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Thermal properties of prototype corium of fast reactor
Nuclear Engineering and Design 322 (2017) 27–31
Contents lists available at ScienceDirect
Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes
Thermal properties of prototype corium of fast reactor N. Mukhamedov a,⇑, M. Skakov b, I. Deryavko b, I. Kukushkin b a b
Shakarim State University, 20a Glinky St., 071400 Semey, Kazakhstan Institute of Atomic Energy of NNC RK, 10 Krasnoarmeyskaya St., 071100 Kurchatov, Kazakhstan
h i g h l i g h t s Developed the technology of manufacture the ingot of the prototype corium of fast reactor. Made the prototype corium of the fast reactor with the sodium coolant. The obtained data on thermophysical properties of prototype corium of fast reactor.
a r t i c l e
i n f o
Article history: Received 14 February 2017 Received in revised form 16 June 2017 Accepted 19 June 2017
Keywords: Prototype corium of a nuclear reactor Thermophysical properties Melting crucible Material carbonization Uranium dioxide Stainless steel
a b s t r a c t The paper is devoted to development and testing of a technology to manufacture the ingot of the prototype corium (resulted from out-of-pile conditions) of fast reactor followed by an experimental determination of the thermophysical properties (TP) (thermal diffusivity a, specific heat capacity Cp, and thermal conductivity k) of such corium at the room temperature (298 K). The data on the thermo-physical properties of corium (melt of structural and fuel materials of the reactor core) will be used to calculate the temperature fields in the modeling the processes of keeping corium inside the power reactor vessel under the conditions of a severe accident. Ó 2017 Elsevier B.V. All rights reserved.
1. Introduction Currently, a great attention is paid to the problem of nuclear reactor safety operation (Alvarenga and Frutuoso, 2015; Fischer et al., 2014; Nguyen et al., 2008; Kang et al., 2014). It is generally agreed that occurrence of an accident accompanied by the fusion of core materials is a rare event. It can occur with a unique combination of circumstances, namely, with the simultaneous rejection of a large number of safety elements, and as a result the operation of cooling systems may break down and loss of the coolant may take place. In this case, the evolving heat of the fission reaction can lead to the destruction of the core geometry and its melting. To fully assess the risk of using reactors and increasing their safety, it is necessary to predict the possible course of an emergency situation, as well as to determine the possible consequences of severe accidents and measures to eliminate them. As is known, the thermophysical properties of the corium (melt of structural and fuel materials of the reactor core), obtained in ⇑ Corresponding author. E-mail address: [email protected] (N. Mukhamedov). http://dx.doi.org/10.1016/j.nucengdes.2017.06.026 0029-5493/Ó 2017 Elsevier B.V. All rights reserved.
experiments simulating severe accidents at nuclear reactors, are extremely important information for identifying mechanisms of severe reactor accidents (Kaity et al., 2012; Skakov et al., 2015a, 2015b, 2017, 2016). That is why the experimental study of the thermophysical properties of the corium is necessary to construct a database that could be used in forecasting the course of severe accidents, and also in computational models. For this purpose, the Institute of Atomic Energy of the National Nuclear Center of Kazakhstan (IAE NNC RK) has been intensively conducting experimental studies on the safety of light water and fast reactors under hypothetically possible severe accidents with melting of core materials. Important results in the study of thermophysical properties of prototype unirradiated corium of light water energy reactors have been obtained in experiments performed on out-of-pile stands of induction heating LAVA-B (Zhdanov et al., 2011) and VCG-135: technology to manufacture both the corium ingot on the stands LAVA-B and VCG-135 and the samples from obtained corium ingots as well as technologies to measure TP of these samples using experimental laboratory facility ‘‘UTFI-2” was developed and tested; the TP data of corium under room temperature was obtained;
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the temperature dependences of the corium TP of various compositions were constructed. Considering that experimental data on TP of corium of energy reactors has huge importance and required at modeling and forecasting the processes of severe accident of a reactor at nuclear power plants (NPP), since 2016 IAE NNC RK conducts similar studies of TP of a prototype corium of a fast reactor with sodium coolant using stand of induction heating VCG-135 and facility ‘‘UTFI-2” for determination of termophysical properties. In this connection the paper provides the first results that were obtained on the development and testing of the technology for manufacture of corium ingot using stand of high-temperature induction heating VCG135, technology to manufacture samples from obtained corium ingot as well as to measure TP of these samples using facility for termophysical studies UTFI-2 under room temperature.
