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Numerical simulation study on the filtration performance of metal fiber filters Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-12 Song Ma, Yanmin Zhou, Zhongning Sun, Haifeng Gu, Yin Wang, Xiang Yu
As an essential component of the ventilation system of nuclear facilities, filters have significant filtration performance for the safety of the environment and personnel. The metal fiber filter has excellent temperature resistance and can still maintain good filtration performance under fire conditions. In the filter design process, it is usually necessary to fold the metal fibrous fabric into a pleated
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Lithium stabilization of amorphous ZrO2 Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-12 Gareth F. Stephens, Jack A. Wilson, Simon F. Curling, Guanze He, P. John Thomas, David W. Williams, Susan Ortner, Chris Grovenor, Michael J.D. Rushton, Aidan Cole- Baker, Simon C. Middleburgh
Small modular reactors (SMRs) are a key option to aid the worldwide net zero targets for carbon emissions. Some pressurised water reactors aim to operate with a boron-free coolant chemistry for simplification in plant design. In the absence of boron, Li has been found to accelerate the corrosion of the zirconium-based alloy fuel cladding under certain conditions and concentrations within pressurised
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A study with PUREX aqueous-organic pair in Taylor-Couette mixing field Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-12 Shekhar Kumar
The hydrodynamics and mass transfer in the Taylor-Couette mixing field are important activities for concurrent process intensification work for miniaturization of the chemical reactor volume and enhancing the energy input per unit volume. In the literature, many studies have been reported for mixing of conventional two-phase systems of industrial interest. However reported studies on mixing of PUREX
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Assessment of thermal shock resistance of refractory magnesia lining under simulated core melt impingement for application to SFR core catcher Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-12 Prabhat Kumar Shukla, Hemanth Rao E, Anish Kumar, Sanjay Kumar Das, Ponraju D, Venkatraman B
Despite a hypothetical event, whole core melt accident in a sodium cooled fast reactor is investigated as a part of plant safety analysis. To prevent breach of primary boundary in such accidents, a Core Catcher (CC) is provided at the bottom of the main vessel for retention of corium in subcritical and coolable state. For future Indian FBRs, refractory magnesia-lined CC is planned for protection against
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Review of thorium-containing fuels in LWRs Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-12 Maria Hendrina Du Toit, Frederik Van Niekerk, Shervineh Amirkhosravi
Stemming from a renewed interest in the suitability of thorium as fertile isotope, this study presents an extensive survey of some of the recent advances on the use of thorium in light water reactors, with the view of informing researchers and policy makers. The isotopic properties of fertile and fissile material are discussed, with an emphasis on the suitability and unique characteristics of thorium
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Utilizing even Plutonium Isotopes as burnable absorbers for controlling the reactivity and power distribution in Pressurized Water Reactors Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-12 Sayed Saeed Mustafa, Esmat A. Amin
In this paper, a new approach is proposed for controlling the reactivity and power distribution in pressurized water reactor cores without the implementation of burnable absorbers. This approach depended on the homogeneous mixing of uranium oxide with even plutonium isotopes; Pu-238, Pu-240, and Pu-242. Three Plutonium fuels; (99% UO + 1% Pu-238), (99% UO + 1% Pu-240), and (98% UO + 0.5% Pu-238 + 1%
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Applying U.S. metal fuel experience to new fuel designs for fast reactors Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-11 Douglas C. Crawford, Douglas L. Porter
With the increasing interest in small modular reactors or microreactors, developers are working to design and submit licensing approval requests of U–10Zr-fueled fast reactors. The developers and their proponents cite prior metal fuel experience (worldwide, but U.S. experience in particular for many developers) as the motivation and justification for their reactor concepts. The experience with metal
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Experimental study on fluid-structure interaction of multiple support barrels in liquid metal fast reactor Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-10 Yu Liu, Dexuan Duan, Chao Deng, Yuxuan Zhu, Yuchao Wang, Yang Zhang, Daogang Lu
