bluuebunny/crossref_metadata_2025 huggingface dataset:
The Emergency Operating Strategy Study of Containment Spray Based on Radiation Level in LBLOCA Condition
Authors: Yi Ke , Li Yuanye
Published: October 2020

Large break LOCA will cause an increase in containment pressure, containment temperature and containment radiation level in nuclear power plants (NPPs). Containment spray is one of the most effective ways to mitigate the consequences of large break LOCA for these following facts, first, with the large space containment design, the containment spray can decrease the pressure peak and keeps containment integrity. Secondly, the containment spray can decrease the aerosol radiation level in containment, iodine in particular, and reduce the risks of radioactive release. Above all, the common strategy of containment spray in NPPs generally includes automatic actuation with high spray flow, in order to achieve good results in relevant accident conditions. Meanwhile, the strategy to shutdown containment spray should be considered as a result of these facts that, a weakened effect in decreasing radiation will occur and negative containment pressure may cause containment integrity damage in post-accident long term operation. For the above considerations, the emergency operating strategy of containment spray based on radiation level in large break LOCA condition and the relevant best estimate work are studied based on one NPP in this paper, in order to achieve reasonable results in containment spray operating strategy, which are able to optimize containment spray and reduce the bad consequences.
Study on Scattering Correction of the 60Co Gantry-Movable Dual-Projection Digital Radiography Inspection System
Authors: Ni Minzi , Li Guangchao , Wang Zhentao
Published: October 2020

Nuclear safety and security is getting more and more attention all over the world. Vehicles in and out of nuclear facilities need to be strictly inspected to prevent carrying of nuclear materials, explosives or other dangerous items. A new-type vehicle inspection system — 60Co gantry-movable dual-projection digital radiography inspection system is developed in Tsinghua University under the support of China’s Nuclear Energy Development Project, which uses two 60Co as radiation sources. The radiation sources are arranged at the bottom and side of the vehicles to be detected and the ionization chamber detectors are set in the two side of the gantry correspondingly. With moving of the sources and gantry synchronously, the system can obtain both side-view and upward-view images of the vehicles simultaneously [1]. However, there is a problem that the scattered rays from source of one projection can enter the detector array of another projection to form an interference signal. Experiments show that this kind of scattering noise can account for 20%, resulting in blurring and even artifacts, especially in thicker areas of the image. This problem needs to be solved urgently.

According to the characteristics of the Compton scattering effect, it is inferred that there is a certain non-linear mapping relationship between the scattering distribution of the detector array of one projection and the mass distribution in the plane of the other projection. This paper attempts to use BP neural network to learn this mapping relationship to quantitatively remove this kind of scattering noise. The results show that this method has certain effects on the removal of artifacts and blur caused by scattering. This method has the advantage of being fast and more targeted, compared with traditional image post-processing methods.
Multiphysics Analysis of Thorium-Based Fuel Performance Under Reactor Steady-State and Transient Accident
Authors: Qiu Chenjie , Liu Rong , Zhou Wenzhong
Published: October 2020

The ThO2 fuel has higher thermal conductivity and melting boiling point than the UO2 fuel, which is beneficial to the fast removal of heat and the improvement of fuel melt margin. In this paper, the material properties and thermodynamic behaviors of thorium-based fuel were firstly reviewed. And then the thermal physical properties and the fuel behavior models of Th0.923U0.077O2 fuel and Th0.923Pu0.077O2 fuel have been implemented in fuel performance analysis code FRAPCON and FRAPTRAN. Finally, the performances of Th0.923U0.077O2 fuel, Th0.923Pu0.077O2 fuel and UO2 fuel under both normal operating conditions and transient conditions (RIA and LOCA) are analyzed and compared. The Th0.923U0.077O2 fuel is found to have lower fuel center-line temperature and the thorium-based fuels are observed to have a delayed pellet-cladding mechanical interaction (PCMI) under steady state. Furthermore, the fission gas release, cladding strain and internal fuel energy under transient conditions are found to be lower too. Lastly, the cladding displacement and temperature under transient conditions are also compared. The thorium-based fuel was found to have a higher safety margin and accident resistance than conventional UO2 fuel under both normal operating conditions and accident conditions.
Stochastic Simulation Method for Reasoning of Dynamic Uncertain Causality Graph (DUCG)
Authors: Nie Hao , Zhang Qin
Published: October 2020

