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Irradiation performance and modeling of HTR-10 coated fuel particles
Nuclear Engineering and Design 236 (2006) 1922–1927
Irradiation performance and modeling of HTR-10 coated fuel particles T.X. Liang a,∗ , H.S. Zhao a , C.H. Tang a , K. Verfondern b a
Institute of Nuclear Energy Technology, Tsinghua University, Beijing 100084, China b Research Center Juelich, 52425 Juelich, Germany
Received 30 November 2004; received in revised form 27 June 2005; accepted 27 January 2006
Abstract The irradiation test of HTR-10 spherical fuel elements was carried out in the Russian IVV-2M research reactor with the irradiation temperature of 1000 ± 50 ◦ C. After the burnup reached 100,000 MWd/t, the irradiation temperature was raised to a higher temperature. The high R/B levels observed during the normal irradiation test were due to manufacture defects of one to four coated particles. Post-irradiation examination indicated that at normal irradiation condition, the pyrolytic carbon (PyC) and silicon carbide (SiC) layers of particles kept their integrity. However, after irradiation at higher temperatures, several types of defects including radial and tangential cracks in SiC layers, cracks in buffer layers, and through coating failure were found, and the failure fraction reached 5.8 × 10−2 . These defects were most likely caused by the higher thermal stresses generated. In this study, PANAMA fuel performance code was used to estimate the heating temperature in the irradiation test. The calculated results showed that when the heating temperature is much higher than 1850 ◦ C, the failure fraction of coated particle can reach the level of 1%. © 2006 Elsevier B.V. All rights reserved.
1. Introduction A characteristic feature of a modular high temperature gascooled reactor (HTGR) is the essential role of the TRISO coated fuel particle acting as a tiny containment and serving as the principal barrier against radionuclide release under normal operation and accident conditions. The quality of the fuel relies on minimizing the number of failures of these particles during reactor operation. The HTR-10 built in China is a modular pebble-bed type HTGR with a thermal power of 10 MW. Spherical fuel elements are employed in the pebble-bed core. The HTR-10 fuel element with 60 mm in diameter consists of matrix graphitic material, in which TRISO coated fuel particles are dispersed. The coated particle is composed of a central UO2 kernel of 500 m in diameter and four layers, which are (1) a low-density porous buffer layer; (2) an inner high-density isotropic pyrolytic carbon (PyC) layer; (3) a silicon carbide (SiC) layer; and (4) an outer high-density isotropic PyC layer. About 20,000 spherical fuel elements were produced for the HTR-10 in 2000 and 2001. Each fuel element contained about 8300 coated
∗
Corresponding author. Tel.: +86 10 89792737; fax: +86 10 69771464. E-mail address: [email protected] (T.X. Liang).
0029-5493/$ – see front matter © 2006 Elsevier B.V. All rights reserved. doi:10.1016/j.nucengdes.2006.01.018
particles. The first two production batches showed a free U fraction of 1.4 × 10−4 and 2.3 × 10−4 (Tang et al., 2002), respectively, this corresponds to one to four defective particles in each irradiated fuel sphere. Four spherical fuel elements were randomly sampled from these two batches for irradiation test. The irradiation test of four HTR-10 spherical fuel elements was carried out in the IVV-2M research reactor at Zarechnyy, Russia. After the burnup of the irradiated fuel elements reached 100,000 MWd/t, a high temperature heating test with the fuel element in capsule 5 was performed in the reactor. Post-irradiation examination indicated that at normal irradiation condition, the PyC and SiC layers of particles kept their integrity. However, after the high temperature irradiation of the fuel element, the failure fraction of coated particles in this sphere reached 5.8%. The heating temperature was expected to be about 1600 ◦ C, but the actual fuel temperature was presumably much higher than 1600 ◦ C because the thermocouple to measure the fuel element failed and it became impossible to record and to control the temperature exactly. One conclusion had been defined that the fuel temperature is much higher than the intended 1600 ◦ C (Tang et al., 2006), but we will ask what is the actual temperature? In this study, the PANAMA fuel performance code was used to find out what the actual temperature during the heating test
T.X. Liang et al. / Nuclear Engineering and Design 236 (2006) 1922–1927
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was by calculating the particle failure fraction under the test conditions and taking the temperature as a parameter. 2. Irradiation samples and irradiation conditions The fabrication technology of irradiation samples includes UO2 kernel preparation by the gel precipitation method, PyC and SiC coating on the UO2 kernels by the CVD process in a fluidized bed, and manufacturing the spherical fuel element through a quasi-isostatic process. Table 1 shows the main characteristics of UO2 kernel and coated particles. Four spherical fuel elements were placed into separate irradiation capsules (nos. 2–5), an in-pile irradiation test was carried out in the Russian IVV-2M research reactor. Each capsule was independently controlled and continuously swept and monitored for volatile fission product release. The irradiation temperature was adjusted by He/Ne mixture gas. The nominal irradiation temperature was 1000 ± 50 ◦ C. Their fast neutron fluence for these four capsules reached 1.10 × 1021 , 1.31 × 1021 , 1.30 × 1021 , and 1.06 × 1021 n/cm2 , respectively. During the irradiation test, the temperature of the fuel element in capsule 3 was increased to 1200 ◦ C for 200 h and to 1250 ◦ C for 200 h, when its burnup reached 38,700 and 57,000 MWd/t, respectively. For the fuel element in capsule 5, the irradiation time was 625 efpd and the irradiation temperature was 1000 ± 50 ◦ C. At the end of the normal irradiation test, the high temperature heating test was carried out in the reactor for 22 h by increasing the thermal neutron flux and adjusting the ratio of He/Ne mixture gas. The post-irradiation examination program included disintegration of the irradiated spheres and determination of the distribution of solid fission products in the matrix material along the sphere diameter, measurement of the failure fraction of the loose coated fuel particles obtained from the ball disintegration by the Irradiated Microsphere Gamma Analyzer (IMGA), and examination of the loose particles by ceramography. 3. Irradiation testing results and discussion The measurement of R/B of Kr85m of the irradiated fuel elements in these four capsules as a function of the burnup is shown in Fig. 1. The free U levels of the fuel elements in the first two
Fig. 1. R/B of Kr85m of the irradiated fuel elements as a function of the burnup (the dotted line represents the R/B of Kr85m for heating test).
