bluuebunny/crossref_metadata_2025 huggingface dataset:
A Study About Safety I&C System Software V&V in Nuclear Power Plant
Authors: Gu Pengfei , Wang Shengchao , Chen Weihua , Yu Suyuan
Published: October 2016

Since the digital instrument and control (I&C) system was used in nuclear power plant (NPP), software Verification and Validation (V&V) is becoming more and more important for ensuring the safety function to be correctly implemented by the safety system. According to the classification of different safety functions of I&C system, with the use of digital technology, software classification needs further discussion to determine. And different classification of software performs different testing. Software V&V processes can be divided into different stages by reference to the standard IEEE 1012-2004, and each stage focuses on different contents. In China, software V&V on the NPP has been started from LingAo Phase II project and strongly done on other CPR1000 projects, which is a Generation II + pressurized water reactor. Based on the practice about YangJiang units 5 and 6 projects, combined with the relevant laws and regulation standards, this study summarizes the characteristics of the independent third party software V&V, analyzes the key points and methods of V&V activities in the software development process. As a result, it is also benefit to the design, operation and maintenance of safety I&C System as technical reference.
Nuclear Systems for a Low Carbon Electrical Grid
Authors: Fathi Nima , McDaniel Patrick , Forsberg Charles , de Oliveira Cassiano
Published: October 2016

The intermittency of renewable power generation systems on the low carbon electric grid can be alleviated by using nuclear systems as quasi-storage systems. Nuclear Air-Brayton Combined Cycle systems can produce and store hydrogen when electric generation is abundant and then burn the hydrogen by Co-Firing when generation is limited. The rated output of a nuclear plant can be augmented by several hundred per cent by Co-Firing. The incremental hydrogen to electricity efficiency can far exceed that of hydrogen in a stand-alone gas turbine.
Verification of the AP1000 Nuclear Power Plant Simulator NPPSim Against RELAP5 for Transient Analysis
Authors: Wu Shifa , Wang Pengfei , Fang Huawei , Wei Xinyu , Zhao Fuyu
Published: October 2016

To thoroughly investigate the AP1000 reactor system, an AP1000 Nuclear Power Plant Simulator (NPPSim) was built and applied to the AP1000 transient and control system analysis in MATLAB/Simulink environment. In this study, a 100%–90%FP load step transient condition of AP1000 PWR was simulated by both NPPSim and RELAP5 simultaneously to verify the NPPSim’s capability of transient analysis. To realize this objective, a RELAP5 model of AP1000 PWR was firstly developed and relevant control systems were implemented in the input deck of RELAP5. Then the transient was performed to conduct this verification under the same conditions. The comparisons show a good agreement in terms of the changing trends. In addition, the deviations of the NPPSim simulation results and the RELAP5 calculating results are in the acceptable range. Therefore, the NPPSim is capable to implement the 100%–90%FP load step transient analysis for AP1000. More transient conditions should be simulated and studied for further verification.
Results of the Loss of Offsite Power Events Analysis
Authors: Volkanovski Andrija , Ballesteros Avila Antonio , Peinador Veira Miguel
Published: October 2016

This paper presents the results of statistical and engineering analysis of Loss of Offsite Power (LOOP) events registered in four reviewed databases. The paper includes events registered in IRSN (Institut de Radioprotection et de Sûreté Nucléaire) SAPIDE and GRS (Gesellschaft für Anlagen- und Reaktorsicherheit mbH) VERA database in time period 1992 to 2011. The Nuclear Regulatory Commission (NRC) Licensee Event Reports (LERs) database and the IAEA International Reporting System (IRS) database are screened for the relevant events registered in period 1990 to 2013.

In total, 228 relevant events were identified in the IRSN database, 190 in GRS, 120 in LER and 52 in IRS. The data include events registered both during the critical (at power) and shutdown operation of the plants. The identified events were classified considering nine different categories.

In the three databases (SAPIDE, VERA, IAEA-IRS) the largest numbers of events are registered for the plant centered category. The largest number of the events in the NRC-LER database is found for switchyard centered events. According to the mode of operation, most events were reported during critical power operation, in all four databases. The “Partial loss of external power” events are the most frequent type of event found in the IRSN and NRC databases while the “Physical loss of electrical busbars” is the main type in the GRS and IAEA databases. The largest number of events in all databases is identified for the switchyard failures followed by the interconnections failures (both lines and transformers). Mainly LOOP event are identified by the fault report in the control room. Electrical deficiency is detected as the main direct cause of events. Environment is registered as the main contributor for the electrical grid deficiency in the French and NRC databases. Electrical failures are dominant contributor to the electrical grid deficiency in the German and IAEA databases. The principal root cause for the LOOP events are human failures with the human errors during test, inspection and maintenance as the largest sub-group. The largest number of the LOOP events resulted in reactor trip followed by the Emergency Diesel Generator (EDG) start. The majority of the reported LOOP events lasted for more than 2 minutes.

