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Cooling disturbances in the core of sodium-cooled fast reactors as causes of fast failure propagation
NUCLEAR ENGINEERING AND DESIGN 14 (1970) 12-22. NORTH-HOLLAND PUBLISHING COMPANY
COOLING
DISTURBANCES
IN THE CORE OF SODIUM-COOLED
AS CAUSES OF FAST FAILURE
FAST REACTORS
PROPAGATION
K. GAST and D. SMIDT
Institut fiir Reaktorentwicklung, Kernforschungszentrum, Karlsruhe Received 4 June 1970
As hypothetical cause of Bethe-Tait type excursions in sodium-cooled fast reactors local cooling failures are considered. Several mechanisms of failure propagation are discussed. The most important one is the thermal interaction between molten fuel and liquid sodium. The initial conditions starting from local failures or subassembly blockage are defined. A criterion for dryout in narrow channels is developed. The importance of sodium boiling superheat is rediscussed in this context. Finally the functional requirements for core instrumentation are given. It is concluded that the probability of failure propagation leading to Bethe-Tait excursions is sufficiently low.
1. Introduction Large nuclear power plants are designed to a standard o f reliability and safety which is unparalleled in other field of modern technology. The necessity for such high safety standards stems from the large hazard potential of the radioactive materials contained in nuclear reactors. This hazard potential is sometimes looked upon without due consideration of the vanishingly low probability of events that could result in the release of sufficient amounts of radioactive materials to cause radiation injuries to the public. For a hypothetical accident in a 1000 MWe reactor the probability of early casualties becomes significant if more than 0.1% o f the volatile fission products are assumed to be released on ground level. Hence, the maximum possible release after any conceivable accident has to be much below this value. For light water reactors this implies, for instance, that following a postulated fracture of the primary system the emergency cooling system must operate and the containment integrity must be maintained. To prove this, the failure probability must be demonstrated to be sufficiently small and the possibility of causal connections of different failures must be considered. However, if the principle of making conservative assumptions is carried to an extreme, unacceptable
situations can be postulated: One example of such hypothetical accidents in the light water reactor is the catastrophic failure of the pressure vessel, which could cause disintegration o f the containment by pressure effects or by missiles. Also the core can be transferred to an undefined geometry for which effective decay heat removal cannot be guaranteed. For sodium cooled fast reactors hypothetical safety considerations are different. The coolant pressure is so small, that a fracture of the primary system in itself cannot cause a large energy release and subsequent failure of the containment. But, since the core of the fast reactor is not in the state o f its largest reactivity, there exists the principal possibility of a fast nuclear exursion, the consequence of which could affect the integrity o f the containment in an analogous way. Therefore, for this type o f accident the same degree of improbability must be attained as e.g. for the fast pressure vessel burst or the primary system fracture combined with containment failure for the light water reactor. Among the most important causes for fast power excursions of the Bethe-Tait type in sodium cooled fast reactors are the partial or total loss of heat removal capability to the core. It results either from a partial core voiding and a positive void reactivity or from core compaction (melt down) in the absence of
COOLING DISTURBANCES sodium [3]. In the following paragraphs, therefore, local coolant disturbances will be studied for their potential as initiation of major damage to the core. In so far as they have primarily local character one has to look for possible mechanisms of a fast failure propagation. Boiling and recondensation shocks, and, most important, the interaction of molten fuel with sodium will be treated as the main modes for failure propagation.
2. Classification of cooling failures 2.1. Cooling disturbances affecting the entire core A cooling disturbance in the total core or in a larger region of it might be generated as a consequence of failure propagation from local effects, by simultaneous loss of flow and failure of the shut-off system or by a loss of coolant. The latter case can be excluded since the primary system has double walls in all lower parts. The second case belongs to the widely discussed type of incidents involving engineered safeguards and their reliability and will not be treated in this context. After a local cooling failure in the core, sodium boiling in the neighbouring subassemblies could occur as consequence of failure propagation. Schlechtendahl [4] has shown that because of the statistical nature of incipient boiling over the subassemblies the reactivity rate by voiding will be relatively low. After signals from different detectors (reactivity, temperature, flow, boiling noise) the reactor is very likely to be shut off before reaching the prompt supercritical condition. However, it has to be ensured that by potential core deformations from mechanical effects ((10) of fig. 2) additional reactivity will not be introduced nor the shut-off rods become ineffective. 2.2. Local cooling disturbances within the bundle Local flow blockages are considered as the most important initiating events, because they are relatively likely to occur, and because they may not be detected by temperature sensors or by flowmeters, before a critical situation such as local or gross sodium boiling is reached. Local cooling deficiencies can be caused by either one or a combination of the following faults: Type A: Blockage of single or several neighbouring cooling subchannels by foreign objects, especially at the spacers.
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Type B: Narrowing of single subchannels by lateral movement, bowing, or swelling of fuel rods. In the general case these disturbances in single channels are more or less statistically independent. In the neighbourhood of a channel, in which the coolant flow is suppressed by movement of a rod, there is another one, where the flow is increased. This type includes especially the effects of thermal bowing according to [1]. Type C: Coolant flow reduction by bursting of cladding tubes. Type D: Reduced heat conductance through the gap between fuel and cladding, caused for instance by a cladding tube inflation from fission gas pressure. Since already a small increase of the gap size results in a large increase of the fuel temperature, this effect cannot be completely excluded. It is counteracted by the radiation induced embrittlement of the cladding material, but in some cases larger strains have been observed, if the short time strength of the material was exceeded at excessive temperatures. The several failure types can cause each other. For instance a local hot spot of tybe B may result in a cladding burst of type C and a subsequent wash-out of fuel particles thereby leading to a disturbance of type A. This again may cause disturbances of type C or D etc. Unless detected, the affected zone may grow in size and eventually reach a critical condition. The growth rate will mainly be determined by the shorttime mechanical properties and fracture mechanics of cladding and grid material at excessive temperatures. The critical condition mentioned above is defined such that it may either lead to local boiling within the bundle, to gross boiling and sodium ejection from the entire subassembly, or to inflation of a sufficiently large number of cladding tubes to allow a considerable amount of fuel to melt because of the increased thermal resistence between fuel and clad. Local boiling can lead to gross boiling and ejection, provided the production rate of sodium vapor is sufficiently large to interfere with the overall flow in the affected subassembly. Fig. 1. shows the hypothetical failure tree connected to local failures. Starting point is always a local cooling disturbance (1) of any of the mentioned types. All branches either end with a reactor shut-down (3) or again in a local cooling disturbance (1) and a new start for a path in the failure tree, or with a propagation event (15).
14
K.GAST and D.SMIDT
(2a) ~
(s) L
{ 2 b ) ~
•
.~" (11) ~ n g j (1) subassernblyoooling ~.~7/" ~ l ~ disturbance Fig. 2 __.~--~__ (12a L detected j : (12b) ~ - ~ t ~ melting typeD[ ~ L .. , I r ~ ~L~,~ shut-off _ ]
~ . - . (5o)~0 SI]
(i)
(3) LShUt-Off J ( 1 3 o ~
(13b)dr~ou~ d
(Sb)F$1
i ;
(14) ~ L~o~.-pu,~,~(5b)
/
Fig. 2
Fig. 1. Failure tree for local cooling disturbances.