production batch was 1.4–2.3 × 10−4 , it means that there might be 1–4 defect coated fuel particles in each irradiated fuel element due to the manufacture. After irradiation, the R/B of Kr85m of fuel elements in capsule 2 and capsule 5 are in the range of 5 × 10−6 to 8 × 10−5 , the comparatively high R/B level was resulted from the higher manufacturing defects. All the release rate curves kept small fluctuation under average fuel temperature, ∼1000 ◦ C. A little ascension of the curves at a point of about 53,000 MWd/t came most likely from the fluctuation of the irradiation temperature. Heating the fuel element in the capsule 3 to 1200 ◦ C for 200 h with the burnup of 38,700 MWd/t, and to 1250 ◦ C for 200 h with the burnup of 57,300 MWd/t caused an increase of about one order of Kr85m release. When the temperature returned to 1050 ◦ C, the release rate was restored to the initial value. Fig. 2 shows the R/B of Kr85m of the irradiated fuel elements as a function of irradiation temperature. It can be seen that the R/B value increased as the irradiation temperature increasing. The irradiating testing of one fuel element in capsule 4 ended up early owing to something wrong in the gas loop of the capsule after irradiated time reached 5349 effective hours (fuel burnup: 37,000 MWd/t). Fig. 3 is the photographs of the irradiated fuel elements. The integrity of the fuel element in capsule 4 was lost. So far the reason of fuel ball damage was not found out. There were not any cracks or corrosive spots in the other three fuel elements, even the fuel element in capsule 5 undergoing heating test in the reactor. The diameters of these three integrity fuel balls perpendicular and parallel to pressing direction were measured before and after the irradiation test, respectively. The shrinkage of irradiated
Table 1 Main characteristics of UO2 kernel and particle coating of HTR-10 first loading fuel UO2 diameter (m) UO2 density (g/cm3 ) O/U ratio Buffer PyC layer thickness (m) I-PyC layer thickness (m) SiC layer thickness (m) O-PyC layer thickness (m) Buffer PyC layer density (g/cm3 ) I-PyC layer density (g/cm3 ) SiC layer density (g/cm3 ) O-PyC layer density (g/cm3 ) I-PyC and O-PyC layer OPTAF
498 10.9 2.00 95 42 37 42 0.98 1.86 3.20 1.87 1.03
Fig. 2. Effect of irradiation temperature on the R/B of Kr85m .
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Fig. 3. Appearance of irradiated fuel elements in (a) capsule 2; (b) capsule 3; (c) capsule 5; and (d) capsule 4.
fuel spheres was about 6% on average. Their diameter change between the perpendicular and the parallel to pressing direction was not different. Fig. 4 shows the microstructure of the coated particles before and after irradiation. The PyC and SiC layers in these particles were kept intact. No cracks and other radiation-induced defects in the layers have been observed. The buffer layers had a sufficient thickness and no damage was found in the I-PyC layers which may have been induced by fission fragments. After the heating test, various types of defects in coated particles like radial and tangential cracks in SiC layers, cracks in buffer layers, and through coating failures have been discovered in the fuel element in capsule 5 (Tang et al., 2006). These defects were most likely caused by the high stress generated by the very high heating temperature. There are many factors conceivable which can lead to the appearance of defects in UO2 kernel and coating layers at high temperature. 3.1. UO2 kernel Point defects, gaseous and solid fission products and oxygen are formed in UO2 kernel under irradiation. Gas bubbles in the fuel lead to a swelling of the kernel and reduce the thermal conductivity of the kernel material. Experimental data suggest a volume increase of 0.6–1.5% V/V produced per 10,000 MWd/t of burnup, thus for 100,000 MWd/t burnup, a 6–15% increase in volume of kernel will be expected (Olander, 1985). A significant
kernel swelling can reduce the void volume in the coated particle and, under certain conditions, cause kernel/coating mechanical interaction. The yield of Xe and Kr gas is about 0.31 per fission. Oxygen liberated from the fission process can form CO gas in the fuel. From the two oxygen atoms which are released per fission, about 1.4 atoms are tied up by rare earth fission product, the remaining oxygen reacts with carbon to produce CO. The empirical relationship for the number of oxygen atoms per fission, O/F, under irradiation conditions is given by the equation (Proksh et al., 1982): O 8500 log + 2 log ti (1) = −10.08 − F Ti where Ti is the irradiation temperature (K) and ti the irradiation time (s). The corresponding equation during heating is O 8500 log = −10.08 − F Ti 1 1 +2 log ti − 4040 − (2) Th Ti + 75 where Th is the irradiation temperature (K). Some fission products, especially Pd can migrate into the coating and may get in contact with the SiC layer, causing it to corrode. From experimental data, a silicon carbide corrosion
Fig. 4. Microstructure of the coated particles from capsule 2 before and after irradiation.
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Table 2 Coated particle failure fractions after irradiation and heating test Capsule no.