Main lessons learned from the analysed events and potential actions for decrease of the number of LOOP events are presented.
Reliability of PCCS in AP1000 Based on Accident Development
Authors: Yu Yu , Wang Shengfei , Niu Fenglei
Published: October 2016

Passive containment cooling system (PCCS) is an important safety-related system in AP1000 nuclear power plant, by which heat produced in reactor is transferred to the heat sink – atmosphere – based on natural circulation, independent of human response or the operation of outside equipments, so the reactor capacity of resisting external hazards (earthquake, flood, etc.) is improved. However since the system operation based on natural circulation, many uncertainty factors such as temperatures of cold and heat sources will affect the system reliability, and physical process failure becomes one of the important contributors to system failure, which is not considered in the active system reliability analysis. That is, the system will lose its function since the natural circulation cannot be established or kept even when the equipments in the system can work well.

The function of PCCS in AP1000 is to transfer the heat produced in the containment to the environment and to keep the pressure in the containment below its threshold. After accidents the steam is injected to the containment and can be cooled and condensed when it arrives at the containment wall, then the heat is transferred to the atmosphere through the steel vessel. So the peak value of the pressure is influenced by the steam situation which is injected into the containment and the heat transfer and condensate processes under the accidents. In this paper the dynamic thermal-hydraulic (T-H) model simulating the fluid performance in the containment is established, based on which the system reliability model is built. Here the total pressure in the containment is used as the success criteria. Apparently the system physical process failure may be related to the system working state, the outside conditions, the system structure parameters and so on, and it’s a heavy work to analyze the influences of all the factors, so only the effects of important ones are included in the model. Monte Carlo (MC) simulation is used to evaluate the system reliability, in which the input parameters such as air temperature are sampled based on their probabilistic density distributions. The pressure curves along with the accident development are gained and the system reliabilities under different accidents are gotten as well as the main contributors. The results illustrate that the system physical process failure probabilities are varied under different climate conditions, which result in the system reliability and the main contributors to system failure changing, so the different methods can be taken to improve the system reliability according to the local condition of the nuclear power plant.
Effect of Waste Slurry Composition on the Radiolytic Stability of Polyethylene for Isolation of Waste From Water Treatment After the Fukushima Daiichi NPS Accident
Authors: Barr Logan D. , Idesaki Akira , Taguchi Mitsumasa , Yamaguchi Tetsuji
Published: October 2016

The gamma radiolysis of simulated waste environments has been investigated to represent the environment of waste containers used following the removal of radionuclides from contaminated water released following the incident at the Fukushima Daiichi Nuclear Power Station in 2011. The effect of the solid flocculants on the yield of oxidative species formed during radiolysis was investigated to determine the potential for polymer degradation induced by oxidative radicals. Changes to the properties of polyethylene induced by gamma radiolysis in the presence of the different waste types have also been investigated. The aqueous environment was shown to alter the hydrogen peroxide yields, in all cases G(H2O2) was reduced relative to the irradiation of pure water. Bulk polyethylene degradation was investigated by gel fraction measurement, the aqueous environment was found to have little effect due to the low rate of diffusion of water into the polymer.
Influence of Wall Sub-Cooling on Local Heat Transfer for Forced Convection Condensation of Steam/Air in a Horizontal Tube
Authors: Xu Huiqiang , Sun Qiunan , Gu Haifeng , Hou Xiaofan , Sun Zhongning
Published: October 2016

For the purpose of analyzing the influence of wall sub-cooling on condensation heat transfer characteristic in the presence of noncondensable gases inside a horizontal tube, experiments for air-cooling and water-cooling at the secondary side outside the condenser tube have been conducted. By comparing the experimental data of different inlet air mass fraction, mixture gases velocity and coolant volume flow rate, the variation of local heat transfer coefficient with wall sub-cooling was obtained. The results show that for annular and wavy flow, the condensation heat transfer coefficient increases with increasing wall sub-cooling but decreases for stratified flow. For annular and wavy flow, the positive influence of wall sub-cooling on condensation heat transfer coefficient is enhanced by the rise of inlet noncondensable gas mass fraction, mixture gases velocity and pressure.
Corrosion of Hot-Dip Galvanized Containment Installations: A Potential Cause for Thermal-Hydraulic Effects After LOCA in PWR?
Authors: Kryk Holger , Harm Ulrich , Hampel Uwe
Published: October 2016

Generic investigations regarding the influence of corrosion processes of hot-dip galvanized PWR containment installations on strainer clogging as well as on the coolant chemistry and possible resulting in-core effects are carried out within joint research projects of the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), TU Dresden (TUD) and Zittau-Görlitz University of Applied Sciences (HSZG). Lab-scale experiments at HZDR and TUD are focused on elucidation of physico-chemical corrosion and precipitation processes as well as resulting clogging effects.

