French LMFBR core thermo-hydraulic studies for nominal and accident conditions

Nuclear Engineering and Design 124 (1990) 403-415 North-Holland

403

French LMFBR core thermo-hydraulic studies for nominal and accident conditions J . M . Seller, G . C o g n e t , E. L e b o r g n e , J. M o r e a u , J . R . P a g e s , J. P a p i n , B. R a m e a u

CEA, Grenoble, France J.P. L a h a y e

Belgonucleawe, Belgtum J. O l i v e

EDF, Parts, France

Ttus paper summarizes the present status of the analytical methods used by the CEA for the study of the thermal behaviour of subassembhes m a LMFBR core durmg normal operation. Particular emphasis is devoted to the effects related to lrra&atmn reduced distortmns. The methods used for the determination of the coolant flow pattern m the mterstmal space between the subassembhes are presented, and the formation and the thermo-hydrauhc consequences of local blockages are analysed. Sodmm boiling phenomena related to loss of coolant accidents are presented and methods are outhned for the estlmatmn of dry-out occurrence m following situations - Loss of flow without scram, - Natural convectmn boiling, - Complete mlet blockage wath coohng through the wrapper. Some observations and results from the SCARABEE m-pile experiments are discussed. A hst of pertinent reference ~s provaded. On pr6sente l'&at de l'art des m&hodes analytiques d6velopp6es par le CEA pour &udler le comportement thermohydrauhque des assemblages darts un coeur de RNR dans les conditions nommales et accldentelles de fonctlonnement On s'mt6resse en paraculier aux cons&luences thermohydrauhques dues h des d~formatlons de g~.om&ne, et h des bouchages locaux. Des m6thodes d6velopp6es pour d&ermlner les ~zoulements dans les espaces entre assemblages sont pr&ent6es. On analyse les ph6nom6nes d'6bullitaon du sodium cons~cutlfs ~ des accidents de perte de refrol&ssement, et on pr6sente des m&hodes pr6dtsant l'ass6chement pour les conditions smvantes: - transltolre de d6blt sans chute de barres - 6bullition en convectmn naturelle - bouchage total en pied d'un assemblage. Des r6sultats provenant d'essms SCARABEE en pde sont &scut6s Une llste de r6f6rences pertmentes est fournle

1. T h e r m a l - h y d r a u l i c operating

studies

of FBR

cores

during

normal

conditions

1.1. Introductton In the past decade m a n y analytical and experimental efforts have been undertaken m order to achieve optim u m design p e r f o r m a n c e for each sub-assembly ( S / A ) a n d for the core as a whole. Design-oriented studies 0029-5493/90/$03.50

were carried out u n d e r the leadership of C E A / D R P in out-of-pile faciliues at C E N Cadarache, C E N G r e n o b l e and in different national university units. C o m p u t a tional codes were developed and improved on the basis of these detailed results. In addition, i m p o r t a n t r e f o r m a t i o n d r a w n b o t h from operating experience and from in-pile-experiments performed on the R A P S O D I E a n d P H E N I X p r o t o t y p e fast reactors, along with the results o f the available

© 1990 - E l s e v i e r S c i e n c e P u b l i s h e r s B.V. ( N o r t h - H o l l a n d )

404

J M Setler et al / French LMFBR (ore thermo-hydrauhcs

post-irradiation examinations of the fuel S / A ' s confirmed the design options for the core and the S / A of the SUPER-PHENIX fast breeder reactor. The objectives of these studies were largely devoted to the investigation of S / A behawour and, more precisely, to the defimtion of the coolant conditions m the fuel pin bundle under normal operating conditions throughout S / A life, with the aim of ensuring the structural integrity of the fuel pm cladding Additional objectives of these studies were concerned with the thermal enwronmental configuration of each S / A pm bundle under real reactor operating condltions, providing typxcal thermal-hydrauhc processes to be compared vath the predicuons of computer codes. The following sections summartze the present status of CEA analysis capabihty to describe the thermal-hydraulic behavlour of the coolant m the core components of fast breeder reactors. Particular emphasis is given to the following topics - the THESEE code and its quahficat~on the F A I D E expertmental programme the determination of the mterstitml flow between subassembhes.

the temperature d~stribut~on of all the pros over the core under any defined operating condmons for the reactor. The computations are based on the use of globally hneanzed parameters derived from adequate destgn code computations. Throughout tins three-step process, winch finally g~ves the clad temperature distribution over the core, attention was focused first on the lndwldual vahdatlon of each code, and then on the calibration of the various models connecting them Winle developing the codes, numerous experiments deahng with different bundle geometries, under various flow conditxons, were performed m the frame of a broad experimental programme. In the light of these out-of-pale experimental results and the in-pile measurements cormng from the commissioning tests of SUPER-PHENIX, the calculations proved to be qmte adequate for normal flow condxtlons, and for low flow conditions up to the core refuelhng operating conditions.