If a local cooling disturbance is not detected, coolant boiling (1 I) may result. There exists the principal possibility that boiling or recondensation shocks (13) may affect the fuel bundle, thereby initiating a propagation to a larger local cooling disturbance, or a cooling disturbance of the whole subassembly (15) (see paragraph 2.3). If also the coolant boiling will not be detected, a dry-out (burn-out) of a fuel element region (13b) and following fuel melting may occur. When molten fuel and sodium eventually get in contact, a sudden release of mechanical energy by the fuel-sodium-interaction [9] (FSI) is possible. Subsequent events could be: a larger local cooling disturbance (1), or a cooling disturbance in the subassembly (15). Also one could think of a fuel melting without preceding sodium boiling after a failure of type D (4). Here, too, the FSI with all consequences could be possible. However, it is highly improbable that in this case larger parts of the fuel elements melt and react. For this reason this branch is more hypothetical with respect to the generation of a fast excursion. It can be shown that local coolant flow deficiencies caused by reduction of the flow path area are unlikely to initiate sodium boiling within the bundle, provided the blockage is not complete and the affected ~z,one occupies only a small fraction of the core height. It seems possible, however, that fuel claddings will swell and finally burst from the internal gas pressure, because the strength of the material is strongly reduced as temperature increases. It is concluded, therefore, that rupture of fuel cans and subsequent
fission gas release may be the first indication of local blockages. Therefore, a fast reacting can leak detection system should be incorporated into the early warning system. 2.3. Cooling disturbances in the entire subassembly This disturbance may occur if either a local disturbance has spread over the whole subassembly, or if the coolant inlet at the subassembly bottom is blocked by some foreign object. By the design of the inlet nozzle [3] in general it can be excluded that the object can block the total flow instantaneously. Fig. 2 shows the relevant failure tree. In many ways it it related to fig. 1. The propagation from FSI (8) finally may result in cavities (10) (positive reactivity formation), caused by the energy release. Also cooling disturbances of the whole core with subsequent overall boiling and control rod defects are possible, which diminish the shut-off capability.
2.4. Consequences Under unfavourable assumptions and excluding counteraction by safety devices, cooling disturbances could lead to fast excursions via boiling and void effect. Also a reactivity input is possible as consequence of local fuel-sodium-interaction. On the other hand a core meltdown in the presence of sodium can be excluded, as already Gast and Schlechtendahl [3] have shown. Undoubtedly it has to be proven that all chains of events which could lead to failure propagation and finally to a nuclear excursion are either in themselves sufficiently improbable, or are so in combination with a suitable instrumentation. For these
subassembly cooling •
disturbance
J
boding or cond ~ ' Ppulses [~ j (3) -shut-off
1
(11)
(lo) ~
- ] " (7) dry out
i
(8)E
]
011 Icor,disturE~
(12)[conlro=rodsinop~m~
excursion lexcurslon___~L.~~~xcunr°~°n ~ curs~ shut-off iT detected
shut-off if detected
i
i_ iexc~°_= ~ , "j n, r~curs~
shut-Off if detected (and reoctivity ¢z/aikat~e
Fig. 2. Failure tree for subassembly cooling disturbances.
COOLING DISTURBANCES reasons first the problem of coolant boiling will be treated, which forms the most important initial condition for an eventual failure propagation. Then failure propagation itself is discussed, and finally the possibilities of instrumentation.
3. Boiling events in sodium Most work on sodium boiling in connection with fast reactor safety has been centered on the process of expulsion. This determines the time scale of reactivity input via the sodium void effect. Only recently the consideration of boiling as an early phase of the possible failure propagation from local causes has been added. Therefore, new points of view have to be considered. The work of Peppler, Schlechtendahl and Schultheiss [5] in this edition gives a survey of our present state of knowledge of sodium expulsion. The characteristic feature is the growth of a single bubble fed by evaporation of a liquid film on the wall, which inhibits the formation of svbsequent bubbles at least at superheats above 30°C. Then follows an oscillating increase and decrease of bubble size, caused by heat transfer to the liquid subcooled sodium above the heated channel. For the aspect of failure propagation on the other hand, in accordance with figs. 1 and 2, the pressure pulses during boiling and the fuel melting after dryout of a channel merit some consideration.
3.1. Pressure pulses during sodium boiling Pressure pulses occur during bubble growth from superheated liquid, or during bubble collapse. The former shall be called boiling pulses, the latter con-
densation pulses. The initial pressure of the boiling pulses correspond to the superheat at boiling initiation. If no special measures are provided, such as artificial boiling nuclei like gas bubbles or wall cavities (for instance between fuel rod and spacers) as already mentioned in [6], under reactor conditions superheats up to about 150°C have to be expected. For local pressures of 2 atm, for instance, this corresponds to an internal overpressure in the bubble of about 3.8 atm, which is effective nearly over the total expulsion phase of about 0.15 sec (see [5], fig. 13) (not to be confused with the pressure pulses measured at the unheated
15
upper end of the test section - [5], table 5). If the boiling occurs only locally in the subassembly and if the disturbed region is surrounded by subcooled sodium, the pressure pulses become shorter and act in a smaller region only. In most cases the subassembly box has a hexagonal cross section of about 10 to 11 cm span with a small distance to the neighbouring boxes, supported on the inside by fuel rods and spacers. If a boiling pulse occurs in one subassembly, it cannot be completely excluded that neighbouring subassemblies will be affected. The effects are not expected to be large because of the short duration of the load and the small amount of energy involved. The order of magnitude may be estimated by comparing the wall stresses of a box of circular cross section and similar size. Here the load would amount to about 1 kp/mm 2, which is in the range of material strength. This is at least a certain indication of what would happen in the actual geometry. Condensation pulses in the total subassembly might be generated by the return flow and collision of two liquid columns, or locally by the collapse of vapour bubbles. In the theoretical model of BLOW 2 ([5], fig. 13) a complete return flow is predicted. But this has not been found in experiments. The reason for this difference is the neglect of vapour pressure drop in the theoretical model. In the experiments condensation pulses were observed only when a first gas bubble, which is submerged in the cold zone, is suppressed by a second one which is still under high pressure. This is shown in [5], fig. 20. Until a nearly complete dryout of the wall film has occured, a complete reentry of liquid sodium will be prevented by the vapour formation. But at this time the fuel may already be partially molten, which is the starting condition for the FSI to be discussed later. The condensation pressure peaks are large but narrow. Energies for a subassembly are in the order of 1 KWsec only. Therefore, no considerable deformations are to be expected. To get a better knowledge of the effects on the core in Karlsruhe the BEVUS experiment will be undertaken in the middle of 1970. It consists of an electrically heated subassembly surrounded by 6 unheated ones in a sodium pool. By a fast-acting pressure release valve an arbitrary superheat can be simulated. The subassemblies have actual dimensions.
16
K.GAST and D.SMIDT
pressure reduction system for initiation of sodium boiling
pressure relRf valv,
expansion tank
z current supply cables
test tank
sodium supply line,
\cover g_as supply lines
core section m o c k
safety tank
heated element
test insert
Fig. 3. BEVUS test facility schematic.
Only the linear rating with 40 W/cm is considerably smaller than in the actual case. Fig. 3 shows the overall arrangement. In a later phase, when the initial conditions of the FSI are better known, the facility will allow relevant experiments in bundle geometry for this case.