In an earthquake, the main equipment support barrels in the pool-type liquid metal fast reactor (LMFR) may experience strong vibrations due to the fluid-structure interaction (FSI) phenomenon. Accurate parameters for FSI characteristics of multiple cylinders are crucial for the seismic design and analysis of these support barrels. Limited experimental data is available for pressure characteristics
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Role of structure and organic contaminants on Cs Sorption by clays Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-10 Rahul Sirvi, Harshala Parab, Nistha Singh, Pranesh Sengupta, Sangita D. Kumar, P.S. Ramanjaneyulu, Uttam K. Bhui
Long-term confinement and isolation of long-lived radionuclides from the biosphere is necessary in view of safety aspects of nuclear reactor activities. Geological materials show favorable sorption characteristics for various radionuclides resulting in their retention for a geological time-scale. In the present study, structural profile of three different clay geosorbents was evaluated using various
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Analysis of loss of flow without scram test in the FFTF reactor – Part II: System thermal hydraulics with point neutron kinetics Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-08 Alexander Ponomarev, Evgeny Nikitin, Emil Fridman
This study presents a benchmark analysis of an unprotected loss of flow transient in a sodium-cooled fast reactor at the Fast Flux Test Facility (FFTF), carried out as part of an IAEA coordinated research project. Three codes, namely Serpent (Monte Carlo), DYN3D (3D nodal diffusion) and ATHLET (system thermal hydraulics), were employed in the benchmark exercise. Two distinct modeling approaches were
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Study on the difference between B4C powder and B4C pellet regarding the eutectic reaction with stainless steel Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-07 Zhenhan Hong, Zeeshan Ahmed, Marco Pellegrini, Hidemasa Yamano, Nejdet Erkan, Avadhesh Kumar Sharma, Koji Okamoto
The development of Generation IV Sodium-cooled Fast Reactors (SFRs) faces a crucial challenge concerning Core Disruptive Accidents (CDAs). The eutectic reaction between boron carbide (BC) and Stainless Steel (SS) can lead to boron migration, resulting in the formation of a eutectic melt that may relocate extensively within the core, thereby affecting neutron balance in the disrupted core. The study
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Analyzing geometric parameters in an inclined wavy-porous cavity filled with magnetic hybrid nanofluid containing a square solid block Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-06 BalaAnki Reddy P, Salah T, M.A Mansour, A.M Rashad, Nabwey HossamA, Shaik Jakeer
Heat transfer through enhanced hydromagnetic mixed convection has the potential to be of long-term benefit in high-performance thermal equipment, hybrid fuel cell technologies, cooling systems for microelectronic devices, and subterranean cable networks. The purpose of this study was to investigate the influence of an inclined magnetic field thermal radiation and a heat source/sink on the flow and
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Modeling interconnections of safety and financial performance of nuclear power plants, Part 2: Methodological developments and case study Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-06 John Beal, Tatsuya Sakurahara, Pegah Farshadmanesh, Seyed Reihani, Ernie Kee, Arden Rowell, Fatma Yilmaz, Zahra Mohaghegh
This research study theorizes and quantifies the interconnections of safety and financial performance of nuclear power plants (NPPs). NPP safety refers to occupational safety and system safety, while financial performance refers to the monetary values associated with operation and maintenance (O&M) strategies. The results of this research study are summarized in three journal manuscripts including
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Review and comparison of neutronic parameters of VVER-1200 and PWR fuel assemblies with different burnable absorbers Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-06 D. Acikgoz, T. Akyurek
Burnable absorbers offer a promising approach for achieving safe and efficient energy generation through the control of fission reactions within nuclear reactors. In this study, the neutronic parameters of Pressurized Water Reactor (PWR) and Water-Water Energy Reactor-1200 (VVER-1200) fuel assemblies were examined and compared using the Monte Carlo N-Particle (MCNP) code with different burnable absorbers
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Numerical investigation on the thermal performance of spirally coiled tubes heat exchanger for nuclear pump under seal injection water interruption Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-04 Xiaoming Chen, Zhiqiang Huang, Peng Shu, Siyuan Chen, Kunyan Liu, Xide Lai
The spirally coiled tube heat exchanger (SCTHE) applied in the nuclear pump is an important heat transfer equipment, of which the main function is to cool leaking fluid from the main loop coolant in case the seal injection water is interrupted. In this study, we have proposed a strategy that simplifies the thermal barrier to conduct numerical simulations for SCTHE. The RNG - turbulence model incorporating
