bluuebunny/crossref_metadata_2025 huggingface dataset:
VINSAP Tool for Visualization of Severe Accident Progress
Authors: Kropik Martin , Duspiva Jiri
Published: October 2017

The contribution provides information about the development of a system for visualization of NPP severe accident progress. This visualization is under development in cooperation of UJV Rez, a.s. and Czech Technical University in Prague. The project is supported by the Technology Agency of the Czech Republic and is planned to be solved from 2015 to 2017. The visualization uses results of an analytical code MELCOR for evaluation of the NPP severe accident progress. The visualization firstly reads MELCOR results, transforms them to a suitable format for quick processing and provides graphical screens with reactor components that could demonstrate the progress of the evaluated severe accident. The visualization can even provide parallel presentation of more different scenarios of the severe accident. The system is planned to be used for training of NPP staff to handle severe accidents. In the first year of the project solution (2015), the software for MELCOR data transformation, next for providing information about transformed data were developed. In the following year (2016), software for creation of graphical screens with reactor components and software for severe accident progress presentation is creating. In the final year of the project (2017), thorough testing is going to be carried out, and the applicability of the visualization for a practical use during a NPP staff training is going to be verified.
Analysis of the Effect of Beam Divergence Angle on Back-Streaming Electron Region in Ion Source for EAST-NBI
Authors: Hu Chundong , Wu Mingshan , Xie Yahong , Wei Jianglong , Yu Ling
Published: October 2017

During the process of beam extraction in positive ion source under high voltage region, a large number of electrons are produced in the gaps of grids. After back-streaming acceleration, these electrons go back to arc chamber or impinge grids and heat electron dump or grids, which are harmful for the safety of ion source. Under the situation of poor beam extraction optics, a large part of the primary beam ions bombard the surface of suppressor grid. And this process produces a large number of electrons. Due to the huge extracted voltage, the secondary electron emission coefficient of the suppressor grid surface is also great, when beam ions bombard on it. As a result, the grids’ current grows. The curvature of ion emission surface and equipotential surface nearby are mainly connected to the perveance and plasma grid geometry. In order to optimize the beam performance of high current ion source and increase the mean arc efficiency, the plasma grid of accelerator is already replaced from circular cross section grid to diamond cross section grid. As a result, the shape of ion emission surface is only connected to the perveance. According the measurement of the current of suppressor grid and the calculation of the perveance of the corresponding shoot, we can analyze the effect of beam divergence angle on back-streaming electron. When the beam divergence angle increases, the number of back-streaming electrons increases rapidly, and grids current changes significantly, especially the current of gradient grid and suppressor grid. The results can guide the parameters operating on the ion source for EAST-NBI and find the reasonable operation interval of perveance and the best one to ensure the safety and stable running of the ion source, which has great significance on the development of long pulse, high power ion source.
Development and Preliminary Verification of the 3D Core Neutronic Code: RCAD
Authors: Chen Xiaosong , Xiong Xin , Peng Lianghui , Gang Zhi , Liu Zhihong , Xu Hong
Published: October 2017

An inherent error occurs in the traditional light water reactor core design calculation, which is generally based on the equivalent homogenization method and diffusion calculation model. Because the boundary conditions and state parameters used in the assembly homogenization calculation is different from the real situation, the equivalent homogenization error is inevitable. This paper introduces the recent progress in the development of the 3D Core Neutronic Code: RCAD. The focus is placed on the equivalent homogenization error correction model and preliminary verification. The preliminary verification result shows that the correction model is effective and RCAD code is qualified for reactor core design for pressurized water reactor (PWR).
Conceptual Design of an Ex-Vessel Melt Cooling Device
Authors: Zhang Li , Yu Mingrui , Guo Qiang , Zhu Yiming , Yuan Yidan , Ma Weimin
Published: October 2017

A helical ex-vessel cooling device is being designed and evaluated to insure retention of the molten corium that may discharge from the failed lower head of a reactor pressure vessel (RPV) during postulated severe accidents of a light water reactor. This conceptual design consists of four functional parts located in the reactor cavity, including an intermediate chamber for temporary retention of corium, an helical channel for corium spreading, cooling pipes embedded in the channel wall and venting system. The heat flux profile and steam generation along the cross section of the spreading channel T are analyzed under several conservative assumptions. The results show the melt coolability is achievable even under Benrenson film boiling regime.
A Development of Human-System Interaction in Digital NPPs
Authors: Lu Jiachuan , Liao Longtao , Feng Bo , Chen Yifen , Hao Junwei , Ye Kai , Zheng Jianxiang
Published: October 2017

