Nuclear Energy Technology Research Articles

Roadmap for the application of ion beam technologies to the challenges of nuclear energy technologies

The use of ion beams to estimate materials performance for nuclear energy applications is advancing at a rapid rate with ion irradiations having been shown to produce radiation effects data of direct relevance for understanding neutron-induced displacement damage. Ion beam irradiation shows considerable promise for assisting in down selecting candidate materials for use in nuclear energy systems. Furthermore, ion beams allow rapid achievement of materials damage levels not accessible by neutron irradiation in test reactors due to cost and time constraints. Compared to test reactor irradiations, the relatively open configuration of ion beam irradiations allows capture of data on the effects of irradiation under very specific conditions of temperature, radiation dose, and radiation dose rate that are difficult or impossible to achieve otherwise.There are still multiple challenges to the deployment of ion beam data in support of reactor materials qualification arising from the lack of a detailed mechanistic understanding of potential difference between ion-induced and neutron induced materials damage. Recently, the Nuclear Science User Facilities presented a roadmap for the development and enhancement of current U.S. ion beam irradiation technologies within university and national laboratory settings, and especially for the deployment of new highly controlled in situ interrogation of materials during irradiation to provide dynamic and mechanistic data for model development. In this presentation, the status and capabilities of relevant U.S. ion beam facilities will be summarized and recommended “best practices” for performing ion irradiations will be described. The potential role of ion beam irradiations to assist the development and deployment of reactor materials will be outlined. Key objectives include developing methods for rapid and cost-effective materials selection and development, characterizing fundamental material response under irradiation, and developing a robust mechanistic understanding of microstructure evolution under irradiation (including development and validation of reliable predictive models for microstructure evolution).

Stability analysis on flow parameters in coolant temperature control system of lead-cooled fast reactor

Lead-cooled Fast Reactor (LFR), one of the promising nuclear energy technologies for future advanced system, is well foreseen in meeting the goals of Generation IV fission reactor. In LFR, Coolant Temperature Control System (CTCS) is one of the critical parts, and the influence of coolant flow parameters on the stability analysis is worth investigating and studying. In this paper, we use the CLEAR-S, an integrated multifunctional non-nuclear test facility, as reference and the stability analysis of CTCS, concerned to coolant flow rate has been completed based on simplified coolant thermal circulation model. Location of eigenvalues and calculation results of real stability radius, both qualitatively and quantitatively reflect the influence of coolant flow rate on stability respectively. The analysis results show that the feed-water flow rate in steady state is a critical parameter for CLEAR-S, because larger feed-water flow rate can improve stability apparently. However, LBE coolant flow rate has little impact on system stability. The results of this work will provide an effective reference for design and operation of LFR.

EWOLUCJA PRAWNEGO STATUSU ORGANÓW NADZORUJĄCYCH BEZPIECZEŃSTWO WYKORZYSTYWANIA ENERGII JĄDROWEJ W POLSCE

Do połowy XX w. z braku wyspecjalizowanych organów państwowych i przepisów prawa użytkownicy źródeł promieniowania bazowali na własnej wiedzy praktycznej i teoretycznej w zakresie wpływu promieniowania na organizm ludzki i możliwości ochrony przed jego negatywnymi skutkami. W latach pięćdziesiątych XX w. powstały pierwsze wprost poświęcone ochronie przed promieniowaniem przepisy rangi państwowej, a w dekadę później wyznaczono organ odpowiedzialny za regulację działalności z wykorzystaniem źródeł promieniowania jonizującego. Przez kolejne dekady organ ten zmieniał nazwę i usytuowanie w strukturze instytucji państwowych, łącząc kompetencje nadzorcze w zakresie bezpieczeństwa ze swoim pierwotnym i głównym zadaniem w postaci stymulacji i koordynacji rozwoju wykorzystania energii jądrowej i technik jądrowych w Polsce. Stopniowo pod wpływem międzynarodowych trendów i prawodawstwa wykształcał się organ odpowiedzialny wyłącznie za bezpieczeństwo jądrowe, ochronę radiologiczną, ochronę fizyczną obiektów i materiałów jądrowych oraz zabezpieczenia materiałów jądrowych. Aktualnie organem tym jest Prezes Państwowej Agencji Atomistyki będący naczelnym organem dozoru jądrowego i jednocześnie centralnym organem administracji rządowej.

