bluuebunny/crossref_metadata_2025 huggingface dataset:
Application of Clearance in Controlling Radioactive Waste of Nuclear Power Plant
Authors: Wu Qiang , Wang LiQinDu , Jiang Chen , Zhou GuangLai
Published: October 2018

Clearance is an effective method to control radioactive waste, but some assessment of types of pollutants may need to be done before clearance in some situation. In fact, there is large volume very low level radioactive materials in nuclear wastes in nuclear power plant, a reasonable method should be built based on both of potential clearance study and clearance method optimization. Some suggestions are presented about the work of radioactive waste disposal for its present condition and development in China.
Application of the Dynamic Rod Worth Measurement Method on WWER
Authors: Zhao Wenbo , Jiang Zhumin , Wang Liangzi , Wang Chenlin , Yu Yingrui , Gong Zhaohu , Zhong Minxiao , Liu Tongxian , Xiang Hongzhi
Published: October 2018

The dynamic rod worth measurement method, which is used widely for PWR with square geometry lattices, is applied on WWER type reactor with hexagonal geometry lattices. RTNP code is developed to calculate the static spatial factors (SSF) and the dynamic spatial factors (DSF) through simulating the 3D space-time neutron dynamics and the response of excore detector during the measurement process. The improved quasi-static method is used for temporal discretization. The time-dependent shape equation is then transferred to a fixed source problem through backward Euler formula. And Multi-group Monte Carlo method is used to solve this fixed source problem. Four tests on Tianwan nuclear power plant (NPP) units 1&2 have been done since 2015. The measured results agree well with the predicted values. It takes about half hour per bank and 5 hours to measure all banks worth. The conventional boron dilution was used for rod worth measurement of Tianwan NPP units 1&2. It took at least 2 hours per bank and produced lots of boron wastes. The boron dilution has been replaced with this method for Tianwan NPP units 1&2.
Calibration of Inspection Strategies in Support of Aging Management Programs: A Probabilistic Approach
Authors: Pandey M. D. , Jyrkama M. I.
Published: October 2018

As existing fleets of nuclear power plants continue to seek long term operation, aging management programs have been developed to mitigate the effects of degradation processes affecting reactor components that are important to safety and generation. Inspection programs are an integral part of aging management, as they provide crucial information to determine the scope of maintenance and replacement programs and guide investment planning of life extension projects. This paper presents a probabilistic framework for calibrating the sample size of a periodic inspection program with the goals of the aging management program, such as limiting the number of potential substandard components below a reasonable threshold.
A Review of Multi-Unit Nuclear Power Plant Probabilistic Risk Assessment Research
Authors: Zhou Taotao , Modarres Mohammad , Droguett Enrique López
Published: October 2018

The events at the Fukushima nuclear power station drew attention to the need for consideration of risks from multiple nuclear reactor units co-located at a site. As a result, considerable international interests and research efforts have been dedicated to addressing the multi-unit risks over the past few years. This paper presents a review of the state-of-the-art multi-unit probabilistic risk assessment (MUPRA) of nuclear power plants. The concept of MUPRA is briefly presented and the relevant activities to address and develop methodologies are summarized including workshops, proceedings, projects and case studies. The paper presents different facets of MUPRA research, including multi-unit event, MUPRA modeling and site-based risk metric. The paper also identifies existing gaps and the need for future research.
iDROP: An Innovative Software Program to Design Nuclear Decommissioning Scenarios
Authors: Chabal Caroline , Perrot Vincent , Ben Mosbah Mehdi , Soulabaille Yves , Thieblemont Jean-Claude , Chaffard Fabien , Chevalier Yann , Chodorge Laurent
Published: October 2018

Many historic nuclear facilities have reached the end of their lifetimes and are now being decontaminated and dismantled. Our institute has been working on an innovative computing solution to simulate intervention scenarios in highly radioactive environments that humans cannot enter. Various treatment techniques can be compared and workers can be trained before operating on site. For the past 10 years, our institute has been developing a Virtual Reality tool that simulates all the key elements of a nuclear project, including the remote handling, and the accessibility of the site for humans, and the dose assessment associated. This tool allows all these parameters to be computed in a single interactive environment, allowing predefined scenarios to be verified and alternative solutions to be designed. This software is presented in the following paper; we here illustrate that it is particularly well-suited to dismantling projects. We also describe a first application of the tool to investigate radiological exposure scenarios. The advantages of this software include its user-friendliness, responsiveness, speed, and usefulness for the preparation of complex dismantling operations. The perspectives of this project include the simulation in Virtual Reality of human interventions.
Extended Ultimate Response Measures for Offshore Nuclear Power Plant Under Barge-Reactor Coupled Conditions
Authors: Wang Jue , Li Longze , Hu Chen , Cong Wang
Published: October 2018

