cc>*t-.-.:..... ~..,.,.T >v.i.-._r:) c,r.... r~.,~

containment spray changes from the RWST to the \\'IC.rmer.d.r.fde1 ~ po0-1,~ The.Avapor r~gion temperature reaches 135°F at about 8.08 days (698,400 sec) in Method B (see Figure 3.10).

Figures 3.11 and 3.12 show the results for the release of a two-phase mixture vlith the energy flow augmented to account for superheating the steam fraction.

Peak pressure Ce:-1, ft'..,* k "'!<!' +--t" a 1,,.eJ.;.-/_dr1c..-

peak 1, t err:p er at u re is µt is about two psi below th~ design pressure, and the*

~-

zg; !". -

about; ~ad:::te4~-a:ttJ.;..-e. Behavior after a few h~ndred seconds is nearly identical for all the A cases and all the B cases.

This is to be e~pected since events are dominated by the absorptive capacity of the heat structures and the effect of the sprays.

The c::---~r~,..-., l--,.~~~-

at~ospheric temperature reaches 135°F at about 8 days.

/\\

In view of all the conservative assumptions shown in Figures 3.11 and 3.12 are sufficiently conservative and meet the Standard Review Plan requirement for superheated-stea~. Since the effluent to the containment will certainly not*be dry steam in view of the *carryover rate

-~.

fraction in the core, this assumption of wet steam with the steam fraction arbitrarily superheated appears to be the maximum which can be justified as realistic.

~'.-r,Ai... > ~,,,:\\atmospheric temperature dropped to 135°F after about 8 days.

So many f-..

conservatisms, including the use of 300°F ECCS water, were ~ade in arriving

(.rn-t""':t1.,"c-** \\

at the re1 eases to th~e1-1 during refl ooq_. an_d past-ref 1 cod t~~\\~_!bese HoweV~t' __.-,

results are definitely conservative...._ )"hey do_ ~meet the SRP mc.ncfate for superheated steam~

4.0 Secondary System Pipe Break Analyses of the containment response to ~ secondary system pipe break were

_also m~d~. For PWR's the most limiting break location is a main steam line break '>'lith pure steam blowdown.

In the case of Palisades the results show

. that a single failure assumption which allmvs both steam generators to blowdown will produce peak pressures and temperaturs which exceed de~ign values.

The model and assumptions that \\,*ere used in *analyzing the main steam line break are given in the following discussion.

t.*.1 Assum?tions A main stea~ ltne break (MSLB) analysis was perf~rmed*by Consumers Pcw~r Company.1 Results given in this reference are u~ed ~or comparison purposes.

In particular, mass and energy release data for the full power MSLB case discussed 2~ the Palisades FSAR i~ provided in Table 1 of reference 1.

I.

The Palisade_s FSAR full povier analysis assumed:

(1) 'A double-ended guillotine rupture of a main steam line inside the containment.

l2)

A reduction in feedwater flow from full flow to zero over the 60 seconds immediately following scram at le-ss than 2 seconds on hi.gh containment pressure; (3)

Both main steam isolation valves would close on low steam generator pres_sure (500 psia) causing the unruptured steam generator to isolate in eight seconds.

(4) Off-site power was available.

• ~

---~

(5)

Only two containment spray pumps were available; no air coolers were available.

. ( 6)

Pt.ire steam b 1 ow down (no moisture carryover).

(7)

A highly conservativ~ containment heat transfer model.

In addition to the results reported in *reference l, a number of analyses for Palisades containment response to the MSLB wer~ made.

The analyses employed RELA?4 to o6tain mass and energy release rates and CONTE~?T-LT/028 to obtain

• containment response.

The RELAP4 mass and energy release rates were obtained using a simplified model based on one volume, one heat conductor, one break junction, and one 1

\\

P~lisades Plant - Automatic Initiation of Auxiliary Feedwater System at Palisade*s Plant, Docket 50-255-License DPR-20, Janury 21, 1980 letter fro:n RoQer W. Huston of Consumers Power Co. to Dennis L. Ziemann of NRR,

!~:.;clear Regulatory Co;n:nission.

