Accepted Manuscripts

Wang Lianjie, Yang Ping, Lu Di and Zhao Wenbo
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037669
An optimization design of CSR1000 conceptual core is proposed. Steady state performance of the proposed core is then studied with the SCWR core steady state analysis code system SNTA. These key parameters such as burnup performance, reactivity control capability, power distribution, maximum fuel cladding temperature and maximum linear power density are analyzed. The relative coolant flow rate of the second flow path which is suited with assembly power is also presented. The study shows that the refueling cycle of CSR1000 core can be extended effectively under the optimization design.
Alberto Sáez-Maderuelo, Michael McTaggart, Xiao Huang and César Maffiotte
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037668
Supercritical Water cooled Reactor (SCWR) was chosen as Generation IV reactor concept in Canada to utilize Canada's expertise and technical capabilities from past research and designs. The conceptual design of Canadian SCWR has a core outlet temperature of 650?C at 25 MPa and a peak cladding temperature as high as 800?C. Corrosion/oxidation resistance is an important factor in material selections and also coating considerations. Most of the reported SCW test data have been obtained at temperatures up to 700?C as no autoclave exists that can operate above 700°C at supercritical pressures and temperatures. Superheated steam is used as a surrogate fluid to SCW in this study to evaluate two coating materials, FeCrAlY and NiCrAl, at 800°C. The results showed that the FeCrAlY became discolored rapidly while NiCrAl still mained some metallic sheen after 600 hours. The weight change results suggest that more oxide formation took place on FeCrAlY than NiCrAl. In particular, grain boundary oxide (Al2O3) formed on FeCrAlY surface upon exposure to steam after 300 hours. Further exposure caused more intragranular Al2O3 to form, in addition to magnetite formation on the grain boundary regions. For NiCrAl samples, NiO formed after steam exposure for 300 hours. Spinel and (Cr,Al)2O3 were also found after 300 hours along with very limited amount of Al2O3. After 600 hours, Al2O3 became well developed on NiCrAl and the coverage of spinel and Cr2O3 on the surface reduced.
Guest Editorial  
Thomas Schulenberg
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037556
This special issue of the Journal of Nuclear Engineering and Radiation Science comprises selected papers from the 8th International Symposium on Supercritical Water-Cooled Reactors (ISSCWR8), held from March 13 to 15, 2017, in the Celebrity City Hotel, Chengdu, China, 17 years after the first International Symposium was held on this topic in Tokyo, Japan. Like in former years, the International Symposium attracted again more than 90 participants from nuclear industry, research centers and universities mainly from China, Canada, Europe, South Korea, Japan and Russia. Within three days, around 90 presentations were given on design, safety issues, materials, thermal-hydraulics and qualification tests of this Generation IV nuclear energy system, providing a worldwide forum for information exchange on innovative nuclear research and technologies.
TOPICS: Water, China, Thermal hydraulics, Generation IV reactors, Nuclear industry, Nuclear engineering, Radiation (Physics), Safety, Nuclear research, Design
Bsat Suzan, Huang Xiao and Penttila Sami
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037324
Concerns with greenhouse gas emissions and the uncertainty of long-term supply of fossil fuels have resulted in renewed interest in nuclear energy as an essential part of the energy mix for the future. Many countries worldwide including Canada, China and EU are currently undertaking the design of Generation IV supercritical water-cooled reactor (SCWR) with higher thermodynamic efficiency and considerable plant simplification. The identification of appropriate materials for in-core and out-of-core components to contain the supercritical water (SCW) coolant is one of the major challenges for the design of SCWR. This study is carried out to evaluate the oxidation/corrosion behaviours of bare Alloy 214 and NiCrAlY coated 214 under SCW at a temperature of 700°C/25MPa for 1000 hours. The results show that chromium and nickel based oxide forms on the bare surface after exposure in SCW for 1000 hours. A dense and adhered oxide layer, consisting of Cr2O3 with spinel (Ni(Cr, Al)2O4), was observed on NiCrAlY surface after 1000 hours in SCW.
TOPICS: Alloys, Corrosion, Design, Water, Supercritical water reactors, Uncertainty, Emissions, China, Fossil fuels, Nuclear power, oxidation, Coolants, Temperature, Nickel
Miltos Alamaniotis and Mauro Cappelli
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037203
Modernization of reactor instrumentation and control systems is mainly characterized by the transition from analog to digital systems, expressed by replacement of hardware equipment with new software driven devices. Digital systems may share intelligence capabilities where except for measuring and processing information may also make decisions. State identification systems are systems that process the measurements taken over operational variables and output the state of the reactor. This paper frames itself in the area of control systems applied to state identification of Boiling Water Reactors (BWRs). It presents a methodology that utilizes machine learning tools, and more specifically, a set of relevance vector machines (RVMs) in order to process the incoming signals and identify the state of the BWR in real time. The proposed methodology is comprised of two stages: in the first stage each RVM identifies the state of the BWR, while the second stage collects the RVM outputs and decides about the real state of the reactor. The proposed methodology is tested on a set of real world BWR data taken from the experimental FIX-II facility for recognizing various BWR LOCA as well as normal states. Results exhibit the efficiency of the methodology in correctly identifying the correct state of the BWR while promoting real time identification by providing fast responses. However, a strong dependence of identification performance on the form of kernel functions is also concluded.