Compared with the land-based nuclear power plant, the operating conditions of offshore nuclear power plant (ONPP) are much more complicated. For example, the barge-mounted platform malfunction, which is as important as the natural events and human events, should be considered in the plant safety analysis,. As a result, a two dimension operating condition coupled with barge and reactor status should be considered in the development of relevant power plant operating procedures. On the other hand, the beyond design basis hazards induced by the combination of unique and unanticipated external events of ONPP may lead to a blind area to both traditional and two dimension procedures mentioned above. Due to the insufficiency of existing operating condition and relevant procedures to tackle with the above events mentioned, an expanded operation strategy, namely the beyond design basis hazards and the extended ultimate response measures, is developed, Injecting sea water into reactor pressure vessel directly after primary system depressurized and venting the containment when necessary, formed the basis of ultimate response measure, which was proposed by Taiwan Power Company after Fukushima Accident. Considering the offshore and barge-mounted features, the ultimate response measure can be extended to include sea water injection into steam generator indirectly through secondary side passive residual heat removal lines and reactor cabin flooding by sea water through Kingston valves, to rebuild a newly, hierarchical one. Finally, the extended ultimate response measures, provided mainly for the plant command staff and operators, are analyzed utilizing thermal-hydraulic integral computer code preliminarily, to prove the effectiveness of the system configuration and operating strategy. It is concluded that injecting sea water into steam generator can remove the decay heat effectively, and the sensitivity study shows that operator intervention is good enough in accident mitigation.
Research on Deposition of Micro-Nano Aerosols in Rising Bubble Under Pool Scrubbing Condition
Authors: Zhou Yanmin , Gu Haifeng , Sun Qiunan , Sun Zhongning , Su Jiqiang , Gao Li , Ma Qianchao , Zhu Gan , Yingzhi Li
Published: October 2018

Aerosols as the main component of radioactive products in migration performance, which is an important factor that a unclear reactor accident present strong diffusion and affects the distributions of source and dose level in reactor containment, and they are therefore expected to be deposited in liquid phase such as in suspension pool and filtered containment venting device. In this paper, the deposition characteristics of micro-nano aerosols in rising bubble under pool scrubbing condition is studied with experiment, the aerosols size in the research range from 20 nm to 600 nm, and the bubble morphology mainly concern homogeneous bubbly flow. The results show that the deposition efficiency and mechanism of aerosol closely relate to gas flow rate, liquid level, particle size and bubbles size and so on. The aerosol deposition near 85nm is proved most difficult because of the convert of deposition mechanisms. In a high liquid level condition, micro-nano aerosol filtration efficiency is enhanced but gradually gradual. Under different gas flow rate, air bubble residence time and the bubble size distributions affect the filtration efficiency of aerosols.
Study on Offsite Emergency Preparedness for the Industry Application of HTR-PM
Authors: Ding Hongchun , Tong Jiejuan , Zhang Liguo
Published: October 2018

The offsite consequence study on High Temperature Gas Cooled Reactor Pebble-bed Module (HTR-PM) shows that its emergency planning zone (EPZ) for whether plume exposure pathway or ingestion exposure pathway can both be limited within the exclusion area boundary (EAB) of nuclear power plant (NPP), which is about a few hundred meters away from the reactor. This conclusion provides theoretical basis and strong technical support for potential industrial applications of HTR-PM. By this premise, the coupling risk for the industrial application of HTR-PM is quantitatively evaluated in this paper. The results indicate that the risk increments for both HTR-PM and industrial facility due to the colocation on one site could ensure they still meet their corresponding risk criteria. On this basis, the current proper management mode is recommended for the ally of HTR-PM and industrial facilities from technical perspective, which is the onsite of this HTR-PM-industry ally can adopt the mode of separated management, while the offsite co-management. Further, the final suggestions on offsite emergency preparedness for this HTR-PM-industry ally are also given out co-considering the current policy, practice and public acceptability. This work will also support the industrial application of other advanced reactors with inherent safety features.
The Design of Nuclear Noise Frequency Spectrum Analysis System and Helium Bubble Movement Behavior Computation for Narrow Channel Core
Authors: Zeng Herong , Yuan Baoxin
Published: October 2018

This paper designs a measure and analysis system of nuclear noise frequency spectrum for narrow channel research reactor core. Based on high-speed synchronization DAQ which is PXI-1031, the hardware design has been carried out. By using Labview, the data acquisition and analysis code has been developed. In order to verify the function of nuclear noise frequency spectrum measure and analysis system, firstly, through using arbitrary waveform generator which is Fluke-282, two sine signals have been tested. Secondly, by using history operation datum, the prompt neutron decay constant has been measured and analyzed. The simulation experimental and operational datum results show that such measure and analysis system of nuclear noise frequency spectrum is successful and reliable design, which can meet the design requirements. Finally, the preliminary research work between helium bubble and its responding frequency for narrow channel core has just started to monitor safety operation or carry out fault diagnosis.
Numerical Study of Water Droplet Heat Removal and Dynamics During its Impact Onto the Micro-Pillar Array at Elevated Temperature
Authors: Shahmohammadi Beni Mehrdad , Xie Shangzhen , Zhao Jiyun
Published: October 2018