1 2 P m bundle thermal-hydrauhc codes

During the past decade, the development of the THESEE code, a three-dimensional distributed parameter code based on a fimte element method, was achieved. Recognizang that the flow has a strongly preferential direction and does not present any reorculatlon, the advectlon effects are found to be dominant. It is thus justified to neglect the diffusion effects in tins direction, and to turn the basis equations into a parabolic system for tins dxrection. Numerically the longatudmal direcuon is processed m a specific way using a s~mple step-by-step integration scheme. The savings thus obtained allow for the computation of the two other d~rectxons m a much more detailed way. In the THESEE code, a fimte element dlscretizatlon, allowing a refmed geometrical description of the bundle sub-channels, was chosen. Associated with a lumping techiuque for the matrix resolutions, this code could afford refined computations of sub-assemblies including up to 91 pins, using in each sub-channel a 25 node d~scret~zauon. Imtially, the mare effort was focused on an accurate representation of the convective effects induced by the helical wire-wrap, setting aside the problems related to the turbulence modelling. At the first stages of development, a very simple model using a given eddy vascos~ty associated with a non-slip condition on the wails was implemented.

-

-

The genenc studies are based on the development and venficatlon of a set of codes designed for the computauon of the temperature field m a fuel bundle These codes can be classified into three different levels, according to an increasing degree of engineering ablhty generally assocmted with a decreasing level m model refinement. 1 2.1. Reference codes These "refined modelling codes" analyse in detail the turbulent convective mixing effects reside and between the sub-channels. Consequently, they can only deal with bundles containing a small number of pins. 1 2 2 Destgn codes

These "averaged parameters codes" deal with a realisuc bundle size but are based on a global descripaon of turbulent rmxlng processes maanly between sub-channels. This global model is calibrated using both results from the reference codes and global experimental resuits. 1.2 3. Management codes These codes constitute the thermal-hydrauhc part of the plant codes system and allow direct computations of

1 3 The T H E S E E reference code (Rouzaud et al, 1981)

J M Seller et a L / French L M F B R core thermo-hydrauhcs

Subsequently, It was reahzed that the code had to be updated to take into account a more refined description of turbulence. The mixing length technique appears to be particularly well matched with the flow nature which, in spite of the effects due to the hehcal wire-wrap, is very smnlar to the usual channel flows. In conjunction with this model, wall functions were also introduced. It is worth noting that this kind of code provides very detailed information on local data which cannot be obtained experimentally with sufficient accuracy. However it is the only way to define the input information needed by the design codes. Looking now at end-of-life conditions, experimental results on the thermal-hydrauhcs of S / A with distorted geometries are needed in order both to qualify design code modelling and to confirm the local hot spots assessments. 1.4 D i s t o r t e d g e o m e t r y analysts - F A I D E e x p e r t m e n t s

The need for accurate evaluation of the temperature distribution inside a fuel bundle with possible changes from initial nominal geometry, has led CEA, m cooperation with its European partners, to imtiate the FAIDE experimental programme at CADARACHE. The first part of this programme, was onented towards fluid velocity measurements inside a 7 pros bundle, with water bemg used as the fluid. Compared to SUPER-PHENIX 1, the geometrical scale is equal to 2.35. The pins and the walls were made of perpex. The triangular pitch (23 mm) and the diameter of the pins (20 mm) did not allow measurement of the velocities from the outside of the bundles. A match of refraction index between the fluid and the perpex could not be aclueved, consequently ~t was necessary to develop a special optical probe which was introduced inside the pins (see Fig. 1). The light of a H e - N e laser is transmitted along 3 m by a fiber up to the head of the probe. There, the beam is split into two parts focused by a lens and directed by a mirror. The lens can be shifted with respect to the mirror, with the aim of moving the measuring volume anywhere in the flow and of rotating the plane of the beams. This allows either the axaal or the transverse component of velocmes to be measured. The backscattered light ff caught by the headprobe and transmitted by a fiber to a photomultlplier. The geometry distortion was achieved by decreasing the distance between the walls of the hexagonal tube, which originated a stress on the bundle, resulting in swelling and twisting of the pins. Knowledge of the distorted geometry was obtained through calculations

LASERHe-Na

405

~

~~@ fiber

photomulllpher

"--~--~'l~m°v'ng i~upper

part of the probe IJ dev,ce . --[~ --j-- beamsphtter

I!t, I

I

_-S"o,,,o0,o0o

IX'measuringvolume Fig. 1. FAIDE. the laser probe.

using a mechanical finite element code The exact location of the centers of the pins and of the measuring volume were derived from these calculations. This approach was validated by some measurements. The salient result of these investigaUons is that bundle swelling and twisting do not significantly affect the geometry and the velocity field in the central channels. This is not the case for the peripheral channels where the wire is not present (channel between pins 2 and 3 and the hexagonal tube): the flow cross section and the velocities in these channels are reduced. Figure 2 reproduces the transverse flowrate profiles at the gap between pins 2 and 3 for the nominal and distorted geometry. The shapes of the profiles are shghtly different but the total flowrate exchanged between 2 parallel channels along a pitch is nearly the same. The local velocities were integrated m order to derive the flowrate through the channels. This work was performed for vahdation purposes. In Fig. 3, a comparison is given between the axaal flowrates measured in the peripheral channels and those calculated by the THESEE code for a nomanal geometry case, showing rather good agreement between calculations and measurements. However, the numerical dlscretization in the axial direction is obviously too large to accurately predact the maximum and minimum values. Reduction of the axaal step of calculation will be undertaken for the future comparisons. This figure also shows the difference between the profiles corresponding to the two geometries. The axial flowrate is much lower for the

406

J M Seder et al / French L M F B R core thermo-hydrauhcs 15 14, 13, 12 11" 10.

