International Conference Nuclear Energy for New Europe 2005 Bled, Slovenia, September 5-8, 2005
Severe Accident Experiments on PLINIUS Platform Results of First Experiments on COLIMA Facility Related to VVER-440 - Presentation of Planned VULCANO and KROTOS Tests
Pascal PILUSO, Eric BOCCACCIO, Jean-Michel BONNET, Christophe JOURNEAU, Pascal FOUQUART, Daniel MAGALLON CEA Cadarache Severe Accident Mastering Laboratory , Building 708 F13108 St Paul lez Durance CEDEX, France [email protected], [email protected], [email protected], [email protected], [email protected], [email protected] Ivan IVANOV, Ivan MLADENOV Technical University,Sofia, Bulgaria [email protected] Stoyan KALCHEV NPP Kozloduy, Bulgaria Pavlin GRUDEV Institute of Nuclear Research and Nuclear Energy, Bulgarian Ac. Sci., Sofia, Bulgaria Hans ALSMEYER, Beatrix FLUHRER Forschungszentrum Karlsruhe,Germany [email protected] Matjaz LESKOVAR Jozef Stefan Institute, Ljubljana , Slovenia, [email protected]
ABSTRACT
In the hypothetical case of a nuclear reactor severe accident, the reactor core could melt and form a mixture of nuclear fuel (UO2+ Fission Products), metallic or oxidized cladding +steel, called “corium”, of highly refractory oxides (UO2, ZrO2) and metallic or oxidized steel, that could eventually flow out of the vessel and mix with the substrate decomposition products (generally oxides such as SiO2, Al2O3, CaO, Fe2O3). The French Atomic Energy Commission (CEA) has launched a R&D programme aimed at providing the tools for improving the mastering of severe accidents. It encompasses the development of models and codes, performance of experiments in simulant and prototypic materials and the analysis of international experiments.
The experiments with prototypic corium (i.e. material containing depleted UO2) are performed in the PLINIUS experimental platform at CEA Cadarache. It comprises the VULCANO facility for 50-100 kg tests (corium-material interactions, corium solidification etc.), the COLIMA facility for smaller scale (~1 kg) experiments, the VITI facility for corium properties measurement and the KROTOS facility for corium-water interaction (a few kg). In the framework of the 5th European Framework Programme, free trans-national access to these facilities has been offered to EU and Associated States researchers. For the first PLINIUS access, COLIMA experiments have been conducted with a Bulgarian Team (TU/SOFIA, BAS/INRNE and NPP/KOZLODUY). This series of tests was devoted to
079.1 079.2 experimental studies on fission products release and corium behaviour in the late phase in a hypothetic case of severe accident in a PWR type VVER-440. The COLIMA experimental results are consistent with previous experiments on irradiated fuels (VERCORS, PHEBUS) with small differences for some fission products and show new results for the remaining corium. For the second visit, scientific users from FZK in Germany were selected to validate the COMET core-catcher with prototypic corium and sustained heating. A third call for proposals is devoted to tests in the KROTOS facility (JSI (Slovenia)).
1 INTRODUCTION
The PLINIUS platform is an experimental infrastructure at CEA Cadarache (France) devoted to the study of prototypic corium, i.e. mixtures including depleted uranium of the same chemical composition as what would be molten during hypothetical severe accident. It consists of 4 corium facilities : VITI, COLIMA, VULCANO, and KROTOS. Through its transnational access to research infrastructures activity, the European Commission finances (travel and subsistence of users and all experimental operations costs) the performance of experiments by European visiting scientists on the PLINIUS platform.
After the first call for proposals, a Bulgarian team was selected for a project in COLIMA facility dealing with fission product aerosols releases from a corium pool in a VVER-440 [1]. In a hypothetic case of severe accident in a PWR type VVER-440, a complex corium pool could be formed and fission products could be released. In order to study aerosols release(mechanisms, kinetics, nature, quantity) and to better precise the source term of VVER-440, two COLIMA-Aerosols experiments CA-U2 and CA-U3 have been performed to study aerosols release coming from a corium pool including non-volatile fission products.