In capsule 2
Failed particles and failure fraction
Number of failed particles in 2117 coated particles
Failure fraction
Number of failed particles in 2014 coated particles
Failure fraction
After irradiate to 100,000 MWd/t After heating test
2
9.45 × 10−4
1
4.97 × 10−4
rates (m/h) was derived (Tiegs, 1982): −1.79 × 105 5 S = 2.613 × 10 exp 8.314T
In capsule 3
(3)
where T is the irradiation temperature (K). Eqs. (1)–(3) indicate that with increasing temperature the gas pressure inside the coated particles increases and degradation of the silicon carbide becomes stronger, both effects leading to an enhanced probability of particle failure. 3.2. Buffer layer Under irradiation, the buffer layer will be densified at early stage, and when the burnup reached larger than 40,000 MWd/t, cracks will be formed. 3.3. I-PyC and O-PyC layer PyC is a brittle material. Under irradiation and thermal impact, its anisotropy factor (BAF) and volume will change. The creep effect can relax internal stress. Due to neutron irradiation, the PyC layer will undergo shrinkage and, as a result, tensile stress is developed. This shrinkage can produce compression stress in the SiC layer that is useful to reduce the tensile stress in this layer. If a crack is generated on the interface of I-PyC/SiC, stress concentration will exist at crack tip on the SiC internal face. The SiC layer will be in tension state if a particle has a cracked I-PyC. 3.4. SiC layer Silicon carbide exhibits a high resistance against irradiation. At high temperature (above 1600 ◦ C), thermal decomposition of the SiC will occur. The IMGA experimental device provides the capability of making statistically accurate failure fraction measurement on irradiated coated fuel particles. A judgment between an intact particle and a failed particle is based on the activity ratio of two isotopes of cesium (Cs134 and Cs137 ) and Ce144 . The actual failure fraction is based on the ratio of the activity of a volatile fission product to a non-volatile fission product. The boiling point of cesium is 678 ◦ C, but 3470 ◦ C for cerium. Thus, in a high temperature irradiation test of one HTR fuel element, cesium will escape much more readily from a defective coating than cerium. Therefore, a measurement of the activity ratio of Cs137 or Cs134 to Ce144 can identify a failed particle.
In capsule 5 Number of failed particles in 2189 coated particles
Failure fraction
127
5.8 × 10−2
After irradiation, the spherical fuel elements were disintegrated by electrolysis to obtain loose coated fuel particles. From the approximately 8300 coated particles in the fuel element, a number of coated particles were sampled and inspected by IMGA. The Cs137 to Ce144 activity ratios of these particles were obtained (Tang et al., 2004). The resulting failure fractions are listed in Table 2. With regard to the free U levels of fuel elements before irradiation and the measured R/B curves in Fig. 1, the one to two particle failures in the spherical fuel elements in capsule 2 and capsule 3 were originated from manufacture. It means that no extra coated particles failed by the normal irradiation. However, in capsule 5, 127 failed particles were caused by the high temperature heating. This heating test was expected to be performed at a temperature of 1600 ◦ C. However, the thermocouple to measure the fuel temperature failed and the actual fuel temperature was presumably much higher than 1600 ◦ C. The question now was: what was the actual temperature in the heating test? 4. Modeling of the irradiation behavior of TRISO coated particles 4.1. PANAMA code The PANAMA model describes the performance of TRISO coated particles during irradiation and under accident conditions (i.e., heating). Two failure modes are being considered. At temperatures below 2000 ◦ C, the failure mechanism based on the pressure vessel model is dominant, whereas at higher temperatures, thermal decomposition of the silicon carbide becomes dominant (Verfondern and Nabielek, 1985). The following equations are cited from Verfondern and Nabielek (1985). The total particle failure fraction is given by Φtotal = 1 − (1 − Φ1 )(1 − Φ2 )
(4)
where Φ1 and Φ2 correspond to the failure fractions due to pressure vessel failure and thermal decomposition, respectively. The SiC layer as the actual retention barrier for fission products represents the pressure vessel wall. In this code, a particle fails when stress in the SiC layer induced by the internal gas pressure exceeds the SiC strength. At time t after the start of the heating test, the failure fraction is given by m σt Φ1 (t, T ) = 1 − exp − ln 2 (5) σ0 σ t is the stress at time t due to the internal gas pressure, σ 0 the remaining strength of the SiC at the end of irradiation. Strength
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Table 3 Failure fraction during the normal irradiation by PANAMA code Irradiation temperature (◦ C) Irradiation time (efpd) Burnup (MWd/t) Failure fraction
1000 625 100,000 2.30E−6
values are scattered in a Weibull distribution, m is the Weibull parameter. σt =
0.5rP d0 1 + dct0
(6)
where r is the average radius of the SiC layer, P the internal gas pressure, c the corrosion rate as a function of temperature, and d0 the original thickness of the SiC layer. P=
BRT (0.31Fd + Of ) Vm Vf /Vk
(7)
where B is burnup, R the gas constant, Fd the relative fraction of fission gas releases, Of the number of oxygen atoms per fission, Vm the molar volume of the kernel, Vk the kernel volume, and Vf the void fraction corresponding to 50% of the buffer layer. At very high temperature beyond 2000 ◦ C, TRISO particles will mainly fail due to the thermal decomposition of the SiC. The failure fraction is given by Φ2 (t, T ) = 1 − exp(−αξ β )
(8)
where α and β are empirical constants, ξ is a so-called action integral which includes the decomposition rate as a function of temperature. For a spherical fuel element, the values for the constants were empirically derived to be α = 1 × 10−4 and β = 4. 4.2. Results One of the most important input parameters for PANAMA is the SiC strength and its Weibull parameter. For HTR-10 coated particles, the SiC strength is 834 MPa and its Weibull parameter is 5.7. 4.2.1. Failure fraction during the normal irradiation Table 3 shows the calculated failure fraction of coated particles at normal irradiation conditions, i.e., irradiation temperature is 1000 ◦ C, irradiation time is 625 efpd. The calculated stresses on each layer of an intact particle (Sawa et al., 1999) have shown that the maximum tensile stress on the SiC layer was almost zero when the burnup was near 50,000 MWd/t. When irradiation starts, a compressive stress acts on the SiC layer by fast neutron-induced shrinkage of the PyC layers. The PANAMA code does not take into account such a behavior of PyC layers, which is a conservative approach. In this case, the calculated result should be higher than that of irradiation result, but in this paper, the calculated result is lower. This is due to the fact that the first production batch had a higher free uranium level (1.4 × 10−4 ). As shown in Fig. 1, the fission gas
Fig. 5. Calculated particle failure fraction vs. heating time for different heating temperatures.
