bluuebunny/crossref_metadata_2025_split huggingface dataset:
A Combined Small Modular Reactor and Gas Turbine Cycle With Reheat
Authors: Stakenborghs Robert , Kramer Gregory 
Published: October 2018

A novel combined small modular reactor (SMR) and gas turbine cycle is presented. This SMR-GT cycle is modeled using fundamental thermodynamic relationships and compared to existing state-of-the-art power generation cycles. The SMR-GT cycle includes an 82 MWe SMR cycle that is combined with a 54 MWe gas turbine cycle. A heat exchanger is used to extract energy from the gas turbine exhaust to create superheated main steam and provide reheat downstream of the LP turbine. This results in a 32 MWe increase in the SMR cycle for total unit output of 136 MWe.
Comparisons of thermal efficiency, heat rate, CO2 emissions, and net generation are made between a stand-alone SMR, a typical combined cycle gas turbine (CCGT), standalone gas turbine and the combined SMR-GT cycles. Several advantages of the SMR-GT cycle are discussed.
In addition, the rapid deployment of a gas turbine allows for a power station to deliver power and earn revenue prior to completion of the more complex SMR portion of the plant. The SMR portion of the cycle also reduces the overall fuel cost volatility associated with gas turbine based power station.
The Nuclear Technology Education Consortium: UK Nuclear Education to Meet the Global Workforce Demand
Authors: Roberts John W. 
Published: October 2018

Networks that foster collaboration have been used for educational purposes for many years as they optimise resources and enhance the student experience by bringing together the very best of the teaching staff available. When nuclear education at UK universities had declined to such an extent that its sustainability was in jeopardy, networking the remaining pockets of expertise at several universities was proposed as the best solution.
The Nuclear Technology Education Consortium (NTEC), was therefore established in 2005, and brings together six preeminent nuclear universities in the UK plus the Nuclear Department of the Defence Academy to provide a Master’s level programme in nuclear science and technology. Following extensive consultation with industry an innovative delivery method of one week modular courses is used for the programme which caters for both full-time and part-time students. Full-time students can complete the programme in one year (September to September) while part-time students complete the programme in three years. As the courses are all delivered in this one-week format they can also be taken individually for continual professional development. The programme can also be completed via web-based distance learning which provides options for the 21st century student wherever they are based.
With many countries embarking on a new build programme, extending the lifetime of their current reactors and planning geological disposal facilities the nuclear workforce will need to expand over the next decade. The flexibility of the NTEC programme allows both the current nuclear workforce to upskill as well as the development of the next generation workforce, providing them with the skills, competencies and professionalism required by the nuclear industry.
Over 200 students have completed the programme on a full-time basis, with 80% either entering the nuclear industry directly, or embarking on further nuclear education programmes, demonstrating that the curriculum of the NTEC programme is matched to the requirements of industry. Nuclear education at UK universities has grown substantially in the last twelve years with NTEC at the forefront, producing graduates to meet the global workforce demand, both in terms of quantity and quality.
Analytical Study on Fragility Evaluation With Uncertainty Against Fault Displacement for Nuclear Power Plant Buildings
Authors: Ishiki Kenshiro , Tsuji Hirokazu , Kanechika Minoru , Mihara Yoshinori 
Published: October 2018

Japan Nuclear Safety Institute had recently reported the pioneering deterministic evaluation approach for nuclear power plant under seismic induced fault displacement. But the uncertainty of fault displacement based on probabilistic hazard analysis is described to be greater than that of other natural phenomena e.g. earthquake ground motions or seismic acceleration vibration in the report. Furthermore, for plant-wide risk assessment against fault displacement hazards beyond design basis displacement level, it is seriously necessary to promote a series of fundamental studies and develop the standard procedures regarding not only accident sequence analysis but also fragility analysis of buildings and structures as well as components and piping systems.
Based on the above background, the objective of this study is focusing to obtain basic fragility data for the aleatory and epistemic uncertainties of structural responses for nuclear power plant buildings against fault displacement. A number of nonlinear soil-structure finite element analyses against relatively large fault displacement are performed with the randomness of soil and building material properties, the uncertainly of contact parameters relating to friction between soil and building, and also the uncertainty of fault hazards such as fault types and geometries. Their quantitative results for fragility data are shown in this paper.
New Needs of Fracture Mechanic Analysis at Design and Operation Level: Status of French Nuclear Mechanical Codes
Authors: Faidy Claude 
Published: October 2018