2. Manufacture of corium ingot and samples for TP measurements When developing the technology for manufacturing a prototype corium of a fast reactor, it was considered that melting of the burden from the mixture of cladding tube (stainless steel) and fuel (uranium dioxide) core material of this reactor in graphite melting crucible will be carried out by high-frequency heating of the crucible inside water-cooled inductor windings in working chamber of the VCG-135 stand. Therefore the technology of manufacturing a prototype corium ingot will be closely related to the melting characteristics of the burden melting in graphite heated up to very high temperature (above the fusion temperature (Tmelt) of UO2, which is 2867 °C (Godin et al., 2008). This feature is that the graphite material has the following significant disadvantage: during heating, it evaporates, and evaporation will more intense if the heating temperature is higher. In this connection, the burden materials in the graphite crucible will begin to carbonized long before the melting starts, and the composition of the resulting corium ingot will contain a significant amount of simple and complex carbides that should not be in the real corium. It is obvious that in order to obtain a ‘‘pure” corium a barrier between the crucible and the melting materials should be created for their protection against carbonization. It should be noted that in the previously developed and tested technology for the manufacture of corium ingots of light water reactors (LWR) (Baklanov et al.), a thin (50 mm) layer of zirconium carbide deposited on the inner surface of the crucible was used as a barrier, which allows obtaining a ‘‘pure” prototype corium of light water reactor at heating the crucible with a burden up to the temperature of 2600 °C. The task of creating a protective barrier on the inner surface of a graphite crucible in the manufacture of a fast reactor corium ingot was solved by placing an insert from the tantalum carbide inside the crucible. The main points of this solving (Skakov et al.) were as follows. First, a glass from a sheet of tantalum of 0.8 mm thick using argon-arc welding was made, and then a crucible with an internal diameter equal to the outer diameter of the glass and a depth of the inner cavity equal to the height of the glass was grinded out from the porous graphite (porosity about 30%) and insert the tantalum glass into the cavity of the graphite crucible. Then a glass was filled with foam graphite and the crucible was closed with a graphite cover, which has a through-hole in the center for viewing with a pyrometer. The assembly prepared by this means (Fig. 1) was thermally insulated with graphite felt and placed inside the inductor of VCG-135 stand (Fig. 2) to perform degassing and carbonization annealing.
Fig. 1. Scheme of preparing the assembly of the melting unit for annealing: 1 – graphite crucible; 2 – tantalum glass; 3 – foam graphite; 4 – crucible cover; 5 – tantalum casing of thermometry system.
The degassing annealing of the assembly is carried out at a temperature of 800 °C in a vacuum with a residual pressure of 0.1 kPa for 30 min, and carbonization annealing is performed in helium at a pressure of 0.13 MPa in two steps: first at a temperature of 2500 °C for 60 min, and then at a temperature of about 2900 °C for 10 min. The temperature of 2500 °C in the first stage of carbonization annealing (where the process of tantalum carbonization begins and ends – the process of transition of tantalum to tantalum carbide by the mechanism of reactive diffusion of carbon into metal) is selected taking into account that it is below the Tmelt of Ta-Ta2C eutectic that is equal to 2830 °C in accordance with the state diagram of the tantalum-carbon system (Hackett et al., 2009), which is shown in Fig. 3. The annealing time at this stage was chosen on the basis of the results of test experiments, from which it followed that after a onehour annealing, the average mass composition of carbidized tantalum, according to X-ray diffraction analysis, is close to TaC0.9. Precisely this composition of the pre-stoichiometric tantalum carbide has a maximum Tmelt for the entire homogeneity area of this compound (Fig. 3). The temperature of 2900 °C in the second stage of carbonization annealing (where basically only the alignment of the carbide composition along the cross-section of the wall of the carbidized article takes place within 10 min) has been selected to control the achieved carbonization effect: the absence of melting of the carbided glass at this temperature will mean the use of a graphite crucible with a protective glass inside, to make an ingot of prototype corium by melting the burden at temperatures of about 3000 °C. Using this technology to create a protective barrier, crucibles with TaC inserts were made to perform an experiment on manufacturing an ingot of a prototype corium (a 10-min carbonization annealing of these crucibles at a temperature of 2900 °C showed a perfectly acceptable state of the protective layer in the crucible). The ingot of the prototype corium was manufactured at the VCG135 stand in one of the graphite melting crucibles with a protective insert (Fig. 4a). The burden loaded into this crucible (Fig. 4b)
N. Mukhamedov et al. / Nuclear Engineering and Design 322 (2017) 27–31
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Fig. 2. Stages of preparing for a high-temperature annealing on the stand of induction heating VCG-135: a) working chamber; b) inductor of the working chamber; c) test assembly inside working chamber.
melting of the burden (the result of the manufacture of a corium ingot). The sequence of sample manufacture from the resulting corium ingot included longitudinal cutting of the crucible in a diametric plane (Fig. 5a), drilling with a tubular drill with an abrasive tip of corium cores in the form of cylinders with a diameter of 11 mm (Fig. 5b), cutting out disc samples of 4 mm with plane-parallel ends, grinding and polishing of the end surfaces of the samples (Fig. 5c). It should be noted that the increased macroporosity of the obtained prototype corium ingot, as can be clearly seen in Fig. 5a, did not allow the production of more than two samples for TP measurements of this corium.