Now Nuclear Power Plants (NPPs) design is moving toward being highly dependent on digital computers in many complex systems, especially microprocessors. As a medium between operators and NPPs for exchange and interaction which ultimate operational decisions still rely on, the Human-System Interaction has been widely concerned and become one of the focuses in NPPs design. So in order to take full advantage of operating experience, human cognitive processing abilities, and progressive technologies, it is critical to plan, design, implement, operate, and maintain a reliable HSIs. The project, funded by the Nuclear Power Institute of China (NPIC), designs and develops a set of typical and comparatively complete technical solution of Human-System Interaction based on instrumentation and control system in actual NPP. To take advantage of the design process and modules as well as templates of this technical solution provided by this project, which take HFE into account, we can achieve the realistic simulation of Human-System Interaction for digital NPPs, making the use of iFIX software, and the Human-System Interaction system can be used to design interfaces for different kinds of NPPs. In this paper, the realization of human-system interaction will be introduced, and the current research status and main challenges of Human-System Interaction are included. And at this stage we have made the processes of the cross-platform data acquisition and monitoring, processing and display of small instrument control systems come true.
Microstructures and Properties of 304L Steel CAP1400 Fuel Bottom Nozzles Prepared by Additive Manufacturing Technology
Authors: Zongshu Li , Wentao Liu , Songtao Yang , Ruotong Hao
Published: October 2017

Nozzles of fuel assembly play an important role in pressure water reactor (PWR) fuel assembly element. For a long time, ordinary processing technologies of nozzles of fuel assembly have the problems of difficult and complicated process, the low availability of material and long the development cycles of manufacturing. However, according to the study these issues can be well settled by using the additive manufacturing technology. This paper studies a nozzle of fuel assembly prepared by this additive manufacturing technology through slow-strain-rate tension (SSRT) test and microstructure observation experiment. The results of SSRT test show that yield strength of the nozzle of fuel assembly is about 401.5MPa, the extensional rigidity is about 673.5MPa and the ductility is about 45.7%. And the SEM fracture results of the SSRT sample indicate that the fracture microstructure contains a large number of dimples, and the way of fracture belongs to plastic. And the metallographic observation consequences manifest that the microstructure of nozzle of fuel assembly prepared by the additive manufacturing technology is composite tissue of both austenite and ferrite, and the grains are settled along the way of laser scanning and there are isometric with some kind of direction. This metallographic microstructure is different from the traditional morphology of the free carbide distributed in the matrix. The dual phase microstructure of austenite and ferrite can improve the mechanical properties of the matrix effectively, and avoid the free carbides which may lead to matrix fragmentation in the tensile deformation process. Moreover, the laser power could affect the microstructure and properties of nozzles of fuel assembly observably, and the high laser power could bring about the ablation of metal. Through the analysis of mechanical properties and microstructure, we have made it possible to make the laser additive manufacturing technology to be used for the fuel assembly nozzle preparation in the nuclear power area. This work not only presents the advantages of the laser additive manufacturing technology in the fuel element processing area of the nuclear power station, but also broadens the application range of the laser additive manufacturing technology. What’s more we provide the new thoughts for the fast and effective preparation of the fuel element especially for the fuel assembly nozzle in the nuclear power station.
The Linear System Solver of Programs Named CORTH and KYLIN2
Authors: Ming Pingzhou , Pan Junjie , Tu Xiaolan , Liu Dong , Yu Hongxing
Published: October 2017

Sub-channel thermal-hydraulics program named CORTH and assembly lattice calculation program named KYLIN2 have been developed in Nuclear Power Institute of China (NPIC). For the sake of promoting the computing efficiency of these programs and achieving the better description on fined parameters of reactor, the programs’ structure and details are interpreted. Then the characteristics of linear systems of these programs are analyzed. Based on the Generalized Minimal Residual (GMRES) method, different parallel schemes and implementations are considered. The experimental results show that calculation efficiencies of them are improved greatly compared with the serial situation.
Influence of Oxygen Concentration on the Initiation of Dissolution Corrosion on 316L Austenitic Stainless Steel at 450°C
Authors: Klok Oksana , Lambrinou Konstantina , Gavrilov Serguei , De Graeve Iris
Published: October 2017

This work presents first results of the study on the influence of the LBE oxygen concentration on the initiation of dissolution corrosion in 316L austenitic stainless steels. 316L steel specimens were exposed at 450 °C to static liquid LBE with controlled and constant oxygen concentration of 10−5, 10−6 and 10−7 mass% for 1000 hours. Corroded specimens were analysed by scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS). Limited oxidation corrosion and no dissolution corrosion was observed in the specimens exposed to LBE containing 10−5 and 10−6 mass% oxygen, while dissolution corrosion with a maximum depth of 59 μm was found in the specimen exposed to LBE containing 10−7 mass% oxygen.
Characterization of FUTURIX-FTA Irradiated Nuclear Fuel Samples
Authors: Andrello Concettina , Freis Daniel , Lo Frano Rosa , Papaioannou Dimitri , Delage Fabienne
Published: October 2017

The amount of spent fuel and high-level waste already available, and which will be produced by the future NPPs operation, calls for the evaluation of any possible technological solution that could minimize the burden of their disposal: reduction of Minor Actinide (MA) content, in addition to the radiotoxicity and radioactivity, and of the generated thermal load (decay heat). In this context, R&D efforts currently focus on the development of methodologies and technical solutions for Partitioning and Transmutation.