A description of the COLIMA facility is first of all given, then the main results of the 2 experiments COLIMA CA-U2 and CA-U3 are described and the post-test analysis of the last test are presented. Different techniques of characterization have been used: SEM/EDS for the corium and impactor and for the W-Thermal Gradient Temperature ICP-AES. At least, an interpretation of COLIMA CA-U3 is given from the point of view of fission products release and remaining corium behaviour. The second access is devoted to the validation with scientific users from FZK (Germany) of COMET core catcher with prototypic corium in the VULCANO facility with sustained heating. The third access to the KROTOS facility is planned to study with the scientific users from JSI (Slovenia) the effect of the presence of fission products and iron oxide in the melt on steam explosion.
2 COLIMA-AEROSOLS EXPERIMENTS
The objectives of these tests are the quantification of the increase of the gaseous radioactive products in severe accidents of nuclear water reactor through a simulation in the PLINIUS Platform in behalf of the Environmental Impact Assessment of the VVER440 plants at NPP Kozloduy. Its aim is to measure and then compute the quantity of the additional gaseous radioactive products by the corium melt, in the hypothetical scenario of a severe accident. Low-volatile Fission Products have been considered for a burn-up of 35 MWd/kgU, 24 hours after shutdown and were inserted in the load as oxides of natural isotopic composition (except for uranium which was depleted). When we compare the rate of the main fission products
Proceedings of the International Conference “Nuclear Energy for New Europe 2005” 079.3 between a VVER-440 and a PWR [3], we can note that the concentration in Ba, Ce, Cs, Mo, Nd, Pr, and Zr is slightly higher for a VVER-440 in comparison with a PWR.
Element VVER-440 core mmol/molU PWR core mmol/molU (moles) VVER-440 core (moles) [3] PWR U 177723 279673 Ba 567 3,2 798 2,9 Ce 929 5,2 1431 5,1 Cs 1083 6,1 1591 5,7 Eu 59 0,3 95 0,3 I 94 0,5 135 0,5 La 446 2,5 638 2,3 Mo 1750 9,8 2480 8,9 Nb 6 0,0 32 0,1 Nd 1368 7,7 1845 6,6 Pr 404 2,3 565 2,0 Rh 214 1,2 287 1,0 Ru 1198 6,7 1878 6,7 Sb 12 0,1 10 0,0 Sr 489 2,7 728 2,6 Te 195 1,1 269 1,0 Y 258 1,5 378 1,4 Zr 1956 11,0 2823 10,1
Table 1: comparison of the main fission products concentration in a VVER-440 and a PWR (from [3]).
2.1 Description of the COLIMA facility
The COLIMA facility uses the hot crucible induction technique. Argon with 2% H2 is the gas medium, but it is possible to chose others gases if needed. The general principles of the hot crucible heating are to use a medium frequency generator (about a few tens kHz) and an inductor in order to heat a susceptor (tunsgten crucible). Then, the susceptor will transmit the heat to the total load (about a few kg) till to reach a temperature higher than the liquidus temperature of the corium. The temperature of the corium pool is estimated thanks to a radiative pyrometer (see Figure 1). For the COLIMA-Aerosols experiments, specific instrumentation has been developed in order to characterize the fission products release. Especially, an impactor has been installed to recover the aerosols coming from the oxidic corium pool and a thermal gradient tube to correlate the fission products deposits with the local temperature (see Figure 2). For COLIMA CA-U2 and CA-U3 tests, a S thermocouple has been placed 10 cm below the W-crucible in order to evaluate the thermal gradient along the W-crucible. For the thermal gradient tube in steel, K thermocouples have been put and for the thermal gradient tube in tungsten, W thermocouples have been put (see Figure 2/inserts).