Dynamic Uncertain Causality Graph (DUCG) is an innovative model developed recently on the basis of dynamic causality diagram (DCD) model, which has been proved to be reliable for fault diagnosis of nuclear power plants. DUCG can represent complex uncertain causal relationship graphically, with both high efficient inference and support of incomplete expression. Therefore, DUCG is often built much larger than Bayesian Network (BN). However, as the scale of real problem is so large, DUCG still has the problem of combination explosion. Stochastic Simulation is a common solution for it. However, it is almost impossible to use traditional sampling algorithms for DUCG because the joint probability of evidences could be less than 10−20. In this paper, the algorithm based on conditional stochastic simulation for the inference of DUCG was proposed. It obtains the probability of evidences by calculating the expectation of the conditional probability in sampling process instead of using the sampling frequency, which overcomes the difficulty. What’s more, this algorithm uses recursive reasoning method of DUCG to calculate conditional probability distributions of node for sampling, which means this process only depends on its parent nodes’ states. As a result, the algorithm features in lower time complexity. In addition, it has the potential of parallelization like other sampling algorithms. In conclusion, this algorithm is promising to provide a new solution to the inference of the DUCG in large-scale and complex state situations.
CFD-DEM Simulations of Graphite Particle Collisions in Opposed Jet Mill
Authors: Peng Sifan , Liu Yujia , Gui Nan , Yang Xingtuan , Tu Jiyuan , Jiang Shengyao
Published: October 2020

Graphite is widely used in nuclear reactors as moderator and structural material. Among present graphite preparation methods, air flow mill is considered to be qualified in the control of particle size and purity, and promising for future mass production. In this work, an opposed jet mill is designed to crush large graphite particles. The opposed jet mill accelerates the particles through two supersonic jet flows in opposite directions, and finally the particles collide in the crushing cavity. In order to estimate the performance of opposed jet mill, it is necessary to solve the coupling calculation of the compressible flow and the collision process of discrete particles. However, the research on calculating the compressible gas solid coupling problems is scarcely rare. In this paper, coupled CFD-DEM model is used to simulate the particle movement process with jet flows and accompanying jet in opposed jet mill. By comparing with experimental results, it is proved that these simulation results of the acceleration process of compressible gas through these nozzles and the collision process of the final two supersonic jet flows in the opposed-jet mill are accurate, with the accuracy model of the coupled CFD-DEM provided. The practice has proved that the contrastive flow mill has a broad application prospect in the production of graphite particles.
Natural-Convective Transport Model of Radon From Building From Building Materials Into Indoor Atmosphere
Authors: Feng Shengyang , Chen Puxin , Cui Yu , Wu Yurong , Li Ce
Published: October 2020

Building material is one of the main sources of indoor radon in China. Radon exhaled from building materials enters the indoor atmosphere and transports in the indoor air space driven by buoyancy-driven airflow. This paper established a numerical model to reveal natural-convective effects on radon transport from building materials into the indoor air space. The building wall is approximated by a rectangular porous medium with uniform porosity and isotropic permeability. The buoyancy-driven flow is assumed to be turbulent and incompressible, ignoring viscous dissipation. The model can predict the dynamic coupling of radon in building materials and indoor atmosphere at different temperature gradient. This paper focuses on the analysis of radon transport with average Nusselt number and Sherwood number of interface between the building wall and the indoor atmosphere, which reveal the mechanism of radon exhalation from building materials into indoor atmosphere.
Study on Natural Circulation Flow Instabilities in Rod Bundle Channel Under Rolling Condition
Authors: Cheng Kun , Deng Jian , Cai Rong , Qian Libo , Qi Peiyao , Ke Bingzheng
Published: October 2020

