Nuclear Engineering and Design最新文献_第5页

PuMMA blind benchmark: Performance of high plutonium content MOX fuel under irradiation

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-03-05 DOI: 10.1016/j.nucengdes.2025.113960

D. Jaramillo-Sierra , M. Stefanowska-Skrodzka , J. Lavarenne , E. Deveaux , E. Brunetto , V. Matocha , A. Magni , K. Sturm , K. Mikityuk , Y. Wang , A. Jiménez-Carrascosa , J. Gado , B. Burger , V. Blanc , V. Dupont , L. Argeles , B. Perrin , G. Michel , A. Scolaro , C. Fiorina , C. Strmensky

This paper presents the results of the first phase of the benchmark exercise in the PuMMA project, aimed at extending the reliability of fuel performance codes for Mixed Oxide (MOX) fuels with high plutonium content. The exercise involved the simulation of three irradiation experiments: (1) CAPRIX (45 % of Pu) carried out in the PHENIX reactor in France, (2) TRABANT1 (45 % of Pu) and (3) TRABANT2 (40 % of Pu) irradiated in the HFR reactor in The Netherlands. Thirteen organizations from nine countries participated, using eight different fuel performance codes. In this phase, besides the fuel specifications, boundary conditions such as linear heat rate, irradiation history, and cladding external temperature were provided. In a “blind” exercise, all codes were free to choose suitable models and correlations for the simulation without access to the experimental results. The main results of the “blind” phase are performed along with a sensitivity analysis to investigate the impact of various uncertainties on measurements, irradiation and input parameters. Significant discrepancies were observed among the different codes, attributed to the use of heterogeneous reference correlations and the absence of suitable models for some MOX-specific phenomena. Finally, insights are offered for the second phase of the benchmark, where the fuel performance codes’ capabilities will be assessed against experimental results from Post-Irradiation Examinations of all three pins.

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引用次数: 0

Aerosol retention in HTGRs: A study on pool scrubbing of graphite and silica particles

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-03-05 DOI: 10.1016/j.nucengdes.2025.113972

Yating Wang , Zhikai You , Zhu Fang , Shumiao Zhao , Xinxin Wu , Libin Sun , Yiyang Zhang

The modular high-temperature gas-cooled reactor (HTGR) is recognized for the inherent safety, i.e., eliminating the possibility of core meltdown. Therefore, the source term becomes the main concern for HTGR accidents, with special focus on the graphite aerosol. However, there is few studies on the active control of graphite aerosol for HTGR. In this study, we propose a preliminary scheme for aerosol retention through pool scrubbing in HTGR water-ingress and loss-of-coolant accidents. The scrubbing characteristics of non-spherical graphite particles are investigated for different gas flow rates and submergences, and compared to the spherical silica particles. For both silica and graphite particles, the trend of decontamination factor (DF) is similar: slowly declining and then rapidly increasing with the increasing particle size. Further analysis with Stokes number indicates that the diffusion dominates below a critical Stokes number, while the inertia impact becomes the main mechanism when above this critical Stokes number. Especially, the DF of graphite particles is significantly lower than that of silica particles in the inertia-controlled regime, due to its irregular shape, porous structure, and hydrophobic nature. A bubble-breaking element is introduced to enhance the retention efficiency of graphite particles, by promoting gas–liquid mixing and more importantly, reducing bubble size in the rising zone. The result shows an average improvement of 135% in scrubbing efficiency, accompanied by a minimal 12.4% increase in pressure drop. This study demonstrates that by introducing an optimized scrubber to the current reactor design, the source term of graphite aerosol can be largely reduced for HTGR accidents with fairly low cost.