3. Research technique and TP measurement results of the corium
Fig. 3. State diagram of tantalum – carbon system (Hackett et al., 2009).
contained 135 g of UO2 (in the form of fine fragments of sintered pellets) and 8.5 g of stainless steel X16N15M3B (in the form of fuel cladding fragments of the BN-350 reactor). Before the melting of the burden, the working chamber was evacuated within 30 min to a pressure of 0.1 kPa at a temperature of 700 °C, after which the burden was heated in argon at a pressure of 1.3 MPa at a temperature of 2850 °C for 10 min. Fig. 4c provides the result of the
TP measurements of the prepared disc samples of the prototype corium were performed on UTFI-2 facility using the technique (Zhdanov et al., 2009), based on the known technique of thermal flare. The main point of the technique consisted in heating one of the ends of a flat disc sample by a short-term action of the heat pulse and recording the time dependence of the temperature T = f(s) at the opposite end of the sample. The scheme of measurement with two samples was used, one of the samples was the main (studied), and the other one was auxiliary (Fig. 6). A disc sample made of sintered UO2 pellet was used as the auxiliary one. Heat impulse was implemented by electric heater located between the studied and auxiliary samples. According to this technique, a thermogram of the test sample (Fig. 7) should be obtained. Such sample parameters as overtemperature DT (the difference between the initial and peak sample temperatures) and the time s1/2 – the time to reach half of the peak surface temperature of the sample, should be determined by the means of this thermogram. These parameters allow calculating required values of thermal diffusivity a, specific heat Cp, and thermal conductivity k of the
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Fig. 4. Stages of manufacture of the prototype corium ingot: a) crucible with TaC-insert; b) crucible with insert and burden; c) crucible with the corium ingot.
Fig. 5. Stages of manufacturing disc samples of prototype corium: a) cross section of crucible with corium; b) bore cores for samples; c) finished samples.
Fig. 7. The typical view of the studied sample thermogram.
Fig. 6. Diagram of samples placement in the UTFI-2 facility: 1 – thermocouple; 2 – heat insulation; 3 – studied sample; 4 – heating element; 5 –auxiliary sample.
studied sample. In fact, to calculate a it is required to know only the values of s1/2 and the sample thickness L, because thermal diffusivity a is equal to 1.38 L2/(p2s1/2), to calculate Cp it is required to know only the values of DT, sample mass M and electrical pulse energy Q, because specific heat Cp is equal to Q/(MDT), and to calculate k it is required to know only sample density q, because thermal conductivity k is equal to the product of aCpq. To determine termophysical properties of the main (studied) sample first of all the calculations of the volumetric density q based on the measurements of its mass M, thickness L and diameter D and then the value calculations of a, Cp, and k is carried out,
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N. Mukhamedov et al. / Nuclear Engineering and Design 322 (2017) 27–31 Table 1 Results of measuring parameters of corium and UO2 samples. Sample identification
L, [mm]
D, [mm]
M, [g]
q, [g/cm3]
e, %
No.1 (main) No.2 (main) UO2 (auxiliary) UO2 (IAEA, 2006)
4.09 ± 0.01 4.13 ± 0.01 4.02 ± 0.01 –
10.70 ± 0.01 10.89 ± 0.07 10.96 ± 0.01 –
3.72 ± 0.01 3.981 ± 0.001 3.87 ± 0.01 –
10.1 ± 0.2 10.00 ± 0.04 10.20 ± 0.02 10.95
4.2 ± 2.5 4.0 ± 1.5 5.2 ± 1.4 –
Table 2 Results of TP measurements of prototype corium samples. 6
m2/s]
Sample
Temperature, [K]
a, [10
No.1 No.2 UO2 UO2 (95% density) (IAEA, 2006)
298 298 298 298
2.8 ± 0.11 3.00 ± 0.02 3.08 ± 0.07 3.08
under the assumption of closeness of geometric and structural parameters of the main and auxiliary samples. The result of TP measurements of each sample is calculated by averaging the processing results of three thermograms, the measurements error is calculated as the standard deviation (mean-square deviation of the individual measurement from the mean value). Table 1 provides the results of thickness, diameter, mass, density and porosity determination of the three samples used for TP measurements as main and auxiliary ones. It is clearly seen from the Table 1 that the condition for the closeness of the geometric and structural parameters of the main and auxiliary samples in the TP measurements is completely fulfilled: the geometric dimensions, masses and densities of the samples of corium and uranium dioxide practically coincide within the limits of the measurement errors. The fulfillment of this condition ensured the maximum reduction of the systematic error in the results of TP measurements of the prototype corium obtained in this procedure under the UTFI2 facility conditions (Table 2) In addition to TP measurements, which were performed in two prototype corium samples, UO2 sample (which previously participated as an auxiliary sample in TP measurements in samples of corium) was also subjected to TP measurements. The results of measurements of the thermal diffusivity a, the specific heat Cp, and the thermal conductivity of this sample were (3.08 ± 0.07) 10 6 m2/s, (236 ± 6) J/(kgK) and (7.4 ± 0.15) W/ (mK) respectively. The obtained values of TP were very close to the TP ones in the samples of the corium, which is naturally related to the closeness of the compositions in the samples of corium and uranium dioxide, because the composition of the prototype corium under studied is uranium dioxide and 6% of stainless steel. 