bluuebunny/crossref_metadata_2025_split huggingface dataset:
Implementation of Solar Salt as Fluid in ASYST4.1 and Validation for a Natural Circulation Loop
Authors: Trivedi A. K. , Novog D. R. , Allison C. 
Published: October 2021

Abstract
Various small modular reactor (SMR) designs such as Micro Modular Reactor (MMR), Stable Salt Reactor - Wasteburner (SSR-W) and the Natrium reactor utilize an intermediate molten salt loop for heat transport and thermal energy storage (TES). This technology is well proven for solar thermal plants and uses solar salt (NaNO3-KNO3 in 60:40 ratio by weight) as a heat transfer and heat storage material. The primary advantages are its thermal stability, large temperature range, low cost, high density, and low vapor pressure. For example, MMR is a prismatic block design which has emerged as one preferred concept of gas cooled reactors (GCRs) and uses helium as the primary coolant and the solar salt as intermediate fluid for transporting the heat to TES system. The current thermal hydraulic system codes can model helium as a dry non-condensable gas but in most cases solar salt properties are not available in these codes. Hence there is a lack of integrated analysis capability across the MMR-type system. An effort is made to collect a robust set of thermodynamic and transport properties and heat transfer correlations for solar salt and incorporate them into the code ASYST4.1. It will allow for the simulation of design basis accident which include the dynamics of the primary system, intermediate loop, and TES.
Reliable computational tools are required for thermal hydraulics design and accident analysis of Small Modular Reactors (SMR) including associated experimental validation. One emerging tool is ASYST (Adaptive SYStem Thermal-hydraulics) - ISA (Integral Simulation and Analysis). It is a new code which combines the capabilities of SCDAPSIM/MOD4 and SAMPSON and is being developed jointly by Innovative Systems Software (ISS), USA and Institute of Applied Energy (IAE) of Japan. The thermal hydraulic module, ASYST-THA, replaces the original US NRC-developed RELAP5 code used in RELAP/SCDAPSIM/MOD3.x and the THA code used in SAMPSON, with new system level hydrodynamic options that include multidimensional, multi-fluid models originally developed by ISS and IAE. ASYST represents an internationallevel collaboration on system code development and benefits from the expertise and experiments available world-wide.
This work presents the details of solar salt implementation into ASYST as well as a benchmarking study on Molten Salt Nitrate Natural Circulation Loop (MSNCL). ASYST calculated results are found to closely match the experimental data. This work has been further extended to quantify uncertainties in the calculated heater outlet/inlet temperature and the flow rate for 1600 kW case using Integrated Uncertainty Analysis (IUA) package of ASYST4.1. The margin in the temperatures is about 35 K which confirms that the ASYST can both overpredict or underpredict the experimental temperatures.
Advances in the Development of a Fluid-to-Fluid Similarity Theory for Fluids at Supercritical Pressure: Results From Sensitivity Analyses
Authors: De Angelis A. , Pucciarelli A. , Kassem S. , Ambrosini W. 
Published: October 2021

Abstract
In the present paper, the capabilities of a similarity theory for heat transfer to supercritical pressure fluids are further investigated with respect to past work, widening the range of the addressed operating conditions. In particular, experimental data for water at 25 MPa are here considered; several cases addressed adopting different simulant fluids (i.e. CO2, NH3 and R23), in order to understand possible dependencies of the proposed rationale on the imposed mass flux and the observed heat transfer conditions. The obtained results confirmed the lessons learned from the recent analyses, though highlighting some possible needs for improvements in both the adopted CFD model and the similarity rationale in some specific operating regions.
The aim of the present paper is to show further results in relation to the capabilities of the proposed theory, which seems sufficiently mature to be now considered for validation against experimental data.
Numerical Simulation of Bubble Shape and Departure in Nucleate Pool Boiling at High Superheat
Authors: Bhati Jyoti , Paruya Swapan , Akhtar Farheen 
Published: October 2021

