bluuebunny/crossref_metadata_2025 huggingface dataset:
Numerical Investigation of Supercritical Water Flow in a 2x2 Rod Bundle Under Non-Uniform Heat Flux
Authors: Rowinski Marcin , Soh Yeng Ch. , White Timothy J. , Chieng Ching Ch. , Zhao Jiyun
Published: October 2016

Generation III/III+ nuclear reactors operate with working fluid under subcritical conditions (Tc = 647K, pc = 22.115MPa). The efficiency, limited by the ratio of source and sink temperatures, is restricted by operating below the critical temperature. The supercritical water reactors (SCWRs) are able to rise efficiency limit while operating at the supercritical conditions. The amount of energy carried by working fluid is higher leading to potential efficiency improvement of nearly 30% above current nuclear stations. Therefore, rendering nuclear energy as one of the most efficient decarbonized electrical energy sources with efficiency of 45% and capacity factor of ca. 90%. Typical capacity factors of competing wind turbines and solar PV cells reaches 45% and 15% while the efficiencies 50% and 45%, respectively.

In a subcritical reactor a uniform heat flux is generated due to relatively constant fuel moderation. However, due to a change of density during transition from sub- to supercritical conditions, the fuel moderation is uneven along the fuel rod and results in a non-uniform heat generation. The literature on SCWR neutronics suggests higher heat generation at the fuel channel entrance. In this paper we simulated for the first time such non-uniform heat flux generated in a SCWR, we analyze the impacts of such flux on the working medium flow and suggest ways to mitigate negative impacts of non-uniform heat flux.

The study was conducted with use of Computational Fluid Dynamics (CFD) software. Obtained results show that the shape of heat flux curve along the channel highly influences the wall temperature distribution along the fuel channel. The differences in maximum wall temperatures can be up to 200K for different curve’s shape. Moreover, the maximum wall temperature is always higher than in default case i.e. when uniform heat flux is applied. It is possible to control the wall temperature distribution by adjusting the shape of heat flux along the axis. Such adjustment can be made by using different enrichment levels along the fuel rod axis, unfortunately any change in power distribution caused rapid temperature increase at the upstream location.
Numerical Study on a Single Square Cylinder Subjected to Vortex-Induced Vibration in HTR-PM
Authors: Xiao Qiankun , Shi Li , Wang Xiaoxin , Luo Xiaowei , Wu Xinxin
Published: October 2016

The heat transfer tube bundles of the steam generator in high temperature reactor pebble bed module (HTR-PM) are subjected to high speed helium flow, which might lead to vortex-induced vibration (VIV). In the present paper, to investigate the vibration of square cylinder under flow effect, vortex shedding phenomena of a stationary square cylinder at a high Reynolds number equal to 6.8 × 104 is simulated by detached eddy simulation (DES) turbulence models. A comparatively close agreement with previous experimental results is achieved. Combining computational fluid dynamics (CFD) and computational structural dynamics (CSD) methods, a fluid-structure-interaction (FSI) model for VIV of the square cylinder is then established, and vibration response perpendicular to flow direction were investigated. Moreover, the safe range of the natural frequency of the square cylinder to avoid synchronized vibration with VIV is analyzed. The results of this paper can provide an important guidance to the design of the heat transfer tubes and their supports in HTR-PM.
Test Analysis of CAP1400 5cm Break Accident Based on the ACME Test Facility
Authors: Zhu Sheng
Published: October 2016

CAP1400 is a large pressurized water reactor based on the passive safety conception. An ACME (Advanced Core-cooling Mechanism Experiment) facility has been designed and constructed in order to validate that the CAP1400 system design is acceptable to mitigate the loss of coolant accident (LOCA). The ACME test facility is an isotonic pressure, 1/3-scale height and 1/54.32-scale power simulation of the prototype CAP1400 nuclear power plant. It contains the main-loop system, passive safety system, secondary steam system and auxiliary system etc. The all of ACME test matrix including 5 kinds 21 cases .In this paper, the test results and the Realp5 prediction of the cold leg 5cm break accident of CAP1400 are compared and analyzed to briefly evaluate the ACME capability. Furthermore, 3 different types of 5cm cold leg break test cases are presented, and the transient process, system responses and key parameters tendency are analyzed based on the test. The results indicate that the passive safety system design can successfully combine to provide a continuous removal of core decay heat and the reactor core remains to be covered with considerable margin for the 3 different 5cm cold leg break accidents.
Numerical Simulation of Flow Regime Transition From Slug to Wavy Flow in Helically Coiled Tubes
Authors: Li Zhaoxu , Zhu Hongye
Published: October 2016

Two-phase flow in helically coiled tubes is becoming the interest of many investigators because of its importance in various applications, such as nuclear engineering, chemical engineering, refrigerating engineering and power engineering. Compared with U-type tubes used in pressurized water reactor (PWR), helically coiled tubes have advantages in size, heat transfer capacity, thermal stress toleration and two-phase stability. Accordingly the helically coiled tubes have been utilized in the steam generators of the next general reactors, such as gas-cooled reactor, fast breeder reactor and integrated pressurized water reactor. In helically coiled tubes the characteristics of momentum and heat transfer are distinct from those in straight tubes due to the presence of centrifugal force, especially for two-phase flow. Meanwhile, the transitions of flow regime, which is the crucial knowledge for the designers to determine the heat transfer rates and flow resistance, are also significantly affected by the centrifugal force.