PANGU code for pebble-bed HTGR reactor physics and fuel cycle simulations

PANGU is a modern computer code for pebble-bed High Temperature Gas-cooled Reactor (HTGR) physics analyses and fuel cycle simulations, developed by Institute of Nuclear and New Energy Technology (INET), Tsinghua University. This paper gives a comprehensive review on PANGU code’s features including its calculation scheme, methodologies, models and capabilities. Numerical tests are presented for typical pebble bed reactor cases, to verify PANGU code’s applicability. It demonstrates that PANGU code is promising for physical designs of traditional and new-conceptual pebble-bed HTGRs.

Positive or negative? Public perceptions of nuclear energy in South Korea: Evidence from Big Data

Abstract After several significant nuclear accidents, public attitudes toward nuclear energy technologies and facilities are considered to be one of the essential factors in the national energy and electricity policy-making process of several nations that employ nuclear energy as their key energy resource. However, it is difficult to explore and capture such an attitude, because the majority of prior studies analyzed public attitudes with a limited number of respondents and fragmentary opinion polls. In order to supplement this point, this study suggests a big data analyzing method with K-LIWC (Korean-Linguistic Inquiry and Word Count), sentiment and query analysis methods, and investigates public attitudes, positive and negative emotional statements about nuclear energy with the collected data sets of well-known social media and network services in Korea over time. Results show that several events and accidents related to nuclear energy have consistent or temporary effects on the attitude and ratios of the statements, depending on the kind of events and accidents. The presented methodology and the use of big data in relation to the energy industry is suggested as it can be helpful in addressing and exploring public attitudes. Based on the results, implications, limitations, and future research areas are presented.

The MOCP code: method of characteristics code for pebble-bed HTR

The method of characteristics which is an effective method for neutron transport calculation with high flexibility in complex geometry has been applied to high temperature gas-cooled pebble-bed reactor geometry. The MOCP (the Method of Characteristics for Pebble-bed high temperature gas-cooled reactor) code was developed in Institute of Nuclear and New Energy Technology, Tsinghua University to achieve this goal. The first stone standing ahead is the geometry modeling for pebble-bed district. In this work, a solid modeling method called constructive solid geometry (CSG) is introduced to model the pebble-bed district. A universal ray tracing technique is developed to arrange tracks in 3-D cylinder geometry. Furthermore, tree search method is implemented into MOCP to improve the ray tracing efficiency. Automatic meshing is implemented to improve the geometry modeling efficiency. A redundant tree was constructed to mesh the interspace among pebbles. The geometry modeling capability of MOCP is verified by fast converging single group benchmarks.

Intensified carbon capture using adsorption: Heat transfer challenges and potential solutions

Up to 25% of the total European Union (EU) CO2 emissions that contribute to global warming are from industry, and while improved energy efficiency and process integration continues to play a role in minimizing these, it is carbon capture (CC) that in future will contribute most to mitigation, until nuclear energy and renewable technologies take over from fossil fuels. One of several CC methods is to use gas-solid adsorption, where the CO2 is adsorbed onto a solid. As with the more common absorption process, regeneration is required, and typically a single bed is employed to adsorb CO2 while regeneration and removal of the CO2 takes place in the second bed – carried out by pressure swing adsorption (PSA) or temperature swing adsorption (TSA). Collaborating in an EPSRC-funded project with Heriot-Watt University, where hydrotalcite-based adsorbents are being synthesised, and Sheffield University, where modelling is being undertaken, Newcastle University is examining the intensification of CC using a TSA-based process involving swirling or toroidal fluidized beds. As well as improving adsorption, it is believed that recovered waste heat could be used for desorption using a similar intensified technology. This paper discusses the potential sources of CO2 that are being addressed, and how they will be integrated with the capture and desorption processes where fluidization will be used for the adsorption process. It also describes preliminary work on fluidization of the particles using additive-manufactured miniaturized fluid beds.