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引用次数: 0

Spent fuel pool boil-off experiments with high power and high time resolution

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-03-05 DOI: 10.1016/j.nucengdes.2025.113958

Christoph Schuster , Oleg E. Stepanov

The occurrence of spent fuel pool (SFP) accidents, in which the water level decreases below the top of the fuel assemblies (FA), may arise due to the loss of active pool cooling or pool leakage. In such scenarios, the ALADIN test facility of the TUD Dresden University of Technology is utilized to simulate a comprehensive boiling water reactor (BWR) FA in a 1:1 scale. This report supplements the existing data base with rod powers between 100 W and 350 W. Boil-off experiments will be analyzed in detail in relation to axial and radial temperature profiles as a function of time and heater power up to a maximum temperature of approximately 800 °C. Additionally, temperature excursion mitigating measures such as spray-cooling and flooding will be discussed.

引用次数: 0

Investigation of decay heat removal systems in the ALLEGRO helium-cooled fast reactor

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-03-04 DOI: 10.1016/j.nucengdes.2025.113952

Zsombor Bali , Gusztáv Mayer

ALLEGRO is a demonstrator reactor of the GFR2400 helium-cooled gas fast reactor selected for future research by the Generation IV International Forum (GIF) in the early 2000s. This 75 MW thermal power demonstrator reactor has been developed in several European projects over the past two decades. The current ALLEGRO design consists of pressurized water on the secondary side of both the main heat exchangers (MHX) and the decay heat removal systems (DHR). Nevertheless, in case of a primary-secondary break at the MHXs or the DHRs, the water coolant may ingress into the gas-cooled reactor core which may cause reactivity and corrosion issues. For this reason, replacing the secondary water in the MHXs and the DHRs is an important topic during the design phase of ALLEGRO. In this paper, the gases of helium, nitrogen and CO₂ are investigated as possible secondary coolants of the DHRs at different pressures using the CATHARE thermal hydraulics system code. The results show that from a cooling performance point of view, CO₂ gas is a promising candidate and by using this gas, a relatively small design change is needed concerning the size of the DHR heat exchangers.

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引用次数: 0

Steam generator model design parameter sensitivity study for small modular reactor system

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-03-03 DOI: 10.1016/j.nucengdes.2025.113973

Palash K. Bhowmik , Congjian Wang , Nicholas Hernandez , Tejas Kedlaya , Piyush Sabharwall

This study focuses on design parameter sensitivity studies pertaining to several Once-Through Steam Generator (OTSG) model cases both with and without a riser using python and advanced risk assessment and optimization tool, i.e. Risk Analysis Virtual Environment (RAVEN) developed at Idaho National Laboratory (INL), to support a Small Modular Reactor (SMR) system. The presented Steam Generator (SG) python-based model is a mathematical representation of a steam-generating unit for a Pressurized Water Reactor (PWR)-type SMR system, including fluid flow and heat transfer equations, models, and correlations. Design studies involve changing the model’s input design parameters (e.g., temperature, pressure, mass flow rate) to observe the resulting effects on the output of the system, such as the Heat Transfer Coefficient (HTC), Reynolds number, Nusselt number, and heat transfer performance. Sensitivity studies analyze the degree to which system output and/or desired parameters (e.g., HTC or heat transfer performance) are sensitive to changes in the input parameters. By using RAVEN, detailed design parametric sensitivity studies. Six input parameters—namely, the pressure, temperature, and mass flow rate for the inlet of the primary-side (hot fluid) and secondary-side (cold fluid) of the SG—were randomly perturbed via RAVEN’s Monte Carlo Sampler module, using uniform distributions (i.e., ±1%, ±5% and ±10 % relative changes) for 600 samples. The analysis results give valuable insights into SG system performance, and provide justification for further research and development such as optimized sensor placement, design verification, validation, and optimization.