Abstract
In this work, the authors reported the dynamics of bubble shape and departure in nucleate pool boiling at the heated surface at a high superheats of 34 K. The numerical results of bubble shape and departure have also been compared with the CFD-results in literature and with our experimental results. The shape of the bubble is influenced by superheats and Bond number Bo. We have improved our numerical model by including the effect of radial variation of film pressure on microlayer evaporation to accurately compute the bubble shape, bubble diameter and growth time. We see that the approximate analytical solution works much better than the direct numerical integration of Young-Laplace equation for non-spherical bubble at high superheat. The Young-Laplace equation seems to include the growth force for capturing the shape of non-spherical bubbles. Interestingly, we find a train of small bubbles at a low Bo, which coalesce to form a big bubble on increasing Bo. While analysing the forces acting on the bubble surface, we also see that the growth force imparting the liquid inertia becomes important for the hemispherical bubbles.
Preliminary Verification Calculation and Sensitivity Analysis on PISAA Code Compared to MELCOR
Authors: Song Mingqiang , Zhang Ningna , Yang Xiaoming , Ma Rubing , Yang Zhiyi , Ding Chao 
Published: October 2021

Abstract
Since the Three Mile Island accident happened in the United States in 1979, the nuclear industry has accelerated the development of calculation codes for simulating severe accident. Especially after the Fukushima nuclear accident, the nuclear industry and nuclear safety regulatory authorities have paid more attention to the study of severe accidents in nuclear power plants. Based on advanced pressurized water reactors in Generation II, typical severe accident scenario, like small break loss of coolant accident (SBLOCA) is selected, and verification calculations on the Program Integrated for Severe Accident Analysis (PISAA), the severe accident integrated code developed by China with MELCOR, is conducted in this paper. The results of flow rates in break, pressure, water level and hydrogen production are obtained, showing that PISAA has the same trend compared with MELCOR reasonably, and PISAA has a completed calculation function which can simulate severe accident phenomena. Uncertainty analysis is also performed, 7 representative parameters related to hydrogen production are selected, and Latin hypercube calculation is performed. Hydrogen production distribution is around 300kg∼520kg, indicating that there is no cliff effect in PISAA. In the future, the researchers plan to do some works to improve the accuracy and functions of PISAA.
Numerical Simulation of Thermo-Hydraulic Characteristics of 7-Pin SFR Test Fuel Bundle With Variable-Pitch Helical Wire
Authors: Li Siyuan , Zhang Aimin , Ji Songtao , Li Yanlin 
Published: October 2021

Abstract
This paper is focused on the applicability of variable-pitch helical wire in a 7-pin SFR test fuel bundle by using CFD method. The differences between the variable-pitch helical wire scheme and the even-pitch helical wire scheme in the pressure loss, velocity, temperature field of the test fuel bundle are compared and analyzed. The simulation results show that it is reliable to carry out the thermo-hydraulic analysis of SFR test fuel bundle by using CFD numerical simulation method. The helical wire mixing ability to the coolant and the pressure drop is influenced by the helical wire pitch. Compared with the even-pitch helical wire scheme on fuel rod surface, the variable-pitch helical wire doesn’t drastically change the flow field structure while the pitch changes, but gradually increases its mixing ability to the coolant along the axial direction, which is more conducive to accelerating balance the thermal inhomogeneity and inhomogeneity ratio of the coolant, solve the problem of temperature rise caused by the coolant congestion on the contact structure between the helical wire and fuel rod surface. The stronger the mixing ability of the helical wire to the coolant, the more effective it is to lower the maximum temperature of the coolant. The application of the variable-pitch helical wire can well match the demand of the coolant for thermal and hydraulic needs, and the safety and economy of the assembly can be correspondingly improved.
Flow Induced Vibration Analysis and Remediation Using a Cartesian Grid Flow Solver
Authors: Boschitsch Alexander , Danilov Pavel , Kaufman Andrew , Bilanin Alan 
Published: October 2021