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15

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3o ~'o 9"0 1~0 1~o 1~o 2102,~02~o3oo 3~o 36o

DISTORTED GEOMETRY ~ ( ~ ) ' ~ / NOMINAL GEOMETRY

Flg.

2

Distorted

geometry Transverse pros 2 - 3 .

flowrate

Tins code presents usual numerical features; Le. It IS based on a finite volume dlscretization using a staggered mesh associated with an explicit transient algorithm carrying out, within each step, a velocity splitting on the pressure unknown which controls directly the mass conservatlon. The turbulence models implemented in ttus code comprise the conventional k - c model for large cavity flows descriptions, but its main features are based on its specific models intended to describe flows through vanous porous media, such as hexagonal and cyhndrical rod bundles. For reactor designs with low level deflectors above the core exit plane, It was found out that a small part of the hot core outlet flow will be deflected down into the gaps between the S / A wrappers. Cold flow also enters the gaps between the wrappers at the diagnd level These two flows may lead to design problems such as uncertainties m the wrapper temperature, and poss~bihtles of thermal striping on the wrapper tube. Through the mechanical behasaour of the whole S / A network, these temperatures may also influence the core reactivity.

between 400

peripheral channels In the distorted geometry than in the nominal one. This first phase of the F A I D E experimental programme, conducted on a 7 pins bundle, wdl be followed by tests on a 37 pins bundle, together with interpretation work for code validation.

<

1 5. lnterstmal flow between subassemblies

2u_

In addition to the S / A bundle thermal-hydrauhc evaluations, which constitute the essential part of the development efforts in core design studies undertaken by CEA during the past decade, other studies were also focused on the determination of the hydraulic and thermal envtronmental conditions of the S / A . Among these various studies, a relevant problem for the thermal conditions of the S / A , and more particularly of its wrapper, is the interstitial flow outside the S / A . This flow, which occurs all over the core region, results from various hydraulic boundary conditions, such as the connection with the hot plenum. It also results from buoyancy effects due to thermal exchanges through the wrapper with the internal flow of each S / A . The first computations used for the investigation of this flow were done with the D I O G E N E code.

380

C ~°E

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360

340.

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320

300

ANGLE (DEGRES) 28O 180 210 240 270 300 330 360 390 420 ,~0 480 510 540

. . . . . . . . DISTOR - -

OMETRY

NOMINAL GEOMETRY

+ THESEE (nominal geometry)

Fig. 3. A x i a l f l o w r a t e m p e r i p h e r a l c h a n n e l s (total f l o w r a t e = 3 l/s)

J.M. Seder et a l / French LMFBR core thermo-hydrauhcs

All these aspects will be analyzed m the framework of a joint European programme initiated under the European collaboratmn agreement on FBR R& D. This programme includes large experimental facilities m the United Kingdom (HIPPO at NRL RISLEY) and in Italy (VENUS at BRASIMONE). The HIPPO model is of a quarter segment of the core and the primary pool of a FBR with the details based on the SUPER-PHENIX 2 design. The model is at half scale of the reactor and uses water to simulate the sodmm coolant. To date, the model is operational and the assocmted experimental programme has produced some early resuits, vahdating the boundary conditmns to be inserted into the codes. Subsequently velocity measurements in the gaps between the wrappers in the core region will be performed using both usual techniques and a specifically designed Laser Doppler Anemometer. The VENUS model, still under constructmn, will be a stxty degree segment of a smaller core including a total of a hundred S/A's. The test section was designed to allow analytical studies such as the investigation of the effect of the similitude parameters, or the hydrauhc impact of the various geometrical designs of the reactor internals.

2. Local

blockages

in subassemblies

The objectwe of the Design Basis Accident Studies is to demonstrate that, taking into account the detection systems, possible defects are not likely to lead to subassembly melting. The studies related to local blockages in a wire wrapped bundle have been separated into two parts: (1) Determination of maximum credible defect (MCD) (2) Determination of maximum allowable defects (MAD), taking into account service limits related to the fuel and cladding. The demonstration to perform ts:

407

Studies of particle transport in the main vessel - Formation of blockages near flow orifices, or w~thin the bundle. These analyses are supported by out-of-pile experiments m the framework of the ABACUS programme (Flonm et al., 1986). Tests have been performed in water and in air using various stzed particles. The conclusion of this programme 1s that important planar blockages are unreahstic in bundles with a wire spacing device. The blockage of two adjacent subchannels is only possible m the peripheral region, along the hexagonal wrapper. The FORBIN computer code describes plugging reception by particles m wire-spaced pin bundles. -

