bluuebunny/crossref_metadata_2025_split 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 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).
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.
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.
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).
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.
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.