Based on ASME Boilers and Pressure Vessels Code the major fracture mechanic analysis is limited to protection of class 1 components to brittle fracture.
All the Operators of future plants have to enlarge the scope of these analyses to different concepts, at design or operation stage:
- brittle and ductile analysis of hypothetical large flaw
- leak before break approach
- break exclusion concept
- incredibility of failure of high integrity components
- end of fabrication acceptable defect
- in-service inspection performance
- acceptable standards in operation
- Long Term Operation (LTO)
All these requirements needs a procedure, an analysis method with material properties and criteria.
After a short overview of each topic, the paper will present how RCC-M, RSE-M French Codes and ASME III and XI take care of all these new modern regulatory requirements.
The International Nuclear Management Academy
Authors: Roberts John W. 
Published: October 2018

The International Nuclear Management Academy (INMA) is an International Atomic Energy Agency (IAEA) framework to support the establishment and sustainability of Master’s level Nuclear Technology Management educational programmes and the development of nuclear technology management professionals.
The INMA framework describes a broad range of competencies across four Aspect Groups of External Environment, Technology, Management and Leadership, that have been identified as the basis for the successful management of nuclear projects. By following the INMA framework these competencies can be achieved by nuclear technology subject matter experts to support their career path into managerial roles or by experienced managers moving into the nuclear sector.
The IAEA in conjunction with worldwide universities with nuclear education programmes have developed an endorsement process to recognise which university Master’s programmes adhere to the INMA framework and can therefore produce graduates with the required competencies. It is also recognised though that the implementation of these competencies can only be fully achieved through on-the-job training or experiential learning. A combination of education and experience is therefore required to be recognised as a nuclear technology management professional.
To date two universities, The University of Manchester and the Moscow Engineering Physics Institute, have received INMA endorsement for their Master’s programmes in Nuclear Technology Management. The University of Manchester programme is part-time while the MEPhI programme is a two-year full-time programme. Several other universities — North West University and University of the Witwatersrand (both South Africa), Texas A&M University and the University of Tokyo having been assessed for endorsement, and many others developing nuclear technology management programmes are entering the process. The IAEA organise an INMA Annual Meeting where universities can meet to express interest in the programme, learn more about what is required for the programme and endorsement, and exchange best practices.
The International Nuclear Management Academy is therefore making significant contributions to improving nuclear technology management competencies leading to improved managerial decision making with the associated benefits to the global nuclear industry.
Research on Coupling Scheme of Monte Carlo Burnup Calculation in RMC
Authors: Li Wanlin , Wang Kan , Yu Ganglin , Li Yaodong 
Published: October 2018

Monte Carlo (MC) burnup calculation method, implemented through coupling neutron transport and point depletion solvers, is widely used in design and analysis of nuclear reactor. Burnup calculation is generally solved by dividing reactor lifetime into steps and modeling geometry into numbers of burnup areas where neutron flux and one group effective cross sections are treated as constant during each burnup step. Such constant approximation for neutron flux and effective cross section will lead to obvious error unless using fairly short step. To yield accuracy and efficiency improvement, coupling schemes have been researched in series of MC codes.
In this study, four coupling schemes, beginning of step approximation, predictor-corrector methods by correcting nuclide density and flux-cross section as well as high order predictor-corrector with sub-step method were researched and implemented in RMC. Verification and comparison were performed by adopting assembly problem from VERA international benchmark. Results illustrate that high order coupled with sub-step method is with notable accuracy compared to beginning of step approximation and traditional predictor-corrector, especially for calculation in which step length is fairly long.
Development of Evaluation Method for Thermal Stratification by Cavity Flow in a Vertical Branch Pipe With Elbow and Horizontal Section: Experimental Results
Authors: Kodama Atsushi , Komuro Yoshiteru , Hori Keiichi , Noguchi Hironori , Kondo Yoshiyuki , Tanimoto Koichi 
Published: October 2018