2.2. R & D concermng M A D ( M a m m u m Allowable Defect)

The first objective was to investigate the thermo-hydrauhc consequences of internal blockages. Therefore EDF performed the SCARLET experiments. 2.2.1 The S C A R L E T experiments (Jolas et al., 1984; Huber and Peppier, 1988) The tests were performed on a sodium loop usmg electrically heated 19 or 37 pin bundles, with geometrical and heat flux characteristics close to SUPERPHENIX nolmnal conditions. 2 2 1.1 S C A R L E T - 2 tests In this expertment, the six central subchannels were blocked (Fig. 4). The blockage was 60 mm high and was located in the middle of the

Blockage

MCD < MAD. 2.1. R & D concerning M C D (Maxtmum Credtble Defect)

The maximum credible defect results from the analysis of: - Defects cause inventory - Probabilistic studies of inventoried defects or applications of line of defense method (Le Rigoleur et al., 1985)

W=re spacer

Heaterpm/ " Fig 4. SCARLET-2 cross section

408

J M Setler et a l / French LMFBR core thermo-hydrauhcs

~ant sodtum

Blockage

Annular s0cfium flow Fig 5 SCARLET-3 cross section

heated length. It was made of smtered titamum spherical particles and its porosity was 32% At sodium flowrates and powers close to S U P E R P H E N I X conditions, the maximum temperature elevation was 300 ° C. The temperature peak was located near the downstream face of the blockage. 2 2 1.2 S C A R L E T - 3 tests The SCARLET-3 subassembly was a 37 electrically heated pm bundle cooled by an external flow of sodium. The blockage was located along the hexagonal wrapper (Fig. 5). Like the SCARLET-2 blockage, it was 60 m m high and its porosity was close to 32%. At nominal condinons, the temperature elevation was equal to 550 ° C 2 2 1.3 S C A R L E T - R tests Single and two-phase natural circulation tests at low power were also performed on a test section sirmlar to SCARLET-3 with a total inlet blockage and no internal blockage. 2.2.2. Improvement o f analytical tools for local blockages The C A F C A codes (EdF, Olive et al., 1984) and the A N D I N code (CEA), have been developed and used m order to predict the temperature distributions m porous blockages. Both codes have been improved by compan-

son w~th the S C A R L E T experimental results (Figs 6 and 7). 2 3. Extrapolanon to reactor condtttons

In the reactor case, blockages are smaller than those tested m the S C A R L E T experiments, and sodium does not reach saturanon temperature. Hence, there ~s no clad or fuel melting for ttus design basis accident

3. Subassembly thermo-hydraulics related to sodium boiling During acctdental sequences, temperatures m the subassembhes can reach saturanon conditions. Sodium bolhng and dry-out are the lmtlating events of a severe core degradanon and, as a consequence, have to be well understood. Basts studtes (B. Rameau and J.M. Seller, 1983, J.M. Seller, 1984). Out-of-pile experiments performed on the C F N a loop at C E A - G R E N O B L E have shown that sodmm boihng develops m 3 penods (see Fig. 8). (1) Hot spot boiling: Stable boiling bubbles appear at hot locations (behind spacers for instance). This type of bolhng only occurs at high heat flux when lntra-subchannel temperature differences are slgmficant. It may

J.M Seder et aL / French LMFBR core thermo-hydrauhcs

lead to early dry-out but does not affect the mean sodium flow. (2) Local boding: is defined when stable boiling spreads over several subchannels but not over the whole cross section of the bundle. Dry out onset during local boiling has been investigated by Bergeonneau and Rameau (1986). (3) Generahzed bothng or "'extended" bolhng: is defreed when stable boiling spreads over the whole bundle cross section. Dry out is likely to occur. Different features related to boiling have been extensively stuched in the last 20 years. (a) Effect of boihng on the flow stability (see Fig. 9) - Steady-state } (Costa, 1977) Slow transients (LOF) Natural convection (Rameau et al., 1984) Fast transients (TOP or fast LOF) (Seller, 1984) -

-

409

The stability analysis for steady-state, slow transients and natural convection boiling in a subassembly rehes mainly on the L E D I N E G G stability criterion. This criterion is expressed for a stable flow by: d APext/d Q < d APmt/d Q,

where: Q is the mass flow rate through the boihng subassembly, APmt(Q ) Is the pressure drop in the boiling subassembly as a function of Q (APmt is called the internal characteristic), APext(Q) is the pressure drop which is imposed by the external circuit when the flow through the subassembly is varied (APex t is called the external characteristic). Figure 9 presents a picture of different situations which may be encountered for LOF or natural convection boihng analyses. (b) Pressure drops related to a sodium boiling flow (Grand, 1978)

temperature axial profile tn the blocked sub-channels •

temperature measurements m the blockage and ~n the spacer w~res ANDIN code results

~_ ,~

CAFCA code results

700 0 -(°C)

650 0

600 o

550 o

500 o

450 0

400 o

ee 350 ( 000

Blockage ,

002

004

006

, 008

010

, 012

0 14(m)

F~g. 6. SCARLET-2 test results and vahdatlon of CAFCA and ANDIN codes.