3.2. The process of dryout of the fuel subassembly The FSI is the more important propagation mechanism. For this first some fuel has to melt, to allow for the necessary intimate mixing. Melting of the fuel is
impossib,e as long as sodium is in contrast with it and cools it by convection or evaporation. This brings up the problem of "burnout", better described as dryout. To get a better understanding of the physical processes we shall analyze the flow o f boiling sodium in a tube heated from the outside. This model, for instance, has been used in the experiments described in [5]. Fig. 4 gives a schematic display of the hypothesized course of events. The liquid sodium enters the heated part of the channel (fig. 4a) with the velocity o i and the tempera-
COOLING DISTURBANCES
la)
the liquid film +cannot be renewed by the liquid plug and the first bubble stays until the liquid wall film begins to dry at some place. After this, cladding and fuel are likely to reach their melting temperatures. The pressure drop is determined by equations (2) to (5).
Ib)
-
(c)
+,'o'+
17
dp -- ~f) -002 - ~ dx + pv do
(2)
(d)
P = P(P)
(3)
rrDq(x-x B) rh -
Ah
(4)
~D 2
rh = vp 4
Fig. 4. Oscillatory boiling in a narrow channel.
ture T i. The heat flux is q. After the saturation temperature is reached at Xs, the first vapour bubble begins to grow at x B driven by an external pressure according to a superheat Tsup - Tsat. In fig. 4b the growing piston type bubble is shown. Sodium vapour is generated from the liquid film remaining on the wall. This, therefore, results in efficient heat removal. In fig. 4c the bubble touches at XA, the upper sodium pool, where rapid condensation takes place. (For pool temperatures of 580°C in [5] condensation fluxes of 500 - 1000 W/cm 2 have been found.) Between x B and x A the vapour flow is subject to a pressure drop ~Pv" If Ap v is smaller than the pressure head Z~gpgenerated by the pumps, the liquid column can partially replace the bubble and at x B a new bubble can be generated. Because of the statistical fluctuation of superheat and the temperature increase in the liquid, x B in reality may vary. In fig. 4d the second bubble displaces the first one while the liquid plug in between renews the film on the heated wall. The orific O in combination with the pumping head inhibits any appreciable reverse flow during the growth phase of the bubble. Therefore, in the framework of this model a sufficient criterion for dryout of the wall film can be formulated. It is
(5)
where m = mass flow; q = heat flux and z ~ = evaporation heat. In eq. (2) it has to be taken into account, that according to experiments [11 ] the flow between x B and x A is not a single phase vapor. It rather is mist flow with a liquid mass fraction in the order of 15%. Accordingly, there is a small liquid flow between O and x B . However, the corresponding pressure drop of the liquid part can be neglected. In fig. 5 the results of the calculation with the criterion ( I ) is compared with results of Peppier [ 11 ].
• Experiments Pept~er [121 pump heocl I,Sat - c,:,l:.uuor, ~;o~-".-.- , ~ .
fordifferent p~m~~
/ / k ( n . m .~. d ~
)
../)..'0h..t ..x [ w , ~ ]
Z~Op < Apv
(1)
If (1) is valid, a second bubble cannot be formed,
0
0
I00
! 200
t 300
i ~.tOO
Fig. 5. Dryout heat flux versus static pressure in the test section parameter : p u m p head.
18
K.GAST and D.SMIDT
Here the dryout heat flux is related to the static pressure in the test section (as described in [5] ). The parameter is the pump pressure head. In the experiments it has been 1.5 atm. The curves represent the theory outlined here. On the right hand side of the respective curve dryout should occur. The points give data of observed dryouts. It has to be noted that the points on the 1.5 atm curve correspond to 5 or 3 measurements, respectively. Only one dryout at the upper left occurred, at a rather low heat flux. Its validity is not quite clear yet. However, it has to be restated, that (1) is a sufficient criterion only, and additional effects may promote an earlier dryout. More experiments are under way. In any case the theory indicates the right order of magnitude. The choice of the friction factor f t u r n s out to be not critical. The dryout conditions discussed here are related in many ways to the well known cases of two-phase flow instability. Applied to the reactor, the calculations show that in long and narrow coolant channels, as in fast reactors, the flow reduction necessary to cause boiling and the nominal flow reduction to cause instability and dryout are not far from each other. The flow regime discussed here and the well known regime of slug flow of classical two-phase flow work have a certain similarity in their appearance. However, there is no relation in their physical natures. While the large bubbles of slug flow are generated by hydrodynamic effects, the single bubbles discussed here are formed by the fast growth of the nucleus from superheated liquid. Their size will be determined by rapid evaporation and condensation fluxes at their surface. In contrast to the classical case, therefore, the phenomenon, which is typical for liquids with superheat, will be called pulsating flow. The thickness of the liquid film remaining on the wall amounts to some tenths of a ram. In [8] some work on its value has been reported. For q = 200 W/cm 2 at 900°C the evaporation rate is 0.7 mm/sec. This means that the film stays for some tenths of a second. Before this time in the pulsating flow the next expulsion must follow if dryout is to be avoided. In the experiments of Schleisiek [7] this is the case. For the same reasons as shown in [5], the single channel results also should be applicable to rod bundles.
3.3. Local boiling While all considerations so far applied to the case of a homogeneous flow reduction in the total subassembly as the starting condition, now the case of a local disturbance within the bundle of one of the types described in par. 2 will be treated. Experimental investigations do not exist so far. Evidently they have to be done in bundle geometry. However, from a qualitative analysis one can derive preliminary conclusions. For instance, if some subchannels are blocked locally by objects at the spacers (type A) behind the blockage there will be a wake. If this occurs in the area of a 7-rod cluster, at the center rod boiling may occur over a length of some cm. The bubble expands to all sides into the subcooled sodium outside of the 7-rod cluster and recontracts very quickly. By this it will transport cold sodium in the direction of the boiling center. Because of the rapid recondensation rate, it probably will collapse completely and then start growing again. Thus dryout at the center rod should be prevented. Even for still larger rod blockages a similar pumping action of the bubble could occur. Comparing the vapour-generating areas with the cross section for the return flow between the rods, the limit for complete refilling should be expected for a cluster in the range of 30 rods. This statement, however, has not yet been experimentally proved. Because of the importance of these events for local dryout, an experimental and theoretical program has been started at Karlsruhe. It includes: a) Experimental determination of the wakes behind blockages in unheated rod clusters with transparent simulation liquids in the one-phase region. From this by theoretical means the temperature distribution of single-phase sodium cooling will be deduced. b) Sodium boiling experiments in an annular channel heated from the outside with local flow restrictions built in. This is named "negative bundle". c) Sodium boiling experiments in rod bundles. For the application to the real case it has to be kept in mind that during or preceding boiling at temperatures around IO00°C the fuel cladding may be destroyed, so that the bundle geometry is changed.
3.4. The present importance of superheat For several years sodium attracted the attention of researchers concerned with safety mainly because
COOLING DISTURBANCES of the rate of reactivity insertion from boiling in large core regions. This is strongly determined by superheat. At present consideration of these events is diminishing compared to the local phenomena in subassemblies or parts of subassemblies, for the reasons discussed in the beginning. The latter events are much more probable, much more difficult to detect, and could be the initiation of dangerous or costly subsequent events. Therefore, neglecting the rather weak boiling pressure pulses, superheat has very little importance here. After boiling has occurred x B and x s very soon will show little difference. For local boiling a larger superheat will increase the growth rate, and, possibly also the subsequent condensation of the bubble, and may even reduce the probability of local dryout. With some reservation, therefore, some superheat of the sodium may even be helpful with respect to local blockages.