release rate curves during in-pile irradiation testing kept small fluctuation under average fuel temperature, ∼1000 ◦ C. It means that no coated particles failed caused by normal irradiation. 4.2.2. Failure fraction under heating test At the end of normal irradiation at 1000 ◦ C for 625 efpd, the fuel element in capsule 5 was expected to expose in a heating test in the reactor to the temperature of 1600 ◦ C for 200 h. Unfortunately, when the in-pile heating test underwent to 22 h (soaking time was about 5 h), R/B value reached to 3 × 10−2 , so the heating test has to be stopped. After lots of discussion, one conclusion that the thermocouples failed was drawn. Because the thermocouples failed, here we estimate that the actual fuel temperature filled in the range of 1600–1900 ◦ C, and use PANAMA code to calculate the failure fraction. The calculation results are given in Fig. 5. It shows that the failure fraction is 5.7 × 10−4 for a heating temperature of 1600 ◦ C. When the heating temperature is higher than 1850 ◦ C, the calculated failure fraction exceeds the level of 1 × 10−2 . A great deal of experimental data indicated that the calculation results of PANAMA code exhibit very good agreement with coated fuel particle irradiation data when the heating temperature is above 1800 ◦ C (Verfondern et al., 2001). Below 1800 ◦ C, the attribution of SiC thermal decomposition to CP fuel failure is not as large as the predication by PANAMA code. Annealing can relax tensile stress in the SiC layer, this is another reason resulted in the overestimation of PANAMA code. It can be concluded that the heating temperature of the HTR-10 spherical fuel elements in the Russian IVV-2M research reactor exceeded 1850 ◦ C. 5. Summary The comparatively high R/B level measured during the normal irradiation test in the IVV-2M reactor with the fuel element of capsule 5 can be traced back to 1–4 manufacture-induced defects of coated particles, which were taken from the first two production batches that had an average free uranium fraction of 1.4–2.3 × 10−4 . Post-irradiation examination indicated that at normal irradiation conditions, the PyC and SiC layers of particles which have
T.X. Liang et al. / Nuclear Engineering and Design 236 (2006) 1922–1927
undergone in-pile test kept their integrity. Any cracks and other radiation-induced defects in the layers were not observed. However, in the heat-up test, several types of defects including radial and tangential cracks in the SiC layers, cracks in the buffer layers and through coating failures were found. These defects were most likely caused by the obviously too high heating temperature. The heating test led to a failure fraction reaching 5.8 × 10−2 . The PANAMA fuel performance code was used to estimate the real temperature during the heating test. The calculated results show that the heating temperature of the HTR-10 spherical fuel elements in the Russian IVV-2M research reactor exceeded 1850 ◦ C. References Olander, D.R., 1985. Fundamental aspects of nuclear reactor fuel element. Technical Information Center, Springfield. Proksh, E., Strigl, A., Nabielek, H., 1982. Production of CO during burnup of UO2 kerneled HTR fuel particles. J. Nucl. Mater. 107, 280–283.
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Sawa, K., Tobita, T., Mogi, H., 1999. Fabrication of the first loading fuel of the high temperature engineering test reactor. J. Nucl. Sci. Technol. 36, 683–686. Tang, C.H., Tang, Y.P., Zhu, J.G., 2002. Design and manufacture of the fuel element for the 10 MW high temperature gas-cooled reactor. Nucl. Eng. Des. 218, 91–102. Tang, C.H., Fu, X.M., Zhu, J.G., 2004. The behavior of HTR-10 fuel under irradiation. In: Proceedings of the 2nd International Topic Meeting on High Temperature Reactor Technology, Beijing, China, 22–24 September. Tang, C.H., Fu, X.M., Zhu, J.G., Liang, T.X., Koshcheyev, K.N., Kozlov, A.V., Karlov, O.G., Degaltsev, Y.G., Vasiliev, V.I., 2006. Fuel irradiation of the first batches produced for the Chinese HTR-10. Nucl. Eng. Des. 236, 107–113. Tiegs, T.N., 1982. Fission product Pd–SiC interaction irradiated coated particle fuels. Nucl. Technol. 57, 389–398. Verfondern, K., Nabielek, H., 1985. PANAMA-Ein Rechenprogramm zur Vorhersage des Partikelbruchanteil von TRISO-Partikeln unter Stierfallbedingungen. FZJ report Juel-Spez-298, Research Center Juelich. Verfondern, K., Sumita, J., Ueta, S., Sawa, K., 2001. Modeling of fuel performance and metallic fission product release behavior during HTTR normal operating conditions. Nucl. Eng. Des. 210, 225–238.
