Impact assessment of the IGR graphite block uneven impregnation with uranium on thermal strength properties
DOI:
https://doi.org/10.26577/RCPh.2022.v82.i3.08Keywords:
IGR, graphite block, neutronic calculations, thermal analysis, structural analysis, ANSYS APDLAbstract
The paper presents the results of a numerical interdisciplinary analysis of the fuel element of a heat capacity type research pulsed reactor (IGR). The fuel element of the IGR reactor is a graphite block impregnated with a solution of uranyl dinitrate. During the operation of the reactor, graphite blocks can be heated to high temperatures in a short period of time, which, together with the uneven impregnation of the block with uranium, which is due to the technological process of its manufacture, leads to the appearance of internal structural stresses. The purpose of this work is to make numerical estimation of the magnitude of thermal stresses arising in a graphite block. To carry out such an assessment, two computational models of a graphite block were built. One model is designed to perform neutron-physical calculations using a verified model of the IGR reactor core and the MCNP code, the other is designed to perform thermal strength analysis in the ANSYS software package. Thermal strength analysis includes two stages of calculations – thermal and structural (strength). Both models developed in the way to have the most similar topology, since this directly affects the correct distribution of the energy release over the volume of the block when transferring the results of the neutron-physical calculation to the thermal model.
The operation of a graphite block as part of the reactor core was simulated during a start-up lasting 4 s at a stationary power of 2 GW, followed by cooling down for 5 s. Numerical values and distribution diagrams of temperature and stresses arising in the volume of the block are obtained. The results of the analysis confirm the effect of the uneven impregnation of the graphite block of the IGR reactor with a solution of uranyl dinitrate on its thermal strength characteristics.
References
2 Stevens, John G., Encyclopedia of Nuclear Energy, 203–9 (2021).
3 V.A. Vityuk, A.D. Vurim, V.M. Kotov and A.V.Pakhnits, At. Energy, 120 (5), 323-327 (2016).
4 G. A. Vityuk, A.D. Vurim and R.A. Irkimbekov, Science and Technology, 3, 31-39 (2016).
5 V.A. Vityuk and A.D. Vurim, Ann. Nucl. Energy, 127, 196-203 (2019).
6 M.K. Skakov, R.A. Irkimbekov, Vurim A.D. et al., Bulletin of KazNITU, 6, 351-357 (2018).
7 V.A. Vityuk, Eurasian Phys. Tech. J., 2, 87–95 (2020).
8 V.M. Kotov, A.M. Kurpesheva and R.A. Irkimbekov, At. Energy, 2, 158–62 (2011).
9 B.Ye. Bekmagambetova, et.al, Bulletin of the NNC RK, 4(76), 60-65 (2018).
10 V.K. Tskhe, et.al., Ann. Nucl. Energy, 108875.
11 Frost and Brian R. T. Nuclear Fuel Elements. (Elsevier, 1982), 284 p.
12 Fuel Pins, Fuel Rods, Fuel Assemblies, and Reactor Cores. Nuclear Engineering Fundamentals, (543-610, CRC Press, 2017).
13 IAEA’s web-accessible database ARIS /Advanced Reactors Information System/ https://aris.iaea.org/sites/overview.html
14 R.E. Nightingale, Nuclear Graphite, 1–20 (2013).
15 IAEA, Thermophysical Properties of Materials for Nuclear Engineering: A Tutorial and Collection of Data, (Non-serial Publications, 2008).
16 T.D. Burchell, Comprehensive Nuclear Materials, 285–305 (2012).
17 Hoffelner and Wolfgang, Materials for Nuclear Plants, 197–254 (2012).
18 MCNP6 Monte Carlo N–Particle Transport Code System, MCNP6.1. LANL, 2013.
19 J.F. Briesmeister, MCNP - a general Monte-Carlo Code for neutron and photon Transport / Los Alamos National Laboratory. J.F. Briesmeister. – April 24, 2003. – 591 с. – LA-7396M, 1997.
20 M.B. Chadwick et al. ENDF/B-VII.0: Next generation evaluated nuclear data library for nuclear science and technology. In: Nuclear Data Sheets 107.12 (Dec. 2006), p. 2931–3059.
21 Ansys Mechanical APDL. Structural Analysis Guide, Ansys Inc., 2020
22 M.K. Thompson and J. M. Thompson, ANSYS Mechanical APDL for Finite Element Analysis (Butterworth-Heinemann, 2017), 889 p.
23 Y. Nakasone, S. Yoshimoto, and T.A. Stolarski, Engineering Analysis with ANSYS Software, 51–142 (2006).
24 Lee, Huei-Huang. Finite Element Simulations with ANSYS Workbench 2021 (SDC Publications, 2021), 600 p.
25 C. Xiaolin, L. Yiijun. Finite Element Modeling and Simulation with ANSYS Workbench, (CRC Press, 2014).
26 A.D. Vurim A.D., V.M. Kotov, R.A. Irkimbekov et.al. Computer model of the IGR reactor for stationary neutron calculations: patent № 2738 of 12.27.16 Republic of Kazakhstan.