In nuclear plant piping, when high-temperature water penetrates from main pipe, a thermal stratification may be formed in a branch pipe with a closed end. The penetrating flow of high-temperature water is called as a cavity flow. If a thermal stratification interface is formed in an elbow of vertical-horizontal branch pipes, thermal fatigue may occur due to periodic temperature fluctuations. Because high cycle thermal fatigue may have significant influence on structure integrity of pipe, it is important to evaluate the position which is called penetration depth of a cavity flow in which a thermal stratification interface is formed. In the present evaluation guideline which was formulated by the Japan Society of Mechanical Engineers, applicable range is limited only to 50 mm diameter of vertical-horizontal branch pipe. Therefore, it needs to expand the applicable diameter range of vertical-horizontal branch pipe.
In this research, 200 mm diameter piping test is conducted to confirm characteristic phenomena, and to build evaluation method of penetration depth in large diameter piping.
This paper presents experimental results and consideration of thermal hydraulics phenomena about thermal stratification. The horizontal piping test and the vertical-horizontal branch piping test are conducted. Pipes are made of acrylic resin for visualization of water flow. The water temperature is less than 60°C. The horizontal piping test reproduces natural circulation flow by installing a heater which imitated heating from cavity flow, and a chiller which imitated heat dissipation from a blockage valve. Temperature profiles are obtained by using thermocouples and optical fibers. Moreover, water flow in piping is visualized by injected ink. From the visualization and the temperature measuring result in the horizontal piping test, flow characteristic of the natural circulation region of large diameter piping is clarified The vertical-horizontal branch piping test reproduces cavity flow penetration from main pipe. Temperature profiles are obtained at cross-section direction and axial direction. Moreover, the amount of mass transfer at the thermal stratification which is formed between a cavity flow and natural circulation is evaluated experimentally by using a lithium tracer method. The second paper[Ref.2] explains modeling of evaluation method for cavity flow penetration depth in large diameter piping.
Development of Evaluation Method for Thermal Stratification by Cavity Flow in a Vertical Branch Pipe With Elbow and Horizontal Section: Modeling of Evaluation Method
Authors: Kodama Atsushi , Komuro Yoshiteru , Hori Keiichi , Noguchi Hironori , Kondo Yoshiyuki , Tanimoto Koichi 
Published: October 2018

In nuclear plant piping, when the high-temperature water penetrates from main pipe, a thermal stratification interface may be formed in branch pipe with a closed end. The penetrating flow of high-temperature water is called a cavity flow. If a thermal stratification interface is formed in an elbow of vertical-horizontal branch pipes, thermal fatigue may occur by a periodic temperature fluctuation. Because high cycle thermal fatigue may have significant influence on structure integrity of pipe, it is important to evaluate the position which is called penetration depth of a cavity flow in which a thermal stratification interface is formed. In present evaluation guideline which was formulated by the Japan Society of Mechanical Engineers, applicable piping is limited only to 50 mm diameter of vertical-horizontal branch pipe. Therefore, it needs to expand the applicable diameter range of vertical-horizontal branch pipe.
In this research, 200 mm diameter piping test is conducted to confirm characteristic phenomena, and to build evaluation method of penetration depth in large diameter piping. This paper presents about modeling of evaluation method for cavity flow penetration depth in large diameter piping. First, the mass transport model at the interface of layer in natural circulation region is arranged with Reynolds number and Richardson number based on visualization and the temperature measuring result. Second, the mass transport model at a thermal stratification interface between a cavity flow domain and a natural circulation domain is built based on time variation of lithium tracer concentration. The evaluation model of cavity flow penetration depth applicable to large diameter piping is proposed by combining new mass transportation models and existing evaluation guideline. In conclusion, predictive accuracy of cavity flow penetration depth in 200 mm diameter piping became the same order as the evaluation result of 50 mm diameter piping by using present evaluation guideline.
Application of Monte Carlo Methods in Reactor Protection System Reliability Research
Authors: Li Duo , Hao Zhaojun , Zhou Shuqiao , Guo Chao 
Published: October 2018