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Experimental investigation on effect of rod bowing on subchannel void fraction in 5 × 5 rod bundles Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-04 Jie Wan, Lei Gao, Wan Sun, Quanyao Ren, Wenxiong Zhou, Luteng Zhang, Longxiang Zhu, Zaiyong Ma, Liang-ming Pan
During operation of a nuclear power plant, the bending deformation of fuel rods usually occurs and has a substantial effect on the heat transfer and flow behavior of coolant. To assess the influence of rod bowing on thermal-hydraulic phenomena in a nuclear reactor, it is crucial to consider the void fraction characteristics in rod bundles. In this study, experiments were conducted using an irregular
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Enhanced PAUT method with pulse compression for heavy-walled CASSs in nuclear power plants Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-03 Yuxuan Wu, Junteng Hu, Cuixiang Pei, Fangjie Shi, Zhenmao Chen
Heavy-walled cast austenitic stainless steels (CASSs) are usually used as the primary circuit pipelines in many nuclear power plants, but detecting the small defects in these materials has been a difficult problem for many years. To improve the defect-detecting ability and signal-to-noise ratio (SNR) of ultrasonic testing for CASSs with serious sound attenuation and scattering, an enhanced phased array
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Experimental investigation on transition boiling during reflooding in a narrow rectangular channel with high wall temperature Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-01 Peng Feng, Hanzhou Liu, Mingjing Chen, Shuhua Ding, Dan Wu, Jian Deng, Haidong Liu, Deqi Chen
During LOCA (loss of coolant accident) of a nuclear reactor, the complex heat transfer during reflooding with high wall temperature of the fuel element is very important to nuclear safety. The transition boiling regime is a specific regime with a transition from film boiling to nucleate boiling during reflooding of the nuclear fuel element. In this study, a bottom reflooding experiment is carried out
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Lightweight optimization of space reactor loop heat pipe radiation radiator based on surrogate models Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-01 Yuandong Zhang, Ren Li, Xiang Zhang, Zhou Tao, Genglei Xia, Xue Du
Loop heat pipes, as a type of two-phase thermal control device, can not only eliminate the reliance on pump transportation technology in single-phase fluid circuits, but also utilize the structural features of gas-liquid pipeline separation to reasonably achieve the evaporation and condensation process. Considering that the loop heat pipes have good application prospects in the field of heat dissipation
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DYSN: Dynamics simulator for the NuScale SMR - A mathematical framework for transient analysis Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-01 A. Fakhraei, F. Faghihi, M. Mohammadi, A. Rabiee
Several innovations have been devised to design a system with lower capital cost and passive safety features in NuScale SMR. The system incorporates primary natural circulation and helically-coiled steam generators, leading to unique dynamical and stability aspects. A mathematical model is developed for both the primary and secondary sides of the reactor, consisting of 56 coupled ordinary differential
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CFD simulation on sub-channel blockage in the fuel plate of RSG-GAS reactor Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-03-01 Sukmanto Dibyo, Surian Pinem, Farisy Yogatama Sulistyo, Veronica Indriati Sriwardhani
The main objective of the paper is to simulate flow blockage in the sub-channel fuel plate in the Indonesian RSG-GAS research reactor. Utilizing SolidWorks CFD Flow Simulation, this simulation investigates coolant and fuel plate temperatures in scenarios involving blockage areas from 60% to 90%. In the first scenario, the blockage is positioned at the inlet sub-channel, while the second scenario involves
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CHF experiments and prediction model under subcooling flow boiling condition in passive IVR-ERVC system Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-29 Shilei Han, Pengfei Liu, Gang Wang, Bo Kuang, Yanhua Yang
In-Vessel Retention (IVR) with External Reactor Vessel Cooling (ERVC) is proved to be the effective strategy for the protection of the Reactor Pressure Vessel (RPV) integrity during the severe accidents. In the prototype reactor, the lower head is cooled by natural circulation of the subcooled coolant in the In-Containment Refueling Water Storage Tank (IRWST). In such flow boiling process, the heat