Results of generic experiments in a lab-scale corrosion test facility suggest that there is a multi-stage corrosion process. The first stage comprises dissolution of the zinc layer in the coolant forming zinc ions and in turn affecting the coolant chemistry. During the second stage, the base material (steel) corrodes forming insoluble corrosion particles, which can subsequently lead to accelerated clogging of fiber-laden strainers within a few hours. The main influences on corrosion were identified as impact of the coolant jet onto the corroding surface, water chemistry and zinc surface / coolant volume ratio.

Furthermore, retrograde solubility of zinc corrosion products in boric acid containing coolants with increasing temperature was observed. Thus, formation and deposition of solid corrosion products cannot be ruled out if zinc containing coolant is heated up during its recirculation into hot downstream components (e.g. hot-spots in core). Corrosion experiments, which included formation of corrosion products at a heated cladding tube, proved that zinc, dissolved in the coolant at low sump temperatures, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces. Due to alternating heating and cooling of the coolant during sump recirculation operation, a cycle of zinc corrosion and zinc borate precipitation may be initiated, which may eventually influence the thermal hydraulics in downstream components during the post-LOCA stage. The results obtained at lab-scale were confirmed by corresponding experiments in semi-technical test facilities of the project partner HSZG.

Based on the experimental results, water chemical measures were tested to reduce corrosion and/or zinc borate precipitation effects.
A High Effective Parallel Method for the Coupling Between Neutronics and Thermal-Hydraulic
Authors: Dong Xiaomeng , Zhang Zhijian , Tian Zhaofei , Li Lei , Chen Guangliang
Published: October 2016

Multi-physics coupling analysis is one of the most important fields among the analysis of nuclear power plant. The basis of multi-physics coupling is the coupling between neutronics and thermal-hydraulic because it plays a decisive role in the computation of reactor power, outlet temperature of the reactor core and pressure of vessel, which determines the economy and security of the nuclear power plant. This paper develops a coupling method which uses OPENFOAM and the REMARK code. OPENFOAM is a 3-dimension CFD open-source code for thermal-hydraulic, and the REMARK code (produced by GSE Systems) is a real-time simulation multi-group core model for neutronics while it solves diffusion equations. Additionally, a coupled computation using these two codes is new and has not been done. The method is tested and verified using data of the QINSHAN Phase II typical nuclear reactor which will have 16 × 121 elements. The coupled code has been modified to adapt unlimited CPUs after parallelization. With the further development and additional testing, this coupling method has the potential to extend to a more large-scale and accurate computation.
Benchmark Accelerated Aging of Harvested Hypalon/EPR and CSPE/XLPE Power and I&C Cable in Nuclear Power Plants
Authors: Duckworth Robert C. , Frame Emily , Fifield Leonard S. , Glass Samuel W.
Published: October 2016

As part of the Light Water Reactor and Sustainability (LWRS) program in the U.S. Department of Energy (DOE) Office of Nuclear Energy, material aging and degradation research is currently geared to support the long-term operation of existing nuclear power plants (NPPs) as they move beyond their initial 40 year licenses. The goal of this research is to provide information so that NPPs can develop aging management programs (AMPs) to address replacement and monitoring needs as they look to operate for 20 years, and in some cases 40 years, beyond their initial, licensed operating lifetimes. For cable insulation and jacket materials that support instrument, control, and safety systems, accelerated aging data are needed to determine priorities in cable aging management programs.

Before accelerated thermal and radiation aging of harvested, representative cable insulation and jacket materials, the benchmark performance of a new test capability at Oak Ridge National Laboratory (ORNL) was evaluated for temperatures between 70 and 135°C, dose rates between 100 and 500 Gy/h, and accumulated doses up to 200 kGy. Samples that were characterized and are representative of current materials in use were harvested from the Callaway NPP near Fulton, Missouri, and the San Onofre NPP north of San Diego, California. From the Callaway NPP, a multiconductor control rod cable manufactured by Boston Insulated Wire (BIW), with a Hypalon/ chlorosulfonated polyethylene (CSPE) jacket and ethylene-propylene rubber (EPR) insulation, was harvested from the auxiliary space during a planned outage in 2013. This cable was placed into service when the plant was started in 1984. From the San Onofre NPP, a Rockbestos Firewall III (FRIII) cable with a Hypalon/ CSPE jacket with cross-linked polyethylene (XLPE) insulation was harvested from an on-site, climate-controlled storage area. This conductor, which was never placed into service, was procured around 2007 in anticipation of future operation that did not occur.