The spray cooling and heat removal efficiency is one of the important aspect of nuclear thermalhydraulics and safety, especially for passive containment cooling after severe accidents. In order to design and optimize these systems effectively, computer modelling of the underlying mechanism of the liquid drop interaction with the hot solid surface would be necessary. Therefore, completeness, accuracy and reliability of the models that are being used in such sensitive areas are vital to the society and environment. Furthermore, the current powerful computer resources need to be fully exploited, so that the precision and the accuracy of the obtained computational results would be further enhanced. Nowadays, Volume-Of-Fluid (VOF) method is widely used in simulating the droplet dynamics, however these models provide estimations that are different in certain extents compare to the experimental results. In present work, we have used the level-set method to study the droplet dynamics and heat removal when the water droplet impact on the surface with different morphologies. The developed model which is based on the finite element method (FEM) has been benchmarked with previously performed experiments regarding the droplet bouncing on a flat hydrophobic surface; these estimations were in a good agreement with the previously published results. Moreover, hot solid surfaces with presence of micro-pillar has been considered to perform sensitivity study for different sizes of the micro-pillars and water droplets. In addition, it has been found that the heat transfer and droplet dynamic behavior would significantly vary in scenarios when the micro-pillars are presents in compare to a flat solid surface; it is observed that a better droplet spreading can be obtained with optimal size of micro-pillars that are present underneath of the droplet axial trajectory. The present study and the model would add valuable information to the field of heat transfer in aspect of spray cooling by investigating the feasibility of using the level-set method for a better estimation of fluid and heat transfer related results.
Modeling of Bubble Behavior in Low Void Fraction Subcooled Flow Boiling
Authors: Sakamoto Shintaro , Ohori Hiroki , Enoki Koji , Okawa Tomio
Published: October 2018

In predicting the void fraction in subcooled flow boiling, accurate evaluation of single bubble behaviors is of considerable importance. In particular, bubble lift-off velocity affects the void fraction significantly since the bubble disappear quickly due to heat transfer with subcooled liquid if the lift-off velocity is high. In this study, the process of bubble lift-off was experimentally investigated to develop mechanistic correlations for the bubble lift-off velocity. In the development of the correlations, it was assumed that the bubble lift-off velocity in the horizontal direction is proportional to the bubble growth rate and that in the vertical direction is determined primarily by the local liquid velocity evaluated at the bubble center position. Then, impact of the bubble lift-off velocity on the void fraction was explored through numerical simulations. In the simulations, the bubble lift-off velocity in the lateral direction was parametrically changed. It was shown that the mean void fraction decreases with an increase in the lateral bubble lift-off velocity since the bubble condensation is enhanced. It was therefore confirmed that accurate evaluation of the bubble lift-off velocity is important for high accuracy prediction of the void fraction.
Evaluation of Core Damage Frequency of High Flux Engineering Test Reactor From Internal Events
Authors: Liu Jinlin , Wang Wanhong , Peng Changhong , Guo Yun
Published: October 2018

PSA is found to be a valuable tool for improving the safety and reliability of nuclear reactors due to intense involvement of human interactions in an experimental facility. It may also be advantageously applied to research reactors under design and construction. This paper demonstrates the study and result of a Level 1 internal initiating event PSA under the power operation mode to the High Flux Engineering Test Reactor (HFETR) which is under design and construction. The occurrence frequency of core damage which may potentially arise in HFETR is evaluated by using RiskSpectrum software, and event trees have been used to study the response of the installation to various initiating events whereas fault trees have been used in the modeling of safety system failures. Calculation results show that core damage frequency for HFETR research reactor is 5.28E−07 per year, and initiating events which have a significant contribution to CDF are loss of coolant water and loss of offsite power.
Experimental Study of Motion-Resistance Force of Hydraulic Cylinder of CRHDM
Authors: Liu Qianfeng , Li Yuzheng , Zhang Huang , Hanliang Bo
Published: October 2018

Hydraulic control rod drive technology (HCRDT) is a newly invented patent owned by Institute of Nuclear and New Energy Technology of Tsinghua University with independent intellectual property rights. Hydraulic cylinder is the core part of this technology, so the seal of hydraulic cylinder directly affects the performance of the hydraulic cylinder and HCRDT. A new experiment table was designed and used to study the leakage and internal cylinder friction at various types of friction coupling between internal cylinder and seal ring. The result shows that Motion-Resistance Force is the smallest when the internal cylinder film is TiN. Furthermore, when seal ring film is DLC (Diamond-like carbon) the leakage is the smallest. At last, the best friction coupling scheme was obtained. The experimental results provide the basis for the establishment of the motion model of hydraulic cylinder, which provides a foundation for further design and optimization of hydraulic drive technology of control rod.
Upgrade and Shakedown Test of a High Temperature Fluoride Salt Test Loop
Authors: Kong Xiangbo , Fu Yuan , Zhang Jianyu , Lu Huiju , Wang Naxiu
Published: October 2018