410

J M Setler et al / French LMFBR core thermo-hydrauhcs

(c) Dry-out occurence - Local boiling (Bergeonneau and Rameau, 1986) Generalized boiling (Rameau and Seiler, 1983) Dry out correlation. The following correlation has been vahdated for forced and natural convection sodium boiling: -

q'oo = 0-42FvTGZ~'(1 - X o o ) D h / / L , where F v = 9 4 / R e v, for laminar flow; F v = 0.316 Re~-°15, for turbulent flow; D h is the hydraulic diameter; Fv is the friction factor; Y is the slip ratio (Lockhart-Martinelh); G is the mass flux; So is the latent heat of vaporization, Re v = G D h / p G q~oo iS the critical heat flux; Xoo is the dry out quahty; L is the heating length. 3 1 Codes Codes for safety analysts (J Papin et al., 1982). A detailed analysis of the phenomena is given by the three-dimensional code T H E B E S (Camous, 1981) de-

rived from the single-phase subchannel code SABRE-2A (UKAEA). The T H E B E S boiling model is based upon an homogeneous two-phase flow model with three equations and in which the void fraction is defined with a " s m o o t h i n g " factor to allow for nonequihbrlum between phases. The two-phase pressure drop is given by a Lockh a r t - M a r t m e l l i formulation with a constant slip ratio of 15 and the clad dry-out is described using the thermal flux correlation established from out of pile experiments. On the other hand, a simpler model, M U L T I C A N A , has been elaborated for use in general accident codes such as P H Y S U R A and P H Y S U R A - G R A P P E , after successfull comparison with the reference code T H E B E S and vahdatlon against experiments. The description is only two-dxmenslonal and is based upon an homogeneous two-phase flow model with thermodynamac equilibrium, in which sodmm compresslblhty is neglected and radial pressure assumed uniform. The void fraction,

temperature axml profile =n the pros located at the center of the blocage •

temperature measurements under the clad ANDIN code results

A A

C A F C A code results for"fuel" temperature

= =

C A F C A code results for "clad" temperature

1000 0 r" (°C)

900 0

800 0

700 0

600 0

500 0

400 0 ~ 300

-

~ ,

O0

0 01

Blockage

~//"//'(/"/////,'f/'7"/'~,/////,f////./'//~' 0 02 0 03 0 04 0 05 0 06

0 07

0 08 (m)

Fig 7 SCARLET-3 test results and valldauon of CAFCA and ANDIN codes

J.M Seder et a L / French LMFBR core thermo-hydrauhcs

{ appropriate - - f e w

Calculahonal { model models ~ o! spacers

I single channel

I saturated

subcooled located boding

flow-pattern { hot-spotboflmg risk of { burn-out by-pass --~

mterconnected channels

////i/// .///i//

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boihng

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Various boding experiments have been performed m out-of-pile facihtles in CEN Grenoble (CFNa and CESAR loops). In-pile experiments are performed m the CABRI and SCARABEE facilities (a hght water pool reactor located in the Cadarache Nuclear Center). Various accidental conditions have been investigated m mneteen and thirty-seven pros bundles with fresh fuel (Moxon et al., 1986)

m

heating I I I

length ~ j 12

14

101

0 99

Ms,~ flow at saturation

Fig. 8 Stages of bolhng m a 19-pin bundle dunng slow-flow transients. two-phase friction pressure drop and dry-out correlations are identical to those of THEBES (Papln and Obry, 1985; Mehs et al., 1985; Meyer-Heine et at., 1986). - -

Pressure drop

internal charactenshc

= = -- : : external characteristic

(AP) "x N

\\,OF

working point ts \ ~ / unstable\ " ~ ~ w o r k m g

~

stat,cpressure head

~

point is stable

hgher powerlevel

-----=.~.~"-'" "i~ 7 -l e p-h -x? s - e-' ~-"/ - ; ~

Project codes. NATREX and M A N D R I N (Grand et al., 1980) are single-channel codes. They are based on a homogeneous flow model with thermodynarmc nonequlllbnum. BACCHUS (Basque et al., 1984) is a 2D bundle boiling code based on a porous medium approach. Homogeneous flow is considered including thermodynarnlc nonequilibnum. SURFASS (Anzleu et al., 1986) ~s used for the calculations of total instantaneous inlet blockage accidents at nominal power level. 3 2 Experiments

0 tW?e; hase I I

h = (M~)

411

~

low power level

I

v.ery Io~t power level " ledlnegg instability ] is not possible natural convection (Q) IP

test sect=on mass flow rate Fig. 9. Qualitative diagram for internal characteristics and related mstabIhty posslblhtles

