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Articoli di riviste sul tema "Htr fuel"

Autore: Grafiati

Pubblicato: 4 giugno 2021

Ultima modifica: 1 febbraio 2022

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1

Marmier, Alain, Michael A. Fütterer, Kamil Tuček, Jim C. Kuijper, Jaap Oppe, Biser Petrov, Jérôme Jonnet, Jan Leen Kloosterman e Brian Boer. "Fuel Cycle Investigation for Wallpaper-Type HTR Fuel". Nuclear Technology 181, n. 2 (febbraio 2013): 317–30. http://dx.doi.org/10.13182/nt13-a15786.

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2

Nabielek, H., W. Kühnlein, W. Schenk, W. Heit, A. Christ e H. Ragoss. "Development of advanced HTR fuel elements". Nuclear Engineering and Design 121, n. 2 (luglio 1990): 199–210. http://dx.doi.org/10.1016/0029-5493(90)90105-7.

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3

Zhang, Hai Quan, Xin Wang, Hong Ke Li, Jun Feng Nie e Ji Guo Liu. "Design and Engineering Verification of HTR-PM Fuel Handling". Advanced Materials Research 621 (dicembre 2012): 317–25. http://dx.doi.org/10.4028/www.scientific.net/amr.621.317.

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Abstract. HTR-PM is a twin-reactor structure pebble bed modular reactor. With the gather-scatter fuel handling system (FHS), handling and circulating function of twin reactors’ fuel elements are performed under a non-shutdown continuous condition. Relying on separate pipeline system of sphere fuel main circulation, FHS achieved automatic operation in the structural model by sphere fuel’s gravity flowing and pneumatic conveying in the twin reactors. The FHS adopted the international experience at design and operation of similar systems, especially based on that of HTR-10. However, some key components and technologies were improved so that fuel handling of HTR-PM becomes more reliable. All of the improved components and technologies will be tested in a full-scale hot testing facility, and some of them were verified and validated with the help of separated cold testing facilities. The functions and design of HTR-PM FHS is introduced in this paper. Design and engineering test of the FHS in HTR-PM demonstration power plant are reviewed.

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4

Wijaya, Rokhmadi, Bebeh Wahid Nuryadin, Khotib Maulani e Topan Setiadipura. "CALCULATION OF PROBABILITY OF TRISO PARTICLE FAILURE USING TIMCOAT AND PEBBED CODE". SIGMA EPSILON - Buletin Ilmiah Teknologi Keselamatan Reaktor Nuklir 24, n. 1 (30 aprile 2020): 17. http://dx.doi.org/10.17146/sigma.2020.24.1.5786.

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CALCULATION OF PROBABILITY OF TRISO PARTICLE FAILURE USING TIMCOAT AND PEBBED CODE. The calculation of the failure probability for fuel particles (TRISO) in the HTGR type reactor has been successfully carried out. This study aimed to estimate the failure probability of the fuel particles in the HTR-10 and HTR-PM, as well as to analyze the fuels of those reactors by varying the SiC thickness. The initial layer thickness of SiC in the HTR-10 and HTR-PM is 35 µm. The PEBBED code was used to simulate calculations resulting in the power distribution data, which is then compared with the results from the TIMCOAT simulation process. The TIMCOAT simulation calculation results, which are based on the SiC thickness variation, showed that the thickness failure is smaller if applied to the HTR-10 and HTR-PM. Based on the comparison between the two reactors, the failure probability of HTR-PM fuel particle has the value smaller than that of the HTR-10 with the difference of 10-5 .Keywords: failure particle, HTR-10, HTR-PM, TRISO, TIMCOAT.

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5

Marmier, A., M. A. Fütterer, K. Tuček, Han de Haas, Jim C. Kuijper e Jan Leen Kloosterman. "Revisiting the concept of HTR wallpaper fuel". Nuclear Engineering and Design 240, n. 10 (ottobre 2010): 2485–92. http://dx.doi.org/10.1016/j.nucengdes.2010.02.043.

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6

de Groot, Sander, Pierre Guillermier, Kazuhiro Sawa, Jean-Michel Escleine, Shohei Ueta, Virginie Basini, Klaas Bakker, Young-Woo Lee, Marc Perez e Bong-Goo Kim. "HTR fuel coating separate effect test PYCASSO". Nuclear Engineering and Design 240, n. 10 (ottobre 2010): 2392–400. http://dx.doi.org/10.1016/j.nucengdes.2010.05.052.