TOPICS: Boiling water reactors, Regression models, Machinery, Control systems, Hardware, Instrumentation, Signals, Computer software
Xu Cheng and Xiaojing Liu
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037117
Supercritical fluids (SCFs) become more and more important in various engineering applications. In nuclear power systems, SCFs are considered as coolant of the reactor core such as the supercritical water cooled reactor (SCWR), superconducting magnets and blankets in the fusion reactors or as fluid in the energy conversion systems of the next generation nuclear reactors. Accurate determination of heat transfer and the temperature of the structural material (e.g. fuel rod cladding) is of crucial importance for the system design. Thus, extensive studies on heat transfer to SCFs have been carried out in the past five decades and are still ongoing worldwide. However, no breakthrough is recognized or expected in the near future. In this paper, the status, main challenges and future R&D needs are briefly reviewed. Three aspects are taken into consideration, i.e. experimental studies, numerical analysis and model development for the prediction of heat transfer coefficient. Several key challenges and also the important subjects of the future R&D needs are identified. They are (a) data base for turbulence quantities, (b) multi solution of wall temperature, (c) extensive RANS method and (d) new prediction method for heat transfer coefficient (HTC).
TOPICS: Supercritical fluids, Heat transfer, Heat transfer coefficients, Supercritical water reactors, Fuel rods, Temperature, Fluids, Fusion reactors, Turbulence, Superconducting magnets, Coolants, Energy conversion, Cladding systems (Building), Design, Engineering systems and industry applications, Numerical analysis, Databases, Nuclear reactors, Model development, Nuclear power, Reynolds-averaged Navier–Stokes equations, Wall temperature, Water
Chuanqi Zhao, Kunpeng Wang, Liangzhi Cao and Youqi Zheng
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037119
Burnable poison (BP) is used to control excess reactivity in Supercritical Water Cooled Reactor (SCWR). It helps reducing the number of control rods. Overall BP designs, the design in which rare-earth oxide mixes with fuel is widely used in SCWR. BP has influence on fuel assembly neutronics performance. After comparing four kinds of rare-earth oxide, Er2O3 is chosen as BP for annular fuel assembly. The effect of different BP loading patterns on assembly power distribution is analyzed. The safety of annular fuel assembly is estimated with different BP containments. Core performance with and without BP is compares. The results have shown that the core radial power peaking factor decreases by introducing BP. It is also shown that the core axial power peaking factor increases and the power peak moves towards the top of the core. The reason of this effect was studied. Two optimizations were given based on this study: decreasing the temperature of lower plenum, and increasing the gradients of axial enrichments. By applying these optimizations, core axial power peaking factor and maximum cladding surface temperature decreased.
TOPICS: Fuels, Design, Water, Manufacturing, Supercritical water reactors, Temperature, Rods, Cladding systems (Building), Safety
Technology Review  
Lembit Sihver and Nakahiro Yasuda
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037116
In this paper the causes and the radiological consequences of the Chernobyl and Fukushima Daiichi nuclear accidents is discussed. The need for better severe accident management (SAM), and severe accident management guidelines (SAMGs), are essential in order to increase the safety of the existing and future operating Nuclear Power Plants (NPPs). In addition to that, improved radiation release barriers, including low corrosive fuel and cladding, should be developed. Stress tests should on a regular basis be performed to assess whether the NPPs can withstand the effects of natural disasters and man-made failures and actions. The differences in safety preparations at the Chernobyl and Fukushima Daiichi will therefore be presented, as well as recommendations concerning improvements of safety culture, nuclear fuel performance, decontamination and disaster planning. The need for a high-level national emergency response system in case of nuclear accidents will also be discussed. The emergency response system should include fast alarms, communication between nuclear power plants, nuclear power authorities and the public people, as well as well prepared and well established evacuation plans and evacuation zones. The experiences of disaster planning and the development of a new improved emergency response system in Japan will be presented together with training and education program, in order to ensure that professional rescue workers, including medical staff, fire fighters, police, etc., as well as the normal populations including patients, have sufficient knowledge about ionizing radiation and are informed about the meaning of radiation risks and safety.
TOPICS: Accidents, Fukushima nuclear disaster, Japan, 2011, Chernobyl Nuclear Accident, Chornobyl, Ukraine, 1986, Safety, Nuclear power stations, Emergency response, Emergency management, Accident management, Radiation (Physics), Evacuations, Cladding systems (Building), Fire, Ionizing radiation, Fuels, Stress, Natural disasters, Failure, Nuclear power, Education, Nuclear fuels, Biomedicine, Decontamination
Alex Matev
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036986
Many researchers are investigating the potential of lead-bismuth cooled fast reactors for producing electricity, as well as for the safe transmutation of minor actinides and the nuclear incineration of long-lived fission products. The paper presents the results from simulating with the RELAP5-3D code of natural circulation in a generic design of a pool-type nuclear reactor with lead-bismuth eutectic alloy (LBEA) as a primary, and water/steam as a secondary coolant. The simulation results provide valuable insights in the evolution of key reactor safety-relevant phenomena and support also the qualified use of system analysis codes as RELAP5-3D for the simulation of transients in pool-type reactor systems.
TOPICS: Simulation, Coolants, Eutectic alloys, Fast neutron reactors, Nuclear reactors, Simulation results, Steam, Water, Transients (Dynamics), Design, Nuclear fission, Systems analysis, Safety
Yan Wang
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4032312
In modeling and simulation, model-form uncertainty arises from the lack of knowledge and simplification during modeling process and numerical treatment for ease of computation. Traditional uncertainty quantification approaches are based on assumptions of stochasticity in real, reciprocal, or functional spaces to make them computationally tractable. This makes the prediction of important quantities of interest such as rare events difficult. In this paper, a new approach to capture model-form uncertainty is proposed. It is based on fractional calculus, and its flexibility allows us to model a family of non-Gaussian processes, which provides a more generic description of the physical world. A generalized fractional Fokker-Planck equation (fFPE) is used to describe the drift-diffusion processes under long-range correlations and memory effects. A new model calibration approach based on the maximum mutual information is proposed to reduce model-form uncertainty, where an optimization procedure is taken.

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