Statistical error propagation in HTR burnup model

Kępisty, G.; Oettingen, Mikołaj; Stanisz, Przemysław; Cetnar, Jerzy

doi:10.1016/j.anucene.2017.03.039

Cited by 10 publications

(7 citation statements)

References 19 publications

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“…The modular high-temperature gas-cooled reactor (MHTGR), as a category of Small Modular Reactors (SMRs), has garnered widespread recognition for its inherent safety and highly efficient energy generation [21,22]. Figure 1 shows a MHTGR module, developed by the Institute of Nuclear and New Energy Technology at Tsinghua University [23][24][25], featuring a helical-coil once-through steam generator (OTSG) and a one-zone pebble-bed MHTGR arranged side by side with the OTSG.…”

Section: Modular High Temperature Gas-cooled Reactormentioning

confidence: 99%

Abnormal Event Detection in Nuclear Power Plants via Attention Networks

Zhang,

Jia,

Guo

et al. 2023

Energies

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Ensuring the safety of nuclear energy necessitates proactive measures to prevent the escalation of severe operational conditions. This article presents an efficient and interpretable framework for the swift identification of abnormal events in nuclear power plants (NPPs), equipping operators with timely insights for effective decision-making. A novel neural network architecture, combining Long Short-Term Memory (LSTM) and attention mechanisms, is proposed to address the challenge of signal coupling. The derivative dynamic time warping (DDTW) method enhances interpretability by comparing time series operating parameters during abnormal and normal states. Experimental validation demonstrates high real-time accuracy, underscoring the broader applicability of the approach across NPPs.

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Section: Modular High Temperature Gas-cooled Reactormentioning

confidence: 99%

Abnormal Event Detection in Nuclear Power Plants via Attention Networks

Zhang,

Jia,

Guo

et al. 2023

Energies

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“…The results showed that the fuel particles' random distribution leads to a slightly larger effective multiplication factor than that of the regular model, and the main reason for the deviation between the two models is the difference in the spatial and angular distribution of the two-particle arrangements [15]. G. Kepisty of AGH University of Science and Technology used 100 independent copies of Monte Carlo burnup simulations to study the statistical error propagation of the full-core HTR model [16]. The results show that the actual and apparent changes of K eff are close to each other from the beginning of irradiation to high burnup.…”

Section: Introductionmentioning

confidence: 99%

Uncertainty Quantification of Engineering Parameters for a Nuclear Reactor Loaded with Dispersed Fuel Particles

Li,

Chen,

Yang

et al. 2024

Energies

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Owing to their high intrinsic safety, dispersed fuel particles are an important fuel pattern for fourth-generation nuclear reactors. Due to the unique cladding layers and the random dispersion characteristics, dispersed fuel particles significantly differ from pressurized water reactors regarding operation-induced uncertainty. This study quantitatively analyzed overall uncertainty while considering a random distribution of dispersed fuel particles, material thickness, and fuel enrichment. The results demonstrated that, for all packing fractions, the uncertainty induced by the random dispersion of dispersed fuel particles was below 0.03%. For every packing fraction, the differences between the results obtained by the regular and the random distribution models increased, and then decreased, until reaching its maximum (1.297%) at 15%. Keff decreased as the radius of the UO2 kernel increased; Keff increased as the thickness of the cladding layer increased; the uncertainty of Keff was 1.003% when a random distribution of particles, material thickness, and fuel enrichment were taken into consideration; the uncertainty of the power distribution of reactor core assemblies was maximized (1.495%) at the edge of the reactor core. Quantitative analysis of uncertainty provides references for the optimization of design and safety margin analysis for reactors.

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“…The two methods are Transmutation Trajectory Analysis (TTA) and Chebyshev Rational Approximation Method (CRAM). In the present study, the CRAM was used based on its matrix exponential solution (Leppänen, 2009a;Jaakko, 2013;Korkmaz and Agar, 2014;Herrero et al, 2016;Kępisty et al, 2017). Furthermore, the cross-section data libraries: radioactive decay and fission yield and also thermal scattering libraries are all used by the Serpent.…”

Section: Introductionmentioning

confidence: 99%

Serpent Code Point Kinetic Parameters Analysis: A Nexus for Nigeria Research Reactor – 1 Core Conversion Using Enriched Uranium Dioxide Fuel

Abubakar¹,

Ismaila²,

Yamusa³

2022

FJS

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The International Atomic Energy Agency (IAEA) requires that all test and research reactors operating on Higher Enriched Uranium (HEU) should be converted to Low Enriched Uranium (LEU) for safety and security purposes. Nigeria having a Miniature Neutron Source Reactor (MNSR) has been long interested in fuel technological research not just to develop the area but also to meet with resolution on the nuclear treaty set out by the global nuclear regulatory body. In this study, reactor kinetic parameters such as effective delayed neutron fractions, prompt neutron lifetime as well as mean neutron generation time were analysed for Nigerian Research Reactor-1 (NIRR-1). Serpent Monte Carlo code 1.17 is used in the analysis. For delayed neutron parameters determination, we used fission probability iteration under one averaged generation time and neutron population rate.The calculated values for delayed neutron were recorded as analogue prompt and implicit prompt neutron lifetime, reproduction time and emission time are in the order of 3×10-7 (s), in agreement with the calculated data from the nuclear data libraries and some literature.The result for delay neutron fraction and other time-based parameters support the fuel core conversion for NIRR-1.The computational and pictorial results obtained from Serpent code simulation described well the transient behavior of the delayed neutron in this reactor.The analytical results also spelled out the relevance and compatibility of low enriched uranium dioxide fuel over higher enriched type.The result of this study conforms with other results obtained from similar reactors but with different Monte Carlo codes and with higher enriched uranium

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Statistical error propagation in HTR burnup model

Cited by 10 publications

References 19 publications

Abnormal Event Detection in Nuclear Power Plants via Attention Networks

Abnormal Event Detection in Nuclear Power Plants via Attention Networks

Uncertainty Quantification of Engineering Parameters for a Nuclear Reactor Loaded with Dispersed Fuel Particles

Serpent Code Point Kinetic Parameters Analysis: A Nexus for Nigeria Research Reactor – 1 Core Conversion Using Enriched Uranium Dioxide Fuel

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