-1]-

f~ :~ ~ ~! ~-1 ~

~

f...... *.. ;; **'-

.~

!-.1 *..J t:... ~... "1 *; ;,.:;.:

~

~* !~1.. : ;"';: ~1 tr

~J tY t! ~Tti Q tl feed~-iater fill junction.

Two break sizes and three feedwater flow assum?ticns were anal)*zed.

• The resu1 ting break fl ow rates are su;r.iilari zed and co;npared to reference 1 results in Figures 4~1 and 4.2 *. For the RELA?4 analyses, the steam generators were assumed to b.e at 770 psi a, with an average water 1

enthalpy of 552 ** 2 Btu/1bm -and -contain 128,456 1bm each.

The primary system was assumed to be held constant during the blowdown with 513.83°F local

• temperature and a 9~2 Btu/hr/ft2 ;°F heat tr an sf er coeff i.ci ent in the steam generator. Figures 4.1 and 4.2 show that the main effect of the feedwater is

. to prolong the time of b 1 ow down period and i ncreas_e the total. mass and energy to the containment.-

The _containment res.ponses for a number of MSLB cases have bce.n c0mpaf"ed. -

The results are given in the following discussion.

4.2 Containment Response Results

.-----~

.. ~-....,~:.. *2::

Case 1 Th~ first case selected for analysis was intended to determine if the CONTEMPT-LT/0~8 ~ode1 used would gi~e results similar to those in referenc~ 1 if simi1 ar assumpt i ans were e:np 1 eyed.

Consequently, t\\-10 CONTEMPT. runs were made, using the reference 1 mass and energy release rates. These CONTEViPT

_ ruris a~sumed two spray pumps and one fan. cooler were available and a TPEM<. of.

eight seconds for the CGNTE-~iiPT Tagami/Uchida heat transfer correlation.

The one run assumed that off-site power was not ava i1 able so the two spray pumps were started at 84 seconds. The other CONTEMPT run assumed off-site power was available so the spray pumps were started at 30 seconds *

. The* containment t:ii stories-for the tv10 Case 1 runs are compared. to the reference 1 pressure history in Figure 4.3P.

The co~parison between the results* with spray after 30 seconds and the reference 1 results i~ close,

~"'\\"t._;,,...... t -.""C indicating an acceptable CONTEJ.:PT model.

• The1...tE:;qeratuie histories for the

~£.e. '

• ~N*.; F.:ede::. are shown i:i Figu,*es 4.3T-A and 4.3T-S. -.

Case 2 The purpose of. the Case 2 analyS*i s was to determine if the' one volume RELAP4 model was adequate for obt~ining mas~ and-energy release rates to the of-t c;:oni.ainmen

• The blm;1d9v;*n mass and energy release rates for various feedw.ater and break -area tdmbinations have been *noted in Figures 4.1 and 4.2 For Case 2, the* ruptured steam generator bl 0\\-1down was si.mul ated by the 36-inch break

. (area= 6.12-ft2) with main feedwater only *. This feedwater flow was*

initially 1650 lb/sec and ramped down to zero.flow at 60 seconds.

The unruptured steam -generator was assumed to isolate _(MSIV closure, not failure) so the mass and energy rel ease rates were obtained from refrence 1.. The mass and energy release rates for the two ste~~ generators were added for input tQ the CONTEMPT model~

The CONTEMPT assumptions for Case 2 were similar to the assumptions for the Case 1 run 'r':ith spray after a 30 second delay.

The containment pressure

• • ~

--**---~ -

and te::r:;:-erature response from.Case 2 is *shown_ in~figures 4.4P and 4*.4T,

. respectively.

The peak pressure is about 65 psia, which is slightly less than for the comparable Case 1 run and the reference 1 value.

Therefore, the.one volume R_ELAP4. model was judged to be adequate for obtaining mass and energy release rates. It is noted that co~plete phase separation i*s modeled in the RELAP4 analyses so that pure stec.:11 blowdown occurs.

Case 3 C~ses 1 and 2 established that the CONTEMPT and RELAP4 models were adequate for obtaining containment response to a MSLB.

