bluuebunny/crossref_metadata_2025 huggingface dataset:
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.
CFD Analysis of Supercritical-Water Flow and Heat Transfer in Vertical Bare Tube
Authors: Zvorykina Anastasiia , Khmil Dmytro , Fialko Nataliia , Pioro Igor , Stryzheus Svitlana
Published: October 2018

In this paper CFD analyses of mixed convection in bare tubes cooled with water at supercritical pressure is presented. The study was carried out using the FLUENT code for upward flow in vertical tubes with a heated length of 4 m and an inside diameter of 10 mm at relatively low water mass flux (G ≈500 kg/m2s) and heat fluxes q (from 239 to 310 kW/m2). Various models of turbulence have been tested. The results of the studies demonstrated a reasonable good agreement between CFD predictions and experimental data on the heat transfer coefficient and internal-wall temperature with use the SST turbulence model. Comparison of the CFD simulation data, which correspond to the presence or absence of the buoyancy forces, was performed. The regularities of the influence of these forces on the damping of turbulent transport, the deformation of the radial profiles of velocity and temperature along the channel length, the reduction of the heat transfer coefficients, etc. were studied. The features of the motion of the pseudo-phase transition within various conditions are presented.
Noise Reduction Treatment and Analysis of Accumulated Radon Concentration in Uranium-Like Rock Based on Wavelet Theory
Authors: Jiang Fuliang , Yang Wenchao , Li Ming , Li Xiangyang , Hong Changshou
Published: October 2018

Based on the principle of similarity, uranium ore samples were prepared from raw materials such as uranium tailings, quartz sand and refined iron powder, and then the samples were treated with different packages for measuring the sample accumulated radon concentration. In the actual measurement process, due to the characteristics of radionuclide decay, instrument reasons and human factors, the data will be a certain deviation. Therefore, the method of wavelet analysis is used to denoise the accumulated radon concentration and obtain radon exhalation rate. The results of the study show: the correlation coefficient of cumulative radon concentration fitted by wavelet denoising is improved greatly, and all of them are above 0.99, the recalculated radon exhalation rate of the single side of the sample is decreased by 0.06Bq · m−2 · s−1, and double-edged is decreased by 0.02Bq · m−2 · s−1. The experiment proved that wavelet theory can be used to correct calculated value of radon exhalation rate of uranium-like rock. At the same time, it provides a new method for further study of uranium mine radiation protection parameters.
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.
Experimental Study of Supercritical CO2 Critical Flow Through Short Tubes
Authors: Zhou Yuan , Fan Xing , Wang Yangle , Chen Jingtan , Huang Yanping , Wang Junfeng
Published: October 2018

The accidental leakage is one of the main risks in supercritical CO2 power conversion system. Complex leakage behavior including critical flow and multiphase jet flow occurs at the leak, which need to be evaluated carefully for safety assessment before deployment of supercritical CO2 power conversion system. A new experimental facility was built to study supercritical CO2 leakage flow. The critical flow and jet flow behavior of supercritical CO2 through short tubes were experimentally investigated. Six sharp-edged short tubes with length between 1.0 and 15.0 mm and diameter of 1.0mm were tested. The upstream temperature was examined ranging from 35 to 100°C. Effects of upstream pressure were tested ranging from 8.1 to 11.0 MPa. Critical mass flow rate of S-CO2 decreases with the increase of upstream stagnation temperature and L/D, and increases with upstream stagnation pressure. An empirical correlation with geometric and thermal parameters is proposed based on a large amount of experimental data. The experimental data of critical flow and empirical correlation could be used for safety assessment and theoretical model validation.
Modelling and Simulation Activities in Support of the UK Nuclear R&D Programme on Digital Reactor Design
Authors: Lindley B. , Allen D. , Lillington J. , Smethurst A. , Smith P. , Bowman D. , Dwyer L. , Lai K. , Levers A. , Vikhorev K. , Litskevich D. , Merk B. , Patelli E. , Patterson E. , Bankhead M. , Peakman A. , de Haas S. , Jackson C. , Draup J. , Galenne E. , Parks G. , Shwageraus E. , Eaton M. D.
Published: October 2018