In this study, two-phase flow regimes in helically coiled tubes are investigated. Computational fluid dynamics (CFD), using fractional volume of fluid (VOF) model, is carried out to simulate wavy and slug flow regimes in helically coiled tubes. The corresponding experiment is also conducted to visualize these flow regimes at different superficial flow velocities. Numerical simulation results actually reflect the influence of centrifugal force on the two-phase flow and show a good agreement with the photographs captured from the experiment. Based on the simulations at different superficial flow velocities, the boundary between the slug and wavy flow regimes is predicted, in addition, compared with that in inclined tubes. The comparison indicates that centrifugal force could induce the appearance of wavy flows in advance and prompt the transition from slug flow to wavy flow.
Development of Transient and Safety Analysis Method for ATF With SiC
Authors: Sebe Fumie , Yamada Masato , Takeuchi Yutaka , Kakiuchi Kazuo , Okonogi Kazunari
Published: October 2016

Based on lessons learned from the Fukushima Daiichi nuclear power plant accident, pursuit of accident tolerant fuel (ATF) has been discussed by many institutions in the world. Toshiba identified a silicon carbide (SiC) ceramic as the most promising material for accident tolerant fuel. Since SiC has less active characteristics in the presence of high temperature water steam (H2O) and is expected to be tolerant of severe accident conditions. Moreover, SiC has a smaller neutron absorption cross-section which is advantageous feature in terms of neutron economy.

Zirconium alloys (Zry) are one of the main structural materials in LWR core. In high temperature H2O environment under severe accident conditions, Zry rapidly reacts with H2O and oxidation reaction accompanied by release of hydrogen gas occurs. Since SiC may inhibit the progress of oxidation reaction compared to Zry metal alloys, hydrogen and heat generation is expected to decrease in the case of core uncovered accident conditions. In order to confirm the advantage of SiC over Zry as core materials, transient analysis and safety analysis are carried out. For transient analysis, analyses of temperature behavior of cladding at plant transient condition are carried out with best-estimate transient analysis code. This analysis confirmed the effect of physical properties differences between SiC and Zry on cladding temperature behavior. Moreover to indicate the effectiveness of SiC under the core uncovered condition with oxidation reaction, safety analysis with latest “MAAP” code is carried out and the whole plant behavior during severe accident sequence is simulated. This analysis showed the effectiveness of SiC to mitigate the oxidation reaction. As the result of these analyses, the advantage of SiC over Zry can be perceived. And also, future challenges of SiC application as ATF can be clarified through these analyses.
Safety Assessment of Fuel Storage Ponds Following Cooling Faults
Authors: Webb Jonathan , Bridgford Charles
Published: October 2016

For spent nuclear fuel stored within a cooling pond, the essential nuclear safety functions of control, cooling and containment are fulfilled by maintaining an appropriate depth of water above the fuel. External cooling systems remove the decay heat generated by the spent fuel stored within the pond, in order to maintain the temperature of the water at a constant level. In the event of a fault within these external cooling systems, there is the potential for a temperature excursion within the pond. Historically the UK nuclear industry has considered that such faults would pose no threat to the structural integrity of the pond containment and hence the only loss of water would be due to evaporation following a loss of cooling. However, more recently, it has been recognised that such temperature excursions may result in through-wall cracking leading to a loss of water and undermining of these essential safety functions. This paper outlines the safety case implications of these realisations and the way in which they are being addressed within the UK’s nuclear power stations. The paper considers the effects of thermal transient faults on the concrete pond structure and the potential nuclear safety issues which may occur as a result of this.

In response to potential pond cooling faults, consideration is given to the requirement for engineered protection systems along with the safety role of the operator in identifying and responding to faults of this kind. Operators provide a versatile mechanism for identifying fault conditions and taking remedial actions, however, the benefit which can be formally claimed for their role within a safety case is generally limited by the availability or reliability of instrumentation to reveal a fault condition. Post fault operator actions may also be limited by the timescales available following a fault, or by other demands on the operators, which may occur in the event of an external hazard which affects multiple site systems. To quantify the timescales available for post fault remedial action, it is necessary to quantify the rate of water loss from the pond, along with the relationship between pond water depth and the radiological consequences both on-site and off-site. This paper investigates the difficulties which may be encountered in quantifying the role of post fault operator actions within such a safety case, and in demonstrating that the overall nuclear safety risk is acceptably low and as low as reasonably practicable (ALARP).
Review of Experimental Studies on the Heat Transfer Behaviour of Volumetrically-Heated Pool With Different Boundary Conditions and the Influence of Crust Formation
Authors: Gaus-Liu Xiaoyang , Miassoedov Alexei , Gabriel Stephan
Published: October 2016