4. Conclusion A set of works on studying the thermophysical properties of the prototype corium of a fast reactor including development and testing of the technique for suppressing the processes of burden carbonization in a graphite melting crucible in the manufacture of corium at the VCG-135 stand, the manufacture of corium samples for UTFI-2 facility conditions and the TP measurement for these samples at room temperature was carried out. According to the results of the performed work, the following general conclusions were made: – for the first time the prototype corium of a fast reactor with a sodium coolant was manufactured; – for the first time data on the thermophysical properties of a such corium at room temperature were obtained;
Cp, [J/(kgK)]
k, [W/(mK)]
230 ± 10 227.1 ± 0.5 236 ± 6 237
6.5 ± 0.17 6.81 ± 0.04 7.4 ± 0.15 7.59
– the results of TP measurements will be used as initial data in the construction of temperature dependences of the termophysical properties of the prototype corium of the fast reactor in the required temperature range. This work was supported by grant of the Committee of Science, Ministry of Education and Science of the Republic of Kazakhstan for 2017 with theme «Investigation of thermal-physical properties of fast reactor core melting». References Alvarenga, M.A.B., Frutuoso, P.F., 2015. Including severe accidents in the design basis of nuclear power plants: an organizational factors perspective after the Fukushima accident. Ann. Nucl. Energy 79, 68–77. Baklanov, V., Skakov, M., Zhdanov, V., et al. Innovative patent of RK No.30667 «Technique of Applying a Protective Barrier Coating of Zirconium Carbide on Inner Surface of Graphite Crucible», Bul. No.12 (I), print. 15.12.2015. Fischer, M., Henning, A., Surmann, R., 2014. Mitigation of severe accidents in AREVA’s Gen 3+ nuclear power plants. Nucl. Eng. Des. 269, 323–329. Godin, Yu., Tenishev, A., Novikov, V., 2008. Physical material science, nuclear fuel materials. MEPhI 2 (6), 604. Hackett, K., Verhoef, Sh., Cutler, R., Shetty, D., 2009. Phase constitution and mechanical properties of carbides in the Ta–C system. J. Am. Ceram. Soc. 92 (10), 2404–2407. Kaity, S., Banerjee, J., Nair, M.R., Ravi, K., Dash, S., Kutty, T.R.G., Kumar, A., Singh, R.P., 2012. Microstructural and thermophysical properties of U-6 wt.% Zr alloy fast reactor application. J. Nucl. Mater. 427, 1–11. Kang, K., Park, R., Hong, S., 2014. An experimental study on layer inversion in the corium pool during a severe accident. Nucl. Eng. Des. 278, 163–170. Nguyen, T., Jaitly, R., Dinnie, K., et al., 2008. Development of severe accident management guidance (SAMG) for the Canadian CANDU 6 nuclear powerplants. Nucl. Eng. Des. 4 (238), 1093–1099. Skakov, M., Mukhamedov, N., Deryavko, I., et al. Petition for Utility Patent RK No.2016/0161.1. «The Technique to Create Tantalum Carbide Inside a Graphite Crucible». Skakov, M., Mukhamedov, N., Deryavko, I., et al., 2015a. Study of thermophysical properties of light-water reactor corium. In: Reports of Inter. Conf. on «Electrical and Electronics Techniques and Applications (EETA), Phuket, August 23–24, 2015», Phuket, p. 75. Skakov, M., Mukhamedov, N., Deryavko, I., et al., 2015b. Temperature dependence of thermophysical properties of light-water reactor prototype corium. In: Reports of Inter. Conf. «Materials and Engineering and Industrial Applications (MEIA), Hong Kong, September 20–21, 2015», Hong Kong, p. 75. Skakov, M., Mukhamedov, N., Vurim, A., et al., 2016. Thermo-physical properties and phase composition of full-scale corium of fast energy reactor. Int. J. ChemTech Res. 12 (9), 725–730. Skakov, M., Mukhamedov, N., Deryavko, I., et al., 2017. Research of structural-phase state of a natural corium of a fast power reactors. Vacuum 141, 216–221. «Thermophysical properties database of materials for light water reactors and heavy water reactors (IAEA-TECDOC-1496)». Printed by the IAEA in Austria, June 2006. Zhdanov, V., Baklanov, V., Sabluk, V., et al., 2009. Technique of determining heattransfer properties of promising fuel samples for VVER. In: Reports of Inter. Conf. «Nuclear Power Engineering of the Republic of Kazakhstan, Kurchatov, Kazakhstan, June 11–13, 2009», Kurchatov, p. 150. Zhdanov, V., Baklanov, V., Bottomley, P.D.W., et al., 2011. Study of the processes of corium-melt retention in the reactor pressure vessel (INVECOR). In: Proceedings of Inter. Congress «Advances in Nuclear Power Plants (ICAPP 2011), Nice, France, May 2–5, 2011», Nice, p. 1300.