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7

Helary, D., O. Dugne, X. Bourrat, P. H. Jouneau e F. Cellier. "EBSD investigation of SiC for HTR fuel particles". Journal of Nuclear Materials 350, n. 3 (maggio 2006): 332–35. http://dx.doi.org/10.1016/j.jnucmat.2006.01.010.

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8

Brähler, Georg, Markus Hartung, Johannes Fachinger, Karl-Heinz Grosse e Richard Seemann. "Improvements in the fabrication of HTR fuel elements". Nuclear Engineering and Design 251 (ottobre 2012): 239–43. http://dx.doi.org/10.1016/j.nucengdes.2011.10.036.

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9

FU, Xiaoming, Tongxiang LIANG, Yaping TANG, Zhichang XU e Chunhe TANG. "Preparation of UO2Kernel for HTR-10 Fuel Element". Journal of Nuclear Science and Technology 41, n. 9 (settembre 2004): 943–48. http://dx.doi.org/10.1080/18811248.2004.9715568.

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10

Sembiring, Tagor Malem, e Pungky Ayu Artiani. "SUBCRITICALITY ANALYSIS OF HTR-10 SPENT FUEL CASK MODEL FOR THE 10 MW HTR INDONESIAN EXPERIMENTAL POWER REACTOR". JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA 20, n. 3 (31 ottobre 2018): 151. http://dx.doi.org/10.17146/tdm.2018.20.3.4630.

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The 10 MW HTR Indonesian Experimental Power Reactor (RDE reactor) is designed identical with the HTR-10 in China, conceptually. However, the review results showed that the spent fuel cask model which is used between two reactors is fully different, such as size and capacity. The proposed cask model in RDE reactor can hold 15 times more fuel pebbles than HTR-10 has. This research activities deal with the subcriticality analysis for the spent fuel cask of RDE reactor if using the HTR-10 cask model. The subcriticality condition is designed to meet the limit of safety value. The objective of this research is to determine the subcriticality value in the normal and accident events for the spent fuel cask when it is in the reactor building and the spent fuel cask room. All calculations were carried out by MCNP6.1 code. The selected external events are the water ingress (reactor room), water flood and the combination event of water flood and earthquake. The calculation results showed that the maximum value of keff (3σ) are 0.47510 and 0.19214 for the cask in the reactor building and in the spent fuel cask room, respectively. This value is far from the limit value of 0.95. The calculation results showed that the spent fuel cask are in the safe condition eventhough in the worst combination events, the cask is flooded and earthquake. The HTR-10 spent fuel cask can be proposed as an alternative for the RDE reactor to get an efficient reactor building.Keywords: spent pebble fuel element, HTGR, subcriticality, MCNP6.1, RDE reactor ANALISIS SUBKRITIKALITAS PENYIMPAN BAHAN BAKAR BEKAS MODEL CASK REAKTOR HTR-10 UNTUK REAKTOR DAYA EKSPERIMENTAL 10 MW TERMAL. Reaktor Daya Eksperimental (RDE) secara konseptual didesain identik dengan reaktor HTR-10 di Tiongkok. Meskipun demikian, terdapat perbedaan yang signifikan untuk desain konseptual cask penyimpan bahan bakar bekas di kedua reaktor seperti dimensi dan kapasitas. Kegiatan penelitian ini berkaitan dengan analisis subkritikalitas cask penyimpan elemen bahan bakar bekas tipe pebble di RDE jika menggunakan model cask yang dipakai di HTR-10. Kondisi sub-kritikalitas didesain memenuhi nilai batas keselamatan. Tujuan penelitian adalah menentukan nilai subkritikalitas dalam keadaan normal atau kondisi kecelakaan di gedung reaktor dan di gudang penyimpan bahan bakar bekas. Perhitungan dilakukan dengan paket program MCNP6.1. Kejadian kecelakaan yang dipilih adalah masuknya air ke dalam cask, cask terendam air dan kombinasi cask terendam air dan kejadian gempa. Hasil perhitungan menunjukkan bahwa nilai maksimum keff (3σ) untuk cask di gedung reaktor dan di gudang penyimpan bahan bakar bekas masing-masing adalah 0,47510 dan 0,19214. Nilai ini masih jauh dari batas 0,95. Hasil perhitungan menunjukkan bahwa cask penyimpan bahan bakar bekas tetap dalam keadaan selamat meski terjadi kombinasi 2 kejadian eksternal.Kata kunci: elemen bahan bakar bekas tipe pebble, HTGR, subkritikalitas, MCNP6.1, RDE

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11

Hrovat, M., H. Huschka, A. W. Mehner e W. Warzawa. "Spherical fuel elements for small and medium sized HTR". Nuclear Engineering and Design 109, n. 1-2 (settembre 1988): 253–56. http://dx.doi.org/10.1016/0029-5493(88)90167-7.

