Neutronic Analysis of the AP1000 Fuel Assembly with Accident Tolerant Cladding Materials

Year 2024, Volume: 37 Issue: 4, 2012 - 2023, 01.12.2024

Wahid Luthfı , Surian Pinem , Farisy Yogatama Sulistyo , Tukiran Surbakti

https://doi.org/10.35378/gujs.1332751

Abstract

An alternative material for fuel cladding was required to prevent oxidation caused by interacting with steam, leading to improvement in core integrity. This study analyzes reactor physics parameters of various cladding material candidates for Pressurized Water Reactor (PWR) such as SS-304 austenitic stainless steel, FeCrAl alloy, APMT alloy, and silicon carbide (SiC) ceramic, as candidate Accident Tolerant Fuel (ATF). The neutronic parameters such as infinite multiplication factor (k-inf), and neutron spectrum, while temperature reactivity coefficient related to fuel temperature (DTC) and moderator temperature (MTC) is also considered, followed by a void coefficient of reactivity (VCR) of each candidate material were then compared with ZIRLO as a standard cladding material of AP1000. k-inf calculated by SRAC2006 is also compared to MCNP for various fuel assembly types. At the beginning of cycle (BOC), the 2.35% UO2 using SiC gives a higher kinf than ZIRLO at 937 pcm, while 4.45% UO2 with 88 IFBA & 9 PYREX at 796 pcm. FeCrAl, APMT, and SS-304 cladding gave a smaller k-inf compared to ZIRLO in the range of 11000-14000 pcm at 2.35% UO2 fuel assembly. The values of DTC, MTC, and VCR were still negative throughout the reactor operation which indicates that the inherent safety feature of alternative cladding was possible for this type of fuel assembly, especially for iron-based cladding material followed by an increase in fuel enrichment.

Keywords

Alternative cladding material, Accident tolerant fuel (ATF), K-inf, Coefficient of reactivity, SRAC2006

Supporting Institution

National Research and Innovation Agency (BRIN)

Project Number

DIPA- 124.01.1.690503/2022

Thanks

Thank you to the head of the Research Center for Nuclear Reactor Technology (PRTRN) Research Organization for Nuclear Energy (ORTN), National Research and Innovation Agency (BRIN) for supporting this research through DIPA 2022 No. DIPA- 124.01.1.690503/2022.