A FLiNaK high temperature test loop, which was designed to support the Thorium Molten Salt Reactor (TMSR) program, was constructed in 2012 and is the largest engineering-scale fluoride loop in the world. The loop is built of Hastelloy C276 and is capable of operating at the flow rate up to 25m3/h and at the temperature up to 650°C. It consists of an overhung impeller sump-type centrifugal pump, an electric heater, a heat exchanger, a freeze valve and a mechanical one, a storage tank, etc. Salt purification was conducted in batch mode before it was transferred to and then stored in the storage tank. The facility was upgraded in three ways last year, with aims of testing a 30kW electric heater and supporting the heat transfer experiment in heat exchanger. Firstly, an original 100kW electric heater was replaced with a 335kW one to compensate the overlarge heat loss in the radiator. A pressure transmitter was subsequently installed in the inlet pipe of this updated heater. Finally, a new 30kW electric heater was installed between the pump and radiator, the purpose of which was to verify the core’s convective heat transfer behavior of a simulator design of TMSR. Immediately after these above works, shakedown test of the loop was carried out step by step. At first the storage tank was gradually preheated to 500°C so as to melt the frozen salt. Afterwards, in order to make the operation of transferring salt from storage tank to loop achievable, the loop system was also preheated to a relatively higher temperature 530°C. Since the nickel-base alloy can be severely corroded by the FLiNaK salt once the moisture and oxygen concentration is high, vacuum pumping and argon purging of the entire system were alternatively performed throughout the preheating process, with the effect of controlling them to be lower than 100ppm. Once the salt was transferred into the loop, the pump was immediately put into service. At the very beginning of operation process, it was found that flow rate in the main piping could not be precisely measured by the ultrasonic flow meter. Ten days later, the pump’s dry running gas seal was out of order. As a result, the loop had to be closed down to resolve these issues.
Design of the Important Area Airborne Radioactivity Monitoring System and Calculation of the Alarm Threshold of HTR-PM
Authors: Li Wenqian , Li Chuan , Cao Jianzhu , Xie Feng
Published: October 2018

Unlike the conventional pressurized water reactor (PWR), helium is adopted as the coolant in the high-temperature gas-cooled reactor pebble-bed module (HTR-PM). Due to the leakage of the primary coolant, in certain process rooms and some important areas, the airborne radioactivity should be monitored for the purpose of personnel radiation protection and emergency control. The system executing above function in HTR-PM is called important area airborne radioactivity monitoring system. This system consists of five separate monitoring channels, in which online and/or sampling monitoring devices will be included. For radiation monitoring system of a reactor, the calculation and setting of alarm thresholds for online devices is an important issue. The set of alarm thresholds requires not only accurate calculation but also feedback from actual operational experience. The design of the high-temperature reactor is different from the PWR, so the set of the alarm thresholds need special considerations. Based on the calculation of source terms and production-removal model, the theoretical radioactive levels of the sampled gases of all the online monitors were calculated and the alarm thresholds adopted in the current design are presented in this paper. The set of the alarm thresholds takes the influences of the background and other factors into account. This paper provides a method for calculating and determining the alarm thresholds of the radiation monitoring system, which can provide references for other nuclear power plants.
Preliminary LOCA Analysis of Heating-Reactor of Advanced Low-Pressurized and Passive Safety System (HAPPY)
Authors: Xing Mian , Meng Zhaocan , Liao Xiaotao , Sun Canhui , Zhang Shuming , Chen Yaodong , Hu Xiao , Sun Peidong , Jiang Huijing
Published: October 2018

SPICRI (State Power Investment Central Research Institute) is developing a new conceptual design of heating-reactor, named Heating-reactor of Advanced low-Pressurized and Passive safetY system (HAPPY), which is targeted for the district heating, desalination of seawater, and other heat applications. It is a 200MWth two-loop low-pressurized water reactor with low thermal parameters. The whole reactor vessel is deployed inside a shielding and cooling pool with thermal insulation measure. The conceptual design of HAPPY is described in this paper, including the design criteria, safety features, main parameters and main components.

A preliminary safety analysis is carried out to provide a reference for the design and optimization of HAPPY. In this paper, four different LOCA analyses are described and compared. The results show that the current design can deal well with all the selected LOCA scenarios and the effectiveness of the safety systems is proved.
Study on the Shaking Dynamic Response of Steam Generator LOCA
Authors: Huang Qian , Yu Xiaofei , Qi Huanhuan , Jiang Naibin , Cai Fengchun , Feng Zhipeng
Published: October 2018

Dynamic response research of steam generator loss of coolant accident (SG LOCA) is essential for the reliability and safety consideration. According to the differences of LOCA loading phenomena, two types of LOCA loads affect the SG: rarefaction wave travels through the primary fluid in the U-tubes, and the SG shakes due to reactor coolant loop(RCL) motions transmitted by the primary loop piping, former loading phenomena evaluation is called SG rarefaction analysis while latter is called shaking analysis. This paper place particular emphasis on shaking analysis.

At present, the published literatures about LOCA mainly focus on RCL LOCA, reactor LOCA and fuel assembly LOCA, few reports concentrate on shaking dynamic response analysis of SG LOCA. Both Westinghouse and AREVA’s methods according to their research reports are to decouple the SG from the RCL: This method results in low computational efficiency as RCL LOCA and SG LOCA are evaluated separately and the decoupling error is uncertainty, meanwhile, in the vicinity of the nodes where the displacements are imposed, distorted reaction forces are usually found.