Irradiation performance and modeling of HTR-10 coated fuel particles T.X. Liang a,∗ , H.S. Zhao a , C.H. Tang a , K. Verfondern b a
Institute of Nuclear Energy Technology, Tsinghua University, Beijing 100084, China b Research Center Juelich, 52425 Juelich, Germany
Received 30 November 2004; received in revised form 27 June 2005; accepted 27 January 2006
Abstract The irradiation test of HTR-10 spherical fuel elements was carried out in the Russian IVV-2M research reactor with the irradiation temperature of 1000 ± 50 ◦ C. After the burnup reached 100,000 MWd/t, the irradiation temperature was raised to a higher temperature. The high R/B levels observed during the normal irradiation test were due to manufacture defects of one to four coated particles. Post-irradiation examination indicated that at normal irradiation condition, the pyrolytic carbon (PyC) and silicon carbide (SiC) layers of particles kept their integrity. However, after irradiation at higher temperatures, several types of defects including radial and tangential cracks in SiC layers, cracks in buffer layers, and through coating failure were found, and the failure fraction reached 5.8 × 10−2 . These defects were most likely caused by the higher thermal stresses generated. In this study, PANAMA fuel performance code was used to estimate the heating temperature in the irradiation test. The calculated results showed that when the heating temperature is much higher than 1850 ◦ C, the failure fraction of coated particle can reach the level of 1%. © 2006 Elsevier B.V. All rights reserved.
1. Introduction A characteristic feature of a modular high temperature gascooled reactor (HTGR) is the essential role of the TRISO coated fuel particle acting as a tiny containment and serving as the principal barrier against radionuclide release under normal operation and accident conditions. The quality of the fuel relies on minimizing the number of failures of these particles during reactor operation. The HTR-10 built in China is a modular pebble-bed type HTGR with a thermal power of 10 MW. Spherical fuel elements are employed in the pebble-bed core. The HTR-10 fuel element with 60 mm in diameter consists of matrix graphitic material, in which TRISO coated fuel particles are dispersed. The coated particle is composed of a central UO2 kernel of 500 m in diameter and four layers, which are (1) a low-density porous buffer layer; (2) an inner high-density isotropic pyrolytic carbon (PyC) layer; (3) a silicon carbide (SiC) layer; and (4) an outer high-density isotropic PyC layer. About 20,000 spherical fuel elements were produced for the HTR-10 in 2000 and 2001. Each fuel element contained about 8300 coated
∗
Corresponding author. Tel.: +86 10 89792737; fax: +86 10 69771464. E-mail address: [email protected] (T.X. Liang).
0029-5493/$ – see front matter © 2006 Elsevier B.V. All rights reserved. doi:10.1016/j.nucengdes.2006.01.018
particles. The first two production batches showed a free U fraction of 1.4 × 10−4 and 2.3 × 10−4 (Tang et al., 2002), respectively, this corresponds to one to four defective particles in each irradiated fuel sphere. Four spherical fuel elements were randomly sampled from these two batches for irradiation test. The irradiation test of four HTR-10 spherical fuel elements was carried out in the IVV-2M research reactor at Zarechnyy, Russia. After the burnup of the irradiated fuel elements reached 100,000 MWd/t, a high temperature heating test with the fuel element in capsule 5 was performed in the reactor. Post-irradiation examination indicated that at normal irradiation condition, the PyC and SiC layers of particles kept their integrity. However, after the high temperature irradiation of the fuel element, the failure fraction of coated particles in this sphere reached 5.8%. The heating temperature was expected to be about 1600 ◦ C, but the actual fuel temperature was presumably much higher than 1600 ◦ C because the thermocouple to measure the fuel element failed and it became impossible to record and to control the temperature exactly. One conclusion had been defined that the fuel temperature is much higher than the intended 1600 ◦ C (Tang et al., 2006), but we will ask what is the actual temperature? In this study, the PANAMA fuel performance code was used to find out what the actual temperature during the heating test
T.X. Liang et al. / Nuclear Engineering and Design 236 (2006) 1922–1927
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was by calculating the particle failure fraction under the test conditions and taking the temperature as a parameter. 2. Irradiation samples and irradiation conditions The fabrication technology of irradiation samples includes UO2 kernel preparation by the gel precipitation method, PyC and SiC coating on the UO2 kernels by the CVD process in a fluidized bed, and manufacturing the spherical fuel element through a quasi-isostatic process. Table 1 shows the main characteristics of UO2 kernel and coated particles. Four spherical fuel elements were placed into separate irradiation capsules (nos. 2–5), an in-pile irradiation test was carried out in the Russian IVV-2M research reactor. Each capsule was independently controlled and continuously swept and monitored for volatile fission product release. The irradiation temperature was adjusted by He/Ne mixture gas. The nominal irradiation temperature was 1000 ± 50 ◦ C. Their fast neutron fluence for these four capsules reached 1.10 × 1021 , 1.31 × 1021 , 1.30 × 1021 , and 1.06 × 1021 n/cm2 , respectively. During the irradiation test, the temperature of the fuel element in capsule 3 was increased to 1200 ◦ C for 200 h and to 1250 ◦ C for 200 h, when its burnup reached 38,700 and 57,000 MWd/t, respectively. For the fuel element in capsule 5, the irradiation time was 625 efpd and the irradiation temperature was 1000 ± 50 ◦ C. At the end of the normal irradiation test, the high temperature heating test was carried out in the reactor for 22 h by increasing the thermal neutron flux and adjusting the ratio of He/Ne mixture gas. The post-irradiation examination program included disintegration of the irradiated spheres and determination of the distribution of solid fission products in the matrix material along the sphere diameter, measurement of the failure fraction of the loose coated fuel particles obtained from the ball disintegration by the Irradiated Microsphere Gamma Analyzer (IMGA), and examination of the loose particles by ceramography. 3. Irradiation testing results and discussion The measurement of R/B of Kr85m of the irradiated fuel elements in these four capsules as a function of the burnup is shown in Fig. 1. The free U levels of the fuel elements in the first two
Fig. 1. R/B of Kr85m of the irradiated fuel elements as a function of the burnup (the dotted line represents the R/B of Kr85m for heating test).