3.3. Investtgatwns of boding and dry out for different kmds of reactor accidents 3.3 1 Loss of coolant flow (LOF) For the French SUPER-PHENIX 1 reactor a loss of flow accident without scram has been considered. As the pumps have great mechanical inertia, the loss of flow is very slow and bolhng conditions are reached about 10 min after the onset of the transient, when the power is between 1/4 and 1/3 of the nominal power level According to the power level in the subassembly, onset of boding can lead either to stable natural circulation boiling or to flow excursion. For low powers The boiling zone extends very little into the fissile zone, and entirely over the upper breeder zone of the pros and the upper neutron shielding at the top of the subassembly. Due to the length of the upper neutron shielding (about 1 m) the spreading of the boihng zone induces very strong buoyancy effects: stable natural circulation boiling with low quality is calculated for power levels up to 10 kW per pin ( > 1 / 4 nominal power level). There should be no dry-out for these boihng conditions but rather a large reactor heatup (Rameau et al., 1984). For high powers. A flow excursion develops (i.e., a LEDINEGG instability), the subassembly flow is decreased to zero in about 15 s (Costa, 1977). This leads to dry-out and consequent material melting. The later events are investigated in the SCARABEE APL experiments. Figure 10 presents the variation of the inlet flow

412

J M Seder et al / French L M F B R core thermo-hydrauhcs

measured m S C A R A B E E and the recalculation with the code P H Y S U R A grappe (model M U L T I C A N A ) . One can dtstlngutsh: the total boiling period with an almost constant mean mass flow rate due to progression of vaporization in the central zone and mternal flow redistribution (central boding zone acting as a blockage). - the generalized boiling spreading up to the periphery w~th inlet flow excursion leading rapidly to clad dry-out (less than 10 s). The interpretaUon of such experiments with the codes underhned the importance of the thermal inertia of the surrounding structure on boihng expansion which lasts several seconds (10-30 s), the importance of rachal heat transfer due to wire-wrap reduced turbulent exchanges. Dry-out at the end of flow reductton occurs with a vapour quahty close to 30%. On the other hand, m s p t t e of flow reduction, sodmm vapour v e l o o t y may be lugh enough to entrain molten clad. -

3 3 2 Transtent overpower (TOP) These accidents are charactertzed by a very rapid reacUwty xnserUon leading to fast temperature ramps (200 ° C / s to 1000 ° C / s ) m the coolant. They have been studied m out-of-pile (Seiler, 1977, 1979) and in-pde (CABRI) experiments (Kiissmaul et al., 1986). Of course, extstmg boihng codes calculate the sodium voiding very well, but it is worthwhile to m e n u o n that for such fast transients the sodium voiding can be calculated very s~mply with a single-phase code by defining the two-phase zone as the r e , o n where the

temperature of the hquld is greater than the saturaUon temperature. It should be noticed that the hydrodynamac effects have very httle real importance on the subassembly voiding which is mamly controlled by the rate of thermal heat-up. This conclusion is also vahd for fast loss of flow transients (Seller, 1980) 3 3 3 Inlet blockages Two types of accidents are considered: - the instantaneous inlet blockage at nominal power the inlet blockage at reactor start-up. Instantaneous mlet blockage at n o m m a l power. In case of a total inlet blockage of a subassembly at normnal power the thermal-hydrauhcs interpretation of the S C A R A B E E BE + tests showed that bolhng progression is mainly due to conductwe heat transfer and external cooling with a neglxgtble effect of natural convectlon The influence of outlet flow instablht~es due to the location of the upper h q m d front near the top of the fissile zone has been pointed out: the resulting axial heat transfer towards upper cold zones affects dry-out onset inside the heating zone. The interpretation of those experiments with the M U T I C A N A model (Moxon, 1986) confirmed its vahdlty in those c o n d m o n s (Fig. 11) * -

3 3 4. Inlet blockage at reactor start-up The prediction of subassembly melting for this kind of accidents is much more difficult. The blocked subas-

* The SURFASS code also descnbes these events

Q (m3/h) 7 6 5

E2

4 3 expenmental,nletf l o w

~

E

3

------ PHYSURAGRAPPEcalculabon

2

flow excursion

1 '

4

'

2b

3b

~o

\ WrW,

Fig 10 SCARABEE APL test - Experimental inlet mass flow rate

t(s)

413

J M. Seder et al / French LMFBR core thermo-hydrauhcs SODIUM BOILING FRONTS

Z(mm) BE+I +300 +200

/

+100

{~

0 \*\

\

central channel mtermedtate channel EXP penpheral channel

-- PHYSURA-GRAPPE

-100 -200

\.,,:

oo

_ _

_

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i

-300 Time(s) .

.

.

.

