bluuebunny/crossref_metadata_2025_split huggingface dataset:
Numerical Study of Tritium Mitigation Strategies for Fluoride Salt-Cooled High-Temperature Reactors
Authors: Zhang Sheng , Wu Xiao , Sun Xiaodong 
Published: October 2020

Abstract
Fluoride salt-cooled High-temperature Reactor (FHR) is one of the advanced non-Light Water Reactor (non-LWR) designs, which adopts a low-pressure fluoride salt as the primary coolant, high working temperatures, coated-particle fuel, and a passive safety system for decay heat removal. However, tritium management is perceived as a critical issue for FHRs since tritium is a radiation hazard when inhaled or ingested and its production rate in FHRs is expected to be significantly higher compared to that in LWRs. To reduce FHR tritium release rates into the ambient, two tritium mitigation options, such as using Double-Wall Fluted-Tube Heat eXchangers (DWFT-HXs) with a tritium carrier or Single-Wall Fluted-Tube HXs (SWFT-HXs) with a tritium barrier, are therefore proposed for key HXs in FHRs, which potentially provide major pathways for tritium release due to their elevated temperatures and large surface areas. Tritium carriers investigated include gases, such as helium, and liquids, such as FLiBe, FLiNaK, and KF-ZrF4, while the tritium barrier investigated in this paper is silicon carbide (SiC) due to its low permeability for tritium. These proposed HX designs are then optimized, using a Non-dominated Sorting in Generic Algorithms (NSGA) optimization approach, for the Advanced High-Temperature Reactor (AHTR), one of the FHR designs with a large power output.
A system-level mass transfer model is developed to evaluate the tritium transport in the two proposed design options for tritium mitigation in FHRs and quantitively analyze the tritium release/leakage rate from the reactor primary system. Our study shows that both the DWFT-HX design with helium as the tritium carrier and SWFT-HX design with SiC coating as the tritium barrier are able to reduce the total tritium leakage rate in FHRs to the same order of magnitude of the typical average tritium leakage rate in LWRs (1.9 Ci/day).
Technology of Phased Array Ultrasonic Testing on Jet Pump Beam Installed in Reactor Pressure Vessel
Authors: Sasaki Shunsuke , Yamamoto Setsu , Takada Toshiaki , Matsunaga Keiji , Uraguchi Kohei , Kameyama Ikuko 
Published: October 2020

Abstract
Jet pumps installed inside Reactor Pressure Vessel (RPV) of Boiling Water Reactor (BWR) have a function to circulate cooling water through the core region in RPV. Jet pump beam is one of the parts in the jet pump assembly. Jet pump beam is made of high nickel base alloy and could have potential of Intergranular Stress Corrosion Cracking (IGSCC) in case of high stress and poor material properties. Especially, it was reported that IGSCC occurred in beam bolt hole region (BB-1), transition region (BB-2) and tapered region (BB-3) of BWRs in US and other countries. Due to this reason, it is important to check the soundness of jet pump beam with BB-1, BB-2 and BB-3. In US, jet pump beam is inspected by Ultrasonic Testing (UT). However, there is some difficulty to perform UT because jet pump beams have some complicated shapes and are installed where it is not easily accessible. Since it causes man-hours increase to take out jet pump beams from RPV for inspection, inspection to be implemented in-vessel is preferable.
In this study, we developed a Phased Array UT (PAUT) technology for flaw detection of jet pump beam without taking it out from RPV. In order to easily detect flaws, immersion non-contact probe was used. Furthermore, appropriate PAUT technique were applied in accordance with the shape of each region of BB-1, BB-2 and BB-3. Two types of UT equipment for jet pump beams were developed: ultrasonic incidence from lateral surface and ultrasonic incidence from upper surface. The UT equipment was installed on jet pump beam bolt grasping jet pump beam trunnion. The UT equipment can be positioned on the basis of the locations of jet pump beam bolts.
We could inspect center cracks that were 5.0 mm length, 1.0 mm depth at BB-2/BB-3, and could inspect corner crack that was 2.5 mm length, 1.0 mm depth at BB-2 by developed PAUT equipment.
Numerical Simulation of Concrete Ablation and Corium Cooling for Molten Corium-Concrete Interaction (MCCI)
Authors: Xu Zhichun , Zhang Yapei , Su G. H. , Tian Wenxi , Qiu Suizheng 
Published: October 2020