Proceedings of the International Conference “Nuclear Energy for New Europe 2005” 079.4
Figure 1: Sketch of the COLIMA facility
Câblage COLIMA CA-U1
Collector- Stage-2
Pyrometer
Flow-meter 1 0 - 50 L/mn Heater Gate 1
Line 1 Diffuser 28 L/mn
Line 2
Impactor
Line 3
Pump Flow-meter 2 0-50 L/mn Collect Pourer
Special thermal gradient painting TC
Thermal Gradient Tube (steel)
Tungsten crucible
Tungsten Tube
Alumina Tube Thermal Gradient Tube (tungsten)
Figure 2: Scheme of COLIMA-Aerosols facility and instrumentation- Inserts: left: Thermal Gradient Tube (TGT), high: impactor, right: corium load and induction system
Proceedings of the International Conference “Nuclear Energy for New Europe 2005” 079.5
2.2 COLIMA preparation load
A load of oxidic corium representative of a severe accident in a VVER-440 plant is considered (86%wt UO2, 13%wt ZrO2, 1%wt FeO). Since the fission product element were inserted as powder in the load, it has been decided not to add the volatile fission products since they are released at low temperatures for which the fuel microstructure controls release. The oxides due to the minor constituents of the cladding (1% Nb) and of the 25X3MФA steel (2.5% Cr, 0.7% Mn,…) were also taken into account. Table 2 gives the initial composition for COLIMA CA-U3. Note that the nuclear fuel is represented by prototypical materials, it is to say UO2 pellets (Figure 3) ; only Fission Products present in the oxidic phase have been taken into account. For such studies, the use of prototypical materials is fundamental to simulate reactor components behavior [4]. The liquidus temperature (Tliq) of this corium mixture has been calculated at thermodynamic equilibrium with GEMINI-2 [5] and nuclear thermodynamic data bases [6]: Tliq = 2575°C.
W-crucible
UO2 pellets
Powder mixture including ZrO2
Figure 3: preparation of the corium load-COLIMA CA-U3
Chemical formula Experimental mass Experimental mass (g) (%) Fe2O3 21.0223 1.06
Cr2O3 0.5748 0.03
ZrO2 260 13.14
Nb2O5 2.7803 0.14
UO2 (pellets) 1665.8 84.18 SrO 97% 1.851 0.09 BaO 3.0101 0.15
TeO 1.191 0.06 2 Nd O 6.1689 0.31 2 3 CeZrO 9.6379 0.49 4 Pr2O3 2.3362 0.12 Y2O3 1.0592 0.05 Rh2O3 0.9599 0.05 La2O3 2.5595 0.13 ratio UO2/ZrO2 6.4 Total mass 1978.9511 Table 2: corium composition COLIMA CA-U3
Proceedings of the International Conference “Nuclear Energy for New Europe 2005” 079.6 2.3 COLIMA CA-U2 and CA-U3 melting experimental procedure
A general experimental procedure has been defined in common with the Bulgarian team to conduct the COLIMA CA-U2 and CA-U3 experiments. In a first step, the load is heated till the liquidus temperature. In a second step, the aerosol forming plateau at liquidus temperature is maintained and in a last step the aerosols are characterized through the impactor. The surface temperature of the corium load is measured thanks to pyrometer. The surface temperature of COLIMA CA-U3 test is shown on Figure4-a, and the flow measurements on lines 1 (impactor), 2 (heating-up) and 3 (heating plateau) on Figure 4-b. A temperature of 2760°C has been recorded during the heating plateau and the FP release phase. This temperature can be taken as representative of the corium melt. At least, we can note the times of each step with different flow-rate on the line 1 (impactor :5 minutes), line 2 heating up) and line 3 (heating plateau : 45 min).