;

I

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FIg 11 Instantaneous inlet blockage at nominal power. SCARABEEexperiment and recalculauon with the code PHYSURAGRAPPE

Neutron shteldmg

Top o| bundle

End of heated length

Slmulat=on of inter-subasser flow

¢/) Fig. 12 Inlet blockage at reactor start-up. Flow configuration as studmd In the SCARABEE BE and ECONA experiments.

sembly can exchange heat either with the surrounding inter-subassembly flow or with the upper hot plenum. The flow m the bundle is controlled by natural convection effects (see Fig. 12). The aim of the out-of-pile ECONA experiments ~s to investigate dry-out occurrence m a totally inlet blocked 37-pin subassembly. Previous experiments (FETUNA 91) were performed m order to investigate sodium thermal-hydraulics under natural and mixed convection in a bundle (Menant et at., 1984). The consequence of a total inlet blockage at reactor start-up has also been studied with the SCARABEE BE test involving natural convection in two-phase flow. Validity of the THEBES model has been confirmed m flus field except when boiling reaches the penphery of the bundle: in this case, the homogeneous boiling model is no longer valid and should allow opposite signs for the two-phase velocities. This later problem has been analysed in (Rameau and N. Bourabaa, 1988). This study has also shown that in such low power conditions, dry-out onset occurs with a higher quahty than in the previous accidental cases. From THEBES calculations it could then be deduced that for a totally blocked ferule subassembly of the SUPER-PHENIX reactor, power should be lower than 550 kW to be sure to avoid clad dry-out.

414

J.M Setler et al / French LMFBR core thermo-hydrauhcs

3 3 5 Ftsston gas release Until now only minor investigations have been made m France about the thermo-hydraullc effects of fission g a s release in a s u b a s s e m b l y . W o r k is g o m g o n in this field. 3 3 6 Inter-subassembly sodtum bothng In the framework of the analysis of the propagation of a n a c c i d e n t i n v o l w n g t h e m e l t i n g o f a s u b a s s e m b l y o n e h a s to i n v e s t i g a t e t h e p o s s l b d i t y o f s o d i u m b o d i n g in t h e r a t e r - s u b a s s e m b l y g a p s T h e d r y - o u t o c c u r r e n c e on the surface of the hexcan of the damaged subassembly is o f special interest. W o r k is g o i n g o n m this field 3.3 7 Control rod wtthdrawal An madvertant control rod withdrawal resultmg from a n o p e r a t o r e r r o r is likely to l e a d to a n i n t e r n a l b l o c k a g e a n d in s o m e c a s e s to s u b a s s e m b l y m e l t m g l m t t a t e d b y c l a d r u p t u r e w i t h m o l t e n fuel ejection T h i s q u e s t i o n is c u r r e n t l y b e i n g s t u d i e d , m p a r t i c u l a r w i t h t h e S U R F A S S c o d e ( A n z a e u et al., 1986).

References [1] P. Anzleu, J.M Le Gouez, C Pourcheresse, S Carller and G Fdlppl, SURFASS a computer code describing the consequences of total instantaneous blockage of a subassembly on the R N R 1500", Proc lnt Conf on Scwnce and Technology of Fast Reactor Safety, Vol 1, Guernesey 12-16 May 1986, pp 447-450 [2] G Basque, L Delaplerre, D G r a n d and P. Mercier, " B A C C H U S a numerical approach of two phase flow in a rod bundle", Nucl Engrg Des 82, 191-204 (1984) [3] P. Bergeonneau and B. Rameau, " C o n & t l o n s for locally stable boding m a p m bundle", Proc Int Conf on Scwnce and Technology of Fast Reactor Safe(v, Vol 2, Guernesey 12-16 May 1986, pp. 445-450 [4] D Blanc and Ph. Rouzaud, "Analyses des r6sultats d'essats r6ahs6s sur la maquette F E T U N A GR91 h l'a~de des codes THESEE et C A D E T " , 1AHR Second lnt Spectahst Meeting on "'Thermal-hydrauhcs m LMFBR Rod Bundles" Roma (1982). [5] F Camous, "Transient boding calculations m rod bundles using the THEBES code", ANS Wmter Meetmg, San Francisco, pp 1017, 1018, 1981, Vol. 39 [6] F Camous, "THEBES: a thermal-hydrauhc code for the calculation of transwnt two phase flow m bundle geometry", Proc Lzqutd Metal Bothng Working Group, October 1982 - Karlsruhe 197, 212 - Avadable from C C R ISPRA [7] J Costa, " C o n t n b u u o n to the study of sodium bolhng during. Slow p u m p coastdown m L M F B R subassembhes", 98 Winter Annual Meeting ASME Atlanta, 27 Novemb e r - 2 December 1977, Vol. 2, pp 155-169.