The effects of rolling condition on the flow instability characteristics of natural circulation (NC) in rod bundle channel were experimentally studied. A 3 × 3 rod bundle channel is used as the testing section. The experimental system pressure range is 0.2 to 0.6 MPa, and the range of inlet subcooling is 10 to 70 °C. The ranges of rolling motion amplitude and period are 10 ∼ 20° and 10 ∼ 30s, respectively. Two typical two-phase flow instabilities in rod bundle channel under rolling condition were found in experiments: (a) the trough-type oscillation caused by the vapor generation at the minimum point of flow fluctuation and (b) the compound oscillation formed by the superposition of the trough-type oscillation and DWOI. Experimental results show that the rolling motion can reduce the threshold heating power of trough-type oscillation and cause the occurrence of NC flow instability in advance. But the rolling motion cannot affect the dimensionless boundary of DWOI in rod bundle channel.
New Control System of VR-1 Training Reactor
Authors: Kropík Martin , Rataj Jan , Reisner Ladislav , Matousek Jiri
Published: October 2020

The VR-1 training reactor with 500W maximum power operates at the Czech Technical University in Prague. Its control system was approaching obsolescence. It was therefore decided to upgrade the control system using up-to-date hardware and software using a new industrial PC with Windows 10 and the RTX real time system. The control system provides control and safety functions. Regarding the control function, the control system involves reactor I&C tests, reactor startup, reactor operation, and automatic power control. The control system receives operational data from the reactor safety (protection) system, computes average reactor power and power rate safety variables, and calculates a third safety variable that provides a percentage deviation of the reactor real power from the given (wanted) power. The control system also provides safety functions; it can shut down the reactor when safety limits are exceeded. The development of the control system software was accompanied by a quality assurance process, and verifications of the software development phases were carried out. After the control system installation, thorough non-active tests of the control system with removal of two fuel elements (because of nuclear safety) from the reactor core were carried out. Input signals for the reactor I&C (neutron chamber pulses and current) were simulated by computer controlled pulse and current generators. All safety functions, reactor start up, and operation were carefully tested. After successful non-active tests, active tests with a standard reactor core configuration were successfully conducted.
Numerical Study of Tritium Mitigation Strategies for Fluoride Salt-Cooled High-Temperature Reactors
Authors: Zhang Sheng , Wu Xiao , Sun Xiaodong
Published: October 2020

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

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

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

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

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

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

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

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

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

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

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

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

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

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.
Thermal-Hydraulic Investigations of a Horizontal Dry Cask Simulator
Authors: Pulido Ramon J. M. , Lindgren Eric R. , Durbin Samuel G. , Salazar Alex
Published: October 2020

Recent advances in horizontal cask designs for commercial spent nuclear fuel have significantly increased maximum thermal loading. This is due in part to greater efficiency in internal conduction pathways. Carefully measured data sets generated from testing of full-sized casks or smaller cask analogs are widely recognized as vital for validating thermal-hydraulic models of these storage cask designs. While several testing programs have been previously conducted, these earlier validation studies did not integrate all the physics or components important in a modern, horizontal dry cask system.

The purpose of this investigation is to produce data sets that can be used to benchmark the codes and best practices presently used to calculate cladding temperatures and induced cooling air flows in modern, horizontal dry storage systems. The horizontal dry cask simulator (HDCS) has been designed to generate this benchmark data and complement the existing knowledge base.

Transverse and axial temperature profiles along with induced-cooling air flow are measured using various backfills of gases for a wide range of decay powers and canister pressures. The data from the HDCS tests will be used to host a blind model validation effort.
FUELPOOL: A Computer Program to Model CANDU Spent Fuel Pool Severe Accident Progression and Consequences
Authors: Song Yong Mann , Jung Jong Yeob , Nijhawan Sunil
Published: October 2020