The UK Department of Business, Energy and Industrial Strategy (BEIS) recently launched an R&D programme in Digital Reactor Design, incorporating the development of a Nuclear Virtual Engineering Capability with an integrated Modelling and Simulation programme. A key challenge of nuclear reactor design and analysis is the system complexity, which arises from a wide range of multi-physics phenomena being important across multiple length scales. This project constitutes the first step towards developing an integrated nuclear digital environment (INDE) linking together models across physical domains and incorporating real world data across all stages of the nuclear lifecycle. Simulation case studies will be developed within the INDE framework, delivering an enhanced modelling capability while ensuring the framework has immediate application. For these case studies have been specified that are relevant to design and operation phases for AGR and PWR type reactors. The AGR case considers the through-life structural performance of graphite bricks. This involves modelling of multi-scale, multi-physics phenomena in the support of reactor operations. The PWR case study is based on core multiphysics modelling, with potential relevance to operating and future PWRs, and in particular in the design of SMRs.
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.
Design and Development of Virtual DCS Debugging and Research Platform Based on NPP Simulation Model
Authors: Zhang Caike , Qi Jingwen , Liu Chun , Xie Chenglong , Liu Peibang , Qu Ming
Published: October 2018

At present, DCS is widely used as the control system for nuclear power plants both at home and abroad, which prompting many companies to research the technology of DCS debugging. In this paper, taking a certain nuclear power plant within China for reference, the virtual DCS debugging and research platform which based on the full-scope nuclear power plant simulation model is developed. It was developed by first establishing a simulation model on the RINSIM Simulation Platform and ordering a customized set of virtual DCS system, then developing a communication program between the simulation model and the virtual DCS system. Users can observe the actual effects and results if following the pre-designed testing procedures after the configuration of control logics, HMI (Human Machine Interface) and I/O communication interfaces. The virtual DCS platform is aimed at assisting with technology research of DCS project for similar nuclear power plants and also can provide professional training for associated personnel of nuclear power plant.
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.
Prediction of Performance of Multi-Stage Orifice Assembly Using CFD Code
Authors: Niyogi Kalyan K. , Anton Stefan , Mitra-Majumdar Debabrata
Published: October 2018

A multistage orifice assembly is often used in fluid systems in nuclear plants in order to achieve a specific flow rate for a large pressure drop. Such an assembly is usually designed with a number of orifices installed in series. Apart from the performance parameters, an important design consideration is to guarantee absence of potential for cavitation inside the assembly. An improperly designed multistage orifice assembly may create cavitation at the outlet of some stages of the assembly when the pressure at the vena contracta reaches a pressure close to the vapor pressure of the fluid at the operating temperature. The cavitation condition makes not only the operation noisy but also results in a significant reduction of equipment life. It is a challenge to the designer to select a set of orifices so that the cavitation condition is simultaneously met at all stages of the assembly. On the other hand, it is also a challenge to design each individual orifice in the assembly so that its influence to the next orifice is minimal. These challenges have been met in design of a replacement multistage orifice design for the centrifugal charging pump mini-flow line of a nuclear plant for a pressure drop of 2,595 psi (17.89 MPa) at 60 gpm (227 l/min) flow-rate. The orifice assembly has eleven stages accommodated within a pipe of 24 inches (600 mm) length, and each stage consists of a 3-dimensional compound orifice. The fluid analysis is performed using the CFD code FLUENT/UNS. The code numerically solves the Navier-Stokes equations of fluid motion to obtain the velocity field and pressure distribution within a modeled 2- and 3-dimensional geometry. Turbulence effects are modeled by relating time-varying Reynolds stresses to the mean bulk flow quantities using the standard k-epsilon model. The design was optimized by fine tuning the geometric parameters of the orifice units to minimize the potential for cavitation and at the same time maintaining a high accuracy in the flow-rate. The full-scale prototype was successfully tested to confirm the performance. Actual flows in the prototype tests were observed to be within a fraction of a gpm of the design flow-rate and pressure drop. After orifice assembly was installed in the plant system, there was absolutely no operating noise, as opposed to the original unit that was extremely noisy. A number of other multistage orifice assemblies have subsequently been designed for different power plants using similar computational technique with great success.
Numerical Simulations on Throttle Characteristic With Large Pressure Drop and Optimal Design of the Orifice Plate
Authors: Chen Hu , Jue Wang , Cong Wang , Zili Gong , Zhen Jia , Anmin Yuan , Lei Chen , Yi Liao
Published: October 2018