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引用次数: 0

Neutronic uncertainty propagation for the SAFARI-1 research reactor benchmark: IAEA CRP T12029

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-03-03 DOI: 10.1016/j.nucengdes.2025.113943

O.E. Montwedi , V.V. Naicker , G.P. Nyalunga

Quantification of full core neutronic and sensitivity analysis of the SAFARI-1 research reactor was performed; the output parameter of interest was the neutron multiplication factor k_eff. Fuel element calculations were also studied to gain insights into the problem. The calculational approach in the Organization for Economic Corporation and Development benchmark for quantification of uncertainties in multiphysics calculations for LWRs was followed. The scope of the work is limited to reactor physics (neutronic calculations). Five core models were built in line with the SAFARI-1 benchmark specifications, of the 2016 IAEA Coordinated Research Programme. Model 1 was a Fuel Element (FE) model with reflective boundary conditions on the sides and void boundary conditions axially. Models 2 – 4 were full core models with control rods at 0 %, 50 % and 100 % insertion in the core. Model 5 was a full core model consisting of only fuel elements.

TSUNAMI-3D, which is part of the SCALE 6.2.3 code package, was used for uncertainty propagation. The uncertainty in the multiplication factor k_eff is calculated with the sensitivity coefficients and the covariance matrix using the so-called “sandwich rule”. It can be seen from the results that the uncertainty in k_eff is 0.490 % for the infinite fuel element model and about 0.640 % (highest value) for the full core model. It is also seen from the sensitivity analysis that the highest contributors to the uncertainty in the FE model are due to the average number of neutrons released per fission and due to the (n, gamma) reactions. For the full core model, the highest contributors are uncertainties due to the average number of neutrons per fission and that due to the fission spectrum, chi. A detailed behaviour of the individual contributors to the uncertainty is also investigated. It was also found that the size of the reactor core influences the value of the individual contributors. To a lesser extent, the heterogeneity of the reactor core also influences the value of the individual contributors. The neutron flux distribution in cores of different sizes was also studied since it contributes to the uncertainty due to chi, as well as the neutron importance, to observe the trend in the individual uncertainty contributions of chi and nubar.

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引用次数: 0

Study on high-resolution activation analysis based on the Monte Carlo global variance reduction method

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-03-01 DOI: 10.1016/j.nucengdes.2025.113961

Leiming Li , Zhenping Chen , Aikou Sun , Chao Yang , Tao Yu

During nuclear reactor operations, neutron activation reactions generate significant quantities of radionuclides from the structural materials, directly impacting shielding design, maintenance planning, and decommissioning strategies. This is a critical component of radiation safety analysis. As advanced nuclear reactor technology evolves, the increasing complexity of reactor geometry, material configurations, and neutron spectra complicates activation analysis. Consequently, there is a pressing need for high-resolution activation analysis of nuclear reactors. This paper presents a high-resolution activation analysis method for large-scale complex structural materials, utilizing the Monte Carlo global variance reduction particle transport technique. A fully automated coupled high-resolution activation analysis program is developed, enabling the calculation of high-resolution decay source distributions and precise evaluation of decay photon sources for extensive complex structural materials. The methodology is benchmarked against the shutdown dose rate benchmark released by the International Thermonuclear Experimental Reactor (ITER) program. Additionally, an application study of high-resolution activation analysis is conducted on a standard pressurized water reactor (PWR). The methodology demonstrated in this paper holds significant engineering value, enhancing the accuracy of activation calculations for large-scale nuclear reactor structural materials and it provides guidance for optimizing shielding design to reduce radiation exposure during operation.

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引用次数: 0

Enhancing uncertainty analysis: POD-DNNs for reduced order modeling of neutronic transient behavior

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-03-01 DOI: 10.1016/j.nucengdes.2025.113969

Yijun Zhang , Wenhuai Li , Sitao Peng , Jinggang Li , Ting Wang , Qingyun He , Tao Wang , Haoliang Lu , Ling Zeng

In reactor safety analysis, sensitivity analyses on critical parameters are essential for ensuring the reliability of safety conclusions, particularly regarding transient behavior, which often requires time-consuming computations. Developing surrogate models presents a promising solution. This paper extends the Proper Orthogonal Decomposition-Radial Basis Function (POD-RBF) framework to the 3D Light Water Reactor Core Transient Benchmark (3DLWRCT) for control rod ejection accidents. The primary aim is to simulate transient behavior under random perturbations in the macroscopic neutronic cross-sections of fuel assemblies.