Through reasonable simplification and equivalence, a detailed nonlinear FEM model of steam generator (SG) of a China 3rd generation nuclear power plant (NPP) is established, this model is then connected with the reactor coolant loop (RCL) to carry out the SG LOCA shaking dynamic response analysis. By calculation, the maximum absolute stresses of SG heat transfer tube bundle and its variation with tube diameter and upper supports reacting forces are obtained. In order to study the effect of SG decoupling from the RCL on shaking dynamic response, a comparative study of decoupling /coupling methods is conducted. Results shows that SG decoupling has a significant impact on the calculation result, the calculation method of coupling is more closer to the real situation and worthy to recommended. Related analytical procedures and calculation results lay the foundation for future SG shaking dynamic analysis and SG design of subsequent power plants.
Thermal and Mechanics Analysis Code of the PWR Nuclear Fuel Performance Based on FEM
Authors: Hui Yongbo , Zhang Bo , Zhang Wenhua , Yun Di , Zhai Peichao
Published: October 2018

Exposed to the high radiation environment, the PWR fuel rod will be swelling. The deformation is quite different azimuthally which caused by the different subchannels coolant temperature and some manufacture issues. And the asymmetry deformation will have a great effect on fuel rod safety for the increasing pellet center temperature and FCMI due to the change of gas gap. An accurate prediction on fuel strain which is associated with the temperature distribution closely should be done to assure the safety of the fuel rod. The conventional finite difference method has deficiency on treating with the deformation and the finite element method (FEM) could satisfy the requirement.

Based on the principle of FEM, the model of thermal conduction, stress-strain analysis is established. The two dimensional model is chosen considering the characteristic of fuel. The analysis code SSTFEM is developed and verified using the experimental data from the opening literature and results of authorized code. After that the code is used to analyze the effect of the asymmetry deformation which is caused by the asymmetry coolant temperature distribution in adjacent subchannels and the manufacture eccentricity between the fuel rod and the cladding. The effect law of deformation and the stress is obtained after a systematic analysis.
Analysis of Flow Blockage of a Single Fuel Assembly in the JRR-3 20MW Research Reactor
Authors: Guo Yuchuan , Wang Guanbo , Qian Dazhi , Yu Heng , Hu Bo
Published: October 2018

The case of flow blockage of a single fuel assembly in the JRR-3 20MW open-pool-type research reactor is investigated without taking into account the effect of the power regulation system. The coolant system and multi-channel reactor core are modeled in detail using thermal hydraulic system analysis code RELAP5/MOD3.4. MDNBR (Minimum Departure From Nucleate Boiling Ratio) and the maximum fuel central temperature are investigated to assess the integrity of fuels. The fuel plates in blocked assembly are not damaged until the blockage ratio exceeds 70%. In addition, the mitigative effect of the assumed 18 MW lower power emergency shutdown operation on the accident is also discussed qualitatively. Results indicate that although the assumed lower power emergency shutdown operation cannot avoid the most severe operating condition, it can obviously mitigate the consequences of the accident. The reactor eventually remains in the long-term safe state when natural circulation is established.
Interior Penalty Schemes for Discontinuous Isogeometric Methods With an Application to Nuclear Reactor Physics
Authors: Wilson S. G. , Kópházi J. , Owens A. R. , Eaton M. D.
Published: October 2018

Until recently, interior penalty methods have been applied to elliptic operators using an approach based on the mass matrix of finite elements that possess a constant Jacobian. In the case of isogeometric analysis, such an approach would not always guarantee coercivity of the bilinear form and thus numerical stability of the solution. In this paper, optimal interior penalty parameters [1] are described and applied to the symmetric interior penalty scheme of the discontinuous Galerkin isogeometric spatial discretisation of the neutron diffusion equation. The numerical accuracy of the proposed method is compared against a standard continuous Bubnov-Galerkin isogeometric spatial discretisation of the neutron diffusion equation. Numerical consistency and order of error-convergence is verified by means of the method of manufactured solutions. Numerical results are also presented for a two-dimensional pin-cell test case, based upon the OECD/NEA C5G7 quarter core MOX fuel assembly benchmark for nuclear reactor physics parameters of interest.
Development of Three-Dimensional Neutron Kinetics Code Based on High Order Nodal Expansion Method in Hexagonal-Z Geometry
Authors: Chao Guo , Yu Liu , Hangxing He , Luguo Liu , Xiaoyu Wang , Sufang Xin , Peiyang Li , Xiaoli Wu , Hongsheng Yuan
Published: October 2018

To solve three-dimensional kinetics problems, a high order nodal expansion method for hexagonal-z geometry (HONEM) and a Runge-Kutta (RK) method are respectively adopted to deal with the spatial and temporal problem. In the HONEM, 1D partially-integrated flux are approximated by using four order polynomial. The two order polynomial is adopted to the approximation of partially-integrated leakages. The Runge-Kutta method is adopted as a tool for dispersing the time term of 3D kinetics equation. A flux weighting method (FWM) is used for obtaining homogenized cross sections of mix node. The three-dimensional hexagonal kinetics code has been developed based on this method and tested with two benchmark problems of VVER which are the control rod ejection without any feedback and with simple adiabatic Doppler feedback. The results calculated by this code agree well with the reference results and the code is validated.
Influence Factors on Moisture Absorption Process of Carbon Material in HTGR
Authors: Li Jun , Yin Huaqiang , He Xuedong
Published: October 2018