Temperature (°C) CA-U3 Induction (%) Flow (l/mn) CA-U3 Induction (%)
3000 70.0 30 70.0
deFbilto fwiltre li neN° 2n° et2 N° 3 impac 3I 2800 and 3 65.0
impac 4I 25 deFlbitow impa lincteeur 2600 60.0 n°1-Impactor 60.0 IMPAC2 % induction 2400 55.0 % induction 20
50.0
C) 2200 50.0 n on i m / l 2000 t 15 45.0 nduct i débi % % induction 40.0
température (° 1800 40.0
10 1600 35.0
Flow line n°1/impactor 1400 30.0 impacteur 30.0 5 Flow lfineiltre n° N°22 Flowfilt ren° 3N °3 1200 25.0
1000 20.0 0 20.0 14:30:00 15:00:00 15:30:00 16:00:00 16:30:00 17:00:00 17:30:00 18:00:00 18:30:00 19:00:00 19:30:00 20:00:00 14:30:00 15:00:00 15:30:00 16:00:00 16:30:00 17:00:00 17:30:00 18:00:00 18:30:00 19:00:00 19:30:00 20:00:00 Time h:min:s heure Time h:min:s heure
Figure 4-a (left): surface temperature COLIMA CA-U3, Figure 4-b (right) : flow measurements on lines 1 (impactor), 2 (heating-up) and 3 (heating plateau)
3 POST-TEST ANALYSES AND INTERPRETATION OF COLIMA CA-U3
3.1 Impactor
During the fission products release phase (line n°1), the impactor was under operation and it has been possible to weigh the different collectors (I-) of each stage of the impactor after the experiment. An example of 3 collectors of the impactor are shown in figure 5. It can be noted that different colors were present according to the stages.
Figure 5: impactor samples from the 1.1-2.1 µm, 0.7-1.1 µm and 0.4-0.7 µm aerodynamic- diameter stages.
Proceedings of the International Conference “Nuclear Energy for New Europe 2005” 079.7 Stages 2,4,5,6,7 and F of the impactor have been analyzed by Scanning Electron Microscope with Energy Dispersion Spectrometry (SEM/EDS) analyses. A part of the filter has been taken and then prepared for such analysis. The synthesis of this results are summarized in Figure 6.
CA-U3 Deposits on impactors
1.8 1.6 tes
u Fe n
i 1.4 Sr m
5
1.2 Rh n
d i 1.0 Te e t i Ba s 0.8 po La
de 0.6 Ce g
m Pr
s 0.4 s U
Ma 0.2 0.0 0.2 0.3 0.6 1.0 1.4 1.9 Geometric diameter (µm)
Figure 6: SEM/EDS and granulometry analysis of the impactor-COLIMA CA-U3
It can be noticed that the main size particles is below 1 µm (geometric diameter). Only some elements have reached the impactor: U,Fe, Ba, Te+ trace Sr Rh Pr.
3.2 Filters
During the 3 phases of the COLIMA CA-U3 experiment (heating-up, high temperature plateau, impactor sampling), the line 1,2 and 3 and their respective filters have been used. Their weighing before and after the experiment are shown in table 3.
Filters Before experiment After experiment Net (g) (g) (g) Filter N°11(impactor) 0.093 0.097 0.004 Filter N°2 (heating-up) 0.095 0.149 0.054 Filter N°3 (plateau) 0.095 0.143 0.048
Table 3: weighing of the filters lines 1,2,3- COLIMA CA-U3
SEM/EDS analysis have been performed on filters 2 and 3. For filter 2 (heating up), comparison of the composition measured on the aerosol side and body show that tungsten, iron, rhodium, lanthanum, cerium and neodymium were more present on the sides, whereas tellurium (and praseodymium) is more concentrated in the aerosol body. For filter 3 (heating plateau), the major species are W and O, the minor species are Te and Fe, and there are traces of Rh, Nd, Cr , Sr, Y, Ba.