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A framework for capturing and representing the process to classify nuclear waste and informing where processes can be automated Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-29 Seonaid Hume, Graeme West, Gordon Dobie
Decommissioning and dismantling of nuclear facilities are complex processes, where an accurate triage of visual and radiological characterisation is an important driver of how this process is executed. In-situ measurements before dismantling are essential for effective, optimized waste management solutions to ensure the safe and secure decommissioning of nuclear installations. Characterising nuclear
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Experimental and numerical research on flow-induced vibration characteristics of Hydraulic Suspended Passive Shutdown Subassembly in SFR Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-29 He Wang, Yu Liu, Daogang Lu, Qiong Cao, Wanmin Zhang
Hydraulic Suspended Passive Shutdown Subassembly (HS-PSS) is a passive nuclear safety technology employed in Sodium-Cooled Fast Reactors (SFR). During Sodium Fast Reactors (SFR) operation, this component is often subjected to stress cycles caused by flow-induced vibration (FIV), which may lead to material failure and potentially threaten the reactor's safety. Therefore, it is necessary to investigate
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Numerical approximation of a fractional neutron diffusion equation for neutron flux profile in a nuclear reactor Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-28 Pradip Roul, Vikas Rohil, Gilberto Espinosa-Paredes, K. Obaidurrahman
This paper focuses on development of an efficient numerical technique for approximation of a fractional neutron diffusion equation with delayed neutrons for neutron flux profile in a nuclear reactor. The method is used to approximate the Caputo time-fractional derivatives in the governing equation. A collocation technique based on quintic B-spline (QBS) basis function is employed for discretization
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Exploring the radiation shielding properties of B2O3-SiO2-ZnO-Na2O-WO3 glasses: A comprehensive study on mechanical, gamma, and neutron attenuation characteristics Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-28 Khalid Alsafi, Yasser A.M. Ismail, Dalal Abdullah Aloraini, Haifa M. Almutairi, Wafa M. Al-Saleh, Kh S. Shaaban
W-doped multicomponent BO-SiO-ZnO-NaO glasses in a range of concentrations (0–6 mol %) have been manufactured using the melt quenching method. The result of replacing NaO with WO on the structure of zinc borosilicate glasses was examined using XRD. The density of the glass system has been determined using Archimedes' method. The substitution of NaO with WO in the glass composition would result in increased
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An FE-MCS based modeling method for event sequence development for the multi-module probabilistic risk assessment Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-28 Hongru Zhao, Jiejuan Tong, Ao Liu, Pengcheng Peng, Jun Zhao
The Multi-Unit Probabilistic Risk Assessment (MUPRA) has been a research hotspot since the Fukushima Daichi accident in 2011. Among the technical elements, event sequence modeling is the most challenging one in MUPRA implementation. During the PRA development for a 600 MW HTR-PM nuclear power plant (HTR-PM600) which consist of 6 reactor modules, we find that the complexity of the obtained event sequence
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Effectiveness of green-synthesized nickel-doped calcium ferrite nanoparticles in the X-ray/gamma radiation shielding applications Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-27 Umashankar raja. R, H.C. Manjunatha, Y.S. Vidya, L. Seenappa, Krishnakanth E., K.N. Sridhar, R. Munirathnam
Nanoparticles comprising calcium ferrite doped with Nickel within a concentration range of 10 to 50 mol% were successfully produced using a solution combustion technique, employing neem leaf extract as a reducing agent. The subsequent calcination process at 500 °C validated the formation of orthorhombic calcium ferrite, featuring distinct Bragg reflections indicative of Nickel doping. An analysis via
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3D LES of natural convection in the side-heated vertical wall with cryogenic helium up to Ra≈1015 Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-27 Songzhi Yang, Raksmy Nop, Alain Girard, Davide Duri, Etienne Studer
The passive safety concept of Small Modular Reactors (SMR) is based on the extraction of residual heat from the reactor to a surrounding water pool. However, the large scale of the reactor vessel (height 15 ) can lead to a rather intensive heat exchange process mostly by natural convection (Rayleigh number . Reliable heat transfer correlations exist to date only up to , with uncertainties in the extrapolation
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How does nuclear energy consumption contribute to or hinder green growth in major nuclear energy-consuming countries? Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-26 Weiming Gao, Sana Ullah, Syeda Maria Zafar, Ahmed Usman