A large amount of carbon material is used in high-temperature gas-cooled reactor (HTGR), with an initial moisture content less than 0.1 %, and under the high temperature condition during reactor’s operation the material would be corroded by chemical corrosion reaction with moisture. The carbon materials in HTGR include carbon material with boron carbide (BC) and graphite IG-110. Based on the moisture absorption experiment of BC, the effective diffusion coefficient D of BC is fitted, and D = 1.35 × 10−9 (m2/s). The experiments shows with the same condition of temperature and humidity, the moisture absorption of BC is higher than that of IG-110. As the environment humidity increases, the moisture absorption increases.
Neutronic Methodological Benchmarks With Simplified Geometries for the Gas Cooled Reactor Group Constant Generating Tools
Authors: Temesvari E. , Batki B. , Gren M.
Published: October 2018

In the ESNII+ EU FP7 project, a reactor physics benchmark aiming at the whole core calculation with the reflectors and detailed description of the structural elements was specified. This benchmark is based on the 2009 CEA concept of the ALLEGRO core. Fixed nominal technological data at nominal reactor state (geometry, composition) were prescribed which had to be modified in specified calculation branches according to different types of the thermal expansion and control rod positions. The parameters of the point kinetic model to be applied in a system thermal hydraulic code had to be determined this way. Static mechanical models of the expansion processes were specified by the benchmark.

The goal of the calculation exercise was to verify the reactor physics codes, namely to get information about the modelling uncertainties and — after — their influence on the calculated results of the safety analyses. The obtained deviations between the participants are characterizing the user effects, the modelling uncertainties and the influence of the nuclear data differences all, without the possibility of their separation because of the complexity of the benchmark problem. A conclusion could be drawn that a step by step procedure starting from simple problems (homogenous material, Wigner-Seitz cell or subassembly in asymptotic approach) is necessary if we wish to identify the reasons of the deviations. For the Doppler effect, a decision was made in this direction already in the ESNII+ project where an infinite regular lattice problem without any leakage had to be solved. This approach of the simplicity is followed by the present benchmarks (one rod and one assembly), but extending the simple benchmarks with burnup calculations and taking into account leakage in asymptotic approximation by neglecting the complicated processes necessary in the reflector regions.
Applying Statistics to Improve the Performance of Nuclear Decommissioning Projects
Authors: Invernizzi Diletta Colette , Locatelli Giorgio , Brookes Naomi J.
Published: October 2018

Nuclear Decommissioning Projects and Programmes (NDPs) are characterized by significant risks, long schedules, and high costs that keep rising. Additionally, due to the NDP complexity and variety, it is extremely hard to understand which are the NDP characteristics that are associated with the NDP performance. This research takes the project management perspective, collects empirical information on NDPs and investigates this relationship between NDP characteristics and NDP performance, both through qualitative cross-comparison and quantitative statistical analysis based on the Fisher’s Exact Test (FET). In this paper, the results from the implementation of the FET applied on a pool of European NDPs are presented and discussed. Key takeaways are that some project characteristics present a stronger relationship with the project performance, while others do not show a significant association. Ultimately, qualitative and quantitative analyses complement each other to support the development of guidelines and improve the selection, planning and delivery of future NDPs.
Numerical Investigation of Buoyancy Effect on Forced Convective Heat Transfer to Supercritical Carbon Dioxide Flowing in a Heated Tube
Authors: Sheng-hui Liu , Yan-ping Huang , Guang-xu Liu , Jun-feng Wang
Published: October 2018

Numerical investigation of buoyancy effect on forced convective heat transfer to supercritical carbon dioxide flowing in a vertical tube was carried out. When the mass flux is low and wall heat flux high, it shows that the buoyancy effect is obvious, which might redistribute the radial and axial velocity, even M shaped distribution in the radial direction. When the zero-velocity-gradient region corresponding to the M shaped velocity distribution appears in the edge of viscous layer, the production and diffusion of eddy will be weakened, resulting in heat transfer deterioration. According to the extended simulations based on experimental data, reducing the wall heat flux, adding the mass flux or raising the inlet temperature can relieve the deterioration of heat transfer caused by buoyancy effect.
Experimental Study on the Safety Injection Pump (SIP) Failure Accompanied by the Steam Generator Tube Rupture (SGTR)
Authors: Park Yusun , Bae Byoung Uhn , Kim Jongrok , Lee Jae Bong , Park Hae Min , Choi Nam Hyun , Kang Kyoung Ho
Published: October 2018

The integral effect test to simulate the safety injection pump (SIP) failure accompanied by a steam generator tube rupture (SGTR), named a SGTR-SIP-01 test, was performed to investigate the thermal hydraulic phenomena during a multiple failure accident.

In this study, a thermal-hydraulic integral effect facility, ATLAS (Advanced Thermal-hydraulic test Loop for Accident Simulation) was utilized to simulate thermal hydraulic phenomenon which can be occurred in the nuclear power plant, as realistically as possible.