Contents lists available at ScienceDirect
Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes
Thermal properties of prototype corium of fast reactor N. Mukhamedov a,⇑, M. Skakov b, I. Deryavko b, I. Kukushkin b a b
Shakarim State University, 20a Glinky St., 071400 Semey, Kazakhstan Institute of Atomic Energy of NNC RK, 10 Krasnoarmeyskaya St., 071100 Kurchatov, Kazakhstan
h i g h l i g h t s Developed the technology of manufacture the ingot of the prototype corium of fast reactor. Made the prototype corium of the fast reactor with the sodium coolant. The obtained data on thermophysical properties of prototype corium of fast reactor.
a r t i c l e
i n f o
Article history: Received 14 February 2017 Received in revised form 16 June 2017 Accepted 19 June 2017
Keywords: Prototype corium of a nuclear reactor Thermophysical properties Melting crucible Material carbonization Uranium dioxide Stainless steel
a b s t r a c t The paper is devoted to development and testing of a technology to manufacture the ingot of the prototype corium (resulted from out-of-pile conditions) of fast reactor followed by an experimental determination of the thermophysical properties (TP) (thermal diffusivity a, specific heat capacity Cp, and thermal conductivity k) of such corium at the room temperature (298 K). The data on the thermo-physical properties of corium (melt of structural and fuel materials of the reactor core) will be used to calculate the temperature fields in the modeling the processes of keeping corium inside the power reactor vessel under the conditions of a severe accident. Ó 2017 Elsevier B.V. All rights reserved.
1. Introduction Currently, a great attention is paid to the problem of nuclear reactor safety operation (Alvarenga and Frutuoso, 2015; Fischer et al., 2014; Nguyen et al., 2008; Kang et al., 2014). It is generally agreed that occurrence of an accident accompanied by the fusion of core materials is a rare event. It can occur with a unique combination of circumstances, namely, with the simultaneous rejection of a large number of safety elements, and as a result the operation of cooling systems may break down and loss of the coolant may take place. In this case, the evolving heat of the fission reaction can lead to the destruction of the core geometry and its melting. To fully assess the risk of using reactors and increasing their safety, it is necessary to predict the possible course of an emergency situation, as well as to determine the possible consequences of severe accidents and measures to eliminate them. As is known, the thermophysical properties of the corium (melt of structural and fuel materials of the reactor core), obtained in ⇑ Corresponding author. E-mail address: [email protected] (N. Mukhamedov). http://dx.doi.org/10.1016/j.nucengdes.2017.06.026 0029-5493/Ó 2017 Elsevier B.V. All rights reserved.
experiments simulating severe accidents at nuclear reactors, are extremely important information for identifying mechanisms of severe reactor accidents (Kaity et al., 2012; Skakov et al., 2015a, 2015b, 2017, 2016). That is why the experimental study of the thermophysical properties of the corium is necessary to construct a database that could be used in forecasting the course of severe accidents, and also in computational models. For this purpose, the Institute of Atomic Energy of the National Nuclear Center of Kazakhstan (IAE NNC RK) has been intensively conducting experimental studies on the safety of light water and fast reactors under hypothetically possible severe accidents with melting of core materials. Important results in the study of thermophysical properties of prototype unirradiated corium of light water energy reactors have been obtained in experiments performed on out-of-pile stands of induction heating LAVA-B (Zhdanov et al., 2011) and VCG-135: technology to manufacture both the corium ingot on the stands LAVA-B and VCG-135 and the samples from obtained corium ingots as well as technologies to measure TP of these samples using experimental laboratory facility ‘‘UTFI-2” was developed and tested; the TP data of corium under room temperature was obtained;
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N. Mukhamedov et al. / Nuclear Engineering and Design 322 (2017) 27–31
the temperature dependences of the corium TP of various compositions were constructed. Considering that experimental data on TP of corium of energy reactors has huge importance and required at modeling and forecasting the processes of severe accident of a reactor at nuclear power plants (NPP), since 2016 IAE NNC RK conducts similar studies of TP of a prototype corium of a fast reactor with sodium coolant using stand of induction heating VCG-135 and facility ‘‘UTFI-2” for determination of termophysical properties. In this connection the paper provides the first results that were obtained on the development and testing of the technology for manufacture of corium ingot using stand of high-temperature induction heating VCG135, technology to manufacture samples from obtained corium ingot as well as to measure TP of these samples using facility for termophysical studies UTFI-2 under room temperature.