Benchmark aging for both jacket and insulation material was carried out in air at a temperature of 125°C or in a uniform 140 Gy/h gamma field over a period of 60 days. Their mechanical properties over the course of their exposures were compared with reference data from comparable cable jacket/insulation compositions and aging conditions. For both accelerated thermal and radiation aging, it was observed that the mechanical properties for the Callaway BIW control rod cable were consistent with those previously measured. However, for the San Onofre Rockbestos FRIII, there was an observable functional difference for accelerated thermal aging at 125°C. Details on possible sources for this difference and plans for resolving each source are given in this paper.
Load-Following Control of Nuclear Reactors Based on Sliding Mode Method
Authors: Liu Yiwen , Xia Hong
Published: October 2016

Load-following control is one of the key techniques that guarantee the safety, stability and efficient operation of the nuclear reactors. In this paper, a variable structure control system based on sliding mode method is designed in order to improve of the pressurized water reactor’s load-following capability. According to the reactivity influential factors during the load-following process, nonlinear point reactor kinetics equations are established. The state observer is introduced to solve the problem that parts of the state variables cannot be measured. Finally, the sliding mode controller is presented via designing a nonlinear sliding mode surface and selecting an appropriate reaching law, which can reduce the chattering phenomenon. It is theoretically proved that this newly-built controller not only provides globally asymptotic closed-loop stability but also is adaptive to the system uncertainty. The simulation results demonstrate that the control system can restrain the overshoot effectively, increase the response speed, and has greater robustness to the external disturbances.
Neutron Up-Scattering Effect in Refined Energy Group Structure
Authors: He Qingming , Wu Hongchun , Li Yunzhao , Cao Liangzhi , Zu Tiejun
Published: October 2016

Aiming at generating a 361-group library, this paper investigated neutron up-scattering effect in the 361-group Santamarina-Hfaiedh Energy Mesh (SHEM). Firstly, the Doppler Broadening Rejection Correction (DBRC) method is implemented to consider the neutron up-scattering effect in Monte Carlo (MC) method. Then the MC method is employed to prepare resonance integral table and scattering matrix for afterward calculation. Numerical results show that the neutron up-scattering affects kinf by ∼200 pcm at most for UO2 pin cell problems in the 361-group SHEM, while the fuel temperature coefficient (FTC) is also influenced by 12∼13%. It has also been found that both of the above two influences acts through scattering matrix rather than self-shielded absorption cross sections. In addition, the self-shielding effect of cladding is studied and it’s been found that it affects kinf by 30∼70 pcm.
Evaluation of Jet Impact Region and Fluid Force Generated From Ruptured Pipes: 3 — Evaluation of Established Standards
Authors: Xu Qiang , Takahashi Shiro , Takamura Noriyuki , Morita Ryo , Uchiyama Yuta , Watanabe Shun
Published: October 2016

Nuclear power plants are designed to avoid damage to their safety installations because of jet impingement when a pipe is ruptured. We have investigated the evaluation method for the design basis of protection of plants against effects of a postulated pipe rupture using the established standard of the American National Standards Institute (ANSI). The steam jet tests using particle image velocimetry and computational fluid dynamics (CFD) analysis were conducted for this verification. Spread angle of steam jet could be visualized. The jet spread angles were lower than 10 deg which was described in the ANSI standard. The ANSI standard was conservative for evaluations of the jet impact region compared to actual test and CFD results. However, it is desirable to use the conservative angle for evaluation of the jet fluid force. We could adequately evaluate the conservative jet fluid forces by the ANSI standard considering the spread suppression region and jet spread angle of 6 deg.
Accurate Estimation of Vaporization Rate in Subcooled Flow Boiling Based on the Results of Visualization Experiment
Authors: Kaiho Kazuhiro , Enoki Koji , Okawa Tomio
Published: October 2016

This research presents experimental approaches to accumulate the data for mechanistic model in subcooled flow boiling. A number of photographic studies have been provided to investigate phenomena of bubble nucleation and condensation process for accurate prediction of void fraction such as bubble detachment diameter, bubble detachment frequency and nucleation site density in a subcooled flow boiling. In this work, a transparent heated surface was used to obtain the data from back side of heated surface to avoid overlapping bubbles by using high speed video camera. It enabled to observe bubble nucleation process and active nucleation sites. The experiment was performed in a vertical rectangular channel at atmospheric pressure and the water was used as test fluid. In generally, the computational analysis for a subcooled flow boing deal with mean bubble size as the size of bubbles produced on heated surface. Although, it was found that mean bubble size can represent bubbles produced at same site because they are almost uniform size. Even though the size of bubbles at same site are almost uniform, the difference of the size of bubbles between other sites are considerable value. Therefore, mean bubble size on the surface should not represent bubbles for all site otherwise the serious error may be caused. It seems that bubble detachment diameter should not be given by correlations of mean bubble detachment diameter for accurate prediction of vaporization rate. Some researchers proposed that bubble size distribution should be considered by Gaussian distribution [1–3]. However, it found that bubble size distribution data accumulated in this work cannot be fitted by Gaussian distribution and there are probability that larger bubbles are neglected due to the configuration features of Gaussian distribution. So, Gamma distribution was used to predict the bubble size distribution and it was evaluated in terms of heat flux, wall superheat, mass flux and liquid subcooling. And then, by the experimental approaches, the important dimensionless parameters are identified such as Nusselt number, Jakob number, Reynolds number and dimensionless subcooling. Furthermore, vaporization rate was calculated by correlations of bubble detachment diameter, bubble detachment diameter and nucleation site density and compared with the data. Finally, the effect of using mean bubble size or bubble size distribution on vaporization rate was investigated.
The Generation of Few-Group Constants for Fast Reactor Analysis
Authors: Du Xianan , Cao Liangzhi , Zheng Youqi
Published: October 2016