In this SGTR-SIP-01 test, a rupture of five steam generator u-tubes on the steam generator hot-side was simulated. Due to the initiation of SGTR, a reactor was tripped by high steam generator level (HSGL) signal.

During the transient simulation for SGTR-SIP-01 test, major thermal-hydraulic parameters such as the system pressures, the collapsed water levels, the flows in the primary loops, and the fluid temperatures, were measured and analyzed. Through this experimental result, insights about the accident management procedure can be provided in the case of the multiple failure accident, such as a SGTR accident with a total failure of SIPs. In addition to that, for improvement of the system code which are now on developing such SPACE or MARS-KS code, this test data can be utilized for validation and verification work.
Strength Limit of Thimble Tube With Material Plasticity Under Bending Moment and Axial Compression Force
Authors: Koike Hisashi , Mori Masaji , Fujiwara Daisuke , Shimomura Takashi
Published: October 2018

The thimble tube, which is made of Zircaly-4, is one of the main components of a PWR fuel assembly. The thimble tube has an important role as a structural member of the skeleton. Another role of the thimble tube is to guide a rod cluster control assembly (RCCA) for insertion during the reactor operation, and the function has to be assured not only in normal operation but in a seismic event.

In a horizontal seismic event, the fuel assembly vibrates laterally, which gives bending moment to the thimble tube. In addition, axial compressive force acts on the thimble tube in a vertical seismic event.

The integrity of the thimble tube has to be maintained while this force and moment act. Mitsubishi has confirmed by the elastic stress analysis that the stress of the thimble tube is lower than the limit value requested for the seismic event. The stress evaluation method is based on the ASME code.

The ASME code also describes the limit analysis which is available when the predicted stress is beyond elastic region of the material. In the analysis, the material is assumed to be elastic-perfectly plastic, and the maximum load that the structure can carry is calculated. For the reason mentioned above, the allowable limit of the thimble tube should be determined as a function between the force and the moment. We are planning to examine the allowable limit experimentally.

As a step before testing, an analytical approach for the limit is discussed in this paper. Firstly, the allowable limit is calculated by a beam model assuming elastic-perfectly plastic material, based on the ASME code.

Secondly, a 3D model analysis with elastic-plastic material is performed to predict the practical strength. Based on the comparison with the analysis using elastic-perfectly plastic material, ASME based limit is considerably conservative compared with the one with the actual stress-strain curve. Conversely, this means there is enough room to rationalize the allowable limit.

As the future work, the experiment will be conducted to obtain the practical limit of the thimble tube and to verify the analysis results.
Conceptual Design and Neutronics/Thermal-Hydraulic Coupling Optimization Analyses of Two Typical Helium Cooled Solid Breeder Blanket Modules for CFETR Phase II
Authors: Cui Shijie , Zhang Dalin , Tian Wenxi , Su G. H. , Qiu Suizheng
Published: October 2018

Chinese Fusion Engineering Test Reactor (CFETR) is a new test Tokamak device which is now being designed in China to make the transition from the International Thermonuclear Experimental Reactor (ITER) to the future Fusion Power Plant (FPP). Breeding blanket is the key component of fusion reactor which is mainly responsible for the tritium self-sufficiency and fusion energy conversion. In the past few years, three kinds of blanket conceptual design schemes have been proposed and tested in parallel for CFETR Phase I, in which the helium cooled solid breeder (HCSB) blanket concept is acknowledged as the most promising one. However, nowadays, the design phase of CFETR has gradually changed from I to II aiming for the future DEMO operation condition, the main parameters of which are quite different from the previous one. Therefore, it’s necessary to perform conceptual design and various analyses for the HCSB blanket under the new working condition. In this work, firstly, a new conceptual design scheme of HCSB blanket for Phase II is put forward. Then, the radial build arrangements, of the two typical blanket modules are optimized by using the NTCOC. This work can provide valuable references for further conceptual design and neutronics/thermal-hydraulic coupling analyses of the HCSB blanket for CFETR Phase II.
Investigation on Breakaway Oxidation Behavior of CZ Alloys
Authors: Gao Chang-yuan , Chen Liu-tao , Shi Lin , Xu Yang , Tan Jun , Deng Yong-jun
Published: October 2018

Breakaway oxidation is a potential fuel failure mechanism during a loss-of-coolant accident (LOCA), especially small-break LOCA. Two kinds of new advanced zirconium alloy, CZ1 and CZ2, were developed by China Nuclear Power Research and Technology Institute of China General Nuclear Power Group. The breakaway oxidation behavior of CZ1 and CZ2 was studied. The outer surface of all samples was examined visually and photographed. After 2730s oxidation in steam, the outer surface of CZ1 alloy sample and CZ2 alloy sample remained lustrous-black. The outer surface of CZ1 sample oxidized in steam at 1000 °C for 4181s was grey, but under the same experimental conditions the outer surface of CZ2 sample was still lustrous black. The hydrogen pickup content of different oxidation time was measured. The samples with grey appearance showed significant hydrogen pickup. The microstructure was observed by optical microscope. Evolution of oxide structure was described, and the mechanism was discussed.
Numerical Simulation of Safety Rod and its Drive Mechanism in Sodium-Cooled Fast Reactor During Scram Action
Authors: Yan Li , Wenjun Hu , Lixia Ren
Published: October 2018