4. The fuel-sodium-interaction
If the boiling events discussed above cause dryout of the liquid film, the following situation develops: a) For the case of boiling in the entire subassembly, after about 0.5 sec the film dries off at some place, whereas the wetted part of the bundle will continue to produce vapour. The dry areas spread, and after about 3 sec the zone of molten fuel has reached the fuel surface. Only after a larger amount of fuel is molten, sodium may eventually return and contact molten fuel. The course of the FSI and its mechanical effects will be influenced by the dynamic and geometrical conditions under which the two liquids get in contact. Since these conditions cannot be determined analytically, a number of simulation experiments with a rather wide variation of parameters will have to be conducted. b) For local boiling conditions, fuel melting is much less probable and the amount of UO 2 would be much less. Theoretical treatment of this problem is at this time restricted to upper bound estimates of the energy released by the FSI on a thermodynamic basis. Hicks and Menzies [9] first analyzed the vapour explosion occurring after mixture of liquid UO 2 and sodium in connection with excursions of the Bethe-
19
Tait type. For upper bound estimates they assumed instantaneous and homogeneous mixing of UO 2 and sodium, and they calculated the available mechanical energy for the optimum mixing ratio of 0.08 g Na per g UO 2 to be a maximum of 200 - 300 J/g UO 2. Using these assumptions the energy from the FSI would exceed the energy from boiling and condensation impulses, even if only 1 g of molten fuel would react with the proper amount of sodium. In reality the reaction is slowed down by the finite size of UO 2 particles, so that their heat is transferred in a finite time. Also, a type of inverse Leidenfrost phenomenon decreases the heat transfer [10]. It should be noted, however, that the FSI in the situation described above is qualitatively different from the one occurring during a nuclear excursion. In the latter case, fragmentation and dispersion of fuel will mainly be effected by fuel vapour pressure, whereas in the former case only driving forces from the relative velocities of the two liquids and of the interaction itself are available, if fission gas release is assumed to be essentially complete by the time the interaction starts. At different places experiments are carried out with UO 2 - Na, UO 2 - H20 , A1 - H20 and Thermit - H 2 0 . All reactions show pressure pulses, but so far no applicable model laws, nor even a sufficient understanding of the processes, could be derived from the experiments. In British investigations for fast heat transfer from hot metal balls in water, different types of heat transfer mechanism, with film boiling in some cases, have been found. If water temperatures are close to the boiling point, the vapour generation can reach an explosive nature. Semeria and Amelard (C.E.N. Grenoble) have found, by pouring molten UO 2 into liquid sodium, that some sodium might be enclosed in a solidified shell of UO 2 filled with liquid UO 2, which by subsequent rapid evaporation of the sodium inside the shell would be dispersed into the surrounding sodium. The amount of finely dispersed fuel increases with increasing sodium temperature, according to these experiments. To get clearer evidence, work on the following areas seems to be especially necessary: a) By the local boiling experiments described above, the starting conditions for dryout and fuel melting should be defined. Local reactions should be simulated, since this case has the greater importance
20
K.GAST and D.SMIDT
top
i
J
j~
bottom i¸
Fig. 6. Vortex generator.
COOLING DISTURBANCES because o f its relatively high probability. Simulation experiments using the thermit technique under sodium with limited reacting masses seem to be best suited. b) Simulation of melting in a total subassembly and sodium return flow into molten UO 2 or penetration of UO 2 melt into the lower blanket. c) In-pile experiments with artificial flow disturbances. For this projects in SEFOR, Cabri, Treat and others are considered. But since in no case all parameters can simulate the actual situation and model laws are missing, one has to be careful in the interpretation of these results. Also these experiments are expensive and time-consuming.
5. The detection of cooling disturbances It is evident that dangerous cooling disturbances must be detected early and result in automatic counteractions. For this reason every subassembly will be instrumented. The detectors relevant to the present problems are: - thermocouples - flow meters - reactivity meters - fission product detectors - boiling noise detectors. Thermocpoules and flow meters allow for detection of cooling disturbances in the total subassembly previous to boiling with a large safety margin in a diverse and redundant manner. The reactivity meters also allow for the measurement of voiding in a reactor zone not yet dangerous in size. For the detection of local disturbances there is not yet complete assurance whether these instruments are sufficient. For instance, for a local channel blockage in 30% of the channels only a 5% flow reduction will result, which is hardly detectable. Until it is demonstrated that a blockage of this size does not present a safety problem in itself, additional detectors are considered necessary. Boiling noise detectors or fission product detectors are sufficient, since some cladding fractures in the disturbed regions can be expected. A local fission gas detector mounted on top of the subassembly has been proposed and developed by K. Gast at the Karlsruhe laboratory. As shown in fig. 6, it consists of a vortex generator of the type originally developed in the UK, which collects the
21
gaseous fission products in the center of a cyclone flow. A heated thermocouple is mounted there and gives a signal by the reduced heat transfer, so only electrical signals have to be transmitted. Tests in water have shown that the cyclone is quite stable and fission gases down to about 3 cm 3 can~be detected. To use the flow meters as a boiling detector for local boiling they have to act quite fast (in the order of 10 msec) to give a distinguishable signal on top of the total flow signal. By these types of measurements the whole region of cooling disturbances can be covered, and a dryout should be prevented. For tl~is a reliable data processing and data evaluation within the safety system is necessary. The decision whether a dangerous local blockage has occurred, should not be entirely left to the operator. The data processing system also has to provide the relevant information on long term drifts by comparing measurements of previous times and measurements in the neighbouring subassemblies originating from local cooling failures. Taking care of this in a suitable manner and with the necessary diversity and redundancy, fast failure propagation can be avoided in the sodium-cooled fast reactor.
Acknowledgement The authors gratefully appreciate the helpful assistance of Mr. P. Wirtz for the calculations necessary for the theoretical dryout model.
References [ 1] M. Fischer and H. Shimamune, Temperature distribution and thermal stability in asymmetrical triangular rod clusters, KFK 724, EUR 4178e (1969). [2] H. K~impf, Allgemeine Spaltgleichung fiir den W~medurchgang Brennstoff-Hiille in Kernbrennelementen mit Tablettenbrennstoff, KFK 604, June 1967. [3] K. Gast and E.G. Schlechtendahl, Schneller natriumgekiihlter Reaktor Na-2, KFK 660 (1967). [4] E.G. Schlechtendahl, Sieden des Kiihlmittels in natriumgekuhlten schnellen Reaktoren, KFK 1020, EUR 4302d (1969) and Theoretical investigations on sodium boiling in fast reactors, Nucl. Science and Eng., to be published. [51 W. Peppier, E.G. Schlechtendahl and G.F. Schultheiss, Investigations on dynamic boiling in sodium cooled fast reactors, Nucl. Eng. Design 14 (1970) 000.
22
K.GAST and D.SMIDT
[61 G.F. Schultheiss, Experimental investigation of incipient boiling superheat in wail cavities, to be presented at the Symposium on Liquid Metal Heat Transfer and Fluid Dynamics, New York City, Nov. 29 - Dec. 3, 1970. [7] K. Schleisiek, Heat transfer and boiling during forced convection of sodium in an induction heated test tube, Nucl. Eng. Design 14 (1970) 000. [8] D. Smidt, P. Fette, W. Peppier, E.G. Schlechtendahl and G.F. Schultheiss, Problems of sodium boiling in fast reactors, KFK 790, EUR 3960e (1968).
[9] E.P. Hicks and D.C. Menzies, Theoretical studies on the fast reactor maximum accident, ANL-7120 (1965). [10] R.S. Hall et al., Inverse Leidenfrost phenomenon, Nature 224, No. 5216 (1969) 266-267. [ 11 ] W. Peppier and E.G. Schlechtendahl, Experimental and analytical investigations of sodium boiling events in narrow channels, to be presented at the Symposium on Liquid Metal Heat Transfer and Fluid Dynamics, New York City, Nov. 29 - Dec. 3, 1970.