The importance of nuclear energy as a possible engine of green development is growing as worldwide worries about climate change and ecological sustainability rise. Previously, some literature has tried to analyze the factors that can impact green growth; however, none of them has shed light on the possible influence of nuclear energy consumption on green growth in major nuclear energy-consuming economies
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Transient analysis and dynamic modeling of the steam generator water level for nuclear power plants Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-23 Xinyu Sun, Fei Song, Jingqi Yuan
Natural circulating steam generators are extensively applied in pressurized water reactor units, where the water level control is a tough challenge. During transients, the water level may change drastically, and its reverse dynamic characteristics may even cause safety concerns. This article studies the dynamic effects of several crucial impact factors on the level, including their gains, time constants
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A goal-oriented emergency countermeasure planning method using graph-based path search and functional reasoning Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-23 Jun Yang, Haoming Ma, Yueming Hong, Shuxin Bi, You Xue, Wenlin Wang
In the paper, we present a novel method for creating mitigation action paths in support of emergency response control and protection of process safety industrial systems. The mitigation action planning is implemented using deductive reasoning for anti-degradation goals and objectives with critical success channel identification in supervisory control of plant processes. First, a conceptual framework
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Fractional Bernstein polynomial method for solving time-fractional neutron diffusion systems Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-23 Yasser Mohamed Hamada
In this research, an innovative numerical method is presented to solve the fractional space-time multigroup neutron diffusion model. The method employs the fractional-order Bernstein polynomials for temporal calculations and higher order finite difference schemes for spatial discretization. Two major drawbacks should be addressed to solve such fractional diffusion model. The first is that the spatial
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Dynamic simulation of a space lithium-cooled reactor system coupled with a He–Xe closed Brayton cycle Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-23 Zongyun Wu, Tiancai Liu, Lin Qi, Lin Sun, Hongwei Yang, Mingyu Wu
The lithium-cooled reactor with Brayton conversion power system mainly consists of fast reactor, intermediate heat exchanger, closed Brayton cycle (CBC) loops filled with Helium–Xenon (He–Xe) binary gas mixtures and heat rejection loop which radiate residual heat through heat pipes to outer space. In this study, a dynamic model of 100-kWe lithium-cooled reactor with Brayton conversion power system
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Analysis of climatic conditions effect on passive containment cooling system reliability in AP1000 for multi-unit nuclear power plant site Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-21 Yu Yu, Wanxin Feng, Guanyu Liu, Fenglei Niu, Yan Dong, Kun Yu
Risk of multi-unit Nuclear Power Plant (NPP) site gets more and more attention in recent years, because more than one unit on the same site can be influenced by the environment. Functional failure is one of the important factors for passive safety system reliability analysis, and the climatic conditions of plant site have significant effect on functional failure evaluation. In this paper, based on
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Improved selectivity of trivalent Am over Cm by modulating donor centres of aza-macrocyclic ligand with soft donors: A theoretical comparison Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-21 Saparya Chattaraj, Arunasis Bhattacharyya
Efficient management of the high-level liquid waste (HLW) generated during the reprocessing of nuclear fuel involves mutual separation and transmutation of the minor actinides like Am and Cm, which are chemically very similar. The structure–activity relationships between the extractants and their separation performances are expected to provide practical approaches for designing proficient ones for
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Review of the representative development history on rod bundle mixing coefficient used in subchannel analysis code of PWR Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-19 Bin Han, Xiaoliang Zhu, Bao-Wen Yang, Aiguo Liu, Yanyan Xi, Yuanyuan Yin, Shenghui Liu, Tianyang Xing
Critical Heat Flux (CHF) is one of the most important indices in determining reactor core safety and economics. As well known,the Mixing Vane Grid(MVG) in the rod bundle reactor core will provide high mixing performance that could mix the coolant and increase CHF. The subchannel analysis code is widely used in predicting CHF and developing CHF correlations. To evaluate the mixing performance of the