Abstract
In a postulated severe accident situation in Light Water Reactors (LWRs), if the core fuel cannot be effectively cooled, the reactor core material will be heated and form a molten corium in the lower head. When the lower plenum of the reactor vessel fails, the molten corium may flow into the cavity under the reactor vessel and react with the concrete. This process, known as Molten Corium Concrete Interaction (MCCI), is characterized by concrete ablation and oxidation of metal in the corium, both of which produce a large amount of combustible and non-condensable gases, threatening the integrity of the containment. Thus in-depth study of the characteristics of concrete ablation and corium cooling have great significance.
In the present study, an MCCI analysis code, MOQUICO (molten corium concrete interaction and corium cooling code, QUI means quintic) has been developed. The MACE M3b and OECD/MCCI CCI-3 tests were analyzed to validate the developed code. The melt temperature, axial and radial ablation depths, upward heat flux were calculated and were in good agreement with the experimental measurements, which proved that the code is capable of simulating MCCI and related phenomena of LWRs. Sensitivity analyses on the factors of decay heat, concrete type and water injection moment were performed and analyzed.
Numerical Simulation of Microparticles Motion in Two-Phase Bubbly Flow
Authors: Yoshida Hiroyuki , Uesawa Shinichiro 
Published: October 2020

Abstract
The radioactive aerosol removal equipment is used as one of the safety systems of nuclear reactors. In this equipment, microparticles of aerosol are removed through gas-liquid interfaces of two-phase flow. The mechanism related to the removal of microparticles through the gas-liquid interface is not precise; a numerical evaluation method of performance of aerosol removal equipment is not realized. Then, we have started to construct a numerical simulation method to simulate the removal of microparticles through gas-liquid interfaces. In this simulation method, a detailed two-phase flow simulation code TPFIT is used as the basis of this method. TPFIT adopts an advanced interface tracking method and can simulate interface movement and deformation directly. Also, to simulate the movement of particles, the Lagrangian particle tracking method is incorporated. By combining the interface tracking method, and the Lagrangian particle tracking method, the interaction between interfaces and microparticles can be simulated in detail. To solve the Lagrangian equations of particles, fluid properties and fluid velocity surrounding aerosol particles are evaluated by considering the relative position of particles and gas-liquid interface, to simulate particle movement near the interface. In this paper, we show an outline and preliminary results of this simulation method.
Development of New Organic Iodine Filter for FCVS
Authors: Tominaga Kazuo , Fukui Sohei , Tanaka Motoi , Hashimoto Tomoharu , Tayama Ryuichi 
Published: October 2020

Abstract
All of the boiling water reactors in Japan were required to install filtered containment venting system (FCVS) for restart after the accident of Fukushima Daiichi Nuclear Power Plant. FCVS is composed of alkaline water and a metal fiber filter (MFF). Alkaline water has the function of reducing aerosol and inorganic iodine (I2). MFF is constituted for the purpose of removing the aerosol which was not able to be removed with alkaline water. With the above system, the released aerosol can be removed with DF1000 and inorganic iodine DF100. On the other hand, organic iodine that cannot be removed by alkaline water and MFF is removed by silver zeolite added downstream of the FCVS because of its high scattering property. Silver zeolite has particle of silver in the pores of the zeolite. Organic iodine is removed by chemical bonding with the silver. In the current system, the silver zeolite is DF50. Hitachi has developed the organic iodine removal system focusing on removing liquid chemicals for the purpose of improving the performance andmaintain ability, and reducing the price. In this presentation, we report the ionic liquid (IL) that has high heat resistance, high radiation resistance and high gas adsorption among liquid chemicals. In the small-scale tests, we found that IL has a higher performance than DF250, and improved maintainability by liquid property that is able to discharge with scrubber water at the same time, and have got the prospect of reducing manufacturing costs.
Analysis of Gas Entrainment Phenomenon From Free Liquid Surface for a Sodium-Cooled Fast Reactor Design: Validation of Velocity Profile and Strouhal Number in a Flow Field
Authors: Uchida Mao , Hirakawa Moe , Sano Aaru , Inoue Keisuke , Sakai Takaaki , Ezure Toshiki , Tanaka Masaaki 
Published: October 2020