This study investigates the heat transfer characters of a volumetrically-heated melt pool in LWR lower plenum under different top and side boundary conditions. The inconsistency in parameter definitions in Nu-Ra correlations is addressed, and the effect of 2D geometry and crust formation are analyzed. A summary of test conditions of the previous experimental studies are given. The average upwards and downward Nusselt number from LIVE tests in 3D and 2D geometries are compared with each other and with other well-known correlations. The differences of heat flux distribution along the vessel wall are analyzed regarding boundary cooling condition and crust formation. This paper provides some explanations about the discrepant among the exiting heat transfer correlations and recommends most suitable descriptions of melt pool heat transfer under different accident management scenarios.
Comparison of the ASME, RCC-M and ISO Requirements for Welding Qualifications and Welding Quality
Authors: Consonni Marcello , Wasylyk Andrew , Leclerc Stéphane , Sperko Walter , Malouines Philippe
Published: October 2016

This paper provides a comparison between the requirements for the qualification of welding procedures, welders and welding operators, as well as for welding supervision/coordination and welding quality according to the ASME Boiler & Pressure Vessel code (ASME BPVC Sections III and IX), the Afcen RCC-M code (France) for nuclear plant components, and ISO standards referenced to by the latter. The work was carried out within an international project that included other major codes and standards applied in the Nuclear sector, ie CSA (Canada), JSME (Japan), KEPIC (Korea) and Rostechnadzor PNAE G-7 (Russia), as part of a larger effort towards harmonization of nuclear pressure-boundary codes and standards. The comparison work aimed at identifying significant differences, as well as areas of current or potential convergence and future harmonization of the abovementioned codes and standards.

A line-by-line comparison was carried out, using the ASME BPVC as the baseline. Corresponding requirements in the different codes and standards were identified and categorized as ‘identical’, ‘equivalent’, ‘technically different’ or ‘not specified’ (when addressed by the ASME BPVC only). The most significant differences are discussed and the main findings are presented, including upcoming code and standard changes that may affect harmonization. Concluding remarks are provided with regard to the future effort required for harmonization of the codes and standards considered, through convergence and/or reconciliation.
Temperature Dependence of Commercially Available Betavoltaic Cells
Authors: Cheu Darrell , Adams Thomas , Revankar Shripad
Published: October 2016

There is an increasing need for devices that can be powered for extended periods of time where it is difficult or impossible to maintain or replace, such as pacemakers, long term space flight or undisturbed sensors for military use. Presently, most portable devices run off a Lithium-Iodide battery, which gives a high amount of power but could only last approximately 2 to 5 years, requiring frequent replacement. However, replacement is unnecessary for betavoltaic cells as they can last at least 20 years. City Labs Inc. received a general license for commercially available tritium betavoltaic cells that were validated at extreme temperatures without permanent degradation. To fully determine the effectiveness of a betavoltaic cell, the electrical performance (I-V curves) of three betavoltaics were evaluated while temperatures were ramped up and down from −30°C to 70°C. Short circuit current, open circuit voltage, maximum power and fill factor were used to compare electrical performance. Results indicated that the open-circuit voltage and maximum power decreased as temperature increased, suggesting that betavoltaic cells are suited for cold environments below 0°C, such as during nightfall when a photovoltaic cell may not be used.
Experimental Investigation of Subcooled Choking Flow in a Steam Generator Tube Crack
Authors: Nguyen Hung , Brown Mark , Revankar Shripad T. , Riznic Jovica
Published: October 2016

Steam generator tubes have a history of small cracks and even ruptures, which lead to a loss of coolant from the primary side to the secondary side. These tubes have an important role in reactor safety since they serve as one of the barriers between radioactive and non-radioactive materials of a nuclear power plant. A rupture then signifies the loss of the integrity of the tube itself. Therefore, choking flow plays an integral part not only in the engineered safeguards of a nuclear power plant, but also to everyday operation. There is limited data on actual steam generators tube wall cracks. Here experiments were conducted on choked flow of subcooled water through two samples of axial cracks of steam generator tubes taken from US PWR steam generators. The purpose of the experimental program was to develop database on critical flow through actual steam generator tube cracks with subcooled liquid flow at the entrance. The knowledge of this maximum flow rate through a crack in the steam generator tubes of a pressurized water nuclear reactor will allow designers to calculate leak rates and design inventory levels accordingly while limiting losses during loss of coolant accidents. The test facility design is modular so that various steam generator tube cracks can be studied. Two sets of PWR steam generators tubes were studied whose wall thickness is 1.285 mm. Tests were carried out at stagnation pressure up to 6.89 MPa and range of subcoolings 16.2–59°C. Based on these new choking flow data, the applicability of analytical models to highlight the importance of non-equilibrium effects was examined.
Numerical Investigation on Heat Transfer of Supercritical Water in Rod Bundle Under Suddenly Decreased Mass Flux Condition
Authors: Cheng Hui , Zhao Jiyun
Published: October 2016