MAs and long-lived fission products are in fact the main contributors to the long-term radiotoxicity of spent nuclear fuel, and their transmutation to short-lived fission products, in fast spectrum nuclear reactors, in transmuters or in Accelerator Driven Systems (ADS), by neutron irradiation of dedicated fuels/targets, is a promising and widely investigated option.

In order to provide substantial input for the safety assessment of innovative nuclear fuels dedicated to MA transmutation, several irradiation tests are being carried out. In some options, the investigated fuels/targets are uranium free, or of low uranium content, to improve the transmutation performance and contain high concentrations of MA and plutonium compounds.

Two molybdenum based CER-MET fuels, called ITU-5 & ITU-6, were prepared at JRC Karlsruhe for the irradiation experiment FUTURIX-FTA (FUel for Transmutation of transURanium elements in phenIX/Fortes Teneurs en Actinide). The experiment performed from 2007 to 2009 in the Phénix reactor, France, in cooperation with CEA. The experiment ended after 235 equivalent full power days (EFPD) at a Linear Heat Rate of circa 130 W/cm and reached burn-ups of 18 %FIHMA and 13 %FIHMA, respectively. Afterwards, the pins were transported to the Hot Cells of JRC Karlsruhe for Post Irradiation Examination.

After a short summary describing the fuel preparation and irradiation conditions of the FUTURIX FTA irradiation experiment, the paper will give an overview on the current status and further planning of the Post Irradiation Examinations of ITU-5 & ITU-6 at JRC Karlsruhe. Finally, the results of the characterisations will be discussed and conclusions on the irradiation performance will be drawn. The results of this experiment will help to increase the knowledge and understanding of the irradiation behaviour of metal based transmutation targets and the qualification and validation of models developed to predict fuel safety performance.
Wet Modal Analysis of Wire-Wrapped Fuel Rod With the Axial Flow
Authors: Liu Haidong , Chen Deqi , Hu Lian
Published: October 2017

Dynamic characteristics of wire-wrapped fuel rod was studied when the structure was coupled with fluid. In this paper, the wet modal analysis of wire-wrapped fuel rod in the axial flow was investigated by using the finite element method, which relied on fluid-structure interaction calculation. This numerical simulation is compared to experiments in the literature and the results agree well. The Lead-Bismuth Eutectic (LBE) was used as the working fluids. The CFD Simulations of a wire-wrapped fuel rod were performed with different boundary conditions. By comparing the modal characteristics of the fuel rod in the LBE and in vacuum, the results showed that the added mass of the fluid has important effect on vibration frequency of the wire-wrapped fuel rod. The different three frequencies and modal shape can be obtained under the different inlet velocity. According to the simulation result, it could be found that the maximal equivalent elastic strain occurred in the constrained end.
The Research on Effects of the Medium Blocking on Motion Performances of Main Steam and Feed Water Isolation Valves With its Driving Device
Authors: Pan Dalei , Zhong Hua , Gu Shenjie , Guo Guangyue , Kuang Hongbo
Published: October 2017

This article presents the study of the medium block effects on the driving device performances of main steam and feed water isolation valves, which are typical electromechanical equipment in nuclear power plants. The equipment simulation model is established, which can calculate motion performances under all working situations. The medium blocking force models during the motion of valves are established and the blocking effects on the driving device performances are calculated and analyzed. The simulation results show that the medium block has little effects on the opening performances of the driving device, while it has important effects on the closing performances, especially on the quick closing, which suggests that the equipment should be lubricated to reduce resistance coefficients and the main parameters of electromechanical equipment should be tuned and calibrated to guarantee the quick closing time.
A Comparative Study of 10 Different Methods on Numerical Solving of Point Reactor Neutron Kinetics Equations
Authors: Zhang Yining , Zhang Haochun , Su Yang , Zhao Guangbo
Published: October 2017

Point reactor neutron kinetics equations describe the time dependent neutron density variation in a nuclear reactor core. These equations are widely applied to nuclear system numerical simulation and nuclear power plant operational control. This paper analyses the characteristics of 10 different basic or normal methods to solve the point reactor neutron kinetics equations. These methods are: explicit and implicit Euler method, explicit and implicit four order Runge-Kutta method, Taylor polynomial method, power series method, decoupling method, end point floating method, Hermite method, Gear method. Three different types of step reactivity values are introduced respectively at initial time when point reactor neutron kinetics equations are calculated using different methods, which are positive reactivity, negative reactivity and higher positive reactivity. The calculation results show that (i) minor relative error can be gain after three types of step reactivity are introduced, when explicit or implicit four order Runge-Kutta method, Taylor polynomial method, power series method, end point floating method or Hermite method is taken. These methods which are mentioned above are appropriate for solving point reactor neutron kinetics equations. (ii) the relative error of decoupling method is large, under the calculation condition of this paper. When a higher reactivity is introduced, the calculation of decoupling method cannot be convergence. (iii) after three types of step reactivity are introduced respectively, the relative error of implicit Euler method is higher than any other method except decoupling method. The third highest is Gear method. (iv) when the higher reactivity is introduced, the relative error of explicit and implicit Euler method are almost coincident, and higher than any other methods obviously. (v) 4 methods are suitable for solution on these given conditions, which are implicit Runge-Kutta method, Taylor polynomial method, power series method and end point floating method, considering both the accuracy and stiffness.
Numerical Prediction of Poly-Dispersed Condensation Using Quadrature Method of Moments and Multi-Fluid Model
Authors: Lankadasu Anjaneyulu , Krumenacker Laurent , Kumar Anil , Tripathi Amita
Published: October 2017