Our results indicate that the traditional POD-RBF approach struggles to accurately reconstruct the highly nonlinear transient system, whether through spatiotemporal folding or spatial data reduction. To overcome these challenges, we enhance the model by integrating Deep Neural Networks (DNNs) and employing the Tree-structured Parzen Estimator for optimal neural network architecture selection. This improved approach significantly increases the accuracy of the surrogate models, demonstrating its feasibility and effectiveness. The integration of DNNs offers a deeper understanding of complex interactions within the reactor core, effectively capturing nonlinearities and yielding reliable predictions even under uncertainty.

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引用次数: 0

VVER long-term operation – A review based on the material studies results from past and ongoing EU-supported research projects

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-03-01 DOI: 10.1016/j.nucengdes.2025.113949

Vladimir Slugen , Jana Simeg Veternikova , Maria Domankova , Matus Gavalec , Jana Petzova , David Slnek , Mykola Dzubinsky , Oleksii Shugaylo , Ildiko Szenthe , Ferenc Gillemot , Maksym Zarazovskii , Jari Lydman , Radim Kopriva , Szabolcs Szavai , Ondrej Srba

Safe and long-term operation (LTO) of nuclear power plants (NPP) with Water-Water Energetic Reactors (VVER) is essential for several central and eastern European countries to keep their energy supply security. During the last decades, several EU-supported projects in framework projects or Horizon2020 schemes have focused on the material degradation of pressurised water reactors (PWRs), especially VVERs. This paper aims to extract the most important results dominantly from EURATOM projects which support or could limit the long-term operation of this Soviet design of PWRs. The accent is given on ongoing projects such as FRACTESUS, STRUMAT-LTO, APAL, and ENTENTE. Actually, running project DELISA-LTO (2022–2026) comprises not only results from previous research in this area but also incorporates wide material databases coming from nuclear power plant V-1 in Jaslovské Bohunice (Slovakia), which was shut down after 28 years of operation and provides wide-range of original service-aged materials specimens for profound material studies having in mind lifetime extension of VVERs. Consolidated knowledge focused mainly on the reactor pressure vessel (RPV) material studies is summarised and supported by relevant references.

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引用次数: 0

[NURETH-20] Evaluation on mitigation performance of flooding safety system under hypothetical loss of coolant accident in Korean i-SMR with MELCOR code

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-02-27 DOI: 10.1016/j.nucengdes.2025.113950

Chang Hyun Song , Jae Hyung Park , JinHo Song , Sung Joong Kim

In response to the growing interest in Small Modular Reactors (SMRs) globally, many countries are actively pursuing the development of high-power SMRs, based on their inherent advantages including enhanced safety, grid flexibility, and potential for hydrogen production. Among these endeavors, an i-SMR with an electrical power output of 170 MWe has been under development since 2021 by Korea Hydro & Nuclear Power Co., Ltd. in Republic of Korea. The i-SMR has established ambitious top-tier requirements, such as core damage frequency less than 1.0 × 10⁻⁹/module-year and large early release frequency less than 1.0 × 10⁻¹⁰/module-year, and emergency planning zone within nuclear power plant site boundary. All of which is extremely challenging and necessitates innovative safety systems with exceptional reliability. In this context, this study proposed a Flooding Safety System (FSS) as a novel safety system, and its mitigating performance under a hypothetical accident scenario was evaluated by using MELCOR code for validating the efficacy of this conceptual approach. To conduct the accident analysis, a MELCOR input model for the i-SMR was developed. The initiating event assumed was the stuck open of depressurization valve, leading to the discharge of coolant from the primary system into the metal containment vessel, which can be deemed as a loss of coolant accident. The findings in this study revealed that timely operation of the FSS can prevent core damage, thus validating its crucial role to assuring the integrity and reliability of the i-SMR.

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引用次数: 0