3.3 Thermal gradient tubes
1 Filter 1 is placed on the diffuser output which is not directed to the impactor.
Proceedings of the International Conference “Nuclear Energy for New Europe 2005” 079.8
A surface chemical attack (mixture of HNO3/HCl) has been performrd to dissolve the fission products till the bulk of the W-tube. All the solutions have a yellow color which means that WO3 has been probably formed. The atomic emission spectroscopy by inductive plasma (ICP-AES) has been used to determine the chemical elements which were present at a given level. Three elements from the load have been found: chromium, iron and uranium. With this technique and estimating the analyzed surface, it has been possible to estimate a superficial concentration in elements without knowing if it was pure element, solutions of elements, oxide phase. Moreover, it appears that the estimated total chromium deposit was above the loaded mass, so the results can only be analyzed as relative values (see figure 7).
CA-U3 Tungsten TGT
300%
250%
200% age density Cr 150% Fe U 100% relative to aver
50% deposit
0% 12345 Level
Figure 7: Relative distribution of deposit density in the tungsten thermal gradient tube Level 1 = bottom – Level 5= top
Nevertheless, Figure 7 indicates that uranium was mainly deposited at the level 3 (around 600°C) while steel vapors were more deposited in the hotter zone. No other fission products have been deposited at this level. For the steel tube, mainly Cr ,Fe, Ba and Te have been deposited
3.4 Corium Two representative samples of the melted corium have been taken to perform SEM EDS analyses. For those analyses, samples are first sectioned using a diamond disk and are then cleaned by ultrasound and dried. The EDS analyses of the corium phases have shown that two families of solid solutions exist (see appendix 20): 1. one enriched in uranium (solid solution U0.8,Zr0.2O2) (the major phase); this phase represents about 95% in volume 2. the other depleted in uranium (solid solution U0.4,Zr06O2) (the minor phase); this phase represents about 5% in volume For the solid solution enriched in uranium (U0.8,Zr0.2O2), it can be noted that La and Ce has been accepted in (U,Zr)O2 solid solution whereas for the solid solution depleted in uranium (solid solution U0.4,Zr06O2), Y, Fe and Sr has been accepted in (U,Zr)O2 solid solution. Comparing the theoretical composition with the measured composition provides information on the behaviour of the various elements
Proceedings of the International Conference “Nuclear Energy for New Europe 2005” 079.9
• Chromium, praseodymium, rhodium, strontium and tellurium has been totally (within analysis precision) released from the melt. • About 50% of the iron (Fe is not a fission product but is present in oxidic corium due to the partial oxidation of steel) has been volatised. • No significant release of barium, cerium, neodymium, niobium, yttrium, lanthanum has been observed. • A small percentage of uranium (and potentially of zirconium) may have released. Even if it is a significant mass, it is below the analysis uncertainties (which is given in percentage of the total mass of the element). From the material analyses, no vapour condensation deposits were found either below the corium crust or on the colder tungsten walls above the free surface.
3.5 Conclusion The main results are presented in Table 4. Fission products have been observed till the impactor, mainly Te, Fe and U. Other aerosols have been deposited on the W-thermal gradient temperature (mainly, Fe, Cr and U). Finally, it has been possible to observe the absence significant amount of Sr, Rh Te and Pr in the corium bulk. Some differences can be noted with a previous evaluation [7] concerning Ba, Sr, Rh and Pr, for the other elements these results are in good agreement.
Element Cr Pr Rh Sr Te Fe Ba Ce Nd Nb Y La U Zr CA-U3 100 100 100 100 100 50 <10 <10 <10 <10 <10 <10 <5 <5 (%) Estimation 10 30 10 100 10- 10 1 10 1 10 10 1 (%) 50
Table 4: CA-U3 experimental results and Ducros’s assessment [7] of the FP and actinides released fraction during a severe accident.