Abstract
Gas entrainment (GE) from cover gas, which is an inert gas to cover sodium coolant in a reactor vessel, is one of key issue for Sodium-cooled fast reactors (SFRs) design to prevent unexpected effects to core reactivity. In this research series, evaluation method has been investigated for surface dimple depth growth of unstable drifting vortex dimples on the liquid surface in the reactor vessel. By using a computational fluid dynamics (CFD) code, analyses have been conducted to estimate the drifting vortex on water experiments in a circulating water tunnel. The unstable drifting flow vortexes on the water surface were generated as wake vortexes behind a plate obstacle. Downward flow velocity was induced by bottom slit flow pass along the flow channel. In the previous study, the onset conditions of the gas entrainment were evaluated based on existing non-dimensional numbers method by using the STREAM-VIEWER code. However, the CFD predication accuracy of the detail flow field itself was not clear especially for vortex frequency in the wake flow and detail velocity profiles in the flow channel. In this study, to clarify the accuracy of CFD analysis, Strouhal numbers of vortex frequency and detail flow velocity profiles were compared with experimental data which were measured by Particle Image Velocimetry (PIV) method. As the results, the Strouhal numbers of the vortex frequency behind the plate obstacle reasonably agreed with experimental data. Prediction accuracy for the velocity profiles in the flow channel were also confirmed by comparisons with measured data by the PIV method.
Time-Dependent Solution of Unsteady Fluid Flow Equations for High Speed Oscillating Compressible Flows and Blast Wave Propagations
Authors: Sinha Ramlala P. 
Published: October 2020

Abstract
A solution of the highly complex unsteady high speed oscillating compressible flow field inside a cylindrical tube has been obtained numerically, assuming one dimensional, viscous, and heat conducting flow, by solving the appropriate fluid dynamic and energy equations. The tube is approximated by a right circular cylinder closed at one end with a piston oscillating at very high resonant frequency at the other end. An iterative implicit finite difference scheme is employed to obtain the solution. The scheme permits arbitrary boundary conditions at the piston and the end wall and allows assumptions for transport properties. The solution would also be valid for tapered tubes if the variations in the cross-sectional area are small. In successfully predicting the time dependent results, an innovative simple but stable solution of unsteady fluid dynamic and energy equations is provided here for wide ranging research, design, development, analysis, and industrial applications in solving a variety of complex fluid flow heat transfer problems. The method is directly applicable to pulsed or pulsating flow and wave motion thermal energy transport, fluid-structure interaction heat transfer enhancement, and fluidic pyrotechnic initiation devices. It can further be easily extended to cover muzzle blasts and nuclear explosion blast wave propagations in one dimensional and/or radial spherical coordinates with or without including energy generation / addition terms.
Investigating Structural Response of Pressure Reducing Valve of Supercritical Steam Generator System Under Cyclic Moments, Thermal Transient, and Pressure Loadings
Authors: Cho Nak-Kyun , Bansal Pavitra , Hurst Antony M. 
Published: October 2020