Accurate prediction of condensation plays an important role in the development of high efficiency turbo-machines working on condensable fluid. Therefore it demands modeling of poly-disperse characteristic of number distribution function while modeling condensation. Two such kind of models are considered in this work and they are namely, quadrature method of moments (QMOM) and multi-fluid method (MFM) models. The vital difference between these two models lies in the method of discretisation of the droplet size distribution. Further, their numerical aspects like ease of implementation in general purpose computational fluid dynamics solvers, accuracy and associated computational cost are discussed. In order to obtain accurate thermodynamic properties, the real gas formulations defined in IAPWS-IF97 are used. These algorithms are applied to the compressible Navier-Stokes solver of Fluidyn MP and tests are carried on Laval nozzle and compared with the experimental measurements.
Code Development for the Neutronics/Thermal-Hydraulics Coupling Transient Analysis of Molten Salt Reactors
Authors: Hu Tianliang , Cao Liangzhi , Wu Hongchun , Zhuang Kun
Published: October 2017

A code system has been developed in this paper for the dynamics simulations of MSRs. The homogenized cross section data library is generated using the continuous-energy Monte-Carlo code OpenMC which provides significant modeling flexibility compared against the traditional deterministic lattice transport codes. The few-group cross sections generated by OpenMC are provided to TANSY and TANSY_K which is based on OpenFOAM to perform the steady-state full-core coupled simulations and dynamics simulation. For verification and application of the codes sequence, the simulation of a representative molten salt reactor core MOSART has been performed. For the further study of the characteristics of MSRs, several transients like the code-slug transient, unprotected loss of flow transient and overcooling transient have been analyzed. The numerical results indicated that the TANSY and TANSY_K codes with the cross section library generated by OpenMC has the capability for the dynamics analysis of MSRs.
Hybrid Nuclear Thermal Energy Storage System Revenue With Design and Market Uncertainty
Authors: Mann W. Neal , Schneider Erich A.
Published: October 2017

A generic thermal energy storage (TES) system is considered as a retrofit to an existing nuclear power plant in Texas. A validated PLEXOS model of the ERCOT electric grid is used to simulate electricity market clearing in 2030. Three scenarios of natural gas price forecasts are used with a coupled capacity expansion model to simulate the deployment of competing technologies. The power, energy capacity and ramp rate of the TES system are varied parametrically in order to calculate the revenues arising from arbitrage as well as ancillary services. Increasing the ramp rate and power increases the storage system net revenue in all cases. Increasing energy capacity has a positive effect on net revenue in a small number of cases but has mixed results overall.
Numerical Study on Hydrogen Flow Behavior for Different Connecting Pipe Structures
Authors: Liu Hanchen , Yuan Kai , Cao Xuewu
Published: October 2017

Hydrogen control in the case of severe accidents has been required by nuclear regulations to ensure the integrity of nuclear containment building. Therefore, getting detailed hydrogen flow and distribution is a key issue to resolve the hydrogen risk in containment and compartments. In this study, local hydrogen behavior has been investigated in a multi-subcompartments using Computational Fluid Dynamics method, concerning the local concentration in the multi-subcompartments. The multi-subcompartments containment is represented by four vessels interconnected by pipe. Hydrogen inlet pipe is located in source vessel. The analysis model is built by 3-dimensional Computational Fluid Dynamics code ANSYS-CFX in Cartesian coordinates. Computational domain is discretized in about 47559 cells, 87484 cells and 126388 cells, respectively. Taking a full consideration of the computational time and accurate, the medium mesh scheme is adopted to simulate the hydrogen transport.

With standard k-ε turbulence model, the effects of the connecting pipe parameters and direction on hydrogen distribution in the multi-compartments are investigated. Local hydrogen behavior showed remarkably different in simulations with the change of the pipe parameter. Increasing the connection pipe diameter and decreasing the connection pipe length are helpful for helium flow between compartments. Compare with the vertical connection pipe, horizontal connection pipe is not conducive to the diffusion of hydrogen from source compartment to non-source compartment.
Modeling the Ambient Condition Effects of an Air-Cooled Natural Circulation System
Authors: Hu Rui , Lisowski Darius D. , Bucknor Matthew , Kraus Adam R. , Lv Qiuping
Published: October 2017

The Reactor Cavity Cooling System (RCCS) is a passive safety concept under consideration for the overall safety strategy of advanced reactors such as the High Temperature Gas-Cooled Reactor (HTGR). One such variant, air-cooled RCCS, uses natural convection to drive the flow of air from outside the reactor building to remove decay heat during normal operation and accident scenarios. The Natural convection Shutdown heat removal Test Facility (NSTF) at Argonne National Laboratory (“Argonne”) is a half-scale model of the primary features of one conceptual air-cooled RCCS design. The facility was constructed to carry out highly instrumented experiments to study the performance of the RCCS concept for reactor decay heat removal that relies on natural convection cooling. Parallel modeling and simulation efforts were performed to support the design, operation, and analysis of the natural convection system.