2. Manufacture of corium ingot and samples for TP measurements When developing the technology for manufacturing a prototype corium of a fast reactor, it was considered that melting of the burden from the mixture of cladding tube (stainless steel) and fuel (uranium dioxide) core material of this reactor in graphite melting crucible will be carried out by high-frequency heating of the crucible inside water-cooled inductor windings in working chamber of the VCG-135 stand. Therefore the technology of manufacturing a prototype corium ingot will be closely related to the melting characteristics of the burden melting in graphite heated up to very high temperature (above the fusion temperature (Tmelt) of UO2, which is 2867 °C (Godin et al., 2008). This feature is that the graphite material has the following significant disadvantage: during heating, it evaporates, and evaporation will more intense if the heating temperature is higher. In this connection, the burden materials in the graphite crucible will begin to carbonized long before the melting starts, and the composition of the resulting corium ingot will contain a significant amount of simple and complex carbides that should not be in the real corium. It is obvious that in order to obtain a ‘‘pure” corium a barrier between the crucible and the melting materials should be created for their protection against carbonization. It should be noted that in the previously developed and tested technology for the manufacture of corium ingots of light water reactors (LWR) (Baklanov et al.), a thin (50 mm) layer of zirconium carbide deposited on the inner surface of the crucible was used as a barrier, which allows obtaining a ‘‘pure” prototype corium of light water reactor at heating the crucible with a burden up to the temperature of 2600 °C. The task of creating a protective barrier on the inner surface of a graphite crucible in the manufacture of a fast reactor corium ingot was solved by placing an insert from the tantalum carbide inside the crucible. The main points of this solving (Skakov et al.) were as follows. First, a glass from a sheet of tantalum of 0.8 mm thick using argon-arc welding was made, and then a crucible with an internal diameter equal to the outer diameter of the glass and a depth of the inner cavity equal to the height of the glass was grinded out from the porous graphite (porosity about 30%) and insert the tantalum glass into the cavity of the graphite crucible. Then a glass was filled with foam graphite and the crucible was closed with a graphite cover, which has a through-hole in the center for viewing with a pyrometer. The assembly prepared by this means (Fig. 1) was thermally insulated with graphite felt and placed inside the inductor of VCG-135 stand (Fig. 2) to perform degassing and carbonization annealing.
Fig. 1. Scheme of preparing the assembly of the melting unit for annealing: 1 – graphite crucible; 2 – tantalum glass; 3 – foam graphite; 4 – crucible cover; 5 – tantalum casing of thermometry system.
The degassing annealing of the assembly is carried out at a temperature of 800 °C in a vacuum with a residual pressure of 0.1 kPa for 30 min, and carbonization annealing is performed in helium at a pressure of 0.13 MPa in two steps: first at a temperature of 2500 °C for 60 min, and then at a temperature of about 2900 °C for 10 min. The temperature of 2500 °C in the first stage of carbonization annealing (where the process of tantalum carbonization begins and ends – the process of transition of tantalum to tantalum carbide by the mechanism of reactive diffusion of carbon into metal) is selected taking into account that it is below the Tmelt of Ta-Ta2C eutectic that is equal to 2830 °C in accordance with the state diagram of the tantalum-carbon system (Hackett et al., 2009), which is shown in Fig. 3. The annealing time at this stage was chosen on the basis of the results of test experiments, from which it followed that after a onehour annealing, the average mass composition of carbidized tantalum, according to X-ray diffraction analysis, is close to TaC0.9. Precisely this composition of the pre-stoichiometric tantalum carbide has a maximum Tmelt for the entire homogeneity area of this compound (Fig. 3). The temperature of 2900 °C in the second stage of carbonization annealing (where basically only the alignment of the carbide composition along the cross-section of the wall of the carbidized article takes place within 10 min) has been selected to control the achieved carbonization effect: the absence of melting of the carbided glass at this temperature will mean the use of a graphite crucible with a protective glass inside, to make an ingot of prototype corium by melting the burden at temperatures of about 3000 °C. Using this technology to create a protective barrier, crucibles with TaC inserts were made to perform an experiment on manufacturing an ingot of a prototype corium (a 10-min carbonization annealing of these crucibles at a temperature of 2900 °C showed a perfectly acceptable state of the protective layer in the crucible). The ingot of the prototype corium was manufactured at the VCG135 stand in one of the graphite melting crucibles with a protective insert (Fig. 4a). The burden loaded into this crucible (Fig. 4b)
N. Mukhamedov et al. / Nuclear Engineering and Design 322 (2017) 27–31
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Fig. 2. Stages of preparing for a high-temperature annealing on the stand of induction heating VCG-135: a) working chamber; b) inductor of the working chamber; c) test assembly inside working chamber.
melting of the burden (the result of the manufacture of a corium ingot). The sequence of sample manufacture from the resulting corium ingot included longitudinal cutting of the crucible in a diametric plane (Fig. 5a), drilling with a tubular drill with an abrasive tip of corium cores in the form of cylinders with a diameter of 11 mm (Fig. 5b), cutting out disc samples of 4 mm with plane-parallel ends, grinding and polishing of the end surfaces of the samples (Fig. 5c). It should be noted that the increased macroporosity of the obtained prototype corium ingot, as can be clearly seen in Fig. 5a, did not allow the production of more than two samples for TP measurements of this corium.