Supercritical water reactor (SCWR) is one of the most promising nuclear reactor system among generation IV reactors thanks to its high thermal efficiency and simplicity. One of the main features of supercritical water is the strong variation of thermal-physical properties in the vicinity of the pseudo-critical temperature, which makes it very hard to predict the thermal-hydraulic behavior near this point. In this paper, CFD is used to investigate heat transfer of supercritical water in a 2×2 rod bundle with SST k-ε turbulence model. Two steady-state and one transient cases are simulated. The results show that there is strong non-uniform temperature distribution around the circumferential direction. Heat transfer deterioration (HTD) is found in the front of the heated section along the axial direction when the bulk temperature is near the pseudo-critical point due to secondary flow. Comparision of transient and steady-state flow shows that when the mass flux is less than 700 kg/m2s, the temperature in transient state is smaller than that in the steady-state, especially when the mass flux is 400 kg/m2s, the temperature difference is more than 10 °C.
Identification Modeling of Accelerator Driven System Based on Growing and Pruning Radial Basis Function Network
Authors: Luo Run , Wu Shifa , Wei Xinyu , Zhao Fuyu
Published: October 2016

An identification method based on growing and pruning radial basis function network (GAP-RBFN) is presented for modeling an accelerator driven system (ADS). Compared with traditional neural networks, GAP-RBFN could automatically adjust the number of hidden neurons to find a suitable network structure by using growing and pruning strategies. In addition, an extended Kalman filter (EKF) algorithm is adopted to update network parameters of neurons in GAP-RBFN, which has a rapid convergence speed during the training process. A numerical calculation code named ARTAP (ADS Reactor Transient Analysis Program) is used to generate data for training GAP-RBFN. After GAP-RBFN is trained by the data, an identification model for ADS is established. The simulation results obtained from the GAP-RBFN model are compared with those obtained from a recurrent neural network (RNN) model. It is shown that the GAP-RBFN model not only has higher prediction accuracy than the RNN model, but also has faster computation speed than the numerical calculation code. Owing to its accuracy, simplicity and fast computation speed, the proposed GAP-RBFN method can be used to model the ADS reactor.
Development of Plenum Temperature Model for Accident Analysis
Authors: Li Wenjie , Jiao Yongjun , Chen Bingde
Published: October 2016

Plenum temperature is an important parameter in the calculation of LWR fuel rod internal pressure, since plenum contributes over 40% of free volume in typical LWR fuel design. In many fuel performance codes the plenum temperature are determined in a simple and empirical method. There is a concern about whether such a simple method can properly reflect the temperature variation during fast transients and accident conditions. Therefore, a detailed simulation of heat transfer process in plenum based on COMSOL Multiphysics software was performed to give a fine resolution of temperature and to investigate the heat transfer mechanisms. Since the COMSOL application is a stand-alone model, a new plenum temperature model which can be coupled with fuel performance code was also developed. Finally the validation of the new plenum temperature model is performed by comparing its results with COMSOL model and FUPAC code.
A Conceptual Multiscale Entropy Production Approach for Assessing Thermal Hydraulics Characteristics in Nuclear Reactor Systems
Authors: Ji Yu , Zhang Hao-Chun , Xu Guo-Qiang , Zhang Yi-Ning , Wang Xu-Wei
Published: October 2016

Thermal Hydraulics (T-H) process exerts important effect on nuclear reactor systems. Due to their complexity, it is necessary to propose an appropriate approach for modeling and investigating the processes, such as multiscale modeling technique. The current work attempts to provide a scheme of conceptual multiscale entropy production approach in view of descriptions of thermal hydraulic characteristics in nuclear reactor system at local, component and system scales. Based on some fundamental principles, the optimal thermal hydraulics design (OTHD) procedure in implementing the approach is established. Moreover, two instances are considered to show the applicability of the current approach, with respect to local and system scale. In addition, the approach proposed would contribute to make deeper understanding of the thermal hydraulic processes.
Modifications of Generic Pressurized-Water Reactor Units to Increase Cycle Efficiency
Authors: Ponomaryov Paul , Zhou Yifeng , Mazza Cristina-Maria , Pioro Igor
Published: October 2016

Currently, Pressurized Water Reactors (PWRs), Boiling Water Reactors (BWRs) and Pressurized Heavy Water Reactors (PHWRs) have the lowest thermal efficiencies compared to those of other nuclear-power reactors and thermal power plants. Therefore, the objective of this paper is to propose modifications to a generic PHWR to yield an increase in overall plant thermal efficiency.