Throughout the testing program, strong influences of ambient conditions were observed in the experimental data when baseline tests were repeated under the same test procedures. Thus, significant analysis efforts were devoted to gaining a better understanding of these influences and the subsequent response of the NSTF to ambient conditions. It was determined that air humidity had negligible impacts on NSTF system performance and therefore did not warrant consideration in the models. However, temperature differences between the building exterior and interior air, along with the outside wind speed, were shown to be dominant factors. Combining the stack and wind effects together, an empirical model was developed based on theoretical considerations and using experimental data to correlate zero-power system flow rates with ambient meteorological conditions. Some coefficients in the model were obtained based on best fitting the experimental data. The predictive capability of the empirical model was demonstrated by applying it to the new set of experimental data. The empirical model was also implemented in the computational models of the NSTF using both RELAP5-3D and STAR-CCM+ codes. Accounting for the effects of ambient conditions, simulations from both codes predicted the natural circulation flow rates very well.
Experimental Investigation in LIFUS5/Mod2 Facility of Spiral-Tube Steam Generator Rupture Scenarios for ELFR
Authors: Pesetti Alessio , Del Nevo Alessandro , Neri Andrea , Cati Stefano , Sermenghi Valerio , Valdiserri Massimo , Giannotti Daniel , Tarantino Mariano , Forgione Nicola
Published: October 2017

In the framework of the European Commission LEADER project, an experimental campaign of seven tests was performed in the LIFUS5/Mod2 facility, at ENEA CR Brasimone, for investigating the postulated Steam Generator Tube Rupture (SGTR) event in a relevant configuration for the Spiral-Tube Steam Generator (STSG) of the European Lead Fast Reactor (ELFR).

The LIFUS5/Mod2 facility is composed by a water tank of 15 L injecting subcooled water up to 200 bar into the reaction tank of 100 L (420 mm of diameter), which is connected by a 3 inch pipe to the dump tank of 2 m3. A dedicated test section was designed, assembled and implemented in the reaction tank. It is composed by 188 tubes, vertically disposed with triangular pitch inside a cylindrical support. This tube bundle is representative of a portion of the STSG of ELFR. The cylindrical support is closed at the lower and upper end by two tube plates and has a perforated lateral shell (300 mm of diameter and 400 mm high). The reaction tank is filled by Lead-Bismuth Eutectic alloy (LBE) at 400°C up to the top tube plate, with an argon cover gas at about 2 bar. The water is injected at about 180 bar and 270°C through the central tube, at middle height of the bundle. The water-LBE interaction is characterised by high quality data acquisition system: 6 fast Pressure Transducers (PTs) working at 10 kHz for precisely characterize the first narrow injection peaks, 70 low constant time Thermocouples (TCs) to understand the vapour evolution path and 13 strain gages (SGGs) for measuring the strain of the bundle and main vessel.

The overall LEADER experimental campaign is constituted by seven tests, divided in three series (B1, B2 and B3), characterized by different injection orifice diameters of 4, 8.9 and 12.6 mm, respectively. This paper presents the experimental results of the first two tests of series B2 (B2.1 and B2.2) having 8.9 mm of injection orifice.

The first test analysed showed a first narrow pressure peak of about 32 bar, some milliseconds after the cap rupture instant. The following pressurization due to the evaporation of water entered into the reaction vessel was of an analogues magnitude for both the tests (about 50 bar) and lasted some tenths of second. The water/LBE interaction lower temperature was reached on the inner ranks of tubes, about 150°C. The outer rank was cooled down to about 300°C. The strain gage measurements showed a decreasing deformation on the tubes toward the outer positions. No ruptures were observed on tubes surrounding the injector. The amount of LBE transported into the dump tank was strongly dependent on the LBE level in the reaction tank at the start of the tests and about 200 kg.
Guided Waves As an Online Monitoring Technology for Long Term Operation in NPPs: New Experimental Results on a Steam Discharge Pipe
Authors: Bertoncini Francesco , Cappelli Mauro , Cordella Francesco , Raugi Marco
Published: October 2017