production batch was 1.4–2.3 × 10−4 , it means that there might be 1–4 defect coated fuel particles in each irradiated fuel element due to the manufacture. After irradiation, the R/B of Kr85m of fuel elements in capsule 2 and capsule 5 are in the range of 5 × 10−6 to 8 × 10−5 , the comparatively high R/B level was resulted from the higher manufacturing defects. All the release rate curves kept small fluctuation under average fuel temperature, ∼1000 ◦ C. A little ascension of the curves at a point of about 53,000 MWd/t came most likely from the fluctuation of the irradiation temperature. Heating the fuel element in the capsule 3 to 1200 ◦ C for 200 h with the burnup of 38,700 MWd/t, and to 1250 ◦ C for 200 h with the burnup of 57,300 MWd/t caused an increase of about one order of Kr85m release. When the temperature returned to 1050 ◦ C, the release rate was restored to the initial value. Fig. 2 shows the R/B of Kr85m of the irradiated fuel elements as a function of irradiation temperature. It can be seen that the R/B value increased as the irradiation temperature increasing. The irradiating testing of one fuel element in capsule 4 ended up early owing to something wrong in the gas loop of the capsule after irradiated time reached 5349 effective hours (fuel burnup: 37,000 MWd/t). Fig. 3 is the photographs of the irradiated fuel elements. The integrity of the fuel element in capsule 4 was lost. So far the reason of fuel ball damage was not found out. There were not any cracks or corrosive spots in the other three fuel elements, even the fuel element in capsule 5 undergoing heating test in the reactor. The diameters of these three integrity fuel balls perpendicular and parallel to pressing direction were measured before and after the irradiation test, respectively. The shrinkage of irradiated
Table 1 Main characteristics of UO2 kernel and particle coating of HTR-10 first loading fuel UO2 diameter (m) UO2 density (g/cm3 ) O/U ratio Buffer PyC layer thickness (m) I-PyC layer thickness (m) SiC layer thickness (m) O-PyC layer thickness (m) Buffer PyC layer density (g/cm3 ) I-PyC layer density (g/cm3 ) SiC layer density (g/cm3 ) O-PyC layer density (g/cm3 ) I-PyC and O-PyC layer OPTAF
498 10.9 2.00 95 42 37 42 0.98 1.86 3.20 1.87 1.03
Fig. 2. Effect of irradiation temperature on the R/B of Kr85m .
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Fig. 3. Appearance of irradiated fuel elements in (a) capsule 2; (b) capsule 3; (c) capsule 5; and (d) capsule 4.
fuel spheres was about 6% on average. Their diameter change between the perpendicular and the parallel to pressing direction was not different. Fig. 4 shows the microstructure of the coated particles before and after irradiation. The PyC and SiC layers in these particles were kept intact. No cracks and other radiation-induced defects in the layers have been observed. The buffer layers had a sufficient thickness and no damage was found in the I-PyC layers which may have been induced by fission fragments. After the heating test, various types of defects in coated particles like radial and tangential cracks in SiC layers, cracks in buffer layers, and through coating failures have been discovered in the fuel element in capsule 5 (Tang et al., 2006). These defects were most likely caused by the high stress generated by the very high heating temperature. There are many factors conceivable which can lead to the appearance of defects in UO2 kernel and coating layers at high temperature. 3.1. UO2 kernel Point defects, gaseous and solid fission products and oxygen are formed in UO2 kernel under irradiation. Gas bubbles in the fuel lead to a swelling of the kernel and reduce the thermal conductivity of the kernel material. Experimental data suggest a volume increase of 0.6–1.5% V/V produced per 10,000 MWd/t of burnup, thus for 100,000 MWd/t burnup, a 6–15% increase in volume of kernel will be expected (Olander, 1985). A significant
kernel swelling can reduce the void volume in the coated particle and, under certain conditions, cause kernel/coating mechanical interaction. The yield of Xe and Kr gas is about 0.31 per fission. Oxygen liberated from the fission process can form CO gas in the fuel. From the two oxygen atoms which are released per fission, about 1.4 atoms are tied up by rare earth fission product, the remaining oxygen reacts with carbon to produce CO. The empirical relationship for the number of oxygen atoms per fission, O/F, under irradiation conditions is given by the equation (Proksh et al., 1982): O 8500 log + 2 log ti (1) = −10.08 − F Ti where Ti is the irradiation temperature (K) and ti the irradiation time (s). The corresponding equation during heating is O 8500 log = −10.08 − F Ti 1 1 +2 log ti − 4040 − (2) Th Ti + 75 where Th is the irradiation temperature (K). Some fission products, especially Pd can migrate into the coating and may get in contact with the SiC layer, causing it to corrode. From experimental data, a silicon carbide corrosion
Fig. 4. Microstructure of the coated particles from capsule 2 before and after irradiation.
T.X. Liang et al. / Nuclear Engineering and Design 236 (2006) 1922–1927
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Table 2 Coated particle failure fractions after irradiation and heating test Capsule no.