Thermal-hydraulics analysis for VVR-KN fuel lead test using PLTEMP code

VVR-KN is one of the low-enriched fuel types to be considered for a new research reactor (RR) of a Centre for Nuclear Energy Science and Technology (CNEST) of Vietnam. This fuel type was qualified by a lead test carried out with three fuel assemblies (FAs) in 6-MWt WWR-K research reactor at the Institute of Nuclear Physics, Kazakhstan. VVR-KN fuel was then used for conversion of the WWR-K reactor core from highly-enriched to low-enriched uranium fuel and the reactor was successfully commissioned in September 2016. PLTEMP is a thermal-hydraulic code with plate and coaxial tube models that seems to be suitable for VVR-KN fuel type. Before using PLTEMP code for thermal-hydraulics analysis of the new RR, a calculation for code validation was performed based on the data of the VVR-KN fuel lead test. First, MCNP5 code was used to calculate the power distribution of WWR-K reactor core with lead test fuel assemblies (LTAs) at the core center. Then, thermal-hydraulics parameters of the LTAs were obtained by using PLTEMP code together with calculated data of the power distribution and the lead test conditions. A comparison between the analytic results and the lead test data was made to confirm the suitability of PLTEMP code for thermal-hydraulics analysis of VVR-KN fuel under forced convection and downward flow conditions.

Measurement on effective thermal diffusivity and conductivity of pebble bed under vacuum condition in High Temperature Gas-cooled Reactor

A full-radius-scale heat test facility has been developed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China to measure the effective thermal diffusivity and conductivity of pebble bed in High Temperature Gas-cooled Reactor (HTGR). In present test, a separate-effect transient experiment under vacuum condition is conducted successfully up to 1200 °C. Through inverse method and experimental temperature data, this paper presents the process of determining these thermal parameters and compares the results with previous researches, which reveals that the results are comparable in low-temperature region and valid in high-temperature region. In addition, it also studies the experimental errors and the results' uncertainties in this test. The correct knowledge on effective thermal diffusivity and conductivity of pebble bed inside the reactor core is significant to ensure the valid design and inherent safety, which will give a better balance between safety and economic competitiveness.

Preliminary phenomena identification and ranking tables on the subject of the High Temperature Gas-cooled Reactor-Pebble Bed Module thermal fluids and accident analysis

The High Temperature Gas-cooled Reactor-Pebble Bed Module (HTR-PM) is a demonstration power plant designed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University of China. In conformance with China’s nuclear safety requirements related to the development, assessment, application and quality assurance of computer software used in the safety analysis of nuclear power plants (NPPs), a phenomena identification and ranking table (PIRT) process was conducted on the subject of the HTR-PM thermal fluids and accident analysis by an expert panel sponsored by INET, on the basis of the latest HTR-PM design scheme. The PIRT objective is to identify safety-relevant phenomena that may occur during normal operation and under accident conditions, and to assess research activities pertaining to code development and assessment, such as the code validation in INET.In the PIRT process, safety-related phenomena were identified by the expert panel for the following HTR-PM scenarios: normal operation (NO), pressurized loss of forced cooling (PLOFC), depressurized loss of forced cooling (DLOFC), air ingress following DLOFC, steam/water ingress and anticipated transients without scram (ATWS). Subsequently, the relative importance of each phenomenon was evaluated according to its effect or influence on the figures of merit established by the expert panel, such as the fuel failure fraction and the maximum fuel temperature. In addition, each identified phenomenon was assigned a knowledge level rank by the expert panel, indicating the prediction capability for this phenomenon based on existing analytical tools and test data. What is more, all of the PIRT elements were well documented by a special report, including PIRT issues, objectives, descriptions of the HTR-PM plant and accident scenarios, evaluation criteria, current knowledge base, ranking scales of relative importance and knowledge level, a set of PIRTs, and expert discussions.

Dynamic Analysis for the Rotor Drop Process and Its Application to a Horizontal Rotor-Active Magnetic Bearing System in Helium Gas

The high-temperature gas-cooled reactor pebble-bed modular (HTR-PM) has been proposed by the Institute of Nuclear and New Energy Technology of Tsinghua University, in which the active magnetic bearings (AMBs) are equipped to support the high-speed rotor in the helium circulator system. In the case of AMB failures, emergencies, or overload conditions, the auxiliary bearing is applied as the backup protector to provide temporary mechanical support and displacement constraint for the dropping rotor. A detailed dynamic model is established to reveal the behavior of the dropping rotor. This model is able to describe the rotor displacement and inclination around each axis. The left and right rotor orbits are revealed. Dropping experiments are also carried out to reveal the actual behavior of the dropping rotor in helium. The predicted and experimental results will benefit further study, design, and application of the auxiliary bearing in HTR-PM helium circulator.