4 PLINIUS/VULCANO experiment
The COMET concept is developed to cool an ex-vessel corium melt by passive injection of coolant water to the bottom of the melt [8]. An advanced version of this concept (CometPCA) uses a porous concrete layer from which the water is supplied to the melt predominantly through a group of porous channels. FZK has successfully performed a series of large scale experiments with simulant corium melts [9]. The COMET cooling concept is under development to arrest and cool ex-vessel corium melts. The concept was designed to be largely independent of different accident scenarios. It is part of previous and current EU Programme (ECOSTAR [9]). A further step in validating the concept is the use of a reactor typical oxide melt (UO2 + ZrO2 + molten concrete). A “unit cell” (see Figure 8) of the cooling device shall be used in the VULCANO facility in a 20 cm ø, 60 cm high crucible with some 40 kg corium melt. In the sequence of the proposed test, the melt is generated and poured from the VULCANO plasma arc furnace into the COMET cooling device, where it is internally heated during the first phase of concrete erosion by induction heating. This establishes the typical initial condition for onset of flooding, when the first layer of concrete is eroded and the passive water injection starts. The next important process is the generation of a permeable porous oxide melt layer by fast evaporation of the injected coolant water, allowing a good contact with the steam/water flow
Proceedings of the International Conference “Nuclear Energy for New Europe 2005” 079.10 from below. Cooling processes until complete solidification will be observed, while induction heating may continue. Post test analysis will characterize the porosity of the solidified melt.
Figure 8: unit cell of the COMET-concept
5 PLINIUS/KROTOS experiment
The KROTOS facility [10] has been transferred from the Joint Research Centre, Ispra to CEA Cadarache. An extension of the PLINIUS platform has been designed to host this facility devoted to Fuel Coolant Interaction studies. KROTOS facility is made of a resistor furnace able to melt about 1 L of corium. After melting, the crucible falls in a transfer tube and is pierced, letting corium flow in the test section. Typical KROTOS test section consists of a water-filled tube installed inside a gas-filled vessel. Pyrotechnic triggers are available to start a steam explosion at a desired instant relative to the melt release.
6 CONCLUSION
PLINIUS is a unique prototypic corium experimental platform open to research groups through FP5 Access to Research Infrastructure activity. A first visit of Bulgarian users lead to the realisation of a fission product release test in the COLIMA facility for the Environmental Impact Assessment of Kozloduy NPP. The result of this aerosol release test by an in-vessel oxidic corium melt is currently under analysis. Thanks to modifications of the COLIMA Facility, it has been possible to perform with success COLIMA CA-U3. Specific instrumentation (Temperature measurement, Thermal gradient temperature, impactor) has allowed to follow the evolution of the corium melt and the departure, transport and deposition of the fission products and also elements of the corium (U,Zr,Fe). Different post-test analyses have been performed (EDS,SEM, ICPMS) to characterize corium and fission products.
Volatile fission products have been observed till the impactor, mainly Te, Fe and U. Other aerosols have been deposited on the W-thermal gradient temperature (Fe, Cr and U). Finally, it has been possible to observe the absence significant amount of Sr, Rh, Te and Pr in the corium bulk.
A test of the COMET concept with prototypic corium and sustained heating is under preparation in the VULCANO facility for German users. The third PLINIUS call, devoted to tests in the KROTOS facility, will be performed with JSI (Slovenia).
Proceedings of the International Conference “Nuclear Energy for New Europe 2005” 079.11 Asides from the scientific and technical outcomes of the individual experiments, this project promotes cooperation between researchers from different EU and Associated countries and provides an unique opportunity for an “hands-on” experience with prototypic corium. For the infrastructure operator point of view, this programme brings new ideas and applications coming from external scientific and technical teams. It is an important networking tool for the building of a Severe Accident European Research Area.
ACKNOWLEDGMENTS
Some results presented in this paper have been obtained in the framework of the European Project "PLINIUS” 5-th Framework Programme (n). The works and efforts of the whole PLINIUS experimental team as well as those of the material analysis team are gratefully acknowledged.
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