A way to generate the few-group cross sections for fast reactor calculation is presented in this paper. It is based on the three steps computational scheme. In the first step, the ultrafine method is used to solve the slowing down equation based on the ultrafine group cross section generated by NJOY. Optional 0D or 1D calculation is used to collapse energy group into broad energy groups. In the second step, the 2D RZ calculation using SN method is performed to obtain the space dependent neutron spectra to collapse broad energy groups into few groups. The anisotropic scattering is well handled by the direct SN calculation. Finally, the full core calculation is performed by using the 3D SN nodal method. The results are compared with continuous energy Monte-Carlo calculation. Both the cross section generated in the first step and the final keff in the last step are compared. The results match well between the three steps calculation and Monte-Carlo calculation.
Experimental Investigation on Analogous U-Shaped Tube Water Seal in Front of Pressurizer Safety Valve
Authors: Bing Tan , Wu Y. W. , Xiang Y. , Shi LeiTai , Tian W. X. , Qiu S. Z. , Su G. H.
Published: October 2016

Water formation test loop was constructed with a scale ratio of 1:1 to investigate the condensation and analogous U-shaped tube water seal in front of pressurizer safety valve. The water seal features were investigated under different pressure and various concentrations of non-condensable gases. The test data reveals that water seal can completely formed even in the presence of non-condensable gases in high content. Water formation time reduced when increased system pressure and refilling of the pipe is much more rapid than the initial one due to a lower content of non-condensable gases remaining in the steam phase. A simple criterion to determine the water seal starting point has been put forward in this paper. The strong natural convection in the water seal keeps the temperature difference lower than 20K.
Evaluation of Jet Impact Region and Fluid Force Generated From Ruptured Pipes: 1 — Numerical and Experimental Evaluation of Affected Region by Steam Jet
Authors: Morita Ryo , Uchiyama Yuta , Watanabe Shun , Takahashi Shiro , Xu Qiang , Takamura Noriyuki
Published: October 2016

Nuclear power plants are designed to prevent the damage of safety installations and human safety due to the jet impingement when a pipe is ruptured. We have investigated evaluation methods for the design basis of protection of plants against effects of postulated pipe rupture using computational fluid dynamics analysis (CFD). The steam jet tests using particle image velocimetry were conducted in order to verify the CFD methods. Spread of steam jet could be visualized. Shapes of steam jet obtained by CFD were almost the same as those by tests. The spread angle of free jet were investigated using CFD, and we have found that the spread angle strongly depends on the inlet pressure. This is thought as the different tendency from Established Standards.
Development and Application of the 3-D Core Fuel Management Transport Program for HFETR
Authors: Zhu Lei , Sun Shouhua , Xu Taozhong
Published: October 2016

The High Flux Engineering Test Reactor (HFETR) is loaded with complex geometry assemblies, taking on strong heterogeneity. These features call for more advanced core analysis and fuel management methodologies compared with traditional PWR. Based on the two-step method, a fuel management simulation platform HEFT was developed, which is a 3-D core fuel management transport calculation program of HFETR. In HEFT, the lattice code HEFT-lat has the abilitiesto describe complex assembly geometriesof HFETR, whereby mitigate errors brought by geometrical approximations.DNTR, a triangular mesh nodal SN method code, is employed to solve the transport equation. This paper introduces the development backgrounds of HEFT, introduces the principle and function of the three main modules of HEFT, that is HEFT-lat?HEFT-core and HEFT-int, the calculation method of lattice parameter and reactor loop calculation method are described. By follow-up calculation of 81-I to 89-II of HFETR core and comparison of intial critical keff, the control rod position during shutting down and the neutron flux, thus obtains the calculation deviation. The result shows the validity of the HEFT code, the lattice and reactor core calculation model. Thus, HEFT has already successfully applied to the fuel management calculation of HFETR.
CFD Analysis of Cross Flow in Rod Bundles Under Rolling Motion
Authors: Liu Di , Tian Wenxi , Qiu Suizheng , Su G. H.
Published: October 2016

A CFD analysis of cross flow in rod bundles in rolling motion was performed to investigate the effect of rolling motion on the flow behavior between the subchannels. The rolling motion was assumed as a sinusoid. The additional forces due to the rolling motion including azimuthal force, centrifugal force and coriolis force were added into the source term in the momentum equation. A transient three dimensional simulation of square rod bundles model was performed in various rolling conditions. In order to precisely predict secondary flow patterns in rod bundles, Reynolds Stress Model was selected as the turbulent model. Effect of various rolling parameters such as rolling velocity and amplitude on the cross mixing was investigated. The results show that cross flow is strongly affected by the rolling motion. The local cross flow field in rolling motion was showed in detail. Also, the mixing coefficients based on CFD results were calculated. Empirical correlations of turbulent mixing were modified to consider the effect of rolling conditions, which can be used in the traditional subchannel thermal hydraulic code.
Numerical Investigation on Critical Submerged Depth During Draining From Cylindrical Tanks
Authors: Shi-wen Liu , Jie Zhou , Wei Huang , Xiang Cheng
Published: October 2016