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12

Wang, Meng-Jen, Jinn-Jer Peir, Shang-Chien Wu, Ming-Hua Li e Jenq-Horng Liang. "ICONE19-43240 EFFECTS OF HOMOGENEOUS GEOMETRY MODELS IN SIMULATING THE FUEL BALLS IN HTR-10". Proceedings of the International Conference on Nuclear Engineering (ICONE) 2011.19 (2011): _ICONE1943. http://dx.doi.org/10.1299/jsmeicone.2011.19._icone1943_102.

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13

Phillips, Jeffrey A., Scott G. Nagley e Eric L. Shaber. "Fabrication of uranium oxycarbide kernels and compacts for HTR fuel". Nuclear Engineering and Design 251 (ottobre 2012): 261–81. http://dx.doi.org/10.1016/j.nucengdes.2011.10.033.

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14

Xiangwen, Zhou, Lu Zhenming, Zhang Jie, Liu Bing, Zou Yanwen, Tang Chunhe e Tang Yaping. "Preparation of spherical fuel elements for HTR-PM in INET". Nuclear Engineering and Design 263 (ottobre 2013): 456–61. http://dx.doi.org/10.1016/j.nucengdes.2013.07.001.

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15

ZHAO, H., T. LIANG, J. ZHANG, Z. LI e C. TANG. "Research on graphite powders used for HTR-PM fuel elements". Rare Metals 25, n. 6 (ottobre 2006): 347–50. http://dx.doi.org/10.1016/s1001-0521(07)60103-x.

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16

Liang, T. X., H. S. Zhao, C. H. Tang e K. Verfondern. "Irradiation performance and modeling of HTR-10 coated fuel particles". Nuclear Engineering and Design 236, n. 18 (settembre 2006): 1922–27. http://dx.doi.org/10.1016/j.nucengdes.2006.01.018.

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17

Liu, Peng, Yanhua Zheng, Wei Xu e Lei Shi. "Nonuniform Oxidation on the Surface of Fuel Element in HTR". Science and Technology of Nuclear Installations 2016 (2016): 1–9. http://dx.doi.org/10.1155/2016/7485602.

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The graphite oxidation of fuel element has obtained high attention in air ingress accident analysis of high temperature gas-cooled reactor (HTR). The shape function, defined as the relationship between the maximum and the average of the oxidation, is an important factor to estimate the consequence of the accident. There are no detailed studies on the shape function currently except two experiments several decades ago. With the development of computer technology, CFD method is used in the numerical experiment about graphite oxidation in pebble bed of HTR in this paper. Structured packed beds are used in the calculation instead of random packed beds. The result shows the nonuniform distribution of oxidation on the sphere surface and the shape function in the condition of air ingress accident. Furthermore, the sensitive factors of shape function, such as temperature and Re number, are discussed in detail and the relationship between the shape function and sensitive factors is explained. According to the results in this paper, the shape function ranges from 1.05 to 4.7 under the condition of temperature varying from 600°C to 1200°C and Re varying from 16 to 1600.

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18

Fang, Sheng, Jianzhu Cao, Wenqian Li, Chen Luo, Feng Yao, Xiaofan Li e Kai Li. "Shielding Design and Dose Evaluation for HTR-PM Fuel Transport Pipelines by QAD-CGA Program". Science and Technology of Nuclear Installations 2021 (3 maggio 2021): 1–6. http://dx.doi.org/10.1155/2021/6686919.