In capsule 2
Failed particles and failure fraction
Number of failed particles in 2117 coated particles
Failure fraction
Number of failed particles in 2014 coated particles
Failure fraction
After irradiate to 100,000 MWd/t After heating test
2
9.45 × 10−4
1
4.97 × 10−4
rates (m/h) was derived (Tiegs, 1982): −1.79 × 105 5 S = 2.613 × 10 exp 8.314T
In capsule 3
(3)
where T is the irradiation temperature (K). Eqs. (1)–(3) indicate that with increasing temperature the gas pressure inside the coated particles increases and degradation of the silicon carbide becomes stronger, both effects leading to an enhanced probability of particle failure. 3.2. Buffer layer Under irradiation, the buffer layer will be densified at early stage, and when the burnup reached larger than 40,000 MWd/t, cracks will be formed. 3.3. I-PyC and O-PyC layer PyC is a brittle material. Under irradiation and thermal impact, its anisotropy factor (BAF) and volume will change. The creep effect can relax internal stress. Due to neutron irradiation, the PyC layer will undergo shrinkage and, as a result, tensile stress is developed. This shrinkage can produce compression stress in the SiC layer that is useful to reduce the tensile stress in this layer. If a crack is generated on the interface of I-PyC/SiC, stress concentration will exist at crack tip on the SiC internal face. The SiC layer will be in tension state if a particle has a cracked I-PyC. 3.4. SiC layer Silicon carbide exhibits a high resistance against irradiation. At high temperature (above 1600 ◦ C), thermal decomposition of the SiC will occur. The IMGA experimental device provides the capability of making statistically accurate failure fraction measurement on irradiated coated fuel particles. A judgment between an intact particle and a failed particle is based on the activity ratio of two isotopes of cesium (Cs134 and Cs137 ) and Ce144 . The actual failure fraction is based on the ratio of the activity of a volatile fission product to a non-volatile fission product. The boiling point of cesium is 678 ◦ C, but 3470 ◦ C for cerium. Thus, in a high temperature irradiation test of one HTR fuel element, cesium will escape much more readily from a defective coating than cerium. Therefore, a measurement of the activity ratio of Cs137 or Cs134 to Ce144 can identify a failed particle.
In capsule 5 Number of failed particles in 2189 coated particles
Failure fraction
127
5.8 × 10−2
After irradiation, the spherical fuel elements were disintegrated by electrolysis to obtain loose coated fuel particles. From the approximately 8300 coated particles in the fuel element, a number of coated particles were sampled and inspected by IMGA. The Cs137 to Ce144 activity ratios of these particles were obtained (Tang et al., 2004). The resulting failure fractions are listed in Table 2. With regard to the free U levels of fuel elements before irradiation and the measured R/B curves in Fig. 1, the one to two particle failures in the spherical fuel elements in capsule 2 and capsule 3 were originated from manufacture. It means that no extra coated particles failed by the normal irradiation. However, in capsule 5, 127 failed particles were caused by the high temperature heating. This heating test was expected to be performed at a temperature of 1600 ◦ C. However, the thermocouple to measure the fuel temperature failed and the actual fuel temperature was presumably much higher than 1600 ◦ C. The question now was: what was the actual temperature in the heating test? 4. Modeling of the irradiation behavior of TRISO coated particles 4.1. PANAMA code The PANAMA model describes the performance of TRISO coated particles during irradiation and under accident conditions (i.e., heating). Two failure modes are being considered. At temperatures below 2000 ◦ C, the failure mechanism based on the pressure vessel model is dominant, whereas at higher temperatures, thermal decomposition of the silicon carbide becomes dominant (Verfondern and Nabielek, 1985). The following equations are cited from Verfondern and Nabielek (1985). The total particle failure fraction is given by Φtotal = 1 − (1 − Φ1 )(1 − Φ2 )
(4)
where Φ1 and Φ2 correspond to the failure fractions due to pressure vessel failure and thermal decomposition, respectively. The SiC layer as the actual retention barrier for fission products represents the pressure vessel wall. In this code, a particle fails when stress in the SiC layer induced by the internal gas pressure exceeds the SiC strength. At time t after the start of the heating test, the failure fraction is given by m σt Φ1 (t, T ) = 1 − exp − ln 2 (5) σ0 σ t is the stress at time t due to the internal gas pressure, σ 0 the remaining strength of the SiC at the end of irradiation. Strength
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Table 3 Failure fraction during the normal irradiation by PANAMA code Irradiation temperature (◦ C) Irradiation time (efpd) Burnup (MWd/t) Failure fraction
1000 625 100,000 2.30E−6
values are scattered in a Weibull distribution, m is the Weibull parameter. σt =
0.5rP d0 1 + dct0
(6)
where r is the average radius of the SiC layer, P the internal gas pressure, c the corrosion rate as a function of temperature, and d0 the original thickness of the SiC layer. P=
BRT (0.31Fd + Of ) Vm Vf /Vk
(7)
where B is burnup, R the gas constant, Fd the relative fraction of fission gas releases, Of the number of oxygen atoms per fission, Vm the molar volume of the kernel, Vk the kernel volume, and Vf the void fraction corresponding to 50% of the buffer layer. At very high temperature beyond 2000 ◦ C, TRISO particles will mainly fail due to the thermal decomposition of the SiC. The failure fraction is given by Φ2 (t, T ) = 1 − exp(−αξ β )
(8)
where α and β are empirical constants, ξ is a so-called action integral which includes the decomposition rate as a function of temperature. For a spherical fuel element, the values for the constants were empirically derived to be α = 1 × 10−4 and β = 4. 4.2. Results One of the most important input parameters for PANAMA is the SiC strength and its Weibull parameter. For HTR-10 coated particles, the SiC strength is 834 MPa and its Weibull parameter is 5.7. 4.2.1. Failure fraction during the normal irradiation Table 3 shows the calculated failure fraction of coated particles at normal irradiation conditions, i.e., irradiation temperature is 1000 ◦ C, irradiation time is 625 efpd. The calculated stresses on each layer of an intact particle (Sawa et al., 1999) have shown that the maximum tensile stress on the SiC layer was almost zero when the burnup was near 50,000 MWd/t. When irradiation starts, a compressive stress acts on the SiC layer by fast neutron-induced shrinkage of the PyC layers. The PANAMA code does not take into account such a behavior of PyC layers, which is a conservative approach. In this case, the calculated result should be higher than that of irradiation result, but in this paper, the calculated result is lower. This is due to the fact that the first production batch had a higher free uranium level (1.4 × 10−4 ). As shown in Fig. 1, the fission gas
Fig. 5. Calculated particle failure fraction vs. heating time for different heating temperatures.