International Conference on Nuclear Technologies and Sciences (ICoNETS 2017)

PREFACEICoNETS, an acronym for International Conference on Nuclear Energy Technologies and Sciences, has been launched in line with the National Nuclear Energy Agency of Indonesia (BATAN) objectives to give researchers and professionals from universities, research institutions and nuclear related industries the opportunities to present their latest studies, excellent ideas and thoughts related to the development of nuclear energy technologies and it’s social culture aspects. It also provides exciting opportunities to all possible stakeholders to amplify their active involvements in future nuclear power plant constructions in Indonesia. ICoNETS2017 is the second in a series of conference on nuclear technologies and sciences. It follows the successful of ICoNETS2015 in Denpasar. ICoNETS2017 in Makassar, once again aims at covering state of the art research and development in all aspects related to nuclear technologies and sciences.ICoNETS2017 in conjunction with the Forth-International Symposium on Smart Material and Mechatronics (ISSMM2017) was held in Makassar and organized by Deputy of Nuclear Energy Technology of National Nuclear Energy Agency (BATAN) and Engineering Faculty of Hasanuddin University. It was also co-sponsored by the Association of Indonesian Nuclear Society (HIMNI), the Indonesian Researcher Union BATAN Branch (HIMPENINDO-BATAN), and the local government of the South Sulawesi Province. The theme of ICoNETS2017 is Nuclear Energy for a Realistic Approach of the National Energy Security and Long Term Sustainability.

Текущие тенденции на мировом рынке ядерной энергетики

The review of the nuclear energy technologies market, namely the main processes of the initial and final stages of the nuclear fuel cycle (NTC) was shown. The authors reveal key players in the markets of natural uranium mining, conversion, enrichment, fabrication of nuclear fuel, direct disposal, and reprocessing as well as determine their market shares. The article shows the fundamental factors influencing the development trends of the global nuclear power market such as: units’ commissioning in China, India, the Republic of Korea and other countries, the restart of the Japanese nuclear power plants, growth of uranium supplies long-term contracting planned for the period up to 2025, volatility of world prices of the NFC initial and final stages, political, economic and environmental reasons for the nuclear power generation choice. The article presents the results of analyses of Russian and world prices on the NFC initial and final stages main processes’ allowing to draw a conclusion about the current competitiveness of Russian nuclear energy technologies

Calculation of neutronic characteristics in different reflector materials with a 15-MWt reactor core using VVR-KN fuel type

VVR-KN is one of the low enriched fuel types intended for a research reactor of a new Centre for Nuclear Energy Science and Technology (CNEST) of Viet Nam. As a part of design orientation for the new research reactor, the calculations of neutronic characteristics in a reactor core reflector using different materials were carried out. The investigated core configuration is a 15-MWt power loaded with VVR-KN fuel assemblies and surrounded by a reflector using beryllium, heavy water or graphite respectively. MCNP5 code together with up-to-date nuclear data libraries were used for these calculations. This paper presents the calculation results of neutron energy spectrum, neutron spatial distribution in the reflector using the above-mentioned materials. Besides, neutronic characteristics calculated for silicon doping irradiation holes in the reflector are also presented and the utilization capabilities of different reflector materials are discussed.

Endurance test of full-scale mock-up helium circulator for HTR-PM

The High-temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM) will be built as the world’s first pebble-bed modular HTGR commercial demonstration plant through the Chinese National Science and Technology program. HTR-PM is expected to generate electricity during 2018 in Shidao Bay, China. Helium circulator is a critical component required for HTR-PM to circulate reactor coolant through the reactor core and steam generators. Since there is no engineering experience with this circulator, a full-scale mock-up was manufactured for tests. The whole parameters of the mock-up were as same as the actual plant circulator for HTR-PM. The mock-up design is a vertical layout with its rotor supported by active magnetic bearings. The whole mock-up including motor and cooler are sealed in a pressure vessel. Since the mock-up is a large mechanical-electrical system and the design includes many new systems, extensive tests are needed to ensure its safety and reliability. An Engineering Test Facility for Helium Circulator (ETF-HC) was then built in Institute of Nuclear and New Energy Technology (INET). Firstly, factory nitrogen tests were conducted by the manufacturer and the customer. After the successful factory tests, the mock-up was delivered to INET to be tested on ETF-HC. One and a half years were spent for the helium endurance tests, including time spent on resolving several helium sealing problems. The mock-up run continuously for 50 h at the rated conditions. Fifty service cycle tests were also conducted. More than 500 operating hours were accumulated during the endurance tests. The final performance shows that the mock-up of helium circulator meets the design requirements and is reliable.