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The spherical fuel elements are adopted in the high-temperature gas-cooled reactor pebble-module (HTR-PM). The fuel elements will be discharged continuously from the reactor core and transported into the fuel transport pipelines during the reactor operation, leading to spatially varying dose outside the pipeline. In this case, the dose evaluation faces two major challenges, including dynamic source terms and pipelines with varying lengths and shapes. This study tries to handle these challenges for HTR-PM through comprehensive calculations using the QAD-CGA program and to design the corresponding shielding of the pipeline. During the calculation, it is assumed that a spherical fuel element stays in different positions of the pipelines in turn, and the corresponding dose contributions were calculated. By integrating the dose contributions at different positions, the dose at the points of interest can be obtained. The total dose is further determined according to the assumed fuel elements transport speed of 5 m/s and total 6000 fuel elements transportation per day. Two types of fuel transport pipelines and two source terms were considered, i.e., the spent fuel element transport pipelines with corresponding spent fuel source term and the different burn-up fuel element transport pipelines with the average burn-up fuel source term. Doses at different points of interest were calculated with no shielding scenario and with lead shielding of different thicknesses scenario. To evaluate the shielding effect, the dose limit of the orange radiation zone of HTR-PM and the radiation damage thresholds from NCRP report No.51 were both adopted. The calculated results show that, for pipelines that transport the spent fuel, a 4 cm lead shielding will be enough. And for pipelines that transport fuel elements with different burn-up, a 5 cm lead shielding will be added. The method and results can provide valuable reference for other work of HTR-PM.

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19

Heath, Paul G., Martin C. Stennett, Owen J. McGann, Russell J. Hand e Neil C. Hyatt. "The Use of High Durability Alumino-Borosilicate Glass for the Encapsulation of High Temperature Reactor (HTR) Fuel". MRS Proceedings 1518 (2013): 3–8. http://dx.doi.org/10.1557/opl.2013.129.

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ABSTRACTThe development of suitable waste forms for waste produced by generation IV reactors is of critical concern for future operations. To date no accepted disposal route for Tri-Structural Isotropic (TRISO) High Temperature Reactor (HTR) fuel exists. Alumino-borosilicate glass has been studied for its ability to encapsulate TRISO particle fuels. This glass was selected for its high aqueous durability. Encapsulation was achieved by cold pressing and sintering of glass powders mixed with HTR fuel. Sintering profiles capable of eliminating interconnected porosity in the composites were developed. The chemical compatibility and wetting of the glass matrix with the fuel were analysed along with the aqueous durability of the sintered glass matrix. Composites sintered under a controlled atmosphere produced unfractured monoliths with minimal chemical interaction between the glass and the TRISO particles. The Product Consistency Test (PCT) durability assessment indicated the sintered alumino-borosilicate glass was approximately an order of magnitude more durable than an equivalent R7T7 borosilicate glass. These results suggest sintered alumino-borosilicate glass-TRISO particle composites may provide a potential disposal route for spent TRISO particle fuel.

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20

Xu, Wei, Yanhua Zheng, Lei Shi e Peng Liu. "Oxidation Analyses of Massive Air Ingress Accident of HTR-PM". Science and Technology of Nuclear Installations 2016 (2016): 1–9. http://dx.doi.org/10.1155/2016/6419124.

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The double-ended guillotine break (DEGB) of the horizontal coaxial gas duct accident is a serious air ingress accident of the high temperature gas-cooled reactor pebble-bed module (HTR-PM). Because the graphite is widely used as the structure material and the fuel element matrix of HTR-PM, the oxidation analyses of this severe air ingress accident have got enough attention in the safety analyses of the HTR-PM. The DEGB of the horizontal coaxial gas duct accident is calculated by using the TINTE code in this paper. The results show that the maximum local oxidation of the matrix graphite of spherical fuel elements in the core will firstly reach3.75⁎104 mol/m3at about 120 h, which means that only the outer 5 mm fuel-free zone of matrix graphite will be oxidized out. Even at 150 h, the maximum local weight loss ratio of the nuclear grade graphite in the bottom reflectors is only 0.26. Besides, there is enough time to carry out some countermeasures to stop the air ingress during several days. Therefore, the nuclear grade graphite of the bottom reflectors will not be fractured in the DEGB of the horizontal coaxial gas duct accident and the integrity of the HTR-PM can be guaranteed.

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21

Hao, Chen, Peijun Li, Ding She, Xiaoyu Zhou e Rongrui Yang. "Sensitivity and Uncertainty Analysis of the Maximum Fuel Temperature under Accident Condition of HTR-PM". Science and Technology of Nuclear Installations 2020 (22 febbraio 2020): 1–21. http://dx.doi.org/10.1155/2020/9235783.