The focus of this paper is primarily on the secondary side of a Nuclear Power Plant (NPP) and is directly dealing with wet-steam High-Pressure (HP) and Low-Pressure (LP) turbine stages and a Moisture Separator and Reheater (MSR). Modifications of the HP and LP turbine stages are based on utilizing moisture removal stages (having up to 60% removal efficiency), which reduce the moisture content as the steam passes through those turbine stages. Reduced energy losses and an increase in mechanical efficiency due to lesser moisture content results in an increase in thermal efficiency. Furthermore, implementing moisture-removal stages in the LP turbine gives the ability to eliminate the reheater in the MSR, thus resulting in an increase of thermal efficiency due to both, the higher mechanical efficiency of an LP turbine and the redirection of live-steam previously used by the MSR to a HP turbine. To be able to show an increase in thermal efficiency based on these modifications of a generic PHWR, the Pickering CANDU-6 nuclear-reactor parameters were used as a reference case in the software, called DE-TOP.

The modifications suggested in this paper can be applied to any NPP that uses a Rankine steam-turbine cycle on the secondary side (PWR, PHWR and/or BWR) and recommended for implementation during planned replacement of LP- and/or HP-turbine rotors for new construction of PWRs and PHWRs.
Neutronics and Thermo-Hydraulic Analysis of Water-Cooled Blanket Based on PWR and SCWR Water Conditions for CFETR
Authors: Cheng Jie , Wu Yingwei , Su G. H. , Qiu Suizheng , Tian Wenxi
Published: October 2016

China Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor being designed in China to bridge the gap between ITER and future fusion power plant. As one of the candidates, a water-cooled solid breeder blanket based on PWR (pressurized water reactor) and SCWR (super-critical water reactor) water conditions were proposed. In the concept, multiplying layers separated by three breeding layers were designed and optimized for higher Tritium Breeding Ratio (TBR) and uniform heat distribution. This blanket uses the Li2TiO3 lithium ceramic pebbles as the breeder, while beryllium pebbles as the neutron multiplier. In this paper, the thermal and fluid dynamic analyses of the optimized blanket on both water conditions were performed by numerical simulation, to discuss thermo-hydraulic performance of the blanket using pressurized water/supercritical water as its coolant. The nuclear heating distribution was obtained from the neutronics calculations by MCNP. The thermal hydraulic behaviors of the first wall (FW), structure material, Li2TiO3 pebble bed and Beryllium pebble bed under normal condition were calculated, respectively. It was found that the temperature on the blanket can be effectively cooled on both water conditions, certified the feasibility of the blanket design with pressurized/supercritical water cooling scheme. It indicated that SCWR case had smaller safety margin than PWR case, but SCWR case would lead higher outlet temperature, thermal conductivity and heat exchange efficiency also. In addition, it was found that beryllium was the dominant factor leading a higher TBR. The results would be important to water condition choice for solid blanket in the future.
Development and Preliminary Validation of a STH-CFD Coupling Method for Multiscale Thermal-Hydraulic Simulations of the MYRRHA Reactor
Authors: Toti A. , Vierendeels J. , Belloni F.
Published: October 2016

MYRRHA (Multi-purpose hybrid research reactor for high-tech applications) is a lead-bismuth eutectic (LBE) cooled research reactor currently under development at SCK•CEN, the Belgian Nuclear Research Centre. The compact design of the pool-type primary system implies the presence of pronounced 3D thermal fluid-dynamic phenomena, which can affect the evolution of certain accidental transients such as loss of flow (LOF). System thermal-hydraulics (STH) codes, conceived to carry out global NPP safety analyses, present severe limitations in taking into account local 3D phenomena including flow mixing, thermal stratification, etc. To overcome this limitation, a promising solution is coupling STH codes with CFD codes, which can calculate complex flow fields but result, on the other hand, in too expensive computational resources for whole-plant simulations.

A domain decomposition method that couples the STH code RELAP5-3D and the CFD code Ansys FLUENT has been developed and implemented. Proof-of-principle tests on simple configurations have been carried out to demonstrate its validity and to identify modeling and numerical issues. The experimental campaign carried out at the test facility TALL-3D, operated by the KTH Royal Institute of Technology in Sweden, has been selected for preliminary verification and validation (V&V) of this method. This paper presents the results of the coupled 1D-3D simulation of a forced-to-natural circulation transient event, whose evolution results to be strongly affected by flow mixing and stratification phenomena. The experimental validation, based on a high-quality set of experimental data, is currently on-going. Further development and validation activities will be carried out in the experimental facility ESCAPE, under commissioning at SCK•CEN, within the recently launched EU project MYRTE (Horizon 2020 programme).
Graphical User Interface With Reverse-Engineering of the Cyclic Voltammetry Method for Electrochemical Processes of Used Nuclear Fuel
Authors: Rakhshan Pouri Samaneh , Phongikaroon Supathorn
Published: October 2016