APPLIED INFORMATICS MODEL ОF KINETICS ACCOUNTING OF CYBERNETIC THREATS IN NPP PHYSICAL PROTECTION SYSTEM

Context. The topical approaches to preventive estimations and mathematical modeling of cyber attacks processes and supply of external technological informational threats were considered, which may be directed on the physical security system of modern nuclear power plant units. Objective – is an adaptation of known mathematical models of cyber attacks kinetics for use in the system of information security and physical protection of nuclear power plants to modern and foreseeable conditions. Method. The method of model dependencies adaptation was proposed, which reflects the kinetics of cyberattacks to the modern operating conditions of protected nuclear facilities was proposed. To improve the model dependencies it was proposed to modify the classic Ferhulst function and, in mathematical modeling of cyber attacks on nuclear power facilities, to use the appropriate transformation of the logistic curve, which has distinctive feature of a well-defined maximum. Results. It was predicted that the model dependence of flow external to NPP cyber threats can be characterized by the alternation of maxima and minima, which indicates the possibility of an oscillatory character of the developing process. Also was put forward and substantiated the hypothesis that the system of physical and, in particular, cybernetic, NPP protection can be characterized by a time lag that in view of possible oscillatory character of information threats admissions process creates the preconditions for the development of oscillatory instability. The results obtained in the research are of applied nature, they can be used to carry out large-scale experiments and forecasting parameters of possible cyber threats aimed at the NPP for the purpose of pre-emption and preventive maintenance of information safety of modern nuclear energy technologies. Conclusions. Recommended for use model dependences were first adapted to the conditions of operating conditions of nuclear powerfacilities, take into account the factors to counter threats and circumstances associated with a specific time delay. It was shown for the first time that there is an opportunity both extreme and oscillatory dependence nature of the cyber attacks intensity, which are focused on software and hardware system of NPP information safety. The new approach allows to assume the reality of the unstable oscillation process in the “attack – defense” system and provides an opportunity of preventive assessment of the stability stocks. The practical significance lies in the fact that the results allow to realistically assess the time of onset of expected cyber attacks current dependence changes and to use this knowledge for organizational and technical preparation of remedies and early anticipation of IT-threats.

Dynamics and thermal analysis of the auxiliary bearing for the helium circulator of HTR-PM

The high-temperature gas-cooled reactor-pebble bed module (HTR-PM) has been proposed by the Institute of Nuclear and New Energy Technology of Tsinghua University. The helium circulator is equipped in the primary loop of HTR-PM to drive high-temperature helium for energy exchange. The active magnetic bearing (AMB), replaced by the ordinary mechanical bearing, is chosen as the perfect supporting scheme for the helium circulator of HTR-PM. Additionally, the auxiliary bearing is applied in case of AMB failures and has become indispensable to provide mechanical supports and displacement constraints for the rotor. When the rotor drops down, the auxiliary bearing needs to sustain fierce impacts and corresponding thermal growth. While due to the special environment in the primary loop of HTR-PM, oil or grease lubrication can’t be applied in the auxiliary bearing. A new design of unlubricated auxiliary bearing is proposed in this paper. This paper mainly deals with the following two aspects to reveal the rotor drop: dynamic analysis including fatigue life prediction, and corresponding thermal growth. Moreover, dropping experiments are carried out in the HTR-PM helium circulator system. The experimental results can verify the auxiliary bearing capacity to endure the dynamic impact of the dropping rotor, and provide suggestions for the auxiliary bearing type selection. The results in this paper will provide theoretical suggestions for further engineering design and application of the auxiliary bearing in the helium circulator of HTR-PM.