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The maximum fuel temperature under accident condition is the most important parameter of inherently safe characteristics of HTR-PM, and the DLOFC accident may lead to a peak accident fuel temperature. And there are a variety of uncertainty sources in the maximum fuel temperature calculations, and thus the contributions of these uncertainty sources to the final calculated maximum fuel temperature should be quantified to check whether the peak value exceed the technological limit of 1620°C or not. Eight uncertainty input parameters are selected for inclusion in this uncertainty study, and their associated 2 standard deviation uncertainties and probability density functions are specified. Then, the DLOFC thermal analyses and uncertainty analysis are performed with the home-developed ATHENA and CUSA. The numerical results indicate that the pebble-bed effective conductivity and the decay heat contribute the most of the uncertainty in the DLOFC maximum fuel temperature while this peak fuel temperature is most sensitive to the initial reactor power and the decay heat. In short, uncertainties in these selected eight parameters lead to the two standard deviation (2σ) uncertainty of ±77.6°C (or 5.2%) around the mean value of 1493°C for the maximum fuel temperature under DLOFC accident of HTR-PM. At the same time, the LHS-SVDC method of CUSA is recommended to propagate uncertainties in inputs and 100–200 model simulations seem to be sufficient to get an uncertainty prediction with full confidence.

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22

Bomboni, Eleonora, Nicola Cerullo e Guglielmo Lomonaco. "Assessment of LWR-HTR-GCFR Integrated Cycle". Science and Technology of Nuclear Installations 2009 (2009): 1–14. http://dx.doi.org/10.1155/2009/193594.

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Preliminary analyses already performed showed that innovative GCRs, both thermal and fast, are very promising candidate to reach the Gen-IV sustainability goal. The integrated LWR-HTR-GCFR basically aims at closing the current nuclear fuel cycle: in principle, thanks to the unique characteristics of Helium coolant reactors, LWR SNF along with DU become valuable material to produce energy. Additionally, burning HMs of LWR SNF means not only a drastic reduction in the demand but also a remarkable decrease in the long-term radiotoxic component of nuclear waste to be geologically stored. This paper focuses on the analyses of the LWR-HTR-GCFR cycle performed by the University of Pisa in the frame of the EU PUMA project (6th FP). Starting from a brief outline of the main characteristics of HTR and GCFR concepts and of the advantages of linking LWR, HTR and GCFR in a symbiotic way, this paper shows the integrated cycle involving a typical LWR (1000 ), a PBMR (400 ) and a GCFR-“E” (2400 ). Additionally, a brief overview of the main technological constraints concerning (Pu+MA)-based advanced fuels is given, in order to explain and justify the choices made in the framework of the considered cycle. Thereafter, calculations performed and results obtained are described.

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23

Zhang, Jingyu, Fu Li e Yuliang Sun. "Physical Analysis of the Initial Core and Running-In Phase for Pebble-Bed Reactor HTR-PM". Science and Technology of Nuclear Installations 2017 (2017): 1–6. http://dx.doi.org/10.1155/2017/8918424.

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The pebble-bed reactor HTR-PM is being built in China and is planned to be critical in one or two years. At present, one emphasis of engineering design is to determine the fuel management scheme of the initial core and running-in phase. There are many possible schemes, and many factors need to be considered in the process of scheme evaluation and analysis. Based on the experience from the constructed or designed pebble-bed reactors, the fuel enrichment and the ratio of fuel spheres to graphite spheres are important. In this paper, some relevant physical considerations of the initial core and running-in phase of HTR-PM are given. Then a typical scheme of the initial core and running-in phase is proposed and simulated with VSOP code, and some key physical parameters, such as the maximum power per fuel sphere, the maximum fuel temperature, the refueling rate, and the discharge burnup, are calculated. Results of the physical parameters all satisfy the relevant design requirements, which means the proposed scheme is safe and reliable and can provide support for the fuel management of HTR-PM in the future.

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Knorr, J., A. Kerber e R. Moormann. "Upgrading (V)HTR fuel elements for generationIV goals by SiC encapsulation". Kerntechnik 77, n. 5 (novembre 2012): 351–55. http://dx.doi.org/10.3139/124.110218.

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Tang, Chunhe, Xiaoming Fu, Junguo Zhu, Hongsheng Zhao e Yanping Tang. "Comparison of two irradiation testing results of HTR-10 fuel spheres". Nuclear Engineering and Design 251 (ottobre 2012): 453–58. http://dx.doi.org/10.1016/j.nucengdes.2011.09.047.