This work focuses on an interactive reverse-engineering program design for the cyclic voltammetry (CV) method to help elucidating, improving, and providing robustness in detection analysis in the absence of complete experimental data sets during an electrorefining process of used nuclear fuel reprocessing. The work has been implemented into a Graphical User Interface (GUI) of the commercial software MATLAB allowing an individual user to directly control and make adjustments to support material detection and accountability. Analyzing and reconstructing the CV plots for uranium (U) in a LiCl-KCl molten salt at 500°C under different scan rates and at 1, 2.5, 5, 7.5, and 10 wt% have been accomplished. These test values provide the current (amp) versus potential (V) and concentration of each species (mol/cm3) versus the operating time (s) graphs under different specified conditions. The computational code uses the electrochemical fundamentals coupling with various experimental values existing in the literature such as the diffusion coefficients, formal potentials, reversible/irreversible time duration for reverse engineering of the CV technique. The user needs to specify only the desired concentration of uranium and the scan rate. All other experimental data sets for each condition have been stored in the code and can be used to interpolate between the existence data. The developed routine can be used to detect the peaks at the reversible and irreversible parts despite deficiencies of experimental data in a very short run time (around one minute) with an adequate selected time interval of approximately 0.08 second. Results indicate that the model can trace the current versus potential graph with a low root-meant-square (RMS) error compared to the experimental reported in literature. The concentration of each species at the reversible and irreversible of anodic and cathodic sides can be calculated and are shown based on increasing time which provided a good view of the whole process.
MOCUM Code Verification and Sensibility Study Using C5G7 Benchmark
Authors: Yang Xue , Borse Rajan , Satvat Nader
Published: October 2016

This work uses the 2-D C5G7 benchmark to verify the accuracy of the MOCUM code, a parallel neutronics program based on the method of characteristics (MOC) for solving arbitrary core geometry. Compared to the MCNP results, MOCUM k-eff, maximum assembly and pin power percentage errors are 0.02%, −0.06%, and 0.64%, respectively. The results demonstrate the high accuracy of the MOCUM code. The calculation uses a total of 56 threads, and the runtime on dual Intel Xeon E5-2699 v3 CPUs is 26 minutes, with speed up higher than 50 times. The sensitivity study of various MOC parameters using the calculation of the C5G7 benchmark problem is also performed. The study reveals that C5G7 requires the usage of 48 or more azimuthal angles. The strong flux gradient and the heterogeneous effects need fine unstructured meshes to resolve. The simulation uses 258 million zones with an average mesh size of 0.016 cm2. The investigation of the polar angle quadrature indicates that Leonard polar angle performs slightly better than Gauss-Legendre and Tabuchi polar angles and more than three polar angles are not necessary. In addition, parameter sensitivity study shows that coarse parameters are prone to introduce error to the neutron flux but not k-eff.
Investigation of the Effects of Hydrogen Atoms Concentration on the Tungsten Sigma 5 (310) Symmetric Tilt Grain Boundary
Authors: Atiku Kabiru , Yang Xue
Published: October 2016

The objective of this work is to investigate the effects of the concentration of hydrogen atoms at the tungsten Σ5 (310) grain boundary (GB) on the GB energy and the GB pulling force. Tungsten (W) is the most preferred plasma facing material (PFM) for the future nuclear fusion reactors such as in the proposed DEMO (demonstration power plant) and ITER (international thermonuclear experimental reactor — which is at the moment the largest Tokamak nuclear fusion reactor under construction in the world. Tungsten is considered as a PFM because of its excellent thermal properties, low-sputtering yield and high melting point. However, hydrogen (H) atoms have an affinity for tungsten grain boundary, and are trapped there permanently. In addition, it makes it prone to failure. W will be exposed to extremely high fluences of H isotopes. The low-energy H isotopes will be retained in the tungsten material leading to the formation of blisters in W and causes degradation of the mechanical and thermal properties of W. Therefore for safety reason, the effect of hydrogen atoms at the tungsten Σ5 (310) symmetric tilt-GB needs to be investigated. This will greatly aid in better designing of fusion wall materials. In addition, the full understanding of the tungsten grain boundary energy and the force required to pull the grain boundary apart will also help in better material selection. Classic molecular dynamics method was used for the investigation.

LAMMPS-a sophisticated, classical atomic and molecular dynamics modelling and simulation software, is a vital tool adopted for the investigation of the effects of hydrogen atoms in tungsten Σ5 (310) symmetric tilt-GB. Tersoff potential for W-H interactions was used for the modelling. The size of the simulation box is 10 × 100 × 10 lattices and it consists of 21262 W atoms. Periodic boundaries were used for all sides of the system. Conjugate gradient method was used for the minimization. The trajectories of the atoms were visualized using visual molecular dynamics (VMD) software. The GB energies are calculated to be −924.060 J/m2, −923.898 J/m2 and −743.414 J/m2 for pure W, W with one H atom and W with 30 H atoms are at the GB respectively. In addition, the forces required to separate the GB apart are 0.0279551eV/Ang for pure W, 0.024789eV/Ang and 0.0185eV/Ang for W with one H and 30 H atoms at the GB respectively.