3. Research technique and TP measurement results of the corium
Fig. 3. State diagram of tantalum – carbon system (Hackett et al., 2009).
contained 135 g of UO2 (in the form of fine fragments of sintered pellets) and 8.5 g of stainless steel X16N15M3B (in the form of fuel cladding fragments of the BN-350 reactor). Before the melting of the burden, the working chamber was evacuated within 30 min to a pressure of 0.1 kPa at a temperature of 700 °C, after which the burden was heated in argon at a pressure of 1.3 MPa at a temperature of 2850 °C for 10 min. Fig. 4c provides the result of the
TP measurements of the prepared disc samples of the prototype corium were performed on UTFI-2 facility using the technique (Zhdanov et al., 2009), based on the known technique of thermal flare. The main point of the technique consisted in heating one of the ends of a flat disc sample by a short-term action of the heat pulse and recording the time dependence of the temperature T = f(s) at the opposite end of the sample. The scheme of measurement with two samples was used, one of the samples was the main (studied), and the other one was auxiliary (Fig. 6). A disc sample made of sintered UO2 pellet was used as the auxiliary one. Heat impulse was implemented by electric heater located between the studied and auxiliary samples. According to this technique, a thermogram of the test sample (Fig. 7) should be obtained. Such sample parameters as overtemperature DT (the difference between the initial and peak sample temperatures) and the time s1/2 – the time to reach half of the peak surface temperature of the sample, should be determined by the means of this thermogram. These parameters allow calculating required values of thermal diffusivity a, specific heat Cp, and thermal conductivity k of the
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Fig. 4. Stages of manufacture of the prototype corium ingot: a) crucible with TaC-insert; b) crucible with insert and burden; c) crucible with the corium ingot.
Fig. 5. Stages of manufacturing disc samples of prototype corium: a) cross section of crucible with corium; b) bore cores for samples; c) finished samples.
Fig. 7. The typical view of the studied sample thermogram.
Fig. 6. Diagram of samples placement in the UTFI-2 facility: 1 – thermocouple; 2 – heat insulation; 3 – studied sample; 4 – heating element; 5 –auxiliary sample.
studied sample. In fact, to calculate a it is required to know only the values of s1/2 and the sample thickness L, because thermal diffusivity a is equal to 1.38 L2/(p2s1/2), to calculate Cp it is required to know only the values of DT, sample mass M and electrical pulse energy Q, because specific heat Cp is equal to Q/(MDT), and to calculate k it is required to know only sample density q, because thermal conductivity k is equal to the product of aCpq. To determine termophysical properties of the main (studied) sample first of all the calculations of the volumetric density q based on the measurements of its mass M, thickness L and diameter D and then the value calculations of a, Cp, and k is carried out,
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N. Mukhamedov et al. / Nuclear Engineering and Design 322 (2017) 27–31 Table 1 Results of measuring parameters of corium and UO2 samples. Sample identification
L, [mm]
D, [mm]
M, [g]
q, [g/cm3]
e, %
No.1 (main) No.2 (main) UO2 (auxiliary) UO2 (IAEA, 2006)
4.09 ± 0.01 4.13 ± 0.01 4.02 ± 0.01 –
10.70 ± 0.01 10.89 ± 0.07 10.96 ± 0.01 –
3.72 ± 0.01 3.981 ± 0.001 3.87 ± 0.01 –
10.1 ± 0.2 10.00 ± 0.04 10.20 ± 0.02 10.95
4.2 ± 2.5 4.0 ± 1.5 5.2 ± 1.4 –
Table 2 Results of TP measurements of prototype corium samples. 6
m2/s]
Sample
Temperature, [K]
a, [10
No.1 No.2 UO2 UO2 (95% density) (IAEA, 2006)
298 298 298 298
2.8 ± 0.11 3.00 ± 0.02 3.08 ± 0.07 3.08
under the assumption of closeness of geometric and structural parameters of the main and auxiliary samples. The result of TP measurements of each sample is calculated by averaging the processing results of three thermograms, the measurements error is calculated as the standard deviation (mean-square deviation of the individual measurement from the mean value). Table 1 provides the results of thickness, diameter, mass, density and porosity determination of the three samples used for TP measurements as main and auxiliary ones. It is clearly seen from the Table 1 that the condition for the closeness of the geometric and structural parameters of the main and auxiliary samples in the TP measurements is completely fulfilled: the geometric dimensions, masses and densities of the samples of corium and uranium dioxide practically coincide within the limits of the measurement errors. The fulfillment of this condition ensured the maximum reduction of the systematic error in the results of TP measurements of the prototype corium obtained in this procedure under the UTFI2 facility conditions (Table 2) In addition to TP measurements, which were performed in two prototype corium samples, UO2 sample (which previously participated as an auxiliary sample in TP measurements in samples of corium) was also subjected to TP measurements. The results of measurements of the thermal diffusivity a, the specific heat Cp, and the thermal conductivity of this sample were (3.08 ± 0.07) 10 6 m2/s, (236 ± 6) J/(kgK) and (7.4 ± 0.15) W/ (mK) respectively. The obtained values of TP were very close to the TP ones in the samples of the corium, which is naturally related to the closeness of the compositions in the samples of corium and uranium dioxide, because the composition of the prototype corium under studied is uranium dioxide and 6% of stainless steel. 