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Yi, DU, WANG Xiangang, XIANG Xincheng e LIU Bing. "Automatic X-ray inspection for the HTR-PM spherical fuel elements". Nuclear Engineering and Design 280 (dicembre 2014): 144–49. http://dx.doi.org/10.1016/j.nucengdes.2014.09.016.

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Zhao, Hongsheng, Tongxiang Liang, Jie Zhang, Jun He, Yanwen Zou e Chunhe Tang. "Manufacture and characteristics of spherical fuel elements for the HTR-10". Nuclear Engineering and Design 236, n. 5-6 (marzo 2006): 643–47. http://dx.doi.org/10.1016/j.nucengdes.2005.10.023.

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Charollais, François, Christophe Perrais, Dominique Moulinier, Marc Perez e Marie-Pierre Vitali. "Latest achievements of CEA and AREVA NP on HTR fuel fabrication". Nuclear Engineering and Design 238, n. 11 (novembre 2008): 2854–60. http://dx.doi.org/10.1016/j.nucengdes.2008.01.020.

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Tang, Chunhe, Ziqiang Li, Yanwen Zou e Xiaoming Fu. "Irradiation testing of matrix material for spherical HTR-10 fuel elements". Nuclear Engineering and Design 238, n. 11 (novembre 2008): 2886–92. http://dx.doi.org/10.1016/j.nucengdes.2008.01.021.

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30

Bolin, J. M., T. D. Dunn, K. Verfondern e M. J. Kania. "Modeling of fission product release from HTR fuel for risk analysis". Energy 16, n. 1-2 (gennaio 1991): 433–39. http://dx.doi.org/10.1016/0360-5442(91)90122-3.

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Bolin, J. M., T. D. Dunn, K. Verfondern e M. J. Kania. "Modeling of fission product release from HTR fuel for risk analyses". Nuclear Engineering and Design 137, n. 2 (ottobre 1992): 207–11. http://dx.doi.org/10.1016/0029-5493(92)90019-r.

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32

Freis, Daniel, Abdel El Abjani, Dragan Coric, Ramil Nasyrow, Joseph Somers, Chunhe Tang, Rongzheng Liu, Bing Liu e Malin Liu. "Burn-up determination and accident testing of HTR-PM fuel elements irradiated in the HFR Petten". Nuclear Engineering and Design 357 (febbraio 2020): 110414. http://dx.doi.org/10.1016/j.nucengdes.2019.110414.

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33

Mazzini, Guido, Eleonora Bomboni, Nicola Cerullo, Emil Fridman, Guglielmo Lomonaco e Eugene Shwageraus. "The Use of Th in HTR: State of the Art and Implementation in Th/Pu Fuel Cycles". Science and Technology of Nuclear Installations 2009 (2009): 1–13. http://dx.doi.org/10.1155/2009/749736.

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Abstract (sommario):

Nowadays nuclear is the only greenhouse-free source that can appreciably respond to the increasing worldwide energy demand. The use of Thorium in the nuclear energy production may offer some advantages to accomplish this task. Extensive R&D on the thorium fuel cycle has been conducted in many countries around the world. Starting from the current nuclear waste policy, the EU-PUMA project focuses on the potential benefits of using the HTR core as a Pu/MA transmuter. In this paper the following aspects have been analysed: (1) the state-of-the-art of the studies on the use of Th in different reactors, (2) the use of Th in HTRs, with a particular emphasis on Th-Pu fuel cycles, (3) an original assessment of Th-Pu fuel cycles in HTR. Some aspects related to Thorium exploitation were outlined, particularly its suitability for working in pebble-bed HTR in a Th-Pu fuel cycle. The influence of the Th/Pu weight fraction at BOC in a typical HTR pebble was analysed as far as the reactivity trend versus burn-up, the energy produced per Pu mass, and the Pu isotopic composition at EOC are concerned. Although deeper investigations need to be performed in order to draw final conclusions, it is possible to state that some optimized Th percentage in the initial Pu/Th fuel could be suggested on the basis of the aim we are trying to reach.

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Fortini, A., F. B. A. Monteiro, M. E. Scari, F. C. da Silva, R. V. Sousa, C. A. M. da Silva, A. L. Costa, C. Pereira e M. A. F. Veloso. "Recent advances on the use of reprocessed fuels and combined thorium fuel cycles in HTR systems". Progress in Nuclear Energy 83 (agosto 2015): 482–96. http://dx.doi.org/10.1016/j.pnucene.2014.09.001.