[8] G L Flonm, R Coudray, G Raffadhac, D Sardaln and B Valentm, "'Research and development supporting the fourth category subassembly accident analysis In the R N R 1500", Conf on Soence and Technology of Fast Reactor Safe(v, BNES London, 1986 [9] D Grand, "Pressure drops in rod bundles", Von K a r m a n Institute for Fluid Dynarmcs, Lecture Series 1978 5 Two-phase flows m Nuclear Reactors, April 17 21, 1978 [10] D Grand, P Mercier and J.M Seller, "Status of the N A T R E X Code", 9th LMBWG Meeting, Rome 4 - 6 June 1980 Pubhshed by C C R Ispra. [11] F Huber and W. Peppier, "Subassembly blockage studies at EDF, PNC, U K A E A and K F K " , Liquid Metal Boding Working Group, 13th Meeting, W m f n t h , September 2729, 1988. [12] P Jolas, C Simeon and J Ohve, "SCARLET-2. experiment on a porous and long central blockage In a 19 wire-wrap p m bundle", Ltqutd Metal Bolhng Working Group, l l t h Meeting, Grenoble, October 23-26, 1984 [13] G KussmauL W Vzith, J Wolff, J Da&llon, M. Haessler and F Sabathler, "'The CABRI project - overall status and acluevements", Proc Int Conf on Scwnce and Technology of Fast Reactor Safety, Guernesey 12-16 May 1986, pp. 103-108 [14] C Le Ragoleur, P Anzleu, G L Flonm, J Moreau and M Boschlero, "Subassembly accidents m the ' R a p l d e 1500' associated programme of R and D", Knoxvdle, April 1985 [15] J C Mehs, R Fablanelh, M Cranga, J Papm and A Meyer-Heine, " P H Y S U R A a code for the interpretation of safety experiments", L M F B R Safety Meeting, Lyon, 1982 [16] B Menant and B Rameau, " S o d m m thermal hydrauhcs under natural and maxed convection bundle geometry", ASME 8 4 - W A / H T - 3 . [17] A Meyer-Heine et al, "'The SCARABEE program and the P H Y S U R A - G R A P P E code a phenomenologlcal approach for subassembly accidents", Conf on Science and Technology of Fast Reactor Safe(v, Guernesey, 1986 [18] D Moxon, J Papm, P. Obry and P Soussan, " S C A R A BEE an interpretation of the p u m p trip and inlet blockage series", Conf on Scwnce and Technology of Fast Reactor Safety, Guernesey, 1986 [19] J Ohve, P. Jolas and S Aubry, "'Description of the multl-&menslonal thermohydrauhc code C A F C A - N a for the analysis of L M F B R fuel assembhes under forced or natural circulation m case of partml or total flow blockages", 11th Ltqutd Metal Bothng Working Group, Grenoble October 23-26, 1984 Avadable from JRC Ispra [20] J Ohve, S Aubry and P M Raboud, " N a t u r a l cxrculatlon of sodium m a 37 electrically heated pin bundle results of the S C A R L E T - R experiments and apphcatlon to decay heat removal in L M F B R fuel a s s e m b h e s ' , A S M E Winter Meeting, Boston, December 13-18, 1987 [21] J Papm, F. Camous, J.M Legouez, Ph Berna and M Fortunato, " U n d e r c o o h n g accidents m a L M F B R ' s subassembly codes and vahdatlon L M F B R " , Safety Meetmg, Lyon, 1982

J M Seder et al / French LMFBR core thermo-hydrauhcs [22] J. Papin and P Obry, "Thermal-hydrauhc studies of undercoolmg accidents in L M F B R safety analysis: codes and vahdatlon", Topwal Meetmg on Reactor Thermal-hydrauhcs, Newport, Rhode Island - USA, 1985 [23] B. R a m e a u and J.M. Seller, "Studies on sodium boding p h e n o m e n a m out of pile rod bundles for various accidental situations in L M F B R " , Course on "thermohydraulics of Llqmd Metals", 30 M a y - 3 June 1983 Von K a r m a n Institute for Fltud Dynarmcs. [24] B. Rameau, J M Seder and K W Lee, " L o w heat flux sodium bolhng - bundle expenmental results and analysis Apphcation to natural convection bolhng m reactor conditions", A SME Winter Annual Meetmg, New Orleans, December 1984. [25] B. R a m e a u and N. Bourabaa, "Sodium boihng natural convection tests m a 37 pin bundle. E C O N A experimental results and analysis", 13th Llqutd Metal Bolhng Workmg Group, September 27-29, W m f n t h 1988. Available from JRC Ispra.

415

[26] Ph. Rouzaud, J Ctunardet, B G a y and R. Verblest, " T h e r m a l hydrauhcs calculations of wtre wrapped bundles using a fimte element method. THESEE Code", Numertcal Methods m Lammar and Turbulent Flows Second lnt Conf., Vemce, 1981. [27] J.M Seller, "Sodium votdlng in an annular channel heated by a single pin in overpower accident simulations", 7th Ltqmd Metal Bothng Working Group, Petten June 1977 Pubhshed by C C R ISPRA. [28] J.M Seder, "Sodium voldmg m overpower accidents", 21th Meetmg of the W A C Group, Brussel, 15 February 1979. [29] J M. Seller, "Progress in the fundamental aspects of sodium bolhng and related posslblhtles for slmphfted calculations", I C H M T lnt Semmar, Dubrovnlk, 1980 [30] J M. Seller, "Stu&es on sodium bolhng p h e n o m e n a m out of pile rod bundles for v a n o u s accidental situations m L M F B R experiments and interpretation", Nucl Engrg Des 82, 227-239 (1984)