release rate curves during in-pile irradiation testing kept small fluctuation under average fuel temperature, ∼1000 ◦ C. It means that no coated particles failed caused by normal irradiation. 4.2.2. Failure fraction under heating test At the end of normal irradiation at 1000 ◦ C for 625 efpd, the fuel element in capsule 5 was expected to expose in a heating test in the reactor to the temperature of 1600 ◦ C for 200 h. Unfortunately, when the in-pile heating test underwent to 22 h (soaking time was about 5 h), R/B value reached to 3 × 10−2 , so the heating test has to be stopped. After lots of discussion, one conclusion that the thermocouples failed was drawn. Because the thermocouples failed, here we estimate that the actual fuel temperature filled in the range of 1600–1900 ◦ C, and use PANAMA code to calculate the failure fraction. The calculation results are given in Fig. 5. It shows that the failure fraction is 5.7 × 10−4 for a heating temperature of 1600 ◦ C. When the heating temperature is higher than 1850 ◦ C, the calculated failure fraction exceeds the level of 1 × 10−2 . A great deal of experimental data indicated that the calculation results of PANAMA code exhibit very good agreement with coated fuel particle irradiation data when the heating temperature is above 1800 ◦ C (Verfondern et al., 2001). Below 1800 ◦ C, the attribution of SiC thermal decomposition to CP fuel failure is not as large as the predication by PANAMA code. Annealing can relax tensile stress in the SiC layer, this is another reason resulted in the overestimation of PANAMA code. It can be concluded that the heating temperature of the HTR-10 spherical fuel elements in the Russian IVV-2M research reactor exceeded 1850 ◦ C. 5. Summary The comparatively high R/B level measured during the normal irradiation test in the IVV-2M reactor with the fuel element of capsule 5 can be traced back to 1–4 manufacture-induced defects of coated particles, which were taken from the first two production batches that had an average free uranium fraction of 1.4–2.3 × 10−4 . Post-irradiation examination indicated that at normal irradiation conditions, the PyC and SiC layers of particles which have
T.X. Liang et al. / Nuclear Engineering and Design 236 (2006) 1922–1927
undergone in-pile test kept their integrity. Any cracks and other radiation-induced defects in the layers were not observed. However, in the heat-up test, several types of defects including radial and tangential cracks in the SiC layers, cracks in the buffer layers and through coating failures were found. These defects were most likely caused by the obviously too high heating temperature. The heating test led to a failure fraction reaching 5.8 × 10−2 . The PANAMA fuel performance code was used to estimate the real temperature during the heating test. The calculated results show that the heating temperature of the HTR-10 spherical fuel elements in the Russian IVV-2M research reactor exceeded 1850 ◦ C. References Olander, D.R., 1985. Fundamental aspects of nuclear reactor fuel element. Technical Information Center, Springfield. Proksh, E., Strigl, A., Nabielek, H., 1982. Production of CO during burnup of UO2 kerneled HTR fuel particles. J. Nucl. Mater. 107, 280–283.
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Sawa, K., Tobita, T., Mogi, H., 1999. Fabrication of the first loading fuel of the high temperature engineering test reactor. J. Nucl. Sci. Technol. 36, 683–686. Tang, C.H., Tang, Y.P., Zhu, J.G., 2002. Design and manufacture of the fuel element for the 10 MW high temperature gas-cooled reactor. Nucl. Eng. Des. 218, 91–102. Tang, C.H., Fu, X.M., Zhu, J.G., 2004. The behavior of HTR-10 fuel under irradiation. In: Proceedings of the 2nd International Topic Meeting on High Temperature Reactor Technology, Beijing, China, 22–24 September. Tang, C.H., Fu, X.M., Zhu, J.G., Liang, T.X., Koshcheyev, K.N., Kozlov, A.V., Karlov, O.G., Degaltsev, Y.G., Vasiliev, V.I., 2006. Fuel irradiation of the first batches produced for the Chinese HTR-10. Nucl. Eng. Des. 236, 107–113. Tiegs, T.N., 1982. Fission product Pd–SiC interaction irradiated coated particle fuels. Nucl. Technol. 57, 389–398. Verfondern, K., Nabielek, H., 1985. PANAMA-Ein Rechenprogramm zur Vorhersage des Partikelbruchanteil von TRISO-Partikeln unter Stierfallbedingungen. FZJ report Juel-Spez-298, Research Center Juelich. Verfondern, K., Sumita, J., Ueta, S., Sawa, K., 2001. Modeling of fuel performance and metallic fission product release behavior during HTTR normal operating conditions. Nucl. Eng. Des. 210, 225–238.