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3D analysis of spray effect on long-term depressurization of VVER-1000 containment during LB-LOCA Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-18 Muhammet Enis Kanik, Omid Noori-kalkhoran, Massimiliano Gei, Kevin Fernández-Cosials
Containment as the last barrier of defence in depth strategy plays the most important role in protecting the environment and public from the release of radioactive material. As a result, analysing its performance in nuclear accidents and the reliability and effectiveness of its engineering safety features (ESFs) in the depressurization of containment is a key step in safety assessment. This study demonstrates
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Physical study of temperature measurement by neutron resonance absorption of tungsten Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-17 Jing Shang, Wei Luo, ChunMing Hu, LongWei Mei, JianFei Tong, Bin Zhou, HaiTao Hu, WenTing Du, ChaoJu Yu
The method of temperature measurement with neutron resonance absorption is a non-contact temperature measurement technique, which makes use of the excellent penetration characteristics of neutrons and the Doppler broadening of the neutron resonance absorption cross section to obtain the static and dynamic temperature parameters as well as the temperature distribution in a closed system. The height
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Experimental and numerical investigation of aerosol penetration through submerged gravel bed scrubber Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-17 Arjun Pradeep, Sanjay Kumar Das, D. Ponraju, B. Venkatraman
Submerged gravel bed scrubbers (SGBSs) are used as air cleaning systems for removing combustion aerosols from carrier air, under the accidental event of sodium fire. In-house experimental investigations show that semiempirical models used till date underpredict sodium fire aerosol removal efficiencies at low air flow rates. This brings in the requirement for advanced computational fluid dynamics (CFD)
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Effect of end wall boundary layer suction on the performance of high load helium compressor cascade Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-16 Yingqi Fan, Zhitao Tian, Jinzhao Zheng, Jianchi Xin, Huawei Lu
The high-load design method of helium compressor effectively solves the challenging of helium compression issue in the high-temperature gas-cooled reactor (HTGR). However, this design method also leads to a more significant airflow turning angle in stators of high-load helium compressor, which makes the separation of working fluid in the corner area of the stators become the main factor affecting the
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232U, 228Th, 227Ac, and 226Ra primary radioisotopes: High-power sources for nuclear batteries Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-16 O.A.P. Tavares, E.L. Medeiros, M.L. Terranova
The present communication deals with the evaluation of the cumulative energy produced by the sequential α- or α- and β-decay processes of U, Th, and Ra isotopes, ending in the β-decaying Pb or Pb isotopes, and by the β of Ac, which ends in the Pb isotope. The obtained results evidence that the overall decay chains greatly contribute to total specific power and total specific energy released by these
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Supercritical carbon dioxide critical flow model based on deep learning Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-15 Yuan Yuan, TianSheng Chen, Yuan Zhou, HaoYang Feng, JunHao Wang, HouZhong Zhai, YuTing Zha, Yukai Meng
The break accident process of a supercritical carbon dioxide (SCO2) reactor system presents a transcritical phenomenon. Operating under high pressure and a wide parameter range, the SCO2 system introduces multiphase characteristics to the critical flow of carbon dioxide (CO2) at various system positions. Nevertheless, current research lacks a comprehensive critical flow model capable of accommodating
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Virtual-Reality training solutions for nuclear power plant field operators: A scoping review Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-15 Pakarinen Satu, Laarni Jari, Koskinen Hanna, Passi Tomi, Liinasuo Marja, Salonen Tuisku-Tuuli
The nuclear power plant (NPP) field operator (FOP) is one agent of a network of activities that ensures safe and efficient plant operation and energy production. One of the key aspects in securing production is continuous and high-quality personnel training. Here, we reviewed existing literature on virtual reality (VR) solutions for FOP training, and how the solutions have been used to develop the
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Neutron sensitivity and uncertainty analysis of rhodium self-powered neutron detectors for reactor monitoring in HPR1000 Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-15 Xiong Wu, Jieqiong Jiang, Tingyu Wu, Shijie Luo