Development of a neutronics analysis code for pebble-bed HTRs

A new code named HINT (High Temperature Reactor Neutronics Tool) is under development in Institute of Nuclear and New Energy Technology (INET). This paper introduces the HINT code’s computational framework and methodologies in lattice physics and core physics computations. Emphases are given to the double heterogeneity treatment and the leakage feedback in broad-group cross-section generation, as these two issues are specific challenges for high temperature reactors (HTR) compared with light water reactors (LWR). Numerical results are presented to verify the HINT code’s capability in HTR fuel pebble and full-core calculations. It demonstrates that the HINT code is promising for pebble-bed HTR neutronics analysis.

PREFACE for the special issue on “progress in novel nuclear energy technologies”

This is an exclusively prepared special issue containing papers selected from two well-established events, namely, International Conference on Emerging Nuclear Energy Systems (ICENES) and International Conference on Nuclear and Renewable Energy Resources (NURER) and some invited papers in order to make a unique source to reflect the progress in the area of novel nuclear energy technologies. ICENES, with a long scientific tradition, started in 1978 as an autonomous, self-organized event by a group of independent nuclear scientists. The goal was to find unconventional new routes for improving the use of the atomic energy beyond the commercial grounds. It was evident that the field of “strong interactions” still has a broad way for further developments to harness the most powerful source of energy of our universe, in a safe way and with better use of natural resources. The Nuclear Field is plenty of opportunities, and no free from risks, and it is the spirit of human vocation to explore those domains where new findings can change some essential aspects of our world, as is the case of Energy. Nuclear energy is considered the ultimate energy to secure the future of the mankind. Hence, considerable research activities and international collaboration are continuing on innovative nuclear energy systems, such as fusion energy, fusion-fission hybrids, GEN-IV reactors, space nuclear reactors, and power systems and accelerator-driven systems. Asian countries are considerably progressing in the construction of new nuclear reactor as well as in the research and the development of innovative nuclear energy technologies. A sort of new international merging is needed to recover capability and potential in this field of Physics and Technology, which should play an important role in coping with the energy problem and its environmental effects. NURER is a relatively younger one, which was initiated in 2010 as an independent event to look inside the long-term future of energy problems, without forgetting to bridge the gap between the present time and the future, parallel to the ICENES series in alternating years. Both are unique conferences in their nature, considering that nuclear energy, and other newly emerging alternative and renewable energy technologies as complementing and not competing directions with each other. On that way, ICENES and NURER conferences represent two alternating fora, where present and future technologies and systems are discussed under the same umbrella and establish a bridge with closer communication and mutual understanding between nuclear and renewable energy communities. The mankind needs both of them. In that sense, both conferences cover a wide range of topics related to nuclear power production; nuclear hydrogen production; hydrogen energy, energy efficiency, and management; solar energy; wind energy; hydrogen production and storage; renewable energy; fuel cells; bio-energy, etc. In closing, this particular special issue aims to make a unique collection on “progress in novel nuclear energy technologies” from the papers selected from the above listed conferences along with some invited papers which have peer-reviewed, improved, enhanced, and accepted for publication in IJER. The guest editor would like to take this opportunity to warmly thank the editor-in-chief, Prof Dr Ibrahim Dincer; Journal Manager, Peter Creaton; and the authors of the special issue papers to make this special issue a fruitful outcome.

Sensitivity analysis of the parameters of the hydraulic cylinder in pressure transients stepping motion for the control rod hydraulic drive system based on Model II

ABSTRACTHydraulic control rod drive technology (HCRDT) is a newly invented patent and Institute of Nuclear and New Energy Technology Tsinghua University owns HCRDT's independent intellectual property rights. The hydraulic cylinder is the key part of this technology, so the performance of the hydraulic cylinder directly affects the function of the integrated valve and the HCRDT. First, the theory Model II which considers the change of inlet velocity has been obtained and the pressure transient has been analyzed. Second, based on the abnormal conditions occurring in the operation of the mechanism, the working performance is obtained and the results are verified by the experiment. Third, the sensitivity analysis of the parameters of hydraulic cylinder was carried out. The results show that the calculated values are in good agreement with the experimental values. The Model II has better numerical stability than the Model I. The influence of various parameters on the operation characteristic of hydraulic cylinder is obtained. The model can be used to analyze and optimize the performance of the HCRDT.