Abstract
Pressure reducing valve (PRV) located before the start-up vessel (SUV) is an essential component that decreases the pressure and temperature of the supercritical state steam by using spray water before it flows into the SUV. The PRV is kept closed during normal operation but opened during start-up and shutdown events, which could initiate thermal fatigue defects due to significant temperature changes. In addition to the thermal shock and internal pressure, system bending and torsional moments may be imparted on the PRV, threatening its integrity. To reinforce these concerns, cracks on the inside surfaces of the PRV have often been reported during planned maintenance activities in nuclear power plants. This research aims at analysing cyclic plasticity of the PRV subjected to cyclic moments, thermal and pressure loadings by means of an advanced direct numerical technique known as the Linear Matching Method (LMM). The cyclic moments are comprised of in-plane and out-of-plane bending, and torsion which are applied to an inlet branch pipe of the PRV. The cyclic thermal load is obtained from the transient heat transfer analysis using real operational data. Two different pressures, which are high and low pressures, are applied to internal surfaces of the PRV body and outlet pipe respectively. The analysed results construct a structural response boundary such as a shakedown limit boundary. The obtained structural response boundary is validated by full cyclic incremental analysis referred to as the step-by-step analysis. The analysed results have demonstrated that the plastic collapse limit is identical to the shakedown limit. Moreover, the results provide engineers with a safe load bearing capacity domain which otherwise requires evaluating structural integrity of the PRV subjected to the complicated cyclic loading condition using detailed assessments and analyses.
Experimental Investigation on Gas Mixing and Stratification in Containment Influenced by External Cooling
Authors: Li Ying , Xue Wei , Cao Xuewu , Tong Lili 
Published: October 2020

Abstract
The distribution of hydrogen inside the containment is a key issue in assessing the evolution of the postulated accident. For safety analysis and codes validation purposes, a large scale comprehensive test facility has been built to investigate the containment thermal-hydraulic characteristics under accident conditions. In this paper, a test was performed to experimentally investigate the distribution of the hydrogen inside the containment and the influence of the external cooling on gas mixing and stratification. The paper presents the experimental results of the integral test performed in this facility. During the experiments, helium was used to simulate hydrogen. Helium and steam are released together and allowed to take additional time to form a relatively stable stratification, then followed by external cooling. The initial pressure of the experiments is around 0.1MPa(a) and the initial Froude number is around 333. The results showed that a helium-enriched stratification emerged in the upper containment due to the density difference after the injection. External cooling caused condensation and intense convective flow. As a result, an overall increase in helium concentration was observed with a decrease in concentration gradient.
Feasibility Study for Multi-Branches and Elbow in Stub Pipe System due to Flow-Induced Vibration Using Computational Fluid Dynamics
Authors: Nguyen Khac-Ha , Kim Won-Tae , Hong Seung-Pyo , Lee Haein , Lee Ahram 
Published: October 2020

Abstract
Acoustic-induced vibration in piping system and other devices leads to premature wear and failure. Especially, in nuclear power plants, very high velocity and temperature gas flows inside pipe systems. Moreover, if a frequency due to the vibration in the piping system is overlapped with a natural frequency of the stud pipe, the magnitude of the amplitude will be increased resulting in severe failure. For example, damage can be considered as flow-induced acoustic resonance at the branch pipes of the safety relief valve in the main steam lines. Specially, the pipe system not only has multi-branches but also includes the elbow that the resonance could occurs making pressure oscillation stronger than that of a single branch because of the interaction between the branches and the elbow. This study has investigated a Computational Fluid Dynamics (CFD) analysis methodology to predict and quantify the vortex shedding frequencies and the pressure pulsation magnitude in the dead-end pipe. The influence of the pressure fluctuation amplitude between each branch, number of branch, and elbow is also investigated.
Time Series Clustering and Classification for Uncertainty Analysis by MAAP5 Code
Authors: Kinoshita Ikuo 
Published: October 2020

Abstract
The MAAP5.04 code uncertainty analysis was carried out for the Power Burst Facility Severe Fuel Damage Test 1-4. Comparisons between experimental data and analysis results were focused on hydrogen generation. The uncertainty propagation analysis was conducted through random variations of input uncertainty parameters of phenomenological models whose ranges were determined by the MAAP5 Zion parameter file. The time series clustering technique using the mean-shift algorithm was applied to the data set generated by the uncertainty propagation analysis. It was confirmed that the code predicted well the hydrogen mass generated and the uncertainty bounds of the analysis included the measured hydrogen generation history. The time series clustering technique demonstrated that the key model parameters could be identified for classifying the uncertainty analysis results using a decision tree classifier. Furthermore, the regression models were constructed which predict the uncertainty of the hydrogen generation from the model uncertainty parameters by using the support vector regressions. The hold-out method of cross validation was applied to the regression models of the hydrogen generation to investigate the training error and the test error for the uncertainty prediction of the hydrogen generation.
Modeling of Low-Temperature Reduction of Metal Oxide in Hydrogen Treatment System for Severe Accidents in Nuclear Power Plants
Authors: Nakamura Kotaro , Tanabe Masashi , Abe Satoru , Mawatari Takashi , Nakagaki Takao 
Published: October 2020