Guided Wave (GW) testing is regularly used for finding defect locations through long range screening using low-frequency waves (from 5 to 250 kHz) [1]-[3]. Magnetostrictive sensors can overcome some issues, which usually limit the application to Nuclear Power Plants (NPPs) [4], like for example, high temperatures [5]-[6], high wall thickness of components in the primary circuit, and characteristic defect typologies. The authors have already shown the basic theoretical background, some simulations and some first experimental results concerning a real steel pipe, used for steam discharge, having a complex structure. Collecting more experimental data with a novel test campaign on the same pipe its complex structure results as a useful benchmark for the application of GWs as Non Destructive Techniques (NDT). Experimental measures using a symmetrical probe and a local probe in different configurations (pulse-echo and pitch-catch) indicate that GW testing with magnetostrictive sensors can be reliably applied to long-term monitoring of NPP components.
Methodology to Investigate Vibration Phenomena Caused by the Steam Condensation at Subatmospheric Condition
Authors: Lo Frano R. , Mazed D. , Aquaro D. , Del Serra D. , Sekachev I. , Giambartolomei G.
Published: October 2017

The phenomenon of steam-water direct contact condensation (DCC) was widely investigated in the past because of conventional and nuclear engineering applications, like in the safety suppression system of BWRs.

When steam is injected into a sub cooled water, different condensation regimes can be observed, such as chugging, condensation oscillation, bubbling condensation oscillation etc. (these have been identified during the experimental campaign on condensation of steam injected into water at sub-atmospheric conditions carried out at Lab. B. Guerrini of DICI - University of Pisa). According to the thermal hydraulic conditions of the condensing jet plume, vibrations may arise.

This paper so deals with the analysis of vibration phenomena that originate during the steam-water direct contact condensation at sub-atmospheric conditions, not treated till today. Numerical investigations (by proper FEM code) as well as experimental activity will be presented and discussed in order to evaluate if vibrations are capable to jeopardise this safety suppression system, which is designed to protect Vacuum Vessel of ITER from pressurizing accidents.
Sensitivity Analysis for Human Errors of a Chemical and Volume Control System at Nuclear Power Plant by Soft Control
Authors: Yuxin Zhang , Kunze Yang
Published: October 2017

The introduction of soft controls into the main control room may provide a more convenient environment for operations, but also may introduce new types of human errors and new risks into the nuclear power plants. In this paper, taking the Chemical and Volume Control System (CVCS) at pressurized water reactor nuclear power plant as an example, the task analysis of operating procedures for starting the CVCS under a normal shut-down operation condition, the potential failure modes during operators executing each step of operating procedures and the dependency analysis of sub-tasks are presented. Furthermore, the sensitivity analyses are conducted for identifying probability importance and critical importance of each model parameter. The countermeasures for preventing and reducing the human errors of soft controls are discussed.
Development of Numerical Simulation Method to Evaluate Heat Transfer Performance of Air Around Fuel Debris: Part 2 — Validation of JUPITER for Free Convection Heat Transfer
Authors: Uesawa Shinichiro , Yamashita Susumu , Shibata Mitsuhiko , Yoshida Hiroyuki
Published: October 2017

A dry method is one of fuel debris retrieval methods for decommissioning of TEPCO’s Fukushima Daiichi nuclear power station. If stopping the water leakage from a damaged primary containment vessel (PCV) is impossible, the dry method may be adopted. Besides, the dry method may be also adopted to retrieve the fuel debris in a pedestal under a reactor pressure vessel (RPV) because accessing and retrieval of the fuel debris from the lateral sides of PCV may be adopted for retrieval of the fuel debris in the pedestal. However, the decay heat must be removed from the fuel debris without water because fuel debris is exposed to the air in the dry method. Thus, we must evaluate the air cooling performance for the fuel debris in the dry method.

Japan Atomic Energy Agency (JAEA) has started the research project to develop an evaluation method by the JUPITER to understand the air cooling performance for fuel debris in the dry method. In the project, free convection heat transfer and the radiation heat transfer experiments for air are also conducted for validation of the JUPITER. In this paper, the free convection heat transfer, the radiation heat transfer, the conduction heat transfer and the Rayleigh number for the free convection in the simplified model were estimated to decide experimental conditions. We confirmed the free convection and the radiation heat transfer were important to estimate heat transfer of the debris exposed to the air in the RPV pedestal. Considering the estimations, the experiment for heat transfer and flow visualization of free convection adjacent to upward-facing horizontal surface was conducted to validate the calculation of the turbulent free convection by the JUPITER. From comparing between the calculation result and the experimental result, we confirmed that the calculation result of the JUPITER was qualitatively reasonable.
Challenges of Evolution for the Risk-Informed Decision Making Standard
Authors: Narumiya Yoshiyuki , Hayashi Kentaro , Kuramoto Takahiro
Published: October 2017

Risk-informed applications for the purpose of optimized decision making in safety related activities, including safety design and operational management of nuclear power plants, can be widely adopted by regulators, manufacturers as well as utilities. Utilization of risk information is expected to contribute to the achievement of safety improvement in operating nuclear power plants in an optimized way. In addition, effective utilization of risk information is essential to cope with the changes in safety related activities reflecting the latest experiences and findings for the continuous improvement of the safety of nuclear power plants into the future.