The result shows that hydrogen acts as a grain boundary embrittler and weakens the GB strength. The GB energy reduces as the concentration of hydrogen at the tungsten GB increases. In addition, the more the hydrogen atoms at the tungsten GB the lesser the value of the force required to pull the GB apart. However, the GB energies and pulling forces starts increasing slowly when H atoms exceed a certain number depending on the H atoms distribution around the W GB.
Guidance for Human Action Evaluations for External Flood Risk Assessment
Authors: Schneider Raymond E. , Visweswaran Srinivasa , Fluehr John , Hackerott H. Alan
Published: October 2016

For many years external flooding hazards have been recognized as significant contributors to plant risk. However, it was not until the events at Fukushima that there was a concerted effort on the part of the utilities to reassess the plant external flood design basis, identify external flood vulnerabilities and take actions to address them. For many plants, resolution of low probability high consequence floods will likely be addressed by a combination of actions involving enhancements to flood protection and hazard mitigation strategies.

Over time, as plants decide on which strategies to apply there is an expectation that the most effective way to develop and justify these strategies will involve probabilistic risk assessment (PRA) concepts. The PRA framework is well suited for performing a human reliability analysis (HRA). Within that framework, HRA evaluations focus on operator and plant staff actions taken in response to plant initiating events (e.g., loss of offsite power, etc.). For many external floods, advance warning of an impending external flood event provides the trigger for pre-emptive manual actions to potentially reconfigure the plant through temporary installation of flood barriers. Unlike the post-initiator actions which tend to be more narrowly focused, these pre-emptive actions are taken in a less controlled environment, may be ad hoc, and may potentially be in competition with site investment protection activities, site evacuation, etc.

The purpose of this paper is to define the challenges in defining an approach for treating external flood actions, identifying external flood timelines, identifying the manual actions/organizational environment during external flooding scenarios and proposing an integrated strategy for quantifying those actions. The proposed quantification process is rooted in management science concepts for evaluating project reliability. The overall methodology identifies flood significant performance shaping factors, and identifies three (3) factors, namely time available for flood mitigation, proper access to plant site following flood and environmental factors, as having an overarching impact on the performance shaping factors affecting each of the flood mitigation tasks.
Experimental Research on Characteristics of Passive Residual Heat Removal System for Small Modular Reactors Under the Station Blackout Accident
Authors: Lu Xiaodong , Peng Chuanxin , Zhang Yan , Bai Xuesong , Zan Yuanfeng , Zhuo Wenbin , Yan Xiao
Published: October 2016

An experimental research on performance characteristics of passive residual heat removal system (PRHRS) for the small modular reactor designed by Nuclear Power Institute of China (NPIC) under the station blackout accident was performed in the CREAS facility, which consists of the primary system, the secondary system, the passive safety injection system, the passive residual heat removal system, the overpressure protection system and the auxiliary system. The experimental results show that, after the station blackout accident, a stable two-phase natural circulation between the steam generators and the heat exchanger in the PRHRS was established with a mass flow of 0.4T/h, thus the heat from the primary system was removed to the water in the containment water tank (CWT). During this period, the core decay residual heat and the sensible heat were removed from the primary system by the PRHRS effectively. The cold water from the core makeup tanks was injected into the reactor pressure vessel for core cooling. The peaked primary pressure was 16.3MPa and less than relief valve opening pressure 16.9MPa. In addition, the average coolant temperature of the reactor core reduced below 483 K, and the reactor operated safely.
Tritium Decontamination of Contaminated System With Tritiated Heavy Water by Drying Treatment
Authors: Kadowaki Haruhiko , Matsushima Akira , Nakajima Yoshiaki
Published: October 2016

Advanced thermal reactor “FUGEN” is a heavy water-moderated boiling light water-cooled pressure tube-type reactor. Because tritium had been generated in the heavy water during the reactor operation, the heavy water system and helium system were contaminated by tritium.

The chemical form of the tritium was water molecule in FUGEN. For the drying treatment of heavy water contaminated by tritium, air-through drying and vacuum drying were applied to the system drying. The air-through drying has an advantage that can reduce the risk of tritium leakage, because the drying equipment can be constructed as closed-circuit. On the other hand, the vacuum drying can dry the whole system because dead end of piping can be aspirated. Helium system, heavy water purification system and rotary type dehumidifier were chosen for the object in this study.

It was demonstrated that both methods were effective for drying treatment of heavy water in system. Helium system, low-contamination and non inclusion, could finish the vacuum drying rapidly. However, Heavy water purification system needed long period for drying treatment. The result showed that it needed long period to dry up if the objects include the adsorbent of water such as alumina pellet, resin and silica gel. But it can be accelerated by replacement absorbed heavy water to light water from the result of drying treatment of the rotary type dehumidifier.
Testing and Reliability Modelling of Safety Software for Digital Reactor Protection System of HTR-PM
Authors: Guo Chao , Xiong Huasheng , Li Duo , Zhou Shuqiao
Published: October 2016

Nuclear safety is one of the key issues for a nuclear power plant (NPP). Digital instrumentation and control (I&C) systems have been employed gradually in the newly-built and upgraded NPPs, while the reliability of software brings great challenges to the Probability Risk Assessment (PRA) of NPPs. Software testing is regarded as one of the most important methods to guarantee the quality of safety software. The testing data can then be adopted to assess the coding quality by reliability modelling. As the variety of digital I&C systems, software modelling methods corresponding to particular I&C system, as well as a model which is suitable in all situations, are both expected.