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35

Bomboni, E., N. Cerullo, G. Lomonaco e V. Romanello. "A Critical Review of the Recent Improvements in Minimizing Nuclear Waste by Innovative Gas-Cooled Reactors". Science and Technology of Nuclear Installations 2008 (2008): 1–18. http://dx.doi.org/10.1155/2008/265430.

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This paper presents a critical review of the recent improvements in minimizing nuclear waste in terms of quantities, long-term activities, and radiotoxicities by innovative GCRs, with particular emphasis to the results obtained at the University of Pisa. Regarding these last items, in the frame of some EU projects (GCFR, PUMA, and RAPHAEL), we analyzed symbiotic fuel cycles coupling current LWRs with HTRs, finally closing the cycle by GCFRs. Particularly, we analyzed fertile-free and Pu-Th-based fuel in HTR: we improved plutonium exploitation also by optimizing Pu/Th ratios in the fuel loaded in an HTR. Then, we chose GCFRs to burn residual MA. We have started the calculations on simplified models, but we ended them using more “realistic” models of the reactors. In addition, we have added the GCFR multiple recycling option usingkeffcalculations for all the reactors. As a conclusion, we can state that, coupling HTR with GCFR, the geological disposal issues concerning high-level radiotoxicity of MA can be considerably reduced.

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36

WANG, Jinhua, Xiang LIU, Bing WANG, Yue LI e Bin WU. "ICONE23-1444 DESIGN OF THE GROUND CRANE AND SHIELDING CASK FOR THE SPENT FUEL CANISTER OF HTR-PM". Proceedings of the International Conference on Nuclear Engineering (ICONE) 2015.23 (2015): _ICONE23–1—_ICONE23–1. http://dx.doi.org/10.1299/jsmeicone.2015.23._icone23-1_204.

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37

Robert-Inacio, F., C. Boschet, F. Charollais e F. Cellier. "Polar studies of the sphericity degree of V/HTR nuclear fuel particles". Materials Characterization 56, n. 4-5 (giugno 2006): 266–73. http://dx.doi.org/10.1016/j.matchar.2005.11.022.

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38

Martin, David G. "Considerations pertaining to the achievement of high burn-ups in HTR fuel". Nuclear Engineering and Design 213, n. 2-3 (aprile 2002): 241–58. http://dx.doi.org/10.1016/s0029-5493(01)00502-7.

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39

Nickel, H., H. Nabielek, G. Pott e A. W. Mehner. "Long time experience with the development of HTR fuel elements in Germany". Nuclear Engineering and Design 217, n. 1-2 (agosto 2002): 141–51. http://dx.doi.org/10.1016/s0029-5493(02)00128-0.

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40

Li, Hua, e Fu Li. "Digital signal processing on the fuel ball counting system for HTR-10". Nuclear Engineering and Design 222, n. 2-3 (giugno 2003): 139–46. http://dx.doi.org/10.1016/s0029-5493(03)00008-6.

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41

van der Merwe, J. J. "Evaluation of silver transport through SiC during the German HTR fuel program". Journal of Nuclear Materials 395, n. 1-3 (dicembre 2009): 99–111. http://dx.doi.org/10.1016/j.jnucmat.2009.09.024.

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42

Guittonneau, Fabrice, Abdesselam Abdelouas e Bernd Grambow. "HTR Fuel Waste Management: TRISO separation and acid-graphite intercalation compounds preparation". Journal of Nuclear Materials 407, n. 2 (dicembre 2010): 71–77. http://dx.doi.org/10.1016/j.jnucmat.2010.09.026.

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43

Tang, Chunhe, Xiaoming Fu, Junguo Zhu, Tongxiang Liang, Konstantin N. Koshcheyev, Alexandre V. Kozlov, Oleg G. Karlov, Yu G. Degaltsev e Vladimir I. Vasiliev. "Fuel irradiation of the first batches produced for the Chinese HTR-10". Nuclear Engineering and Design 236, n. 1 (gennaio 2006): 107–13. http://dx.doi.org/10.1016/j.nucengdes.2005.11.001.

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44

Yang, Hui, Hong-sheng Zhao, Zi-qiang Li, Xiao-xue Liu, Kai-hong Zhang, Tao-wei Wang e Bing Liu. "Review of oxidant resistant coating on graphite substrate of HTR fuel element". Journal of Central South University 26, n. 11 (novembre 2019): 2915–29. http://dx.doi.org/10.1007/s11771-019-4224-2.