Abstract
At the Fukushima Daiichi nuclear power plant, zirconium in the fuel rod cladding reacted with water vapor at elevated temperatures due to a loss of cooling water, resulting in the production of a large amount of hydrogen. This hydrogen leaked from the reactor vessel and accumulated in the top of reactor building, eventually leading to an explosion. A hydrogen treatment system that re-oxidizes hydrogen to water vapor is one of the effective methods to prevent such an explosion. A prominent re-oxidation method is via a fixed bed reactor packed with metal oxide pellets. The advantages of this method are its relatively fast oxidation rate without external oxygen/air injection. In this study, experiments and complementary numerical calculations were performed on the hydrogen re-oxidation reaction by metal oxides. The oxidation of hydrogen by copper oxide is modeled by 5 interacting, elementary reactions consisting of 6 chemical species. Experiments were performed using two packed bed set-ups, with measurement of inlet/outlet gas composition and pre/post-analysis of solid composition used to determine constants of the individual reaction rates for numerical calculations. From these reaction constants, the temporal behavior of the outlet gas was predicted.
Thermal-Hydraulic Design Support and Safety Analyses of SEALER UK Demo
Authors: Zwijsen Kevin , Uitslag-Doolaard Heleen , Roelofs Ferry , Wallenius Janne 
Published: October 2020

Abstract
SEALER (SwEdish Advanced Lead Reactor) is a passively safe lead-cooled reactor designed for commercial power production, under design by the LeadCold company. The reactor is modular in design, allowing for factory production and reduction in investment risk compared with new-build of large Light Water Reactors. Furthermore, its core is designed such that it can generate power for up to 25 years without the need of on-site fuel-cycle operations. The SEALER UK model has specifically been designed to produce base-load power on the UK grid. In the design and safety evaluation process, NRG is currently providing support to LeadCold Reactors with respect to thermal-hydraulic safety analyses utilizing Computational Fluid Dynamics (CFD) competences. The current paper gives a comprehensive description of a 3D CFD model created of SEALER UK Demo, which is a scaled-down demonstrator of SEALER UK. The geometry of the CFD model of SEALER UK Demo as well as the modelling approach and numerical settings are presented here. Assumptions were made in order to make it computationally feasible to perform simulations. These are discussed as well. Subsequently, the 3D CFD model is used to perform steady-state analyses of SEALER UK Demo operating under nominal conditions. Main parameters such as mass flow rates, temperatures and core pressure drops coming from the model match the design values well, with differences being at most a couple percent. Also, it is found that the margin to lead freezing with the current design parameters is more than 50K.
MELCOR Demonstration Analysis of Accident Scenarios at a Spent Nuclear Reprocessing Plant
Authors: Wagner Kenneth C. , Louie David L. Y. 
Published: October 2020

Abstract
The work presented in this paper applies the MELCOR code developed at Sandia National Laboratories to evaluate the source terms from potential accidents in non-reactor nuclear facilities. The present approach provides an integrated source term approach that would be well-suited for uncertainty analysis and probabilistic risk assessments. MELCOR is used to predict the thermal-hydraulic conditions during fires or explosions that includes a release of radionuclides. The radionuclides are tracked throughout the facility from the initiating event to predict the time-dependent source term to the environment for subsequent dose or consequence evaluations. In this paper, we discuss the MELCOR input model development and the evaluation of the potential source terms from the dominated fire and explosion scenarios for a spent fuel nuclear reprocessing plant.