The Reactor Protection System (RPS) in High Temperature Gas-Cooled Reactor - Pebble bed Module (HTR-PM) is the first digital RPS designed and to be operated commercially in China. As the designer, we also took part in the software testing work of this digital I&C system. In this paper, we gave a comprehensive introduction to the software testing and reliability modelling research of RPS in HTR-PM, including the objective, tools, methods, testing strategy, organizational structure, and the implementation phases.

During the testing experience of safety software of RPS in HTR-PM, we collected the software abnormal reports which could be employed for the reliability analysis to evaluate the quality of the safety software. We introduced the data mining and reliability modelling research according to the abnormal reports. Different characteristics of faults could be used for software reliability modelling, such as software version, fault severity, test stage, submission date, debugging data, and so on. In the end we introduced a software modelling method based on severity analysis of the abnormal reports.

The work we showed in this paper can contribute to improve the process of testing and reliability analysis for other digital I&C systems in NPPs.
In-Plant 3D Positioning System Using Point Cloud Data for Remote Decontamination Machine
Authors: Satoh Yoshinori , Aikawa Tetsuro , Kawaguchi Yuji , Sakamoto Naoya , Saito Masahiro , Kaneda Masayuki , Sakai Hitoshi
Published: October 2016

Toshiba Corporation, a member of International Research Institute of Nuclear Decommissioning (IRID), has contributed to decontamination works throughout Fukushima Daiichi Nuclear Power Plant from outside ground to inside of the buildings. Speedy decontamination works allow workers to access and stay inside of the contaminated buildings for many hours, and as a result, all decommission works can be accelerated. Some remote decontamination machines to decontaminate the inside of the reactor buildings from a remote safe building have been developed for workers not to be radiated by high-level radiation. Conventionally, operators have remotely controlled the decontamination machine through multiple views of some remote surveillance cameras mounted on it. Because the position data such as GPS data is not available in the buildings, it was hard for operators to detect its absolute position and orientation in the building, and it took much time to recognize targets to be decontaminated. In order to reduce positioning time and make operation works easier, we constructed 3D positioning system to automatically detect the absolute 3D position of the decontamination machine in the reactor buildings from a remote safe building. Moreover, we can also keep records of decontamination works easily by tracking 3D position of the decontamination machine.

This paper shows the overview of our approach of 3D positioning and a result of examinations in the mock-up facility that simulates a part of the inside of the reactor building at Fukushima Daiichi NPP.
Numerical Study of Fluid Flow and Heat Transfer in Plate-Fin Heat Transfer Exchanger of HTGR
Authors: Peng Wei , Chen Tao , Yang Xiaoyong , Zhao Gang , Wang Jie
Published: October 2016

This paper numerically study the flow and heat transfer characteristics of fluids within the plate-fin heat exchangers with plain and serrated fins. The calculation result is steady-state solution. Effects of the parameters such as inlet velocity and fin type on heat exchange performance are analyzed. The result indicates that under the same inlet velocity, the fluid temperature would be more uniform for the serrated fins. Whereas for the plain fins, the stratified distribution of the fluid temperature is more obvious. Being different from the plain fins, the serrated fins can destroy the boundary layer periodically and strengthen the disorder of the flow field, so its heat exchange performance is better than the plain fins’, however, the pressure drop with serrated fins increases correspondingly.
Ultrasound Reflector Recognition and Tracking Technique for Two-Phase Flow
Authors: Povolny Antonin , Kikura Hiroshige
Published: October 2016

Two-phase liquid-gas flow occurs in many safety systems of nuclear reactors as well as in reactor cores. To further improve both safety and commercial performance of nuclear reactors, it is important to improve numerical codes and deepen the understanding of two-phase flow with experiments on gas behaviour in liquids. Among several available measurement methods, ultrasound based methods are affordable and easy to use even for high pressure/temperature flows in non-transparent pipes. Ultrasound Reflector Recognition and Tracking Technique (URRTT) has been developed as a new technique. It uses an ultrasound transducer, which emits ultrasound beam into the liquid with gas bubbles. The phase interface reflects the beam and because of that, the phase interface can be recognised in the reflected signal and the distance (from the transducer) can be calculated. The core of this technique is the tracking algorithm that can separate data of different bubbles from each other and obtain their one dimensional trajectories along the measurement line. Trajectories measured simultaneously by more transducers (at different positions or from different directions) can be combined. That means trajectory of the bubble interface from one transducer can be connected to trajectory from a different transducer and by doing so, a secondary data can be obtained using the information that those trajectories belong to the same bubble. As an example, the average two dimensional velocity between two parallel measurement lines can be obtained. Another example is the measurement of the bubble size using one measurement line with two oppositely oriented transducers. Experiments have been conducted to prove the concept of URRTT. Results have been validated to data obtained by the image processing of footage taken by a high speed camera. The results obtained by URRTT can be of high value since each detected bubble is measured individually and thus, difference in the bubble behaviour based on the size, velocity or history of the bubble can be described.