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45

Proksch, E., A. Strigl e H. Nabielek. "Carbon monoxide formation in UO2 kerneled HTR fuel particles containing oxygen getters". Journal of Nuclear Materials 139, n. 2 (giugno 1986): 83–90. http://dx.doi.org/10.1016/0022-3115(86)90025-5.

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46

Mehner, A. W., W. Heit, K. Röllig, H. Ragoss e H. Müller. "Spherical fuel elements for advanced HTR manufacture and qualification by irradiation testing". Journal of Nuclear Materials 171, n. 1 (aprile 1990): 9–18. http://dx.doi.org/10.1016/0022-3115(90)90341-j.

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47

Bertram, W., M. Fleischer, H. Haas, H. J. Kampffmeyer, R. Lison, E. Sigismund e J. Zeumann. "Verschweißen von Transport- und Lagerbehältern für abgebrannte HTR-Brennelemente / Weld-tightening of transport and storage containers for irradiated HTR fuel elements". Kerntechnik 53, n. 2 (1 febbraio 1988): 165–69. http://dx.doi.org/10.1515/kern-1988-530219.

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48

Chen, Fubing, Yujie Dong e Zuoyi Zhang. "Temperature Response of the HTR-10 during the Power Ascension Test". Science and Technology of Nuclear Installations 2015 (2015): 1–13. http://dx.doi.org/10.1155/2015/302648.

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Abstract (sommario):

The 10 MW High Temperature Gas-Cooled Reactor-Test Module (HTR-10) is the first High Temperature Gas-Cooled Reactor in China. With the objective of raising the reactor power from 30% to 100% rated power, the power ascension test was planned and performed in January 2003. The test results verified the practicability and validity of the HTR-10 power regulation methods. In this study, the power ascension process is preliminarily simulated using the THERMIX code. The code satisfactorily reproduces the reactor transient parameters, including the reactor power, the primary helium pressure, and the primary helium outlet temperature. Reactor internals temperatures are also calculated and compared with the test values recorded by a number of thermocouples. THERMIX correctly simulates the temperature variation tendency for different measuring points, with good to fair agreement between the calculated temperatures and the measured ones. Based on the comparison results, the THERMIX simulation capability for the HTR-10 dynamic characteristics during the power ascension process can be demonstrated. With respect to the reactor safety features, it is of utmost importance that the maximum fuel center temperature during the test process is always much lower than the fuel temperature limit of 1620°C.

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49

Hong, Jhao-Yang, Shin-Rong Wu, Shang-Chien Wu, Der-Sheng Chao e Jenq-Horng Liang. "Burnup computations of multi-pass fuel loading scenarios in HTR-10 using a pre-generated fuel composition library". Nuclear Engineering and Design 374 (aprile 2021): 111063. http://dx.doi.org/10.1016/j.nucengdes.2021.111063.

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50

Setiawati, Evi, Hammam Oktajianto, Jatmiko Suseno, Choirul Anam e Heri Sugito. "THE EFFECTIVENESS ANALYSIS OF FUEL BALL (KERNEL) DIMENSION SIZE AND URANIUM ENRICHMENT TO REACTOR REACTIVITY". International Journal of Engineering Technologies and Management Research 4, n. 2 (31 gennaio 2020): 1–9. http://dx.doi.org/10.29121/ijetmr.v4.i2.2017.70.

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Reactor reactivity does not only depend on reactor diameter but also radius and enrichment of fuel ball (kernel) to operate reactor optimally. This research analyses effectiveness of kernel radius and enrichment to achieve critical reactor condition. The HTR in this research adopts HTR-10 China and HTR of pebble bed. The calculations are performed by using MCNPX code in each kernel radii of 320-350 µm and enrichments of 5-10% Uranium. Kernel is composed of Uranium Dioxide coated by four outer layers: Carbon, IpyC (Inner Pyrolytic Coating), SiC (Silicon Carbides) and OpyC (Outer Pyrolytic Coating). It is called TRISO and it is distributed in pebble-bed ball using Simple Cubic Lattice whereas pebble-bed and moderator balls are distributed in the core zone using a Body Centred Cubic (BCC) lattice by ratio of 57:43. The research results are obtained that the reactor will be effective to achieve critical condition in kernel radius of 325-330 µm at 9% Uranium enrichment and will be in supercritical condition if the reactor uses more than 330 µm of kernel radius and 9% enrichment of Uranium but the reactor will be subcritical if Uranium enrichment is 5-8%.

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