CANDU PHWR spent fuel pools (SFPs), smaller than a tennis court, contain up to 38,000 or more (49,000 in Wolsong)fuel bundles in geometries not replicated in any other power reactor. Therefore, the phenomenological issues, accident progression pathways and effectiveness of mitigative actions are somewhat different. This requires a dedicated approach in progression and consequence assessments of potential accidents and development of mitigation measures. The SFPs house densely packed fuel bundles stacked in about a hundred vertical stainless steel tray towers, each containing 24 spent fuel bundles in each of the 16 or more (19 in Wolsong) horizontal fish basket style steel trays. Some of theupto 10 year worth of the on-line refuelled bundles in the SFP are at relatively high decay powers as fuel trays are prepped for the towers in near daily basis. In addition, there is a provision (see Figure 1) that a full core of bundles 20 days after being at full power can be transferred to the spent fuel bay at any time. About 4.5m of additional water layer on top of the tray towers provide radiation protection and a healthy margin to small rate of fluid loss.
Fatigue Risk Evaluation of a Pressure Vessel Plug Subject to Flow Induced Vibration
Authors: Wang Robert , Chang Louis , Hurst Tom , Hurst Antony
Published: October 2020

The steam generator (SG) channel head forms part of the reactor coolant pressure boundary and is of high nuclear safety duty. The channel head failure is considered intolerable and there are no reasonably practicable physical provisions available to prevent its failure. Therefore the channel head is classified as either an Incredibility of Failure (IoF) or High Integrity (HI) component and it requires additional analyses and assessments beyond the design code requirements to achieve and demonstrate its structural integrity.

The hydrodynamic effects of the primary coolant in the annulus around the drain plug of the SG channel head are not very well understood, but are exacerbated by high flow rates in its immediate vicinity due to typical design details. Vibration of the drain plug due to coolant flow may result in fatigue induced failure of the channel head drain penetration weld. In the study presented here, random vibration analyses and a fatigue assessment have been carried out for a submerged drain plug in a pressurized water reactor (PWR) SG channel head. A finite element (FE) model of the drain plug submerged in water coolant has been developed. Modal analyses confirmed that the natural frequency of the submerged drain plug is significantly reduced by the large hydrodynamic added mass from the surrounding fluid. The fatigue evaluation undertaken using ASME III fatigue curve concluded that the fatigue life usage due to the vibration of the drain plug is negligible even after an extended plant life. Therefore the coolant flow-induced drain plug vibration is not a threat to the channel head integrity.
Development of the Fuel Behavior Analysis Code for Mechanical Fuel Cladding Failure During Reactivity Insertion Event in PWR
Authors: Higashi Yuma , Murakami Nozomu , Yodo Tadakatsu , Yamamoto Teruhisa
Published: October 2020

Pellet-Cladding Mechanical Interaction (PCMI) failure is the one of failure mode which must be evaluated in the nuclear fuel safety. PCMI is caused by the mechanical load to cladding due to the fuel pellet expansion. Under the high fuel burnup condition, the fuel cladding may become degraded by embrittlement under the neutron-induced irradiation and hydrogen accumulation due to the waterside corrosion. In order to consider the further deterioration of the material with higher burnup, the evaluation using mechanical indicators, e.g. strain and stress, might be required.

In this study, a transient model is developed, which can mechanistically evaluate the PCMI behavior, in particular, for fuel rods under the higher burnup condition. The model is incorporated in a fuel behavior analysis code and verified by benchmarks with other similar codes. The PCMI predictability of this code is validated using the experimental test data.
Investigation on Velocity Distribution in the Subchannels of Pin Bundle With Wrapping Wire: Evaluation of Reynolds Number Dependence in 3-Pin Bundle
Authors: Aizawa Kosuke , Hiyama Tomoyuki , Nishimura Masahiro , Kurihara Akikazu , Ishida Katsuji
Published: October 2020

A sodium-cooled fast reactor is designed to attain a high burn-up core in commercialized fast reactor cycle systems. In high burn-up fuel subassemblies, the deformation of fuel pin due to the swelling and thermal bowing may decrease local flow velocity in the subassembly and influence the heat removal capability. Therefore, it is important to obtain the flow velocity distribution in a wire wrapped pin bundle. In this study, the detailed flow velocity distribution in the subchannel has been obtained by PIV (Particle Image Velocimetry) measurement using a wire-wrapped 3-pin bundle water model.

The test section consisted of an irregular hexagonal acrylic duct tube and fluorinated resin pins which had nearly the same refractive index with that of water and a high light transmission rate. This enables to visualize the inner subchannel through the outer pins. Flow velocity conditions in the pin bundle were set from 0.036 m/s (Re = 270) to 1.6m/s (Re = 13,500).