The advanced Gen-III pressurized water reactor (PWR) HPR1000 utilizes rhodium self-powered neutron detectors (SPNDs) for reactor monitoring to guarantee safety operation. However, the neutronic characteristics of rhodium SPND result in its sensitivity being dependent on both thermal neutrons and epithermal neutrons and thereby introducing uncertainty. In this study, a computational model based on the
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A system diagnostic and prognostic framework based on deep learning for advanced reactors Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-13 Andy Rivas, Gregory Kyriakos Delipei, Ian Davis, Satyan Bhongale, Jason Hou
To meet the projected energy demand in the next 30 years, advanced reactor designers are looking to maximize system capacity factor to increase economic competitiveness. To maximize capacity factor, operators must minimize the system downtime due to forced shutdowns from transients. To accomplish this, the objective of this work is to develop a System level Diagnostic/Prognostic (SDP) framework based
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Uncertainty quantification study of the physics-informed machine learning models for critical heat flux prediction Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-13 Congshan Mao, Yue Jin
Accurate prediction of critical heat flux (CHF) is important for enhancing the reliability and efficiency of systems used in the energy and power engineering sectors. Traditional physical and mathematical models are constrained by current theoretical maturity. While machine learning (ML) techniques offer alternative ways for CHF modeling, its 'black-box’ nature poses risks of producing outcomes that
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Development and validation of a neutron transport solver with SP3 method based on OpenFOAM Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-13 Xingguang Zhou, Dalin Zhang, Lei Zhou, Wenqiang Wu, Xisi Zhang, Wenxi Tian, Suizheng Qiu, Guanghui Su
A deterministic neutron transport solver with method based on OpenFOAM has been developed and validated in this paper. The physical models, numerical methods and schemes, boundary conditions, and solving strategy are discussed in detail. The approximate Marshak boundary condition of this solver is validated by the ISSA benchmark problem. Pin-by-pin lattice and simplified core neutron transport benchmark
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Transient analysis of the 1970 Windscale nuclear criticality incident Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-12 J.R. Daniels, M.M.R. Williams, M.D. Eaton
This paper describes a novel methodology for the analysis of transient nuclear criticality in layered aqueous-emulsion-organic plutonium nitrate systems. The presented methodology includes point neutron kinetics equations coupled with phenomenological one-dimensional nuclear thermal hydraulics models, which describe the variation in mass, power, reactivity, temperature and voidage within the system
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Surging parametrization for gas-liquid centrifugal pumps Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-12 L.E.M. Carneiro, G.S.O. Martins, C.M.P. Rosero, J.B.R. Loureiro, A.P. Silva Freire
The present work investigates the dimensional and similarity representations of two-phase flows in centrifugal pumps. A new parametrization scheme is advanced for the capacity () and head (/) coefficients in terms of the non-dimensional parameters (no-slip volume gas fraction) and (= , = mixture velocity in the inlet pipe). The new parametrization yields universal performance curves that capture the
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NMR-based comparative study of gas permeability and pore structure of GMZ bentonite Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-10 Hongyang Ni, Jiangfeng Liu, Zhipeng Wang, Qier Sa, Xiangyu Zhang
The gas permeability of buffer/backfill materials is one of the key scientific issues in deep geological disposal. It is closely related to the pore structure. The current study investigates the pore structure and its effect on gas permeability by combining gas permeability tests and NMR tests on GMZ bentonite. It was found that the gas permeability decreases with the confining pressure. The variation
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LBM study on the heat and mass transfer characteristics of the droplet in pressurizer Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-10 Qianglong Wang, Yue Li, Linrong Ye, Mingjun Wang, Wenxi Tian, Suizheng Qiu, G.H. Su
Pressurizer is widely used in the industrial systems, which can maintain the pressure of the system in a certain range. Especially, as the system pressure rises, the spray droplets interacts with the saturated steam in pressurizer to reduce system pressure. In this paper, a heat and mass transfer model of droplet and steam in pressurizer is established through the lattice Boltzmann method (LBM), and