Cases 3-and 4 were.

. designed to determine the respons~ of the Palisades containment to the MSLB for bJm.,.dovm of both steam generators, with off-site power available. Case 3 assumed that* each steam generator would bl ow down through a 24-inch break (3.06 ft2).

Case 4 assumed that the ruptured stea:n generator \\':ould blow

t.

~'I *. '

~

c,

~ '" h *l : !

~*

~ ~.. ; r~:~ f.; :~ 1.:~ D

,.A ~l : : ~~. r~ l-'

1.-;;

l~*;r ;: \\l.'.1 ?.:. J

~o~n through the 36-inch (6~12 ft2) break and the unruptured stea~ generator would blow down through a 24-inch break.

The assumptions used for Case 3 include:

(1)

If off-site power is avai~_able, the spray pumps will be acti_vated at 30 s~ndsafter high containment pressure {5 psig).

High containment pressure occ~rs in about 1.7 seconds.

A conservative value of 33 seconds was used in the analysis.

(2)

Single failure is the Main Steam Isolation Valve (MSIV) failure causing both steam generators to blow down *.

( 3)

Ruptured steam generator blows down through o'ne-h a1f the maximum areas, or 3.06 ft2, as areas larger than this would not give* a pure steam blowdown.

(4). !sol ated steam generator blows do\\-m through one-ha 1f the steam-1 ine area becuase of MSIV flow area restrictions, thus through _3.06 ft2 *.

(5)

A1l three spray pumps and a11 four-.fa~olers will be.aya)lable.

The si ng1 e failure is assumed in the MS IV.

(6). The CONTEMPT time T?EAK for-the Tagami/Uchida heat transfer correlation was changed to 99 seconds to correspond with th~ end ~f bl owdown.

(7)

Main feedwater is available to each steam Qenerator at 1650 lb/sec iriitially and ramps down to zero.flow at 60 sec.

The ~esul ti:ng containment pressure and temperature hi story is sho*,.,rn in Figures 4.5P and 4.5T, respectively. The peak pressure is about 107 psia,

• which is substantially greater than the 55 psig design pressure. r a

u 1-.:. A f"-1 The results show that the worst single failure assumption is a MSIV failure which would allow both steam generators to blowdown.

This is possible s-il1e the closure of Palisades MSIVs is only in the forward direction due to the nature of the MSIVs (check valves held open by air

~gainst the normal flow of steam). This makes it possible for a steam

~ine break to give rise to the blowdo~n of its associated steam generator plus the blowdown of the second steam generator through the failed MSIV,*

connecting header, and reversed flow through the closed MSIV.

This produces the maximum pressure and temperatures which are higher than.

design va.lues.

Case 4 The assumptions used for Cas~ 4 were identical to those used in Case 3 except tha: the ruptured stea:11 generc.tor \\':as al la*.. :ec! to b10\\"1dovm thrcugh the maxirnu~ area of 6.12*ft2.

The resulting containment pressure and terr.;:ierature predictions are shown in Figures 4.6P and 4.6T, respective~/. The peak pre~sure. is about 106 psi a.

Case 5 Cases 3 and 4 assumed that bff-site power was available. Case 5 was run to inveitigate the containment response for the loss of off-site power assumption.

Case 5 is similar to Cases 3 and 4 except for two assumptions.

First, because off-site power is lost, the spray pumps are not available until 84 seconds.

In addition, the loss of off-site power results in a cornple:te* and immediate loss of feedwater.

l I -

Case 5 was based on each steam generator blowing down through a 24-inch di a'ileter. break.

The.pressure and temperature response is shown in*

Figures 4.7P and 4.7T, respectively.

The peak pressure is about *gs psi a and a

{c.... "\\o..;""'c~ c-.lrri~ht.riC,...

peakAtemperature of 4&5°F *at 10 seconds~

-*-~

Analyses have also been. performed a~suming a fix which would prevent the blowdown of both-steam generators *. In this case the single fai-lure assumption is loss-of-offsite power with a failure of one diesel

• generator *. The mass and energy release data used in the analysis is for

• the full power MSLB with one steam generator blowdown.