The orifice throttle pipelines with large pressure drop suffer damages from two aspects: high frequency vibration caused by cavitation and low frequency vibration caused by high flow velocity. In order to solve the vibration phenomenon of typical throttling pipe with large pressure drop in the system of nuclear power plant, the key hydraulic characteristics such as pressure drop, flow velocity, streamline and eddy current were simulated and analyzed for single-stage orifice throttling pipe. The negative pressure area was found in the downstream of the orifice, where cavitation occurred. Eddy current was formed due to the large local velocity caused by the orifice plate jet. Then, the throttle performance of multi-stage concentric orifice plates was evaluated by means of blockage pressure drop method. The cavitation damage was relieved greatly. But it was not eliminated, especially in the last stage orifice. The expanding type five-stage orifice plate was designed according to pressure drop stage-decreasing principle, with which the possibility of cavitation was eliminated but large pressure drop resulted in large flow velocity at the downstream of the first orifice plate. Multi-stage eccentric orifice plate was designed with the consideration both in cavitation characteristics and velocity distribution, which could eliminate the harm of cavitation and reduce the low frequency vibration caused by large flow velocity to a maximum extent. As a result, multistage eccentric orifice plates could be recommended as an optimized design scheme for the vibration control of the orifice piping with large pressure drop.
Parallel Computation of the Point Neutron Kinetic Equations Using Parallel Revisionist Integral Deferred Correction
Authors: Cai Yun , Peng Xingjie , Li Qing , Zhang Zhizhu , Jiang Zhumin , Rui Guo
Published: October 2018

The point kinetics is very important to the safety of the reactor operation. However, these equations are stiff and usually solved with very small time step. These equations are solved by Revisionist integral deferred correction (RIDC), which is a parallel time integration method. RIDC is a highly accurate method, and it reduces the error by iteration. Based on C++ and MPI, a four-core fourth-order RIDC is implemented and tested by several cases, such as step, ramp, and sinusoidal reactivity insertion. Compared with other methods, the time step of RIDC in the step reactivity insertion case is smaller, but it’s larger in the case of the sinusoidal reactivity insertion. RIDC can keep high accuracy while the time step is appropriately large. The numerical results also show that the speed-up ratio can achieve 2 when 4 processors are used.
EPR: Burnable Absorber Optimization
Authors: Zeman Michal , Škoda Radek , Závorka Jiří
Published: October 2018

The research of burnable absorber (BA) is a very important issue, due to BA influence on regulation and control of reactivity in the reactor. Burnable absorbers compensate initial excess reactivity in the reactor and thus allow longer fuel cycle with higher fuel enrichment. This paper deals with the depletion calculation of Evolutionary Power Reactor (EPR) nuclear fuel with different types of BA, and with optimization of the BA. The main objective of this paper is to calculate burnable absorber composed of 2 elements. The distribution of BA in the fuel is here assumed uniform. Based on depletion calculation, elements are divided into two groups, fast and slow BAs. Several combinations of two BA elements are then calculated based on this division. The evaluation of each BA combination is made, and the most appropriate one is selected.
Experimental Study on Two-Phase Natural Circulation Flow Instability in Rod Bundle Channel Under Low Pressure Condition
Authors: Cheng Kun , Tan Sichao , Liu Zheng , Meng Tao
Published: October 2018