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Hydrogen risk analysis in small containment with a pressure suppression tank Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-08 Zhiqiang Zou, Hang Zhang, Hongping Sun, Yuejian Luo, Jian Deng, Huanhuan Peng, Ming Zhang, Zhifang Qiu, Xiaowei Jiang
The total volume of the small containment of a small modular reactor (SMR) is significantly reduced, and the pressure in the containment may rise rapidly when the non-condensable gas and water steam jet into the small volume containment space during a loss of coolant accident (LOCA). SMR has adopted the pressure suppression tank (PST) as a rapid and efficient measure to mitigate the containment pressure
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Research on power regulation and distribution of the improved NHR200-II low temperature nuclear heating reactor Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-08 Zongyang Li, Wentao Hao, Wenwen Zhang, Weihua Li, Xingtuan Yang
Tsinghua University's low-temperature nuclear heating reactor (NHR) primarily focused on heating supply in its initial design. However, utilizing the reactor solely as a heat source limits its economic efficiency and competitiveness. To further broaden the application scope of the NHR200-II reactor, several modifications were implemented based on the original design. These include replacing the vertical
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Bubble characteristics on FeCrAl surface in subcooled boiling flow Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-07 Bing Tan, Jiejin Cai, Songbai Cheng
Iron–Chromium–Aluminum(FeCrAl), recognized as one of the most promising materials for accident-tolerant fuel (ATF), bears substantial commercial potential owing to its advantageous attributes. To gain a deeper insight into the subcooled boiling properties of FeCrAl, a series of boiling experiments were conducted on FeCrAl material surface for the evaluation of bubble characterization, bubble population
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Comprehensive analysis of the effects of Mo and Co on the synthesis, structural, and radiation-shielding properties of TiO2 based composites Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-05 K.A. Mahmoud, Mazen Binmujlli, Mohammad Marashdeh, M.I. Sayyed, Mamduh J. Aljaafreh, Hanan Akhdar, Islam G. Alhindawy
In the present work, three materials-based nanoscale TiO compound materials were directly fabricated via hydrothermal synthesis methods to be used in gamma-ray shielding applications as ceramics and paints. X-ray diffraction and energy dispersive X-ray techniques were utilized to characterize the formation of TiO, Mo–TiO, and Co–TiO nanoparticles. Additionally, transmission electron microscopy was
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Solidification performance and mechanism of typical radioactive nuclear waste by geopolymers and geopolymer ceramics: A review Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-05 Jiarui Liu, Yidong Xu, Wensheng Zhang, Jiayuan Ye, Rui Wang
The radioactive waste generated from nuclear fuel cycles poses significant risks to the natural environment and the health of surrounding residents. Thus, the utilization of appropriate solidification materials is essential for their safe disposal. Geopolymers, as a novel type of cementitious material, exhibit effective immobilization properties for radioactive nuclear waste. Moreover, the amorphous
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Experimental study on air-water countercurrent flow limitation in horizontal pipes with different types of obstructions Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-05 Xixi Zhu, Chende Xu, Mingzhou Gu, Shuai Tang, Naihua Wang
In a nuclear power plant, gas-liquid countercurrent flow limitation (CCFL) may occur in the liquid level measurement system where an orifice obstruction is installed in the horizontal pipeline, which will introduce errors into the system and affect the reliable operation of the system. The investigations on the CCFL phenomenon in the flow channel with obstruction are important for nuclear reactor safety
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Analysis of the radioactivity in bio-shield of Kori NPP unit 1 via computational simulation Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-05 Hyeonmin Lee, Jaeho Lee, Woo Nyun Choi, Seungbin Yoon, Hee Reyoung Kim
This study derived neutron flux and radioactivity distribution of the bio-shield to furnish fundamental datasets for evaluating its radiation degradation performance within decommissioning-bound nuclear power plants in Korea. Concrete within nuclear power plants degrades due to continuous radiation exposure. Particularly, neutrons exert a preeminent influence on radiation degradation, inciting changes
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Adsorptive removal of uranium from aqueous solution using graphene oxide reinforced and Mg-doped hydroxyapatite composite Prog. Nucl. Energy (IF 2.7) Pub Date : 2024-02-04 Wenjun Wu, Jianlong Wang