This is discussed

(.Jo. t by Consumer Power Company in ref. 1 and provided in Table -t." The assumptions made in the mass and energy release analysis are the fo 11 owing:

1.

A double-ended guillotine rupture of-a main steam line inside the containment.

2.

A reduction in feedwater flow from full flow to zero over the 60 seconds imnediately following seram at less than 2 seconds on hi9h containment pressure *

3.

6oth rr.c.instea:n isolation valves,... ould close on--10.,.,* s::am generator pressure (500 psia) caus~ng the unrupture steam

• generator to isolate in eight seconds.

4.

Off-s.ite power.,.,*as available to maximize the rate of energy transfe~ from the primary to secondary.

5.

Pure steam blowdo~n (no moisture carryover).

For the containment response calculation the follo~ing assumptions were made:

1.

Loss-o1-offsite power and failure of on~ diesel generator.

2.

2 of 3 contain~en~ spray pump~ available.

3.

Containment-spray initiation at 36.7 seconds (200 gpm) and fuil flow at 52-.5 seconds (2680 gprn)~{ rz~;cfl.. (.L z..).

A

4.

1 of 4 air coolers available at 23 sec-ands.

5.

Tagami/Uchida heat tr an sf er correl at~ with Tc.gami peal ~iJme..at end of blowdown (6S ~econds).

The results of th.is ar.=.1ysis are the pressure and te;;iper-ature u..2P

<.f.~i responses shown in Figures ?--and f:"' The c_alculated peak pressure is 68.5 psi*a reached at 67 seconds~ This is 1.2 psi below design.

The calculated peak temperature is 413°F reached at 37 seconds. Therefore, based on this an~lysis, a fix which would prevent the blowdown of both steam generators would limit the calculated peak pressure to 1.2 psia below design.

Pc...l 1-~c.. A'1~ f'{ ~l"lt - -.t>rofe->2ef

'Tet.l..,..,,~~

Sr.<2.c.fl*u-...-t"1~nJ.

c..kt.,.,f-Rt!IC\\..t"ccl

-to lc""fc..\\*~... or1..t". <;.rrd...'j 2.....-~lt'<-t,?:,..

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t>PR -

7-'~,, N'owrr-.ber :1..4->

1'1S~

2.et~~....

-9YofT\\

D. r=>. H.off,.,....,,._""

of lC>r>.$LlY'~er.: "Pa-<J2..r

Conclusion Based on the results of Case 1 and 2, the CONTEMPT and RELAP4 models are adequate for obtaining containment response to a ~SLB. The results of Cases 3, 4) and 5 showed that the blowdown of both steam generator will result in containment pressure 0t1d telJlµera+ure:~ ***i:iich-exceed design values. This is regardless of whether off-site power is available. Tiu. Y"~s,,Jf~ of &:..£~ G

. 5 f...:we.d +-Lex c..

e-la..s ir d,,,..n(f-.w.L.-... L..wru(d. 1 r"ev-*,..f -;r-/..i b/~wd ~" of

/;.r;rl, s1'""" ft"<>~tors

~ol.4..Ld

-4et2..f --Hu:.

c. o""+t..~'r\\~v;\\ -rcJ fr-.G~<;.(.ar~

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I

Table3.1 P.0f3~-;-F-?

-Ji ~i }r f: l~ ~j Palisades Double-Ended Suction Leg 6reaK;_'.J;.iG'...:~f;-r;_~~

1.

Slowdown Energy Balance (Million Btu) -

0 Seconds Primary System Coolant Inventory 253.7 Steam Generator Coolant Inventory 140.0 Secondary Flow to T~rbine{l)

Accumulator System Inventory( 2) 19.5

• care Stored Heat 18.9 Conductor Stored Heat(3) 111.0 Decay and Fission Heat 543.l Note:

-- ---~-

Decrease 20.4 Seconds 246.6 7.1 2.0 138.0

-9.2 0.7 18.8 4.6 14.3 3.7 107.3 5.0 253.4 285.5

• -:.*~

(1) Flow continues until valuve is fully closed at six seconds after the break.