COOLING
DISTURBANCES
IN THE CORE OF SODIUM-COOLED
AS CAUSES OF FAST FAILURE
FAST REACTORS
PROPAGATION
K. GAST and D. SMIDT
Institut fiir Reaktorentwicklung, Kernforschungszentrum, Karlsruhe Received 4 June 1970
As hypothetical cause of Bethe-Tait type excursions in sodium-cooled fast reactors local cooling failures are considered. Several mechanisms of failure propagation are discussed. The most important one is the thermal interaction between molten fuel and liquid sodium. The initial conditions starting from local failures or subassembly blockage are defined. A criterion for dryout in narrow channels is developed. The importance of sodium boiling superheat is rediscussed in this context. Finally the functional requirements for core instrumentation are given. It is concluded that the probability of failure propagation leading to Bethe-Tait excursions is sufficiently low.
1. Introduction Large nuclear power plants are designed to a standard o f reliability and safety which is unparalleled in other field of modern technology. The necessity for such high safety standards stems from the large hazard potential of the radioactive materials contained in nuclear reactors. This hazard potential is sometimes looked upon without due consideration of the vanishingly low probability of events that could result in the release of sufficient amounts of radioactive materials to cause radiation injuries to the public. For a hypothetical accident in a 1000 MWe reactor the probability of early casualties becomes significant if more than 0.1% o f the volatile fission products are assumed to be released on ground level. Hence, the maximum possible release after any conceivable accident has to be much below this value. For light water reactors this implies, for instance, that following a postulated fracture of the primary system the emergency cooling system must operate and the containment integrity must be maintained. To prove this, the failure probability must be demonstrated to be sufficiently small and the possibility of causal connections of different failures must be considered. However, if the principle of making conservative assumptions is carried to an extreme, unacceptable
situations can be postulated: One example of such hypothetical accidents in the light water reactor is the catastrophic failure of the pressure vessel, which could cause disintegration o f the containment by pressure effects or by missiles. Also the core can be transferred to an undefined geometry for which effective decay heat removal cannot be guaranteed. For sodium cooled fast reactors hypothetical safety considerations are different. The coolant pressure is so small, that a fracture of the primary system in itself cannot cause a large energy release and subsequent failure of the containment. But, since the core of the fast reactor is not in the state o f its largest reactivity, there exists the principal possibility of a fast nuclear exursion, the consequence of which could affect the integrity o f the containment in an analogous way. Therefore, for this type o f accident the same degree of improbability must be attained as e.g. for the fast pressure vessel burst or the primary system fracture combined with containment failure for the light water reactor. Among the most important causes for fast power excursions of the Bethe-Tait type in sodium cooled fast reactors are the partial or total loss of heat removal capability to the core. It results either from a partial core voiding and a positive void reactivity or from core compaction (melt down) in the absence of
COOLING DISTURBANCES sodium [3]. In the following paragraphs, therefore, local coolant disturbances will be studied for their potential as initiation of major damage to the core. In so far as they have primarily local character one has to look for possible mechanisms of a fast failure propagation. Boiling and recondensation shocks, and, most important, the interaction of molten fuel with sodium will be treated as the main modes for failure propagation.
2. Classification of cooling failures 2.1. Cooling disturbances affecting the entire core A cooling disturbance in the total core or in a larger region of it might be generated as a consequence of failure propagation from local effects, by simultaneous loss of flow and failure of the shut-off system or by a loss of coolant. The latter case can be excluded since the primary system has double walls in all lower parts. The second case belongs to the widely discussed type of incidents involving engineered safeguards and their reliability and will not be treated in this context. After a local cooling failure in the core, sodium boiling in the neighbouring subassemblies could occur as consequence of failure propagation. Schlechtendahl [4] has shown that because of the statistical nature of incipient boiling over the subassemblies the reactivity rate by voiding will be relatively low. After signals from different detectors (reactivity, temperature, flow, boiling noise) the reactor is very likely to be shut off before reaching the prompt supercritical condition. However, it has to be ensured that by potential core deformations from mechanical effects ((10) of fig. 2) additional reactivity will not be introduced nor the shut-off rods become ineffective. 2.2. Local cooling disturbances within the bundle Local flow blockages are considered as the most important initiating events, because they are relatively likely to occur, and because they may not be detected by temperature sensors or by flowmeters, before a critical situation such as local or gross sodium boiling is reached. Local cooling deficiencies can be caused by either one or a combination of the following faults: Type A: Blockage of single or several neighbouring cooling subchannels by foreign objects, especially at the spacers.
13
Type B: Narrowing of single subchannels by lateral movement, bowing, or swelling of fuel rods. In the general case these disturbances in single channels are more or less statistically independent. In the neighbourhood of a channel, in which the coolant flow is suppressed by movement of a rod, there is another one, where the flow is increased. This type includes especially the effects of thermal bowing according to [1]. Type C: Coolant flow reduction by bursting of cladding tubes. Type D: Reduced heat conductance through the gap between fuel and cladding, caused for instance by a cladding tube inflation from fission gas pressure. Since already a small increase of the gap size results in a large increase of the fuel temperature, this effect cannot be completely excluded. It is counteracted by the radiation induced embrittlement of the cladding material, but in some cases larger strains have been observed, if the short time strength of the material was exceeded at excessive temperatures. The several failure types can cause each other. For instance a local hot spot of tybe B may result in a cladding burst of type C and a subsequent wash-out of fuel particles thereby leading to a disturbance of type A. This again may cause disturbances of type C or D etc. Unless detected, the affected zone may grow in size and eventually reach a critical condition. The growth rate will mainly be determined by the shorttime mechanical properties and fracture mechanics of cladding and grid material at excessive temperatures. The critical condition mentioned above is defined such that it may either lead to local boiling within the bundle, to gross boiling and sodium ejection from the entire subassembly, or to inflation of a sufficiently large number of cladding tubes to allow a considerable amount of fuel to melt because of the increased thermal resistence between fuel and clad. Local boiling can lead to gross boiling and ejection, provided the production rate of sodium vapor is sufficiently large to interfere with the overall flow in the affected subassembly. Fig. 1. shows the hypothetical failure tree connected to local failures. Starting point is always a local cooling disturbance (1) of any of the mentioned types. All branches either end with a reactor shut-down (3) or again in a local cooling disturbance (1) and a new start for a path in the failure tree, or with a propagation event (15).
14
K.GAST and D.SMIDT
(2a) ~
(s) L
{ 2 b ) ~
•
.~" (11) ~ n g j (1) subassernblyoooling ~.~7/" ~ l ~ disturbance Fig. 2 __.~--~__ (12a L detected j : (12b) ~ - ~ t ~ melting typeD[ ~ L .. , I r ~ ~L~,~ shut-off _ ]
~ . - . (5o)~0 SI]
(i)
(3) LShUt-Off J ( 1 3 o ~
(13b)dr~ou~ d
(Sb)F$1
i ;
(14) ~ L~o~.-pu,~,~(5b)
/
Fig. 2
Fig. 1. Failure tree for local cooling disturbances.