An experimental investigation was conducted in a natural circulation (NC) loop to study the characteristics of two-phase flow instability under low pressure condition. A 3 × 3 rod bundle channel was used as the test section. The effects of heating power, inlet subcooling degree and system pressure on the two-phase NC flow instability types and stable boundaries were studied. The experimental results show that three typical flow conditions can occur in rod bundle channel under NC condition, which are single-phase NC flow, subcooled boiling NC flow oscillation and density wave oscillations (DWO). The oscillation amplitude and period of DWO can be enlarged by increasing the heat flux. Increasing the inlet subcooling degree can increase the marginal heating power of flow instability in NC system. The occurrence of DWO can be suppressed by increasing the system pressure. The flow instability boundary presented by the subcooling number and phase change number was also obtained in present work.
Preliminary Design Considerations of He-Xe Mixture Cooled Space Nuclear Reactor
Authors: Meng Tao , Tan Sichao , He Yuhao , Yuan Dongdong , Cheng Kun
Published: October 2018

The space nuclear reactor has been widely studied since 60s in the last century. However, upon the signing of the Treaty on the Non-Proliferation of Nuclear Weapons (NPT) and the ending of the Cold War, the space nuclear technology gradually faded out the public and researchers view. In recent years, due to the proposal of the United States’ Moon and Mars Project and China’s Deep Space Scientific Project, space nuclear technology are welcoming new opportunities for research and development. In this paper, the Prometheus project design has been analyzed and thereafter been based for multi-kilowatt He-Xe cooled space nuclear system preliminary design. Parameters like system efficiency, compressor ratio, temperature are given and neutron calculations are conducted in order to evaluate its physical performance and provide guidelines for future optimization. A computer program, which can calculate performance of heat pipe radiator, is also coded and thereafter used. At last, some consideration guidelines are concluded for larger power space reactor design.
The Qualitative Analysis of Vertical Seismic Acceleration Effect on a Single Nuclear-Coupled Boiling Channel Natural Circulation Loop
Authors: Lee Jin-Der , Lin Yuh-Ger , Chen Shao-Wen , Pan Chin , Peir Jinn-Jer
Published: October 2018

Based on the external force method, the present study integrates the nonlinear dynamic model developed previously by the authors with vertical seismic accelerations to investigate the seismic-induced effect on the single nuclear-coupled boiling channel natural circulation loop. The natural frequencies of the states in the stable region are widely explored through nonlinear analysis. The results indicate that the natural frequency of initial state tends to increase as the increase in the phase change number (operating power) or as the decrease in the subcooling number (inlet subcooling). As supposing that a real seismic acceleration is directly imposed on the system, some parametric effects on the seismic-induced oscillations are performed in the present natural circulation loop system. The seismic-induced oscillations are found to be consistent with the combined results of the inherent system stability characteristics and the impact of external seismic acceleration. The system with a larger outlet loss coefficient of the heated channel, or a longer heated channel, would destabilize the seismic-induced oscillations, while the inlet loss coefficient of the heated channel has a stable effect on the system. A much stronger resonance oscillation could be induced by the increase in the core inlet subcooling. Notably, the enlargement and contraction of the heated channel diameter would move the natural circulation system closer to type-I and type-II instability regions, respectively. These both generate unstable effects on the seismic-induced oscillations due to the inherent stability characteristics of the initial states.
Research on the Physical Modelling for the Subchannel Analysis of PWR Core
Authors: Chen Guangliang , Dong Xiaomeng , Li Lei , Hu Peizheng , Appah Thompson , Tian Zhaofei , Zhang Zhijian
Published: October 2018

To obtain a better subchannel analysis tool for PWR core, an investigation on the physical modelling was done via CFD analysis in this paper. A detailed discussion on the TH status was carried out to consider the suitability of the physical modelling in the subchannel analysis. The counteracted effect on the balance of the mass, momentum, and energy was found. Its influence on the physical mechanism was analyzed for the TH status in PWR core. Then it was found that the subchannel analysis based on each flow channel is very necessary for the accuracy of the 3D engineering TH analysis for the whole PWR core.
Multiphysics Modeling of Thorium-Based (Th, U)O2 and (Th, Pu)O2 Fuel Performance in a Light Water Reactor
Authors: Liu Rong , Cai Jie-Jin , Zhou Wen-Zhong , Wang Ye
Published: October 2018