Gas ingestion arises from free surface dip during draining from liquid storage tank does serious damage to the nuclear engineering practices, which should be considered in the design work. In this article, numerical simulations of transient process of draining from cylindrical tanks have been conducted with volume-of-fluid (VOF) method, by using the commercial CFD code FLUENT14.5. The relation between critical submerged depth and the Froude number has been investigated, then the influence of different factors on critical submerged depth also has been discussed. The results show that the Jain’s empirical formula are available to estimate the critical submerged depth flow with weak vortex in the condition of high diameter ratio and high initial water level. Moreover, the initial strong vortex significantly increase the critical submerged depth, while the surface tension has pretty small effects on the critical submerged depth when the Froude number is large.
FMEDA-Based NPP I&C Systems Safety Assessment: Toward to Minimization of Experts’ Decisions Uncertainty
Authors: Yasko Alexander , Babeshko Eugene , Kharchenko Vyacheslav
Published: October 2016

There are many techniques for the Nuclear Power Plants Instrumentation and Control (NPP I&C) systems safety assessment. Each of them is suitable for specific types of I&C systems and life cycle stages. Though general procedures of techniques application are specified by standards and described by guides, there is no universal solution that could be unambiguously applied to any NPP I&C system.

The Failure Modes, Effects and Diagnostics/Criticality Analysis (FME(D/C)A) is the one that is most often used. Using this technique, the process of assessment is not trivial because of dimensionality problem that is especially critical for modern NPP I&C systems that contain many complex electronic components. Another challenge is the need of compliance to varying requirements of standards. Furthermore, modern I&C systems are based on different platforms (FPGA, microcontrollers).

Most of safety and reliability assessment techniques, including mentioned FME(D/C)A, are based on expertise and thereby results are dependent on experts’ decisions very much. This could be a serious challenge, because it is very difficult to find universal experts that have sufficient experience to be equally qualified in different electronic components (FPGA, microcontrollers etc.) used in modern I&C systems.

The goal of this paper is to analyze the ways of automation of FMEDA-based NPP I&C systems safety assessment and minimization of uncertainty degree of experts’ decisions.

Possible experts’ errors and the uncertainty degree of their decisions are analyzed. We propose integration of all existing FMEA-based techniques into XME(D/C)A that includes Functional FMEA, Design FMEA, Software FMEA, Hardware FMEA etc. Such approach allows performing more comprehensive analysis. Developed tool AXMEA (Automated XMEA) represents an integrated solution that provides the automation of stages of FMEDA technique applied to NPP I&C.

The case study is the application of proposed technique and tool during SIL3 certification of the modular RadICS™ platform.
Computer Simulation of Flow and Heat Transfer in Bare Tubes at Supercritical Parameters
Authors: Zvorykin Alexander , Aleshko Sergey , Fialko Nataliia , Maison Nikolay , Meranova Nataliia , Voitenko Artem , Pioro Igor
Published: October 2016

This paper deals with CFD predictions for flow and heat transfer in supercritical water in a bare tube. Studies were performed using the software FLUENT for upward flows in vertical tubes with heated length of 4 m and an inner diameter of 10 mm at high values of mass flux (G > 1000 kg/m2s). Turbulence models verification data for the given physical conditions are presented. Besides the testing of different turbulence models that are presented in modern catalog of these models is carried out. Namely, the models related to the following three groups: High–Reynolds number k-ε models with wall functions, k-ω models and Low-Reynolds number k-ε models were considered. On the basis of performed studies the best compliance of known experimental data with computer simulation results fits the k-ω SST turbulence model is shown.
Control Policy for Shared Objects in HTR-PM: Toward the Collision and Deadlock Avoidance
Authors: Zhou Shuqiao , Huang Xiaojin , Qianqian Jia
Published: October 2016

High Temperature gas-cooled Reactor Pebble-bed Module (HTR-PM) is a high temperature gas-cooled reactor demonstration plant with a structure of two modules feeding one steam turbine. Compared with the structure of a single reactor feeding a turbine, there are more devices shared between these two modules. When they are operated, the shared components are prone to introduce collisions or even logical deadlocks for different technical processes. The future commercial HTR-PM plants are supposed to comprise more modules for a larger turbine, thus the collision problem and potential deadlocks introduced by the shared components will become severer, which may impact the efficiency or even the functions of the related systems. Therefore, how to design suitable control policies in the distributed control system (DCS) to relieve the collisions and avoid the deadlocks during using these shared devices is a new and also a very important problem. In this paper, the classifications of the shared devices are first addressed, and then how to identify the shared objects of nuclear power plant (NPP) is proposed. Furthermore, a general model for the control logic design is proposed, taking into consideration the collision avoidance, time delay and fairness. Moreover, a scheme for deadlock avoidance is proposed. The example of how to apply the schemes to relieve the conflicts and deadlocks in the processes of using the shared devices in fuel element cycling system is illustrated.
Reliability Analysis and Availability Enhancements for Redundant Systems With Reparable Components
Authors: Zhou Shuqiao , Li Duo , Guo Chao
Published: October 2016