In this regard, the Standards Committee of the Atomic Energy Society of Japan is working on the establishment of the new standard specifying the basic requirements regarding risk-informed decision making (hereafter, called “RIDM”). This paper introduces the policies toward the establishment of the RIDM Standard.
Emergency Strategy Research of Loss of Coolant Accident for Small Modular Reactor
Authors: Yuqi Wang , Aimin Yu , Qingming Yang
Published: October 2017

This paper researched the behavior of 20mm break Loss of Coolant Accident (LOCA) which is located in the Direct Vessel Injection (DVI) line of the integrated small modular reactor (SMR) in case of full power with RELAP5-3KEYMASTER simulation system. The response of passive safety systems is analyzed and compared with the Primary Safety Analysis Report (PSAR) post-accident. Tendency for the variation of main parameters after the accident agree well with the PSAR, which validates the accuracy and rationality of the model, and solves the new problems in the process of modeling and provides an important tool for the research and development of SMR. Cooling and depressurization are calculated post-accident. The variation of main parameters post-accident and the accident advancement and results have been analyzed. Operation intervention is given and the effects with it are discussed. And the emergency strategy for development and verification of Emergency Operating Procedures (EOP) is given.
Development and Validation of Evaluation Method on Hypothetical Total Instantaneous Flow Blockage in Sodium-Cooled Fast Reactors and its Application to a Middle Size SFR
Authors: Fukano Yoshitaka
Published: October 2017

Local faults (LFs) have been considered to be of greater importance in safety evaluation in sodium-cooled fast reactors (SFRs) because fuel elements were generally densely arranged in the subassemblies (SAs) in this type of reactors, and because power densities were higher compared with those in light water reactors. A hypothetical total instantaneous flow blockage at the coolant inlet of an SA (HTIB) gives most severe consequences among a variety of flow blockages.

Although an evaluation on the consequences of HTIB using SAS4A code was also performed in the past study, SAS4A code was further developed by implementing analytical model of power control system in this study. An evaluation on the consequences of HTIB in Monju by this developed SAS4A code was performed in this study.

Furthermore SAS4A code was newly validated using an in-pile experiment which simulated HTIB events. The validity of SAS4A application to safety evaluation on the consequence of HTIB was further enhanced in this study.

It was clarified by the analyses considering power control system that the reactor would be safely shut down by the PPS triggered by either of 116% over power or DND trip signals. Therefore the conclusion in the past study that the consequences of HTIB would be much less severe than that of ULOF was strongly supported by this study.
The Sensitivity Simulation of Different Geometric Pretreatment in Method of Characteristics
Authors: XiaoRui Tian , Tao Zhou , Zichao Li , Tao Yu
Published: October 2017

In reactor core physics analysis,the research about the pre-processing of Method of Characteristic (MOC) including the generation and storage of characteristic line,the progress of calculation and the choosing of different quadrature set.In addition,doing some simulations,which is based on OpenMOC code and C5G7-MOX benchmark,about different parameters (including the track spacing,azimuthal angles and polar angles) and calculated its impacts on the computational efficiency and accuracy.the simulation results are as following:setting the track spacing as 0.1 cm or the azimuthal angle number as 4,the simulation results have better accuracy. Whether choosing the Leonard’s optimum quadrature set or the Tabuchi-Yamamoto quadrature set,the number of polar angles have tiny impact on accuracy.
High Temperature Oxidation of Zry-4 in Oxygen-Nitrogen Atmospheres
Authors: Negyesi Martin , Amaya Masaki
Published: October 2017

Zry-4 fuel cladding tubes were exposed in mixtures of oxygen and nitrogen at temperatures of 800–1380 °C. The influence of various flow rates of oxygen and nitrogen as well as specimen height on the weight gain was examined. Metallographic observations were carried out and residual mechanical properties were assessed employing ring compression tests (RCT). The overall weight gain was substantially affected by both the applied flow rates and the height of specimens. The oxidation kinetics in air was assessed based on the results of weight gain measurements. A transition in the kinetics was observed at 800 and 1000 °C. The kinetics in the post-transient regimes was rather accelerated than linear. The equation proposed in this study for air condition was in good agreement with the Leistikow-Berg correlation for air condition and the Baker-Just correlation for steam condition. Prior β-phase shrinked when the oxide scale along with the α-Zr(O) layer progressed. Eventually, both the specimen plastic strain and maximum load decreased due to the shrinkage and increasing embrittlement of the prior β-phase.
Numerical Investigation of Melt Jet Breakup With Different Shapes in Water Pool
Authors: Cheng Hui , Zhao Jiyun
Published: October 2017