If a local cooling disturbance is not detected, coolant boiling (1 I) may result. There exists the principal possibility that boiling or recondensation shocks (13) may affect the fuel bundle, thereby initiating a propagation to a larger local cooling disturbance, or a cooling disturbance of the whole subassembly (15) (see paragraph 2.3). If also the coolant boiling will not be detected, a dry-out (burn-out) of a fuel element region (13b) and following fuel melting may occur. When molten fuel and sodium eventually get in contact, a sudden release of mechanical energy by the fuel-sodium-interaction [9] (FSI) is possible. Subsequent events could be: a larger local cooling disturbance (1), or a cooling disturbance in the subassembly (15). Also one could think of a fuel melting without preceding sodium boiling after a failure of type D (4). Here, too, the FSI with all consequences could be possible. However, it is highly improbable that in this case larger parts of the fuel elements melt and react. For this reason this branch is more hypothetical with respect to the generation of a fast excursion. It can be shown that local coolant flow deficiencies caused by reduction of the flow path area are unlikely to initiate sodium boiling within the bundle, provided the blockage is not complete and the affected ~z,one occupies only a small fraction of the core height. It seems possible, however, that fuel claddings will swell and finally burst from the internal gas pressure, because the strength of the material is strongly reduced as temperature increases. It is concluded, therefore, that rupture of fuel cans and subsequent
fission gas release may be the first indication of local blockages. Therefore, a fast reacting can leak detection system should be incorporated into the early warning system. 2.3. Cooling disturbances in the entire subassembly This disturbance may occur if either a local disturbance has spread over the whole subassembly, or if the coolant inlet at the subassembly bottom is blocked by some foreign object. By the design of the inlet nozzle [3] in general it can be excluded that the object can block the total flow instantaneously. Fig. 2 shows the relevant failure tree. In many ways it it related to fig. 1. The propagation from FSI (8) finally may result in cavities (10) (positive reactivity formation), caused by the energy release. Also cooling disturbances of the whole core with subsequent overall boiling and control rod defects are possible, which diminish the shut-off capability.
2.4. Consequences Under unfavourable assumptions and excluding counteraction by safety devices, cooling disturbances could lead to fast excursions via boiling and void effect. Also a reactivity input is possible as consequence of local fuel-sodium-interaction. On the other hand a core meltdown in the presence of sodium can be excluded, as already Gast and Schlechtendahl [3] have shown. Undoubtedly it has to be proven that all chains of events which could lead to failure propagation and finally to a nuclear excursion are either in themselves sufficiently improbable, or are so in combination with a suitable instrumentation. For these
subassembly cooling •
disturbance
J
boding or cond ~ ' Ppulses [~ j (3) -shut-off
1
(11)
(lo) ~
- ] " (7) dry out
i
(8)E
]
011 Icor,disturE~
(12)[conlro=rodsinop~m~
excursion lexcurslon___~L.~~~xcunr°~°n ~ curs~ shut-off iT detected
shut-off if detected
i
i_ iexc~°_= ~ , "j n, r~curs~
shut-Off if detected (and reoctivity ¢z/aikat~e
Fig. 2. Failure tree for subassembly cooling disturbances.
COOLING DISTURBANCES reasons first the problem of coolant boiling will be treated, which forms the most important initial condition for an eventual failure propagation. Then failure propagation itself is discussed, and finally the possibilities of instrumentation.
3. Boiling events in sodium Most work on sodium boiling in connection with fast reactor safety has been centered on the process of expulsion. This determines the time scale of reactivity input via the sodium void effect. Only recently the consideration of boiling as an early phase of the possible failure propagation from local causes has been added. Therefore, new points of view have to be considered. The work of Peppler, Schlechtendahl and Schultheiss [5] in this edition gives a survey of our present state of knowledge of sodium expulsion. The characteristic feature is the growth of a single bubble fed by evaporation of a liquid film on the wall, which inhibits the formation of svbsequent bubbles at least at superheats above 30°C. Then follows an oscillating increase and decrease of bubble size, caused by heat transfer to the liquid subcooled sodium above the heated channel. For the aspect of failure propagation on the other hand, in accordance with figs. 1 and 2, the pressure pulses during boiling and the fuel melting after dryout of a channel merit some consideration.
3.1. Pressure pulses during sodium boiling Pressure pulses occur during bubble growth from superheated liquid, or during bubble collapse. The former shall be called boiling pulses, the latter con-
densation pulses. The initial pressure of the boiling pulses correspond to the superheat at boiling initiation. If no special measures are provided, such as artificial boiling nuclei like gas bubbles or wall cavities (for instance between fuel rod and spacers) as already mentioned in [6], under reactor conditions superheats up to about 150°C have to be expected. For local pressures of 2 atm, for instance, this corresponds to an internal overpressure in the bubble of about 3.8 atm, which is effective nearly over the total expulsion phase of about 0.15 sec (see [5], fig. 13) (not to be confused with the pressure pulses measured at the unheated
15
upper end of the test section - [5], table 5). If the boiling occurs only locally in the subassembly and if the disturbed region is surrounded by subcooled sodium, the pressure pulses become shorter and act in a smaller region only. In most cases the subassembly box has a hexagonal cross section of about 10 to 11 cm span with a small distance to the neighbouring boxes, supported on the inside by fuel rods and spacers. If a boiling pulse occurs in one subassembly, it cannot be completely excluded that neighbouring subassemblies will be affected. The effects are not expected to be large because of the short duration of the load and the small amount of energy involved. The order of magnitude may be estimated by comparing the wall stresses of a box of circular cross section and similar size. Here the load would amount to about 1 kp/mm 2, which is in the range of material strength. This is at least a certain indication of what would happen in the actual geometry. Condensation pulses in the total subassembly might be generated by the return flow and collision of two liquid columns, or locally by the collapse of vapour bubbles. In the theoretical model of BLOW 2 ([5], fig. 13) a complete return flow is predicted. But this has not been found in experiments. The reason for this difference is the neglect of vapour pressure drop in the theoretical model. In the experiments condensation pulses were observed only when a first gas bubble, which is submerged in the cold zone, is suppressed by a second one which is still under high pressure. This is shown in [5], fig. 20. Until a nearly complete dryout of the wall film has occured, a complete reentry of liquid sodium will be prevented by the vapour formation. But at this time the fuel may already be partially molten, which is the starting condition for the FSI to be discussed later. The condensation pressure peaks are large but narrow. Energies for a subassembly are in the order of 1 KWsec only. Therefore, no considerable deformations are to be expected. To get a better knowledge of the effects on the core in Karlsruhe the BEVUS experiment will be undertaken in the middle of 1970. It consists of an electrically heated subassembly surrounded by 6 unheated ones in a sodium pool. By a fast-acting pressure release valve an arbitrary superheat can be simulated. The subassemblies have actual dimensions.
16
K.GAST and D.SMIDT
pressure reduction system for initiation of sodium boiling
pressure relRf valv,
expansion tank
z current supply cables
test tank
sodium supply line,
\cover g_as supply lines
core section m o c k
safety tank
heated element
test insert
Fig. 3. BEVUS test facility schematic.
Only the linear rating with 40 W/cm is considerably smaller than in the actual case. Fig. 3 shows the overall arrangement. In a later phase, when the initial conditions of the FSI are better known, the facility will allow relevant experiments in bundle geometry for this case.
3.2. The process of dryout of the fuel subassembly The FSI is the more important propagation mechanism. For this first some fuel has to melt, to allow for the necessary intimate mixing. Melting of the fuel is
impossib,e as long as sodium is in contrast with it and cools it by convection or evaporation. This brings up the problem of "burnout", better described as dryout. To get a better understanding of the physical processes we shall analyze the flow o f boiling sodium in a tube heated from the outside. This model, for instance, has been used in the experiments described in [5]. Fig. 4 gives a schematic display of the hypothesized course of events. The liquid sodium enters the heated part of the channel (fig. 4a) with the velocity o i and the tempera-
COOLING DISTURBANCES
la)
the liquid film +cannot be renewed by the liquid plug and the first bubble stays until the liquid wall film begins to dry at some place. After this, cladding and fuel are likely to reach their melting temperatures. The pressure drop is determined by equations (2) to (5).