Redundancy is widely used for enhancing a system’s overall availability. As an HTR demonstrated plant, a high temperature gas-cooled reactor pebble-bed module (HTR-PM) now is under construction in China and the construction will be completed around 2017. In HTR-PM, there are many devices and device groups used in a redundant way to guarantee the high availability of the related functions, especially the functions shared by two reactors during the entire life time. It is very important and necessary to determine their reliabilities as well as how to make a decision about the related maintenance policies to enhance their availabilities. In this paper, typical redundant styles in the HTR-PM are summarized and demonstrated. Accordingly, the theoretical models, which are able to describe the reliabilities of the redundant systems, are proposed based on Markov chain model. Moreover, for a specific redundant structure, the relationship between the availability and the maintenance period is analytically addressed. Based on the model, we address that: as the digital monitoring and control technologies are widely used in nuclear power plants, monitoring methods targeting for decreasing maintenance costs and meanwhile increasing the availabilities for different redundant styles are very beneficial.
Effect of Bolt Clamping Force and Constraint Arrangement on Structural Strength of Bolted Joint
Authors: Mao Jianfeng , Bao Shiyi , Lijia Luo , Gao Zengliang
Published: October 2016

In this article, the effect of bolt clamping force and constraint arrangement on structural strength of bolted joint was investigated by finite element method (FEM) prior to hardware tests. This study developed a numerical simulation to predict the deformation behavior and detect potential failure modes. In achieving it, a three dimensional (3D) detailed model of bolted joint was constructed. FE dynamic simulation was used to simulate the structural behavior of the bolted joint by gradually applying tension force on the ends. The numerical simulations were conducted with the torque of 0.3N.m, 3.0N.m, 6.0N.m at different tensile levels in several frictional states between contact plates. In order to determine the critical friction force on the plates, three kinetic frictions between bolt and plate hole were employed in the FE calculation to detect the shear stress on the bolt. Finally, the structural behavior of the bolted joint was analyzed in terms of stress distribution, deformation state by varying clamping force and frictional coefficient.
The Influence of the Wall Stored Energy on the Natural-Circulation Mode in the Core Make-Up Tank Test Facility
Authors: Zhu Mingzhang , Chang Huajian , Wang Han , Xu Xi , Shi Yang
Published: October 2016

Scaled down thermal-hydraulic test facility is widely used in the nuclear reactor safety analysis. The wall stored energy is one of the common problem in a scaled down test facility, which will cause distortions in the simulation of the transients or local phenomena. For the natural-circulation mode in the Core Makeup Tank (CMT), the cold wall will work as heat sink and absorb heat from the heated water, and then has an influence in energy balance on the simulation of the prototype phenomena with scaled down test facility. In order to study the influence of the wall stored energy on the natural-circulation mode in the CMT test facility, this paper established the CMT test facility model with RELAP5/MOD 3.4, and studied the influence caused by the wall stored energy distortion. The simulation results show that the natural-circulation process in the CMT test facility can be divided into two stages: 1. Before the thermal stratification reaches the outlet, the natural-circulation flow rate decreases gradually. 2. Once the thermal stratification reaches the outlet, the natural-circulation flow rate has an apparent turning point and decreases faster. That is because the heated water reaches the outlet and fills the discharge line quickly, which reduces the density difference between the hot leg and cold leg, thus causing the natural-circulation flow rate decrease faster. Besides, before the turning point, the wall stored energy basically has no influence on the natural-circulation process, and the influence can be neglected in the design of test facility. However, after the turning point, the wall stored energy distortion results in a slower descent speed of natural-circulation flow rate, which is not conservative. Therefore, we can conclude that the scaled down CMT test facility can only simulate the prototype properly and conduct the natural-circulation simulation experiment before the thermal stratification reaches the exit, while afterwards, the simulation of the test facility won’t be accurate and conservative.
Experimental Investigation on the Thermal and Flow Characteristics in the IRWST Under Spraying of the Steam From Automatic Depressurization System in AP1000 Plant
Authors: Lu Daogang , Zhang Yuhao , Wang Zhongyi , Wu Guanghao , Cao Qiong
Published: October 2016