Energy value is net loss for stea.11 and feed1*1ater flo'rr'S.

(2) Accumulators and lines at 900f.

(3) Conductors include all metal transferring heat to the primary coolant system except for the fuel rods.

(; 5 Table 3.2 Palisades Double-Ended Suction Leg Break Reflood Energy Balance Reactor Coolant System tnventory Safety Injection Tank Water(l)

Safety' Injection Pump Flow(l)

Core Stored Heat(2)

Decay and Fission Heat Primary Vessel Walls Primary Vessel Internals Primary Loop Metal Steam Generator Inventory (I. L. 69.8 - 59.1 = 10.7)(3)

Steam Generator Tube Metal (I. L: 7.6 - 6:3 *= Li)( 3)

Approximate Break Flow Energy S. G. Si de Pump Side

• Total 72 (105) Btu 81 (106) Btu 153 (106) Btu Reference Temperature is 320F

• . Notes:

(Million Btu) 20 Seconds Decrease 22.4

-3.l

.77.5 13.5 15.2 11.0 12.3 59.3 3.3 12.4 3.5 28.4 1.0 137.9 33.7 14.3 3.5

=

TOTALS 290.9 156.4 120 Seconds 25.5 5.2 56.0

. 8. 9

.27.4 104.2 10.7

~

~

.... 237. 9 (1) The S. I. water temperature was 3000F to prevent numerical instabilties.

Actual value should be llOOF.

( 2) Based on ANS + 20% decay heat curve.

(3) Energy from intact loop steam generator.

(...*

>:'."\\ --~ :'* r.;:::;~

\\

• -\\~ ';; Q..,

1,;

~

(1?.,., (-. ~--

~

Table 3.3

~ ;,,-~~"',~....

~

l:~ r.: ::;.~ !:. ~. ~:

~t ES:i ti ~ !.! i !!

~

Heat Str~ctures in Palisades Containment Model

• Structure

1.

Tanks arrd piping {.~53 inc~)

2.

DLlcts (.10 inch)

3.

Reactor crane (2.35 inch)

4.

Internal concrete {33 inch)

5.

Gf~~ings and trusses

6.

Cont a i n:nent dome

7.

Containment dome.base

8. Contain:nent side wall

9.

St6rage pool floor.and shielded walls

. 10. Containment base slab

11. Biological shield wa11

• 12. Structural support st~el.

~*

Area ft2 19,332 20,072 6,973 9,401 20, 996 7,270 11,000 50,600 4,456 8,229

. 2' 340 26,320

---~

Deck received had 2.25, which was in error.

1, c

-~A-Thickness ft

.0378

.0083

.1958

.2. 75*

.0144 3.0217

. 7. 7~

3.5217

. 4. 35 12.44

7. 8572.

.45

• 't" t *

• lo~

• Table 3.4 Palisades Contain~ent - Initial Conditions Outside air temperature f

Outside air pre~s~re Relative humidity-of outside air Volume of primary capable of holding liquid_

Temperature of primary system vapor region Temperature of primary system liquid region

~-

Volume of~'~ tol"ic...\\~\\~"c~'\\

• ii'\\ Co..... T~1r-......,~...._l c;.14.......,p

• Volume of liquid pool ~ry.1011 floor

\\

to"' "'i e... \\ "Tnl:°"t Temperature of dl:.y."!-8-l-l.vapor region t I)" 1Cr.\\Y'l\\'1'\\LJ\\\\

Temperature of a....,y,*,*el1 liquid.region Pres sure in ~y.trll tc 1.;t~~"h\\en"\\

Re 1 ut i ve bm i dity in.dgr.w.e-+-4 u*-:tc.. ;¥\\MC. ~... t Horizontal cross-sec ti ona l area of d.l:;ywe+l lo V\\.i(.., 1°1"'1\\-r\\C-h \\

~

---

~

.8 a/ 95°F 14.7 psia 0.60 3050. 3 f t 3 250°F 2so°F

1. 6E6 ft3 10 ft3 120°F 120°F 14.7 psia

1. 0 8, 229 ft2

~-I TABLE J!1--*

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1. 530E07 1201. 8 1.3

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