Abstract
Flow through pipe junctions produces unsteady vortex shedding processes that can generate acoustic pressures whose modal amplification contributes to structural fatigue. Such flow-induced vibration (FIV) is an enduring concern in nuclear power plants (NPPs) and has led to delays in extended power uprate licensing, produced strong fluctuating loads in piping, and contributed to cracks in low pressure turbine blades. Modern computational fluid dynamics (CFD) analysis can analyze FIV phenomena, but the practical challenges of setting up and adequately meshing multiple candidate designs continue to deter routine and reliable CFD-based FIV assessment. Recently, CFD solvers implemented on Cartesian grids and offering autonomous mesh generation about complex geometries have made significant strides in addressing these challenges. Here, the successful application of a CFD solver (CGE) used in combination with subscale testing to rectify FIV in NPP piping, is described. The article reviews the role of CFD in FIV analysis and design, summarizes the principal analysis components of CGE, outlines the 1/8th subscale testing program used in conjunction with CFD to assess FIV, and concludes with postinstallation assessments of the modification. The concepts contained here have helped address FIV in several NPPs resulting in significant reductions in operation & maintenance (O&M) costs.
Production of Cyclotron-Based Gallium-68 With Low Energy Protons: Preliminary Target Design and Cyclotron Shielding Considerations
Authors: Salas-Tapia Luis Fernando , Zhang Tian 
Published: October 2021

Abstract
Nuclear reactors, cyclotrons and generators are being used to produce nuclear medicine radionuclides. Radiopharmaceuticals based on Gallium-68 are widely used both experimentally and clinically in Positron Emission Tomography (PET). Currently Ga-68 is mainly produced via 68Ge/68Ga generators, which are expensive. An alternative method is by cyclotron using high enriched Zinc-68 via the 68Zn(p,n)68Ga reaction. Investigations have been conducted for producing Ga-68 using a solid target impacted with a proton beam in energies between (13–14) MeV or higher. The main interest of this work is to produce Ga-68 using energies within a lower range, therefore Ga-68 can be accessed to locations with cyclotrons of lower energy. In this article, an approach to a novel target design to get Ga-68 with low energy protons is conducted. Adequate shielding to avoid extra doses outer the cyclotron is also considered. Monte Carlo code FLUKA is used to simulate the processes of interaction of radiation with matter and the radioactive isotope production. For a beam energy of 11 MeV and current on target of 40 μA the simulated produced activity is about 69.1 GBq. The results show the feasibility of producing Ga-68 with low energy protons and self-shielding system guarantees the regulation-based dose exposition outer the cyclotron.
Experimental Study of Characteristics of Flow Field in Rod Bundle Channel Under Blocking Conditions
Authors: Yu Xiaoyong , Zhang Yonghao , Qi Peiyao , Liu Yusheng , Qiao Shouxu , Tan Sichao 
Published: October 2021

Abstract
The rod bundle fuel is characterized by compact structure and narrow flow passage. The fragments and corrosion products, flowing with the coolant, can cause local blockage accident, threaten the integrity of the fuel cladding. Therefore, it is necessary to use the Particle Image Velocimetry (PIV) to visualize and measure the flow fields downstream of the blockages. The results show that partial blockages will cause flow reversal. In the backflow zone, vortices are generated downstream of the blockage, causing increase in the resistance. The length of backflow zone increases with the increase of the Reynolds number. The wake area formed downstream of the blockage presented periodic changes with the time and the period is about 0.8s. For the blockage of the interior subchannels, in the backflow zone, two rows of asymmetrically distributed vortices, and the vortices interfere with each other and cause squeeze deformation. For the blockage of the side and corner sub-channels, the formed vortices have irregular shape and nonuniformed distribution, and the flow field is more complex and changeable. This is believed to be caused by the high intensity turbulence and the influence of the wall.
Experimental and Numerical Investigation on Debris Bed Quenching With Additional Injection of Non-Condensable Gas
Authors: Petroff Markus , Kulenovic Rudi , Starflinger Jörg 
Published: October 2021

