bluuebunny/crossref_metadata_2025_split huggingface dataset:
Sensitivity and Uncertainty Information Incorporated Loss of Flow Accident Analyses for Research Reactors
Authors: Liu Tao , Wu Zeyun 
Published: October 2020

Abstract
This paper outlines a system level safety analysis procedure for research reactors incorporating sensitivity and uncertainty components. The protected loss of flow (LOF) accident was selected as an exemplified design basis accident to demonstrate the analysis procedure. The conceptual NIST (National Institute of Standards and Technology) horizontally split-core based research reactor was adopted as a research reactor model in the study. Two system level dynamics codes, RELAP5-3D and PARET, were employed in this work in a comparison study manner. The primary objective of the present work is to demonstrate the analysis capability of integrating sensitivity and uncertainty information in addition to traditional predictions of the system code models for the study of the thermal-hydraulics (T/H) safety characteristics of research reactors under accidental transient scenarios. The canonical transient predictions on the LOF accident yielded from the two system codes mentioned above have demonstrated some noticeable yet acceptable discrepancies. To better understand the discrepancies observed in the simulations, sensitivity and uncertainty analyses were performed by coupling the RELAP5-3D model and the data analytic engines provided by the RAVEN framework developed by INL. The sensitivity information reveals the significances of key figure of merits such as the peak cladding temperature varies with different boundary and initial parameters in both normal operation and design basis transients. The uncertainty analysis informs the deviations of the responses contributed by the errors of various input components. Both the sensitivity and uncertainty information will be incorporated into a safety analysis framework as part of the safety characteristic predictions delivered by the framework.
Discussion of Water Quality Factors Affecting 137Cs Adsorption and the Relationship Between 137Cs-Kd and Flow Rate in Freshwater
Authors: Wang Jiaxin , Zhang Liuguo , Qu Jingyuan , Wu Guohua 
Published: October 2020

Abstract
The discharge water of Inland Nuclear Power Plant (INPP) is freshwater. Freshwater has a lower environment capacity and more complex environmental impacts comparing with the salt water (oceans), which makes the public more sensitive and stricter. 137Cs, with a long half-life of 30.17 years, is likely to exist as particle form in freshwater, which increases the difficulty to be removed and also becomes a long-term potential radiation hazard. The factors affecting the behavior of 137Cs in freshwater could be divided into water quality factors and hydraulic factors. For water quality factors, the paper summarizes factors that have significantly influence on the adsorption of 137Cs: size and concentration of suspended particles, competing cations, pH, concentration of organic matters and another two environmental factors, time and extreme meteorological events such as floods, heavy rain. We describe the qualitative relationship that tells how the factors take effect on adsorption process. In terms of hydraulic factors, this paper focuses on a representative factor, flow rate, and takes the original measured distribution factor (137Cs-Kd) and flow rate data of 9 freshwaters from Asia and Europe for analysis. The study finds that there are two segmented index relationships between 137Cs-Kd and flow rate above the threshold of 1 m3/s, the experiences of other five rivers that had not participated in data fitting verify the accuracy and generic character of the index relationships. The qualitative summary of water quality factors and the quantitative conclusions between 137Cs-Kd and flow rate can provide advice for evaluating the behavior of 137Cs in freshwater, and then help in the site selection process for INPP.
Experimental Investigation on Temperature Characteristics of Direct Contact Condensation in a Natural Circulation System
Authors: Sun Jianchuang , Ding Ming , Mi Zhengpeng , Zhang Zhuohua 
Published: October 2020

Abstract
Direct contact condensation (DCC) is a common physical phenomenon appearing in the nuclear power plants and other industrial applications. The current research on DCC focuses on steam-water counterflow or jet flow under forced flow conditions. For some natural-circulation passive safety systems in floating nuclear power plants, the heated section is connected with the heat sink by the horizontal pipes. The heat sink is usually the ocean. In such natural circulation systems (NCSs), the steam produced in the heated section will inevitably contact the subcooled water in the ocean and result in DCC event. In this paper, the fundamental characteristics of two-phase flow were described in detail. In addition, the influences of the subcooled water temperature on the flow rate, outlet temperature, steam bubble behaviors, and pressure surge were emphatically discussed. The experimental results revealed that the subcooled water in the water tank could reversely flow into the pipe, which was able to cause condensation induced water hammer (CIWH) event and flow oscillation. Further research showed that the subcooled water was evidently prevented from reversely flowing into the pipe with the increase in the subcooled water temperature. The position where the DCC event occurs transferred from the pipe to the water tank, and the pressure peak obviously reduced, even disappear when the subcooled water temperature is larger than 61 °C.
Study on Irradiation Damage of RPV Steels Based on Coupling Cluster Dynamics and Crystal Plasticity Finite Element Method
Authors: Wang Xiaotong , Luo Ying , Dong Yuanyuan , Yao Weihua 
Published: October 2020