From the PIV results, it was confirmed that the normalized flow velocity near the wrapping wire in low Re number condition was decreased relatively compared to that in high Re number condition. In the region away from the wrapping wire, the maximum flow velocity was increased by decreasing the Re number. In particular, the maximum flow velocity was more than twice the cross section average velocity in the laminar flow (Re = 270). Moreover, the PIV measurements by using the 3-pin bundle geometry without the wrapping wire were conducted. From the results, the effect of the wrapping wire on the flow field in the subchannel was understood. There experimental results useful not only for understanding of pin bundle thermal hydraulics but also code validation.
Flow Characteristics of Upward Two-Phase Flows in a Rod Bundle Geometry
Authors: Han Xu , Shen Xiuzhong , Yamamoto Toshihiro , Nakajima Ken , Hibiki Takashi
Published: October 2020

In the present paper, the local two-phase flow parameters were measured with a four-sensor optical probe in adiabatic upward air-water two-phase flows in a 6 × 6 vertical rod bundle with rod diameter of 10 mm, pitch of 16.7 mm, square channel box side length of 100 mm and hydraulic equivalent diameter (DH) of 18.7 mm. The local measurements were performed in an octant triangular region of the rod bundle cross-section at the axial position with height-to-diameter ratio (z/DH) of 149 under a total of 16 flow conditions. The local void fraction, interfacial area concentration (IAC), bubble diameter and bubble velocity were obtained in the four-sensor probe measurements. Both of the measured void fraction and IAC show their radial local distributions with the core-peaking and wall-peaking shapes which are closely linked with the superficial velocity of two phases. The distribution shapes tend to change from the core-peaking to the wall-peaking when the superficial liquid velocity () increases and the superficial gas velocity () decreases. The measured diameters of local bubbles keep the similar values in the measuring cross-section, increase when the  increases and decrease when the  increases. The bubbles behave with the velocities whose main flow direction component keeps a typical radial power-law (core-peaking) profile and whose cross-sectional velocity components show a significant trend of bubbles migrating from the center to the wall region of the channel box especially under high  conditions. The area-averaged results of void fraction and IAC were obtained by a cross-sectional area-averaging scheme. The resultant area-averaged void fraction and IAC were used to check the void fraction predicting capability of two drift-flux correlations and the IAC prediction performance of two IAC correlations respectively. The applicability of these correlations to the rod bundle geometry was discussed and concluded finally in this paper.
Development of Design Support System for Piping Route and Differential Pressure Flowmeter by Three-Dimensional Fluid Analysis
Authors: Miura Takatsugu , Igarashi Kingo , Hosaka Tomoyuki , Kitagawa Takumi , Yashiki Tatsurou , Itabayashi Yuki , Suzuki Hirotsugu , Nakahara Takashi , Kitamura Junichi , Sano Tadashi
Published: October 2020

In power plants that becoming more compact, it will expend much time and effort to satisfy the requirement for the differential pressure flow measurement according to ISO’s standards. Therefore, it is difficult for engineers in the design phase to completely remove the potential for large errors in flow measurement. This paper presents the 3D fluid analysis system that is a lower cost than the conventional method to confirm the soundness of such measurement in the phase of piping route design.

This system has the function to automatically generate the analysis models from general 3D piping CAD data. The analysis program is written by the open source code to reduce a license fee. Also, this system has the function of calculating the swirl strength along the pipe axis as one of the means for efficiently supporting the design change. In order to verify and validate the analysis system, we analyzed several flow paths, confirmed the response of the swirl strength and flow rate indication value of the differential pressure flowmeter model. The analysis result well simulated the increase or decrease swirl strength in the complex flow path, and fluctuation of the flow rate indication value. Also, the system supports to set the flowmeter in the appropriate position by providing visualization of the swirl strength along the pipe axis. In the flow path analysis in this validation, it took about one month to visualization of the swirl strength along the pipe axis from the generation of the analysis models.

The 3D fluid analysis system collaborative with 3D piping CAD design system has been developed. This system enable to confirm the effects of swirl strength on flow measurement and the soundness of the differential pressure flow measurement at a lower cost in comparison with conventional method.