ThO2 has been considered as a possible replacement for UO2 fuel for future generation of nuclear reactors, and thorium-based mixed oxide (Th-MOX) fuel performance in a light water reactor was investigated due to better neutronics properties and proliferation resistance compared to conventional UO2 fuel. In this study, the thermal, mechanical properties of Th0.923U0.077O2 and Th0.923Pu0.077O2 fuel were reviewed with updated properties and compared with UO2 fuel, and the corresponding fuel performance in a light water reactor under normal operation conditions were also analyzed and compared by using CAMPUS code. The Th0.923U0.077O2 fuel were found to decrease the fuel centerline temperature, while Th0.923Pu0.077O2 fuel was found to have a bit higher fuel centerline temperature than UO2 fuel at the beginning of fuel burnup, and then much lower fuel centerline than UO2 fuel at high fuel burnup. The Th0.923U0.077O2 fuel was found to have lowest fuel centerline temperature, fission gas release and plenum pressure. While the Th0.923Pu0.077O2 fuel was found to have earliest gap closure time with much less fission gas release and much lower plenum pressure compared to UO2 fuel. So the fuel performance could be expected to be improved by applying Th0.923U0.077O2 and Th0.923Pu0.077O2 fuel.
Validation of Film Evaporation Model in GASFLOW-MPI
Authors: Yabing Li , Han Zhang , Jianjun Xiao
Published: October 2018

A dynamic film model is developed in the parallel CFD code GASFLOW-MPI for passive containment cooling system (PCCS) utilized in nuclear power plant like AP1000 and CAP1400. GASFLOW-MPI is a widely validated parallel CDF code and has been applied to containment thermal hydraulics safety analysis for different types of reactors. The essential issue for PCCS is the heat removal capability. Research shows that film evaporation contributes most to the heat removal capability for PCCS. In this study, the film evaporation model is validated with separate effect test conducted on the EFFE facility by Pisa University. The test region is a rectangle gap with 0.1m width, 2m length, and 0.6m depth. The water film flowing from the top of the gap is heated by a heating plate with constant temperature and cooled by countercurrent air flow at the same time. The test region model is built and analyzed, through which the total thermal power and evaporation rate are obtained to compare with experimental data. Numerical result shows good agreement with the experimental data. Besides, the influence of air velocity, wall temperature and gap widths are discussed in our study. Result shows that, the film evaporation has a positive correlation with air velocity, wall temperature and gap width. This study can be fundamental for our further numerical study on PCCS.
Simulation of HTR-10 Anti-Compton HPGE Gamma-Ray Spectrometer With Geant4
Authors: Mao Cui , Liu Yibao , Zhang Liguo , Tong Jiejuan , Xia Bing , Yin Zaizhe
Published: October 2018

The efficient and accurate burn-up measurement of the spherical fuel element is the key component of the operation of the pebble bed high temperature gas-cooled reactor. The accuracy of the method that determine burnup by the activity of Cs-137 degrades due to operation characteristics of HTR-10. HTR-10, as an test reactor, operated on and off during the past years. It stayed shutdown more than power operation. In order to improve the measurement accuracy of Cs-137 activity and enhance the possibility to detect radionuclides with low activity, which can be used to correct the classic burnup assay method, a new measurement system is now discussed using anti-coincidence technology, which suppresses the Compton plateau. In this paper, Geant4 is used to simulate the anticoincidence measurement process taking high purity germanium γ-ray spectrometer as main detector and plastic scintillator as the annular detector. By analyzing the signal to noise ratio in different detection scenarios with all kinds of shape parameters of the annular detector, the annular detector with the best anti-coincidence effect are optimaized. The above research results provide an important theoretical basis for the construction of online burn-up measurement system based on anti-Compton technology.
VVER 1000 Pressurizer System and Control Modelling in Dymola
Authors: Fořtová Anna , Ježek Filip
Published: October 2018