Digital Reactor Protection System (RPS) is one of the most important systems in instrumentation and control systems of Nuclear Power Plants (NPP). The reliability analysis of RPS plays an important role both in theory and engineering application. Traditional reliability methods, such as fault tree analysis and Markov chain theory, have many limitations in the research of RPS reliability, since the number of system states increases exponentially with the growth of system complexity.
Aiming at the reliability analysis of complex system like RPS, the Monte Carlo method simulates the system behaviors and obtains the reliability calculations through a large number of simulations. This paper takes a preliminary research of RPS reliability based on Monte Carlo Methods, including static reliability analysis based on Monte Carlo simulation of the behavior of every equipment in the RPS, and dynamic characters of the RPS based on the simulation of RPS period tests.
DEM-CFD Study on Fluid Flow Through Radial Layered Composite Packed Beds
Authors: Guo Zehua , Sun Zhongning , Zhang Nan , Ding Ming , Bian Haozhi , Meng Zhaoming 
Published: October 2018

The radial porosity generally have a higher value at the container wall than that in the core part. Consequently, the fluid flows are mal-distributed in packed bed with significant bypass flow at the wall, which lowers the convective heat transfer performance inside the packed bed. To overcome this drawback of packed bed, we developed an effective way to construct the radial layered composite packed bed, which can easily realize placing small particle at the near wall region and large spheres in the core region. Therefore, smaller pores forms close to the container wall and larger channels presented in the core part. This could result in a much homogenous radial porosity distribution, which is benefit to restrain the bypass flow near the wall.
In present paper, the packing procedure is simulated by the discrete element method (DEM). Radial layered composite packed bed and traditional packed bed with uniform spheres are compared on radial porosity distribution. By altering the size of spheres ratios in the near wall region different radial layered packings are also generated and compared. Then, the geometries of these packed bed are imported into the computational fluid dynamics (CFD) simulation. The fluid flow inside the packed bed is investigated. It finds that the radial layered packed bed has a lower pressure drop with the ordered packing structure. And a much homogenous fluid flow distribution is obtained than the traditional, which is benefit for the heat removal inside the packed bed. This would be useful for the optimum design of packed bed in industry applications.
Development and Application of Level 3 PSA for the UK ABWR Generic Design Assessment
Authors: Grindon Elizabeth , Harman Neil , Niculae Carmen , Ang Ming Leang , Iwanami Hironobu , Hashimoto Tomoharu 
Published: October 2018

A full scope Probabilistic Safety Assessment (PSA) was provided as an integral part of the safety case for the UK ABWR Generic Design Assessment (GDA) and this included a Level 3 PSA. The main objectives of the Level 3 PSA for GDA are to provide a demonstration of the compliance, for a single unit UK ABWR, with numerical risk targets defined in the UK Safety Assessment Principles and to support the ALARP assessment. This paper includes:
• An overview of the methodology, PSA model development and illustration of some results.
• A summary of the key assumptions made during the model development for the GDA phase of the project. Compliance with the numerical risk targets has been investigated through assessments against the individual off site risk from the facility (Target 7), facility dose bands (Target 8) and off site societal risk (Target 9).
• Some conclusions of a peer review against the draft ASME/ANS standard for trial use (ASME/ANS RA-S-1.3, Feb 2016 for Level 3 Probabilistic Risk Assessment (PRA)), which was a key aspect of this study.
Transient Boiling and Cross Flow in 5×5 Rod Bundle With Rapid Heating
Authors: Takiguchi Hiroki , Furuya Masahiro , Arai Takahiro , Shirakawa Kenetsu 
Published: October 2018