Abstract
Severe accidents of light water reactors with core degradation can lead to the formation of a so-called debris bed inside the reactor cavity. In the scenario of depleted residual water, the bed can partially melt and interact with the concrete underneath generating non-condensable gases (NCG) at the bottom of the particle bed, which will flow through the debris bed. The impact of additional gas on the quenching process can be considered in thermal-hydraulic system codes like ATHLET. However, there is still a need for experimental validation of respective models or verification of corresponding simulation results. In the frame of this work, specific quenching experiments on a monodispersed particle bed with various NCG injection rates and initial bed temperatures are carried out employing the FLOAT test facility to extend the existing experimental database. Furthermore, numerical simulations are carried out with COCOMO-3D for the prediction of the thermo-hydraulic conditions inside the debris bed, the quench front propagation as well as the total quenching times.
Design of a Novel Test Section for the Lead Fast Reactors Development: The CIRCE-THETIS Facility
Authors: Lorusso Pierdomenico , Di Piazza Ivan , Martelli Daniele , Musolesi Andrea , Tarantino Mariano 
Published: October 2021

Abstract
Within the R&D program of Generation IV lead fast reactors and Accelerator-Driven Systems, a new project named PATRICIA (EC – H2020) has been launched by the European Commission to support innovative solutions for the development of MYRRHA ADS concept and lead fast reactor technologies in general. In this framework, the ENEA Brasimone Research Centre (Italy) supports such research activities through experimental campaigns, involving CIRCE, a large scale pool-type facility using lead-bismuth eutectic as the primary coolant and pressurized water as the secondary fluid.
A new test section named THETIS is currently under design and it will be installed in the main vessel of the CIRCE facility. The test section will include a vertical mechanical pump for the primary coolant circulation and a new prototypical helical coil steam generator (HCSG). This type of steam generator turns out to be very interesting for nuclear power plants, since the helical geometry is very compact and it assures a high power removed, taking up the minimum amount of space.
The experimental tests will aim at: (i) investigating the thermal-hydraulic behavior of the system in steady state operation (forced circulation regime), during operational and accidental transients (postulated scenarios) and in natural circulation regime, considering as heat sink the HCSG (acting as Decay Heat Removal system, DHR) and the reactor vessel auxiliary cooling system, in stand-alone or coupled operation, (ii) characterizing the performances of the helical coil steam generator.
The present paper aims at presenting the preliminary layout of the THETIS test section, describing in details the main components, the instrumentation installed and the main operating conditions of the facility.
Study on Water Cooler Performance for High Temperature Helium Experimental System
Authors: Hu Qingxiang , Peng Wei , Zhao Gang , Wang Jie 
Published: October 2021

Abstract
Very-high-temperature gas-cooled reactor (VHTR) has good inherent safety and is one of the six reactor technologies of Generation IV nuclear energy systems. The ultra-high core outlet temperature makes VHTR suitable for process heat applications such as hydrogen production. The intermediate heat exchanger (IHX) is the key equipment for energy transfer between VHTR and the process heat application system. To verify the high temperature performance of IHX, it is generally necessary to set up a high temperature helium test system to conduct a series of performance tests. In this experimental system, a well-behaved water cooler needs to be designed to cool down the high-temperature helium in the experimental loop to ensure the safety of the experimental system. In the present study, the effect of the cooling water flow rate G, the cooling water injection position L, and the lower heat exchange tubes material were analyzed. The results show that increasing the flow rate of the cooling water, the maximum temperature of the tube sheet Tmax decreases and the temperature gradient in the flow direction ∇Ty increases. Tmax increases first and then decreases with the increasing L. There exists an optimal cooling water injection position between L = 175mm and 225mm. The variation of ∇Ty is the opposite. The use of ceramic instead of steel reduces Tmax by about 295K and reduces ∇Ty by half. Subsequent research and optimization on structure parameters of the water cooler are expected to further improve the cooling performance.
Study on Flow Characteristics of Double Loop Natural Circulation System Under Asymmetric Conditions
Authors: Wang Shuang , Li Xin , Liu Yongchao , Qiao Shouxu , Tan Sichao 
Published: October 2021