Safety rod and its drive mechanism is one of the shutdown systems in sodium-cooled fast reactor, which must be quickly inserted into the core to achieve emergency shutdown in the event of an accident. Therefore, it is necessary to study the falling process of safety rod. In this paper, the numerical simulation method is used to analyze the falling process of safety rod and its drive mechanism in China Experimental Fast Reactor. According to the flow path of the safety rod and its drive mechanism, the pipe system hydraulic method is used to model the safety rod and its drive mechanism and calculate the hydraulic force of the safety rod and its drive mechanism during the falling process. The relationship between time, displacement, velocity and acceleration is presented. The drop time of safety rod is calculated, which is compared with the experimental results. The factors that affect the drop process are analyzed and a sensitivity analysis is presented.
Experimental Study of Aerosol Behavior During Pool Scrubbing: Part 2 — Decontamination of Aerosol Particle in Two Phase Flow
Authors: Kikuchi Wataru , Fujiwara Kota , Nakamura Yuki , Saito Shimpei , Yuasa Tomohisa , Kaneko Akiko , Abe Yutaka
Published: October 2018

In order to decrease the leakage and diffusion of Fission Products (FPs) in a severe accident (SA) of nuclear power plants, BWR have a suppression chamber filled with water to decontaminate the polluted air from the reactor vessel when emergency ventilation is done. It’s called pool scrubbing having a function of decontamination effect that transfer particles from gas to liquid phase by blowing gas containing FP into water. decontamination factor (DF) which is an index of decontamination performance evaluation of pool scrubbing is predicted and calculated by using a flow model in the existing analysis model such as MELCOR code. Evaluation of the DF by pool scrubbing is important in analyzing SA progress of nuclear power plants. For this reason, validation of flow model and decontamination model is important. It is necessary to validate whether the analysis result of each model matches the actual phenomenon. Despite of these needs, generic experimental data on gas-liquid two phase flow including aerosol is insufficient in confirming the validity of existing model. Therefore, the aim of this study is to investigate aerosol behavior in air jet and its relation to bubbly flow structure during pool scrubbing. This is to obtain the flow structure when air - aerosol released to the pool part and are obtained. Transition behavior of particles from the gas phase into the liquid phase was observed by high-speed camera. In addition, the effect of aerosol on void fraction in the pool was measured by wire mesh sensor from the nozzle exit to the water surface ware also observed. Ejecting air-aerosol from the nozzle to the pool part at various flow rates, visualization of bubble diameter, bubble aspect ratio, void fraction and the gas phase velocity were done by using image processing and a wire-mesh sensor. From these results, we compared the parameters by the presence or absence of aerosol. DF is also compared with the model used in the MELCOR code.
Rewetting Analysis of Hot Moving Surface by Round Water Jet Impingement
Authors: Sharma Avadhesh Kumar , Modak Mayank , Sahu Santosh K.
Published: October 2018

Impinging jets are commonly utilized in the run-out table (ROT) cooling in the hot rolling process in steel manufacturing industries. The phenomenon of rapid cooling of a sufficiently hot surface is termed as the quenching. The present paper reports the rewetting behavior of 0.15 mm thick hot moving stainless steel foil (SS-304) by circular impinging jet from bottom side through experimental investigation. The transient temperature of the hot foil is recorded by using thermal imaging camera (A655sc, FLIR system). Tests are performed for a varied range of Reynolds number (Re = 2500–10000), nozzle to plate distance (z/d = 6), moving plate velocity (0–40 mm/s) and initial surface temperature 500±10 °C. Transient temperature obtained from thermal imaging camera is used to evaluate rewetting time and rewetting velocity. Based on the experimental investigation correlation has been proposed to predict non-dimensional rewetting velocity as a function of various parameters, namely, Reynolds number, non-dimensional axial distance and moving plate velocity.
A Review of Brayton Helium Gas Turbine Cycles for GFR and VHTR Generation IV Nuclear Power Plants
Authors: Gad-Briggs Arnold , Pilidis Pericles , Nikolaidis Theoklis
Published: October 2018

Studies are currently on-going on the cycle performance of Generation IV (Gen IV) Nuclear Power Plants (NPPs) for the purpose of determining optimum operating conditions for efficiency and economic reasons. For Gas-cooled Fast Reactors (GFRs) and Very-High Temperature Reactors (VHTRs), the cycle layout is predominantly driven by the choice of components, the component configuration and the coolant. The purpose of this paper to present and review the cycles currently being considered — the Simple Cycle Recuperated (SCR) and the Intercooled Cycle Recuperated (ICR). In all cases, the cycles utilise helium as the coolant in a closed Brayton gas turbine configuration. Comparisons between the cycles are made for Design Point (DP) and Off-Design Point (ODP) analyses to emphasise the benefits and drawbacks of each cycle. The paper also talks about future trends which include higher Core Outlet Temperatures in excess of 1000 degrees Celsius and the proposal of a simplified cycle configuration which eliminates the need for the recuperator.