Abstract
Irradiation hardening is one of the most important aging effects of reactor pressure vessel (RPV) steel in long-term service. A number of studies have indicated that irradiation hardening is mainly caused by irradiation induced defects, such as dislocation loops and precipitates. In this paper, we have simulated the irradiation damage of low-copper reactor pressure vessel (RPV) steel. The generation of Mn-Ni-Si precipitates is simulated by the cluster dynamics model based on rate theory. On this basis, the crystal plasticity finite element method based on modified crystal plasticity model is used to simulate the effect of Mn-Ni-Si precipitates on the mechanical properties of RPV steels. The simulated results has been compared with the experimental results from the literature. By coupling the cluster dynamics and the crystal plastic finite element method, we suggest a multi-scale simulation method to simulate and predict irradiation damage of RPV steel.
Comparative Thermal Analyses Between Theoretical Mode and RELAP5 Code Simulation for OTSG of a Small PWR
Authors: Jiang Baihui , Zhou Zhiwei , Xia Zhaoyang , Sun Qian 
Published: October 2020

Abstract
Due to the low nuclear safety risk, low initial investment cost and short construction period, integrated small nuclear reactors have received wide attention from all over the world. As advanced new type of nuclear reactors, the technologies of integrated small nuclear reactors are in the process of exploration and development. Steam generators are used as the heat transfer system for energy exchange between the primary and secondary circuit in reactors, and their heat transfer analysis is very important for reactor design and development. Due to the simple structure, strong heat exchange capacity and timely load following, Once-Through Steam Generators (OTSGs) are the mainly used steam generators in the design of integrated small nuclear reactors.
RELAP5/MOD4.0 is a commercial software developed by Innovative System Software, LLC for transient analysis of light water reactors (LWR). After years of development and improvement, RELAP5 has been a basic tool for analysis and calculation of various simulators of nuclear power plants. However, when RELAP5 models steam generators, only structural models related to straight pipes can be established, which is very inconvenient for the heat transfer research of Once-Through Steam Generators.
In this paper, Once-Through Steam Generators with specified structural parameters are taken as the research object. The heat transfer calculation is performed on the simplified inclined tube models by RELAP5 code and the theoretical calculation of the spiral tube heat transfer models is also carried out. Comparing the steam outlet temperature on the primary and secondary sides, the heat exchange power, the average heat transfer coefficient and the tube length of different heat exchange zones under given primary and secondary side inlet fluid conditions, it is confirmed that the RELAP5 heat transfer calculation is verified for simplifying Once-Through Steam Generators with inclined tube models.
Impact Analysis of NPP H4 Connections Design Improvement on Emergency Operation
Authors: Yuqi Wang , Ke Yi 
Published: October 2020

Abstract
After the loss of coolant accident (LOCA), the safety injection system injects water into the reactor coolant system (RCS), and the residual heat rejects from the break. The containment spray system is operating in recirculating cooling mode to ensure that the containment is cooldown. This state must be maintained for several months. After the accident, in order to respond the design extension conditions (DEC) of failure of two containment spray pumps or two low pressure safety injection pumps, the design of the original H4 connections was improved, and the H4 procedure (loss of containment spray pumps or low pressure safety injection pumps) was developed. H4 procedure demands to put into service 2 permanent (one for each train) interconnections of containment spray system and safety injection system, called “H4 connections”. Through the design improvement of the H4 connections, the mutual backup function of safety injection and containment spray can be realized implemented. The manual valves of the H4 connections in the original design were changed to electric valves, which ensured the accessibility of operator and avoided the radiation of high radioactivity level to operator after the accident. In addition, the improved H4 connections enable mutual backup of safety injection and containment spray in the early stage after the accident to be implemented, which fully improves the ability to respond to accidents and safety design level of the nuclear power plant (NPP). This also makes it possible to intervene early after the accident.
The Emergency Operating Strategy Study of Containment Spray Based on Radiation Level in LBLOCA Condition
Authors: Yi Ke , Li Yuanye 
Published: October 2020