The pressurizer is responsible for controlling the pressure and temperature in the primary circuit of the nuclear power plant. It is basically a pressure vessel, filled partially with water and partially with steam at saturation state. The controlling process can be described by two regimes, either by the self-control regime or by the automatic control regime. This paper is describing the simplified automatic control regime on the model of the pressurizer in the primary circuit of VVER 1000 made in Dymola software with the ClaRa library. Reactor power change and corresponding steam generator power change are the actuators in the simulation. The behaviour of the coolant level and pressure in the pressurizer is simulated in the model and it is then compared with data provided by the supplier of VVER 1000.
Research on Structural Design and Analysis of S-CO2 Turbine Impeller
Authors: Wu Jun , Ma Can , Dai Chunhui , Zhao Zhenxing , Dai Lu , Liu Zhouyang
Published: October 2018

The Brayton cycle with supercritical carbon (S-CO2) as working medium is one of the most promising new nuclear power systems. Turbine is the key device during the working process in the Brayton power cycle. The turbine structural presents small size and extremely high rotational speed for the special physical properties of S-CO2, which increase the difficulty for the structural design and strength safety significantly. According to the aerodynamic design and optimization results of 200 kW S-CO2 radial inflow turbine, this paper proposes a detail structural design and analysis method for turbine impeller. Based on the three-dimensional blade profile data and meridional planes data, key structural design parameters are chosen and the parametric geometry model is established by CAD tools. On this basis, numerical simulation models of turbine are established to analyze the structural strength in detail. Then the influence of parameters on the turbine impeller strength is studied by a series of finite element numerical procedures. The influence mechanisms of key structural design parameters on impeller strength are discussed. Moreover, the final model of turbine impeller is obtained by parameter comparison and selection. The results show that for the initial model, the maximum von-Mises equivalent stress is 400.10 MPa, the maximum radial deformation is 0.0333 mm and the maximum axial deformation is 0.0770 mm. For the final model, the maximum von-Mises equivalent stress is 294.26 MPa, the maximum radial deformation is 0.0279 mm and the maximum axial deformation is 0.0769 mm. The maximum von-Mises equivalent stress and maximum radial deformation of structural decreases 26.45 % and 16.22 % respectively compared with the initial model. As a result, the impeller structural strength safety margin is obviously improved by the parameter analysis.
Study on Shutdown Fire PRA for Nuclear Power Plant
Authors: Liu Meiru , Zhao Qingnan , Deng Wei , Du Jinyan , Sun Lin
Published: October 2018

Fire Probabilistic Risk Assessment (PRA) is one of the main methods of fire safety analysis for nuclear power plants (NPPs). At present, the fire PRA under the at-power condition has been widely studied, while the research on the low power and shutdown condition (LPSD) is quite limited. Therefore, in this paper, a second generation NPP on the east coast of China is taken as the research target, and the analysis methods are based on the latest LPSD fire PRA theory in report NUREG/CR-7114. This paper studies the initiating events and ignition frequencies of fire PRA considering the real conditions in LPSD, and established LPSD Fire PRA model, finally obtained the quantitative risk result of the core damage caused by the fire According to the results of this LPSD fire PRA, the fire risk-significant sources and fire risk weakness are found out and the improvement suggestions have been promoted.
Structural Analysis and Manufacturing of HL-2M Vacuum Vessel and Support Structures
Authors: Huang Yuncong , Ran Hong , Cao Zeng , Hou Jilai , Song Binbin
Published: October 2018

The construction of the HL-2M tokamak (an upgraded machine of HL-2A) is in proximity to the finalization. As one of the main components of the HL-2M tokamak device, the vacuum vessel is designed to a D shape double-thin-wall welded sturcture. The whole structure is divided into 20 sectors in the toroidal direction including 4 sectors that are used to compensate for manufacturing and assemble errors during the construction period. Considering to the thermal expansion of the vacuum vessel and the limited space in this tokamak device, there are five support rods which are made of Inconel 718 locating in the midplane of the vacuum vessel. The structural analysis has been performed on magnetic loads induced by a plasma disruption and thermal loads during a plasma discharge. The maximum stress is obtained within allowable stresses of vacuum vessel material Inconel 625, it is concluded that this type of vacuum vessel structure is suitable for HL-2M. Up to now, the manufacturing work of 16 sectors and all the support rods have been finished. The entirely assembly and related text work (dimension measurement, leak detecion and pumping experiment) will be finished before June 2018.
Study on Methods for HCLPF Value of Nonlinear Supports System of Steam Generator
Authors: Cai Feng-chun , Ye Xian-hui , Huang Qian , Zhang Wenzheng
Published: October 2018