In AP1000 plant, the automatic depressurization system (ADS) works to discharge the high-temperature and high-pressure steam from the Reactor Coolant System (RCS) primary side to the In-containment Refueling Water Storage Tank (IRWST) in the LOCA conditions. In particular, for the AP1000 plant, both the IRWST and ADS spargers are specially designed, and the ADS spargers are located in one corner of the IRWST. All the factors lead to the special and complicated thermal and flow behavior in the IRWST, which in turn have great influences on the condensation effects of the ejected steam. In the present work, an overall scaled-down IRWST and ADS sparger models are built to study the condensation and mixing phenomena in the accidental depressurization events in AP1000. Thermocouples matrix with more than 200 T-type sheathed thermocouples are utilized to measure the three dimensional temperature in the large tank. The Particle Image Velocimetry (PIV) is employed for the measurement of the natural convection flow velocity. Based on the experimental data, the local spraying flow patterns, flow behavior, and thermal stratification phenomena in IRWST etc. are analyzed. The results indicate that the spraying steam condensation flow patterns are closely related to the subcooling degree in the IRWST. In addition, the stratification number is developed to evaluate the thermal stratification extent in the IRWST, which indicates that only part of the fluid are used efficiently for condensing the spraying steam directly.
The Preliminary Experimental Study Based on Failure Mechanism of the RPV Lower Head
Authors: Meng XianKe , Fei LiKai , Zhang Aijing , Xiong SiJiang , Cui Lei , Chen YaoDong
Published: October 2016

In-Vessel Retention is a key severe accident management strategy for reactors such as AP/CAP series reactors. The IVR success evaluation criterion is whether the RPV is melted through or not at the final RPV state. Once the RPV lower head melt through, the liquid corium will flow into the reactor cavity and will lead to complex phenomena, such us steam explosion and the reaction between the corium and concrete. These will make temperature and pressure of the containment vessel rise quickly and is a threat to the integrity of the containment vessel. When the wall surface of RPV lower head heating condition exceed the critical heat flux, the temperature rises rapidly, it is generally assumed that the RPV lower head in this state will inevitably melt through. This is the so-called IVR failure.

In order to study the possible failure modes and mechanism of RPV lower head under the IVR measures, an experimental facility called TRECT is built. By measuring the parameters such as temperature, flow of the test section to study the influence to CHF by the parameters such as flow velocity and angle. All of these can provide reliable basis to the effectiveness appraisal and model development on the area of severe accident mitigation measures (IVR). To be specific, the test section is rectangular channel whose section is 50 × 20 mm. The upper surface is the heat surface and using a direct current heating mode to supply heat power. The heat flux can reach 1.5MW/m2. We use this upper surface heated rectangular channel to simulate RPV ERVC channel. By adjust the angle of test section to simulate the different circum ferential location of RPV lower head. And the Adjusting range can be 0° to 90°. The experimental results show that flow rate was reduced by 11% in the experiments, the critical heat flux density increased by 4.5%. Inclined angle increased from 16° to 29°, CHF increased by 7.9%.
The Future of Nuclear Power Generation
Authors: Panchal Raj , Pioro Igor
Published: October 2016

Electrical power is a resource humans heavily rely on, and it has become a basic human need. Today, the major sources of electricity generation are fossil fuels, renewable energy, and nuclear power. This paper concentrates on electricity generated through nuclear power and compares it to the other electricity generation technologies. The objective behind this paper is to discover the impact that nuclear power has on the total electricity generated in Canada, and in addition on a global scale. The paper presents the current role that nuclear power plays in the global electricity generation, and also the expansions that need to be made in the nuclear power industry to fulfill the future electrical power demands. A number of projections have been made based on the current rate of nuclear reactors being put into operation, which is approximately 4 reactors per year, and current term of reactor operation, which is 45 years. These projections were made for the nuclear power in the world. A major outcome of this analysis projects that between 2030 and 2035, the number of operating nuclear reactors in the world can drop by 50%. If this dangerous trend is not addressed, we can lose a viable, and reliable source of energy. The datasets that were analyzed during the process were taken from multiple open literature sources such as journals, reports, and online databases. The paper presents a comparison between nuclear power and other energy sources, and the positive impact nuclear power can have on the world if needed advancements were made in building new nuclear power plants.
An Experimental Study of Closed Loop Two-Phase Thermosyphon for Spent Fuel Pool Passive Cooling
Authors: Wang Minglu , Zheng Mingguang , Ye Cheng , Qiu Zhongming , Xiong Zhenqin
Published: October 2016

The study reported here examined a closed loop two-phase thermosyphon (CLTPT) of evaporator length 7.6m and internal diameter 65mm used to cool the spent fuel pool. This experimental study investigates the thermal performances and heat and mass transfer characteristic of CLTPT by examining the thermodynamic cycles and overall thermal resistances with ammonia, R134a and water as the working fluid. Measurements of temperature and pressure distributions of the fluid around the loop were made under various conditions. Results show that this loop operates with low filling ratio, low mass flow rate, and high heat-transfer coefficient and the CLTPT has the ability to cool the spent fuel pool. The working fluid flowing through the heat pipe evaporator section generally experienced a subcooled zone, pool boiling zone and high gas quality two-phase region. The average heat transfer coefficient of evaporator reaches 450 W/m2•°C using R134a as working fluid. The thermal resistance of R134a is always smaller than ammonia but the thermal resistance of water is largest at small temperature difference while is smallest when temperature difference is large.