During a severe accident in nuclear power plant, core damage may occur due to decay heat and molten fuel can pour into and interact with water resulting in steam explosion. The energetics of steam explosion strongly depends on the initial premixing stage during which the molten fuel undergoes a coarse fragmentation process, which determines the surface area for fuel-coolant contact and heat transfer. Extensive research has been done to understand the premixing stage, however, most of the studies are focused on the cylindrical jet interaction with water. In fact, during core melt, the molten fuel may pour near the edge of core, so the shapes and size of melt jet may differ significantly based on specific conditions. In this paper, numerically study on the melt jet breakup with different shapes in pool water are conducted, such as elliptical shape with VOF method. Firstly, the deformation of molten jet under the same conditions in 2D model is compared with 3D model and shows that the breakup of 3D model is quite different from 2D model, the integration of 3D model is maintained much better than 2D model. Then the characteristics of breakup of elliptic cylindrical melt jet are analyzed and compared with cylindrical melt jet. The results shows that the interface surface area of elliptic cylindrical jet is nearly twice the cylindrical jet.
Feasibility and Implementation of the Modified Clay Technique in Control of Phaeocytstis Globosa Blooms in the Water-Intake Area of the Fangchenggang Nuclear Power Plant
Authors: Yu Zhi-ming , Wang Kai , Cao Xi-hua , Wang Bai-zhong , Song Xiu-xian , Xu Bo-tao , Yu Fan , Xu Xin
Published: October 2017

In the current study, we demonstrated the effectiveness of using modified clay as an emergency treatment for controlling Phaeocystis globosa blooms in terms of the removal mechanism, removal rate in both the laboratory and field, and the engineering required. Based on these results, we developed a series of blueprints, including the monitoring and early warning of P. globosa blooms in the water-intake area of the Fangchenggang nuclear power plant (NPP), the manufacture of the modified-clay emergency disposal facilities, and the operation schemes in treating different scale of blooms. From December 2014 to February 2015, modified clay was applied to successfully control P. globosa blooms and to effectively protect the cold-water source of the NPP. Thus, a set of engineering solutions using modified clay was established and confirmed to be a viable approach for treating harmful algae blooms that threaten the cold-water sources of NPPs.
Screening Key Parameters for Passive System Based on Morris Method
Authors: Jiang Lizhi , Cai Qi , Zhang Yongfa , Shi Hao
Published: October 2017

Thermal-hydraulic reliability (TH-R) analysis for passive system requires repeated runs of thermal-hydraulic models or codes to get enough input/output data. Thermal-hydraulic models or codes of passive system are costly in terms of running time which results in slow computational speed and low estimation efficiency. To improve the estimation efficiency, surrogate models (or response surface) can be used to replace the real thermal-hydraulic model. During response surface construction, there are many input parameters, some undetected interactions or nonlinear effects involved in thermal-hydraulic model, which debase the accuracy and validity of the response surface. As a global sensitivity analysis method, Morris method is to determine factors which are important, nonlinear and non-additive, or involved in interactions with other factors. In this paper, screening key parameters for passive system is studied based on Morris method. The results of Morris method are insightful to select a reasonable surrogate model according to the interactions or non-linear characteristics of the model in the process of response surface construction. As a result, estimation efficiency of TH-R analysis for passive system can be improved greatly while ensuring the response surface’s accuracy.
A Study of Power Decrease Strategy on CPR1000 Full Scope Simulator
Authors: Zhang Zhenhua , Zhang Bo , Wang Qingsong , Wang Chengzhi , Xing Dan
Published: October 2017

Normally, nuclear power plants operate at 100% full power level. However, in some special cases, nuclear power plants need to decrease power and operate at lower power level. For example, because of the typhoon, test and equipment maintenance requirements etc., nuclear power plants decrease power to specific power level and stay for several hours at this power level. Because there is inserted time limit of the power compensation rod (G rod) in the technical specification, inserting control rod to decrease power cannot always be chosen, and the strategy that decreases power by boron injection should be taken. In this paper, powers down strategies are introduced. And the similarities and differences between decreasing power by inserting control rod and decreasing power by injecting boron are also introduced. Before reducing power, reactivity equilibrium calculation should be performed and the predicted amount of boron injection can be calculated. Then according to the power down rate, Boron injection rate can be estimated. Finally, Boron injection operation can be performed. During the power reducing, R rod and some important equipment associated with boron injection should be monitored closely, especially the drain tank water level of condensate extraction system (CEX) and the response of feedwater flow control system (ARE). A key control of power down is to control the axial power deviation (delta I) that is one of the most difficult technical problems. Delta I must be kept in the functional area and it is inhibited that the operating point goes outside limits (left limit and right limit) of region II in any situations. It is performed on CPR1000 full scope simulator (FSS) to decrease power by inserting control rod and boron injection in this paper. Several curves obtained during power reducing are analyzed in detail. According to analysis results, the key procedure, operation risks and operation control during power decreasing are proposed as reference. The purpose of them is to meet the requirements of nuclear safety during power down operation.
Accessibility Analysis for Staff After Accident of PWR Unit
Authors: Zhang Ye , Wang Xiaoxia , You Wei , Li Zhuoran , Zhang Liying , Mi Aijun
Published: October 2017

Severe accident has become one of the main directions of research since the crisis at Fukushima plants in Japan, including release of radioactive material, accessibility analysis for staff and evaluation of consequence. This paper, mainly for design basic accident and severe accident, makes calculation of migration and release of radioactive material after accident by considering the different building, also combine the on-site operation requirement of worker after accident, analyzes dose rates of typical zone for staff and evaluate the exposure caused by radioactive material. The main results of the paper supply reference and basis for person accessibility research after design basis accident and severe accident.