High confidence of low probability of failure (HCLPF) values of equipment, representing the seismic capacities of the equipment, are the fundamental ingredient in seismic probability safety assessment (SPSA) and seismic margin analyses (SMA). In this paper, two methods for calculating the HCLPF values of equipment were investigated, fragility analysis, and conservative deterministic failure margin (CDFM). These methods are linear methods. Based on these methods, HCLPF value of equipment can be computed conveniently by scaling the results of the existing seismic analysis. For a nonlinear systems, the HCLPF values based on these linear scaling methods are unrealistic. For a complicated nonlinear equipment or structure, a detail nonlinear model was used to derive the seismic capacity. The results by this method are realistic, but cost calculation time. In this paper, a nonlinear model of reactor coolant system coupled reactor building was built. This model includes the steam generator and considers the nonlinear factors of steam generator such as gap in the supports, plasticity of hot leg and cold leg. Forced motion was applied to the base of reactor building. And seismic response of the steam generator was calculated iteratively by scaling the ground motion level step by step. Based on these calculations, a curve of load on the supports VS peak ground acceleration (PGA) can be obtained. Then based on these curves and allowable load of supports of steam generator, which derived from stress analysis on support of steam generator, seismic capacity of the supports of steam generator was determined. Then the HCLPF Value of the supports of steam generator was obtained by this nonlinear time history analysis and was compared with the results based on the CDFM. The two results were different. Therefore, the HCLPF seismic capacity of equipment with nonlinearity, such as gap nonlinearity, should be calculated by nonlinear time history method.
Experimental Study and Analysis of the Deformable Pipe of CRHDM
Authors: Li Yuzheng , Zhang Huang , Liu Qianfeng , Hanliang Bo
Published: October 2018

Hydraulic Control Rod Drive Technology (HCRDT) is a newly invented patent owned by Institute of Nuclear and New Energy Technology Tsinghua University with independent intellectual property rights. The hydraulic cylinder which is connected by the deformable pipe is the key part of this technology, so the performance of the deformable pipe directly affects the function of the hydraulic cylinder and the HCRDT. Based on the conditions occurring in the operation of the Control Rod Hydraulic Drive System, a kind of new deformable pipe is designed and has been tested by the experiment and analyzed by ANSYS. The result shows that the longer deformable pipe is the longer its lifetime is firstly. Second, lifetime of the deformable pipe which has net is longer. Furthermore, the design of HCRDT can be optimized.
Development and Validation of a Correlation for Wet Resuspension Simulation
Authors: Jankowski Tobias , Koch Marco K.
Published: October 2018

In case of a postulated severe accident scenario in a nuclear power plant coolant might be released from the cooling circuit into the containment atmosphere. Due to the pressure difference between the cooling circuit and the containment, the fluid will be released gaseous and might contain fission products, depending on the severe accident scenario. The effluent vapour might condensate on the colder containment structures and flow down to the floor, where water pools will be formed. Because of the decay heat of the fission products or due to a pressure decrease in the containment, which might be induced by the operation of filtered containment venting systems, the water pools could start boiling. During this boiling process gas bubbles will rise to the water pool surface and release former solved fission products by bubble bursting into the containment atmosphere. Those released fission products will dry due to their decay heat and remain as very fine particles for a long time in the containment atmosphere. In the context of filtered containment venting system operation a detailed knowledge of those particles is among others of interest for the development of e. g. accident management measures. A correlation for the estimation of the so-called droplet re-entrainment, which is defined as the ratio of the droplet mass flow, released from the bursting bubbles, and the gas flow, which streams through the effective pool surface, is developed and will be validated based on several single effect tests. The considered tests were conducted in test facilities of different scale and investigated the influence of different parameters on the release of aerosols by the effect of droplet re-entrainment. The calculation results are in a good agreement with the experimental data, indicating that parameters with a relevant influence on the droplet re-entrainment are considered in the developed correlation with a suitable impact.