References

• [1] Liang, Y., Lan, B., Zhang, Q., Seidl, M., Wang, X., “Neutronic analysis of silicon carbide Cladding-ATF fuel combinations in small modular reactors”, Annals Nuclear Energy, 173: 109120, (2022).
• [2] Younan, S., Novog, D., “Assessment of Neutronic Characteristics of Accident-Tolerant Fuel and Claddings for CANDU Reactors”, Science Technology Nuclear Installation, (2018).
• [3] Chen, S., Yuan, C., “Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U3Si2-FeCrAl”, Science Technology Nuclear Installation, (2017).
• [4] Alrwashdeh, M., Alameri, S.A., “Preliminary neutronic analysis of alternative cladding materials for APR-1400 fuel assembly”, Nuclear Engineering Design, 384: 111486, (2021).
• [5] Sembiring, T.M., Pinem, S., “The validation of the NODAL3 code for static cases of the PWR benchmark core”, Jurnal Nuklir Sains Teknologi Ganendra, 15(2): 82–92, (2012).
• [6] Pinem, S., Sembiring, T.M., Liem, P.H., “NODAL3 Sensitivity Analysis for NEACRP 3D LWR Core Transient Benchmark (PWR)”, Science Technology Nuclear Installation, 1–11, (2016).
• [7] Sembiring, T.M., Pinem, S., Liem, P.H., “Analysis of NEA-NSC PWR Uncontrolled Control Rod Withdrawal at Zero Power Benchmark Cases with NODAL3 Code”, Science Technology Nuclear Installation, (2017).
• [8] Pinem, S., Sembiring, T.M., Liem, P.H., “The verification of coupled neutronics thermal-hydraulics code NODAL3 in the PWR rod ejection benchmark”, Science Technology Nuclear Installation, 1–9 (2014).
• [9] Pinem, S., Sembiring, T.M., Tukiran, Deswandri, Sunaryo, G.R., “Reactivity Coefficient Calculation for AP1000 Reactor Using the NODAL3 Code”, Journal of Physics: Conference Series, 962: 1, (2018).
• [10] Sembiring, T.M., Susilo, J., Pinem, S., “Evaluation of the AP1000 delayed neutron parameters using MCNP6”, In: International Conference on Nuclear Technologies and Sciences (ICoNETS), Journal of Physics: Conference Series, p. 012030, (2018).
• [11] Galahom, A. A., Abdel-Rahman, M. A. E., Mohsen, M. Y. M., Hakamy, A. “Investigation of the possibility of using a uranium–zirconium metal alloy as a fuel for nuclear power plant AP-1000”, Nuclear Engineering and Design, 406: 112257, (2023).
• [12] Pino-Medina, S., François, J. L.”Neutronic analysis of the NuScale core using accident tolerant fuels with different coating materials”, Nuclear Engineering and Design, 377: 111169, (2021)
• [13] Cozzo, C., Rahman, S., “SiC cladding thermal conductivity requirements for normal operation and LOCA conditions”, Progress Nuclear Energy, 106 (March): 278–83, (2018).
• [14] Snead, L. L., Zinkle, S. J., White, D. P. “Thermal conductivity degradation of ceramic materials due to low temperature, low dose neutron irradiation”, Journal of Nuclear Materials, 340: (2-3), 187–202, (2005).
• [15] Katoh, Y., Snead, L. L., Nozawa, T., Kondo, S., Busby, J. T. “Thermophysical and mechanical properties of near-stoichiometric fiber CVI SiC/SiC composites after neutron irradiation at elevated temperatures”, Journal of Nuclear Materials, 403: (1-3), 48–61, (2010).
• [16] Lipkina, K., Hallatt, D., Geiger, E., Fitzpatrick, B.W.N., Sakamoto, K., Shibata, H., et al, “A study of the oxidation behaviour of FeCrAl-ODS in air and steam environments up to 1400 °C”, Journal Nuclear Material, 541: 152305, (2020).
• [17] Shrestha, T.,” Metal urgency kanthal APMT for nuclear-energy application”, (August): 34-6, (2022).
• [18] Hartanto, D., Alshamsi, A., Alsuwaidi, A., Bilkhair, A., Hukal, H.A., Zubair, M., “Neutronics Assessment of Accident-Tolerant Fuel in Advanced Power Reactor 1400 (APR1400)”, Atom Indonesia, 46(3): 177–86, (2020).
• [19] Abrefah, R.G., Atsu, P.M., Sogbadji, R.B.M., “Neutronic safety analysis of proposed reactor technologies for ghana’s nuclear power plant using the MCNP code”, Nuclear Technology Radiation Protection, 34(3): 238–42, (2019).
• [20] Reda, S.M., Gomaa, I.M., Bashter, I.I., Amin, E.A., “Effect of MOX Fuel and the ENDF / B-VIII on the AP1000 Neutronics Parameters Calculation by Using MCNP6”, Nuclear Technology Radiation Protection, 34(4): 325–35, (2019).
• [21] Schulz TL, “Westinghouse AP1000 advanced passive plant”, Nuclear Engineering Design, 236(14–16): 1547–57, (2006).
• [22] Okumura, K., Kugo, T., Kaneko, K., Tsuchihashi, J., “SRAC2006: A Comprehensive Neutronics Calculation Code System”, Tokai JAEA, 44–64, (2007).
• [23] Pinem, S., Sembiring, T.M., Surbakti, T., “PWR Fuel Macroscopic Cross Section Analysis for Calculation Core Fuel Management Benchmark”, Journal of Physics: Conference Series, 1198(2), (2019).
• [24] Luthfi, W., Pinem, S., “Calculation of 2-Dimensional PWR MOX/UO2 Core Benchmark OECD NEA 6048 With SRAC Code”, Jurnal Teknologi Reaktor Nuklir Tri Dasa Mega, 22(3): 89, (2020).
• [25] Luthfi, W., Pinem, S., “Validation of SRAC Code System for Neutronic Calculation of the PWR MOX/UO2 Core Benchmark”, Urania Jurnal Ilmu Daur Bahan Bakar Nuklir, 27(1): 47–56, (2021).
• [26] George, N.M., Terrani, K.A., Powers, J.J., “Neutronic analysis of candidate accident-tolerant iron alloy cladding concepts”, Transaction American Nuclear Society, 109(PART 2): 1506–8, (2013).
• [27] George, N.M., Powers, J.J., Maldonado, G.I., Terrani, K.A., Worrall, A., “Neutronic analysis of candidate accident-tolerant cladding concepts in light water reactors”, Transaction American Nuclear Society, 111: 1363–6, (2014).