Rapid thermal elevation in nuclear reactor is an important factor for nuclear safety. It is indispensable to develop a three-dimensional nuclear thermal transient analysis code and confirm its validity in order to accurately evaluate the effectiveness of the running nuclear safety measures when heating power of reactor core rapidly rises. However, the heat transfer characteristics such as reactivity feedback characteristics due to moderator density and the technical knowledge explaining the uncertainty are insufficient. In particular, the cross propagation behavior of vapor bubble (void) in cross section of fuel assembly is not grasped.
This study evaluates the cross propagation void behavior in a simulated fuel assembly at time of rapid heat generation with a thermal hydraulic test loop including a 5 × 5 rod bundle having the heat generation profile in the flow cross sectional direction. In this paper, the branching heat output condition of transient cross propagation was investigated from visualization of high speed video camera and void fraction measurement by wire mesh sensor with the inlet flow rate 0.3m/s and the inlet coolant temperature 40°C, which are based on the transient safety analysis condition. In addition, we applied the particle imaging velocimetry (PIV) technique to measure liquid-phase velocity profile of the coolant in the transient cross flow and experimentally clarified the relationship with the cross flow.
Experimental Study on Spray Pattern of Pressure-Swirl Nozzle in Reactor Containment
Authors: Bian Jiawei , Zhang Dalin , Sun Rulei , Wu Yingwei , Tian Wenxi , Su G. H. , Qiu Suizheng 
Published: October 2018

Spraying system plays an important role in the safety of PWR. To ensure homogeneous spraying of the containment, the layout of nozzles on the spray header was taken into consideration in design. In this paper, an experimental study was conducted to obtain spray characteristics data, including spray cone angle and 2-D spray flux distribution for the purpose of achieving optimal design of the spraying system.
According to the specialty of the spray field involved, a testing loop with four pressure-swirl nozzles was established for the study. Spray cone angles were obtained by photograph method. The volume flux distribution was measured by collecting the spray droplet along the cross-section diameters. Based on the experimental data, typical spray flux distributions were obtained. The flux distribution results were used to build 3-D coverage models. Then these models were used to calculate the overall spray coverage in the containment. The present work introduces the experimental study of spray behavior of a typical pressure-swirl nozzle in containment and the method to evaluate spray coverage through building 3-D spray flux distribution models. The work is expected to be helpful for the optimization design of spraying systems.
Development and Verification of a New Nuclear Data Processing System NECP-Atlas
Authors: Xu Jialong , Zu Tiejun , Cao Liangzhi , Wu Hongchun 
Published: October 2018

To process the evaluated nuclear data file (ENDF) libraries and generate the cross section data library for neutronics calculations, a new nuclear data processing system NECP-Atlas was developed by Nuclear Engineering Computational Physics Lab. of Xi'an Jiaotong University. Meanwhile, some flaws of the current widely used nuclear data processing systems were made up.
Some new methods and techniques were proposed and integrated into NECP-Atlas. NECP-Atlas could process ENDF and generate point-wise evaluated nuclear data file (PENDF) and the multigroup cross section data library in WIMS-D format. Verification of NECP-Atlas was carried out by comparing the keff values for WLUP benchmark cases and benchmark experiments in the ICSBEP handbook using cross section data libraries processed by NECP-Atlas with those by NJOY2016. The results showed that NECP-Atlas processes the ENDF correctly and generates more reliable cross section data libraries.