Ib)
-
(c)
+,'o'+
17
dp -- ~f) -002 - ~ dx + pv do
(2)
(d)
P = P(P)
(3)
rrDq(x-x B) rh -
Ah
(4)
~D 2
rh = vp 4
Fig. 4. Oscillatory boiling in a narrow channel.
ture T i. The heat flux is q. After the saturation temperature is reached at Xs, the first vapour bubble begins to grow at x B driven by an external pressure according to a superheat Tsup - Tsat. In fig. 4b the growing piston type bubble is shown. Sodium vapour is generated from the liquid film remaining on the wall. This, therefore, results in efficient heat removal. In fig. 4c the bubble touches at XA, the upper sodium pool, where rapid condensation takes place. (For pool temperatures of 580°C in [5] condensation fluxes of 500 - 1000 W/cm 2 have been found.) Between x B and x A the vapour flow is subject to a pressure drop ~Pv" If Ap v is smaller than the pressure head Z~gpgenerated by the pumps, the liquid column can partially replace the bubble and at x B a new bubble can be generated. Because of the statistical fluctuation of superheat and the temperature increase in the liquid, x B in reality may vary. In fig. 4d the second bubble displaces the first one while the liquid plug in between renews the film on the heated wall. The orific O in combination with the pumping head inhibits any appreciable reverse flow during the growth phase of the bubble. Therefore, in the framework of this model a sufficient criterion for dryout of the wall film can be formulated. It is
(5)
where m = mass flow; q = heat flux and z ~ = evaporation heat. In eq. (2) it has to be taken into account, that according to experiments [11 ] the flow between x B and x A is not a single phase vapor. It rather is mist flow with a liquid mass fraction in the order of 15%. Accordingly, there is a small liquid flow between O and x B . However, the corresponding pressure drop of the liquid part can be neglected. In fig. 5 the results of the calculation with the criterion ( I ) is compared with results of Peppier [ 11 ].
• Experiments Pept~er [121 pump heocl I,Sat - c,:,l:.uuor, ~;o~-".-.- , ~ .
fordifferent p~m~~
/ / k ( n . m .~. d ~
)
../)..'0h..t ..x [ w , ~ ]
Z~Op < Apv
(1)
If (1) is valid, a second bubble cannot be formed,
0
0
I00
! 200
t 300
i ~.tOO
Fig. 5. Dryout heat flux versus static pressure in the test section parameter : p u m p head.
18
K.GAST and D.SMIDT
Here the dryout heat flux is related to the static pressure in the test section (as described in [5] ). The parameter is the pump pressure head. In the experiments it has been 1.5 atm. The curves represent the theory outlined here. On the right hand side of the respective curve dryout should occur. The points give data of observed dryouts. It has to be noted that the points on the 1.5 atm curve correspond to 5 or 3 measurements, respectively. Only one dryout at the upper left occurred, at a rather low heat flux. Its validity is not quite clear yet. However, it has to be restated, that (1) is a sufficient criterion only, and additional effects may promote an earlier dryout. More experiments are under way. In any case the theory indicates the right order of magnitude. The choice of the friction factor f t u r n s out to be not critical. The dryout conditions discussed here are related in many ways to the well known cases of two-phase flow instability. Applied to the reactor, the calculations show that in long and narrow coolant channels, as in fast reactors, the flow reduction necessary to cause boiling and the nominal flow reduction to cause instability and dryout are not far from each other. The flow regime discussed here and the well known regime of slug flow of classical two-phase flow work have a certain similarity in their appearance. However, there is no relation in their physical natures. While the large bubbles of slug flow are generated by hydrodynamic effects, the single bubbles discussed here are formed by the fast growth of the nucleus from superheated liquid. Their size will be determined by rapid evaporation and condensation fluxes at their surface. In contrast to the classical case, therefore, the phenomenon, which is typical for liquids with superheat, will be called pulsating flow. The thickness of the liquid film remaining on the wall amounts to some tenths of a ram. In [8] some work on its value has been reported. For q = 200 W/cm 2 at 900°C the evaporation rate is 0.7 mm/sec. This means that the film stays for some tenths of a second. Before this time in the pulsating flow the next expulsion must follow if dryout is to be avoided. In the experiments of Schleisiek [7] this is the case. For the same reasons as shown in [5], the single channel results also should be applicable to rod bundles.
3.3. Local boiling While all considerations so far applied to the case of a homogeneous flow reduction in the total subassembly as the starting condition, now the case of a local disturbance within the bundle of one of the types described in par. 2 will be treated. Experimental investigations do not exist so far. Evidently they have to be done in bundle geometry. However, from a qualitative analysis one can derive preliminary conclusions. For instance, if some subchannels are blocked locally by objects at the spacers (type A) behind the blockage there will be a wake. If this occurs in the area of a 7-rod cluster, at the center rod boiling may occur over a length of some cm. The bubble expands to all sides into the subcooled sodium outside of the 7-rod cluster and recontracts very quickly. By this it will transport cold sodium in the direction of the boiling center. Because of the rapid recondensation rate, it probably will collapse completely and then start growing again. Thus dryout at the center rod should be prevented. Even for still larger rod blockages a similar pumping action of the bubble could occur. Comparing the vapour-generating areas with the cross section for the return flow between the rods, the limit for complete refilling should be expected for a cluster in the range of 30 rods. This statement, however, has not yet been experimentally proved. Because of the importance of these events for local dryout, an experimental and theoretical program has been started at Karlsruhe. It includes: a) Experimental determination of the wakes behind blockages in unheated rod clusters with transparent simulation liquids in the one-phase region. From this by theoretical means the temperature distribution of single-phase sodium cooling will be deduced. b) Sodium boiling experiments in an annular channel heated from the outside with local flow restrictions built in. This is named "negative bundle". c) Sodium boiling experiments in rod bundles. For the application to the real case it has to be kept in mind that during or preceding boiling at temperatures around IO00°C the fuel cladding may be destroyed, so that the bundle geometry is changed.
3.4. The present importance of superheat For several years sodium attracted the attention of researchers concerned with safety mainly because
COOLING DISTURBANCES of the rate of reactivity insertion from boiling in large core regions. This is strongly determined by superheat. At present consideration of these events is diminishing compared to the local phenomena in subassemblies or parts of subassemblies, for the reasons discussed in the beginning. The latter events are much more probable, much more difficult to detect, and could be the initiation of dangerous or costly subsequent events. Therefore, neglecting the rather weak boiling pressure pulses, superheat has very little importance here. After boiling has occurred x B and x s very soon will show little difference. For local boiling a larger superheat will increase the growth rate, and, possibly also the subsequent condensation of the bubble, and may even reduce the probability of local dryout. With some reservation, therefore, some superheat of the sodium may even be helpful with respect to local blockages.