4. Conclusion A set of works on studying the thermophysical properties of the prototype corium of a fast reactor including development and testing of the technique for suppressing the processes of burden carbonization in a graphite melting crucible in the manufacture of corium at the VCG-135 stand, the manufacture of corium samples for UTFI-2 facility conditions and the TP measurement for these samples at room temperature was carried out. According to the results of the performed work, the following general conclusions were made: – for the first time the prototype corium of a fast reactor with a sodium coolant was manufactured; – for the first time data on the thermophysical properties of a such corium at room temperature were obtained;
Cp, [J/(kgK)]
k, [W/(mK)]
230 ± 10 227.1 ± 0.5 236 ± 6 237
6.5 ± 0.17 6.81 ± 0.04 7.4 ± 0.15 7.59
– the results of TP measurements will be used as initial data in the construction of temperature dependences of the termophysical properties of the prototype corium of the fast reactor in the required temperature range. This work was supported by grant of the Committee of Science, Ministry of Education and Science of the Republic of Kazakhstan for 2017 with theme «Investigation of thermal-physical properties of fast reactor core melting». References Alvarenga, M.A.B., Frutuoso, P.F., 2015. Including severe accidents in the design basis of nuclear power plants: an organizational factors perspective after the Fukushima accident. Ann. Nucl. Energy 79, 68–77. Baklanov, V., Skakov, M., Zhdanov, V., et al. Innovative patent of RK No.30667 «Technique of Applying a Protective Barrier Coating of Zirconium Carbide on Inner Surface of Graphite Crucible», Bul. No.12 (I), print. 15.12.2015. Fischer, M., Henning, A., Surmann, R., 2014. Mitigation of severe accidents in AREVA’s Gen 3+ nuclear power plants. Nucl. Eng. Des. 269, 323–329. Godin, Yu., Tenishev, A., Novikov, V., 2008. Physical material science, nuclear fuel materials. MEPhI 2 (6), 604. Hackett, K., Verhoef, Sh., Cutler, R., Shetty, D., 2009. Phase constitution and mechanical properties of carbides in the Ta–C system. J. Am. Ceram. Soc. 92 (10), 2404–2407. Kaity, S., Banerjee, J., Nair, M.R., Ravi, K., Dash, S., Kutty, T.R.G., Kumar, A., Singh, R.P., 2012. Microstructural and thermophysical properties of U-6 wt.% Zr alloy fast reactor application. J. Nucl. Mater. 427, 1–11. Kang, K., Park, R., Hong, S., 2014. An experimental study on layer inversion in the corium pool during a severe accident. Nucl. Eng. Des. 278, 163–170. Nguyen, T., Jaitly, R., Dinnie, K., et al., 2008. Development of severe accident management guidance (SAMG) for the Canadian CANDU 6 nuclear powerplants. Nucl. Eng. Des. 4 (238), 1093–1099. Skakov, M., Mukhamedov, N., Deryavko, I., et al. Petition for Utility Patent RK No.2016/0161.1. «The Technique to Create Tantalum Carbide Inside a Graphite Crucible». Skakov, M., Mukhamedov, N., Deryavko, I., et al., 2015a. Study of thermophysical properties of light-water reactor corium. In: Reports of Inter. Conf. on «Electrical and Electronics Techniques and Applications (EETA), Phuket, August 23–24, 2015», Phuket, p. 75. Skakov, M., Mukhamedov, N., Deryavko, I., et al., 2015b. Temperature dependence of thermophysical properties of light-water reactor prototype corium. In: Reports of Inter. Conf. «Materials and Engineering and Industrial Applications (MEIA), Hong Kong, September 20–21, 2015», Hong Kong, p. 75. Skakov, M., Mukhamedov, N., Vurim, A., et al., 2016. Thermo-physical properties and phase composition of full-scale corium of fast energy reactor. Int. J. ChemTech Res. 12 (9), 725–730. Skakov, M., Mukhamedov, N., Deryavko, I., et al., 2017. Research of structural-phase state of a natural corium of a fast power reactors. Vacuum 141, 216–221. «Thermophysical properties database of materials for light water reactors and heavy water reactors (IAEA-TECDOC-1496)». Printed by the IAEA in Austria, June 2006. Zhdanov, V., Baklanov, V., Sabluk, V., et al., 2009. Technique of determining heattransfer properties of promising fuel samples for VVER. In: Reports of Inter. Conf. «Nuclear Power Engineering of the Republic of Kazakhstan, Kurchatov, Kazakhstan, June 11–13, 2009», Kurchatov, p. 150. Zhdanov, V., Baklanov, V., Bottomley, P.D.W., et al., 2011. Study of the processes of corium-melt retention in the reactor pressure vessel (INVECOR). In: Proceedings of Inter. Congress «Advances in Nuclear Power Plants (ICAPP 2011), Nice, France, May 2–5, 2011», Nice, p. 1300.