Abstract
The load distribution and coupling feedback between parallel loops is one of the issues in the multi-loop nuclear reactor system. In the present study, a dual-loop reactor natural circulation experimental system with typical structural features was designed and built based on the characteristics of a floating nuclear power plant. The system is used to study the system response characteristics under the conditions of heat load disturbance, resistance disturbance and system response time. The results show that the flow of the primary loop will be redistributed when the load disturbance was introduced into the natural circulation loop. The primary flow rate changes quickly when the heat load disturbance reaches the critical point. When the heat load disturbance is lost on one side, the total natural circulation flow rate of the loop decreases to 60% of the value in the steady-state flow. The response of the system after introduction of resistance disturbance is similar to the introduction of the load disturbance. The response of the system can be divided into two stages, i.e., the rapid response period and the slow adjustment period after the introduction of the disturbance. With the increase of disturbance, the negative slope of stage 1 increases, and the slope of stage 2 gradually changes from positive to negative. With the experimental, a model is developed to predict the response of the system. A characteristic number S is defined to characterize the symmetry of the system. Based on the experimental data, it is found that when S is between 0.5–2, the external disturbance will not cause asymmetric response of the system.
Study on Collapsed Cross Section for Radial Reflector in LWR
Authors: Shibano Ryosuke , Kawano Tatsuya , Takeda Satoshi , Kitada Takanori , Masaoka Yoshitada , Nagano Hiroaki , Hyoudo Hideaki , Kodama Yasuhiro 
Published: October 2021

Abstract
Few-group cross sections of fuel and reflector region are generally used for the core design of LWR. The power distribution and reactivity depend on the few-group cross sections in the core design. The few-group cross sections depend on the fuel adjacent to the reflector region. Therefore, it is necessary to calculate the few-group cross sections. However, the effect of changing fuels adjacent to the radial reflector has not been clarified. Therefore, the purpose of this study is the evaluation of the effect of changing fuels adjacent to the radial reflector on the power distribution and the reactivity in order to clarify the effect of changing fuels adjacent to the radial reflector on the radial reflector cross sections. First, the effect of changing fuels adjacent to the radial reflector on the power distribution is evaluated and is compared to the effect of changing temperature or boron concentration which are generally taken into account in the conventional model of the radial reflector cross sections. In this evaluation, the effect of changing fuels adjacent to the reflector on the radial reflector constant is evaluated by using UO2 and MOX fuels. Furthermore, the effect on the radial reflector constant is evaluated based on the perturbation theory. The results show the effect on power distribution is about 0.3 % at maximum, and the effect on the reactivity is quite low (&lt; 1 × 10−7Δk/k).
Research on Application of Additive Manufacturing Technology In Nuclear Fuel Assembly Field
Authors: Li Hua , Yue Ti , Zhu Fawen , Peng Yuan , Li Yun , Huang Chunlan , Zhang Youjia 
Published: October 2021

Abstract
As one of the key technologies of the fourth industrial revolution, Additive Manufacturing (AM, also called 3D printing), which is the basis of innovation design with its high degree of manufacturing freedom, has been widely used in the construction field, aerospace field, automobile manufacturing field, biomedical field, etc. With the increasing maturity of additive manufacturing technology, it has been gradually used in the field of nuclear energy in order to realize design reform. Public reports have shows that additive manufacturing for the pressure vessel fuel, some kinds of fuel pellets, parts of fuel assembly and related core components has been realized. In this paper, the advantages and challenges of additive manufacturing technology used in nuclear fuel assembly design and manufacturing are analyzed, the application status of additive manufacturing technology in the field of nuclear fuel assembly design and manufacturing is summarized, and the application development trend of the additive manufacturing technology for the fuel assembly is also clarified. In general, although the application of the additive manufacturing technology in the field of nuclear fuel assembly design and engineering is still at the stage of research, the application scope of this innovative technology will be greatly expanded with the continuous progress and improvement of the additive manufacturing technology in the near future.
Applicability Assessment of Accident Analysis Codes and Determination of Testing Facility for Validation of the CAP1400
Authors: Cai Xiaoyu , Shi Guobao , Yan Jinquan , Fan Pu , Zhao Dongjian 
Published: October 2021

Abstract
The approach in uprating the AP1000 to the CAP1400 was to increase the power capability of the plant while retaing the creadibility of proven components and substantial safety margins. Therefore, the CAP1400 retains the AP1000 licensing basis. A comparison of the CAP1400 major design with the AP1000 is provided in Tab. 1.