4. The fuel-sodium-interaction
If the boiling events discussed above cause dryout of the liquid film, the following situation develops: a) For the case of boiling in the entire subassembly, after about 0.5 sec the film dries off at some place, whereas the wetted part of the bundle will continue to produce vapour. The dry areas spread, and after about 3 sec the zone of molten fuel has reached the fuel surface. Only after a larger amount of fuel is molten, sodium may eventually return and contact molten fuel. The course of the FSI and its mechanical effects will be influenced by the dynamic and geometrical conditions under which the two liquids get in contact. Since these conditions cannot be determined analytically, a number of simulation experiments with a rather wide variation of parameters will have to be conducted. b) For local boiling conditions, fuel melting is much less probable and the amount of UO 2 would be much less. Theoretical treatment of this problem is at this time restricted to upper bound estimates of the energy released by the FSI on a thermodynamic basis. Hicks and Menzies [9] first analyzed the vapour explosion occurring after mixture of liquid UO 2 and sodium in connection with excursions of the Bethe-
19
Tait type. For upper bound estimates they assumed instantaneous and homogeneous mixing of UO 2 and sodium, and they calculated the available mechanical energy for the optimum mixing ratio of 0.08 g Na per g UO 2 to be a maximum of 200 - 300 J/g UO 2. Using these assumptions the energy from the FSI would exceed the energy from boiling and condensation impulses, even if only 1 g of molten fuel would react with the proper amount of sodium. In reality the reaction is slowed down by the finite size of UO 2 particles, so that their heat is transferred in a finite time. Also, a type of inverse Leidenfrost phenomenon decreases the heat transfer [10]. It should be noted, however, that the FSI in the situation described above is qualitatively different from the one occurring during a nuclear excursion. In the latter case, fragmentation and dispersion of fuel will mainly be effected by fuel vapour pressure, whereas in the former case only driving forces from the relative velocities of the two liquids and of the interaction itself are available, if fission gas release is assumed to be essentially complete by the time the interaction starts. At different places experiments are carried out with UO 2 - Na, UO 2 - H20 , A1 - H20 and Thermit - H 2 0 . All reactions show pressure pulses, but so far no applicable model laws, nor even a sufficient understanding of the processes, could be derived from the experiments. In British investigations for fast heat transfer from hot metal balls in water, different types of heat transfer mechanism, with film boiling in some cases, have been found. If water temperatures are close to the boiling point, the vapour generation can reach an explosive nature. Semeria and Amelard (C.E.N. Grenoble) have found, by pouring molten UO 2 into liquid sodium, that some sodium might be enclosed in a solidified shell of UO 2 filled with liquid UO 2, which by subsequent rapid evaporation of the sodium inside the shell would be dispersed into the surrounding sodium. The amount of finely dispersed fuel increases with increasing sodium temperature, according to these experiments. To get clearer evidence, work on the following areas seems to be especially necessary: a) By the local boiling experiments described above, the starting conditions for dryout and fuel melting should be defined. Local reactions should be simulated, since this case has the greater importance
20
K.GAST and D.SMIDT
top
i
J
j~
bottom i¸
Fig. 6. Vortex generator.
COOLING DISTURBANCES because o f its relatively high probability. Simulation experiments using the thermit technique under sodium with limited reacting masses seem to be best suited. b) Simulation of melting in a total subassembly and sodium return flow into molten UO 2 or penetration of UO 2 melt into the lower blanket. c) In-pile experiments with artificial flow disturbances. For this projects in SEFOR, Cabri, Treat and others are considered. But since in no case all parameters can simulate the actual situation and model laws are missing, one has to be careful in the interpretation of these results. Also these experiments are expensive and time-consuming.
5. The detection of cooling disturbances It is evident that dangerous cooling disturbances must be detected early and result in automatic counteractions. For this reason every subassembly will be instrumented. The detectors relevant to the present problems are: - thermocouples - flow meters - reactivity meters - fission product detectors - boiling noise detectors. Thermocpoules and flow meters allow for detection of cooling disturbances in the total subassembly previous to boiling with a large safety margin in a diverse and redundant manner. The reactivity meters also allow for the measurement of voiding in a reactor zone not yet dangerous in size. For the detection of local disturbances there is not yet complete assurance whether these instruments are sufficient. For instance, for a local channel blockage in 30% of the channels only a 5% flow reduction will result, which is hardly detectable. Until it is demonstrated that a blockage of this size does not present a safety problem in itself, additional detectors are considered necessary. Boiling noise detectors or fission product detectors are sufficient, since some cladding fractures in the disturbed regions can be expected. A local fission gas detector mounted on top of the subassembly has been proposed and developed by K. Gast at the Karlsruhe laboratory. As shown in fig. 6, it consists of a vortex generator of the type originally developed in the UK, which collects the
21
gaseous fission products in the center of a cyclone flow. A heated thermocouple is mounted there and gives a signal by the reduced heat transfer, so only electrical signals have to be transmitted. Tests in water have shown that the cyclone is quite stable and fission gases down to about 3 cm 3 can~be detected. To use the flow meters as a boiling detector for local boiling they have to act quite fast (in the order of 10 msec) to give a distinguishable signal on top of the total flow signal. By these types of measurements the whole region of cooling disturbances can be covered, and a dryout should be prevented. For tl~is a reliable data processing and data evaluation within the safety system is necessary. The decision whether a dangerous local blockage has occurred, should not be entirely left to the operator. The data processing system also has to provide the relevant information on long term drifts by comparing measurements of previous times and measurements in the neighbouring subassemblies originating from local cooling failures. Taking care of this in a suitable manner and with the necessary diversity and redundancy, fast failure propagation can be avoided in the sodium-cooled fast reactor.
Acknowledgement The authors gratefully appreciate the helpful assistance of Mr. P. Wirtz for the calculations necessary for the theoretical dryout model.
References [ 1] M. Fischer and H. Shimamune, Temperature distribution and thermal stability in asymmetrical triangular rod clusters, KFK 724, EUR 4178e (1969). [2] H. K~impf, Allgemeine Spaltgleichung fiir den W~medurchgang Brennstoff-Hiille in Kernbrennelementen mit Tablettenbrennstoff, KFK 604, June 1967. [3] K. Gast and E.G. Schlechtendahl, Schneller natriumgekiihlter Reaktor Na-2, KFK 660 (1967). [4] E.G. Schlechtendahl, Sieden des Kiihlmittels in natriumgekuhlten schnellen Reaktoren, KFK 1020, EUR 4302d (1969) and Theoretical investigations on sodium boiling in fast reactors, Nucl. Science and Eng., to be published. [51 W. Peppier, E.G. Schlechtendahl and G.F. Schultheiss, Investigations on dynamic boiling in sodium cooled fast reactors, Nucl. Eng. Design 14 (1970) 000.
22
K.GAST and D.SMIDT
[61 G.F. Schultheiss, Experimental investigation of incipient boiling superheat in wail cavities, to be presented at the Symposium on Liquid Metal Heat Transfer and Fluid Dynamics, New York City, Nov. 29 - Dec. 3, 1970. [7] K. Schleisiek, Heat transfer and boiling during forced convection of sodium in an induction heated test tube, Nucl. Eng. Design 14 (1970) 000. [8] D. Smidt, P. Fette, W. Peppier, E.G. Schlechtendahl and G.F. Schultheiss, Problems of sodium boiling in fast reactors, KFK 790, EUR 3960e (1968).
[9] E.P. Hicks and D.C. Menzies, Theoretical studies on the fast reactor maximum accident, ANL-7120 (1965). [10] R.S. Hall et al., Inverse Leidenfrost phenomenon, Nature 224, No. 5216 (1969) 266-267. [ 11 ] W. Peppier and E.G. Schlechtendahl, Experimental and analytical investigations of sodium boiling events in narrow channels, to be presented at the Symposium on Liquid Metal Heat Transfer and Fluid Dynamics, New York City, Nov. 29 - Dec. 3, 1970.