Abstract
Large break LOCA will cause an increase in containment pressure, containment temperature and containment radiation level in nuclear power plants (NPPs). Containment spray is one of the most effective ways to mitigate the consequences of large break LOCA for these following facts, first, with the large space containment design, the containment spray can decrease the pressure peak and keeps containment integrity. Secondly, the containment spray can decrease the aerosol radiation level in containment, iodine in particular, and reduce the risks of radioactive release. Above all, the common strategy of containment spray in NPPs generally includes automatic actuation with high spray flow, in order to achieve good results in relevant accident conditions. Meanwhile, the strategy to shutdown containment spray should be considered as a result of these facts that, a weakened effect in decreasing radiation will occur and negative containment pressure may cause containment integrity damage in post-accident long term operation. For the above considerations, the emergency operating strategy of containment spray based on radiation level in large break LOCA condition and the relevant best estimate work are studied based on one NPP in this paper, in order to achieve reasonable results in containment spray operating strategy, which are able to optimize containment spray and reduce the bad consequences.
Study on Scattering Correction of the 60Co Gantry-Movable Dual-Projection Digital Radiography Inspection System
Authors: Ni Minzi , Li Guangchao , Wang Zhentao 
Published: October 2020

Abstract
Nuclear safety and security is getting more and more attention all over the world. Vehicles in and out of nuclear facilities need to be strictly inspected to prevent carrying of nuclear materials, explosives or other dangerous items. A new-type vehicle inspection system — 60Co gantry-movable dual-projection digital radiography inspection system is developed in Tsinghua University under the support of China’s Nuclear Energy Development Project, which uses two 60Co as radiation sources. The radiation sources are arranged at the bottom and side of the vehicles to be detected and the ionization chamber detectors are set in the two side of the gantry correspondingly. With moving of the sources and gantry synchronously, the system can obtain both side-view and upward-view images of the vehicles simultaneously [1]. However, there is a problem that the scattered rays from source of one projection can enter the detector array of another projection to form an interference signal. Experiments show that this kind of scattering noise can account for 20%, resulting in blurring and even artifacts, especially in thicker areas of the image. This problem needs to be solved urgently.
According to the characteristics of the Compton scattering effect, it is inferred that there is a certain non-linear mapping relationship between the scattering distribution of the detector array of one projection and the mass distribution in the plane of the other projection. This paper attempts to use BP neural network to learn this mapping relationship to quantitatively remove this kind of scattering noise. The results show that this method has certain effects on the removal of artifacts and blur caused by scattering. This method has the advantage of being fast and more targeted, compared with traditional image post-processing methods.
Multiphysics Analysis of Thorium-Based Fuel Performance Under Reactor Steady-State and Transient Accident
Authors: Qiu Chenjie , Liu Rong , Zhou Wenzhong 
Published: October 2020

Abstract
The ThO2 fuel has higher thermal conductivity and melting boiling point than the UO2 fuel, which is beneficial to the fast removal of heat and the improvement of fuel melt margin. In this paper, the material properties and thermodynamic behaviors of thorium-based fuel were firstly reviewed. And then the thermal physical properties and the fuel behavior models of Th0.923U0.077O2 fuel and Th0.923Pu0.077O2 fuel have been implemented in fuel performance analysis code FRAPCON and FRAPTRAN. Finally, the performances of Th0.923U0.077O2 fuel, Th0.923Pu0.077O2 fuel and UO2 fuel under both normal operating conditions and transient conditions (RIA and LOCA) are analyzed and compared. The Th0.923U0.077O2 fuel is found to have lower fuel center-line temperature and the thorium-based fuels are observed to have a delayed pellet-cladding mechanical interaction (PCMI) under steady state. Furthermore, the fission gas release, cladding strain and internal fuel energy under transient conditions are found to be lower too. Lastly, the cladding displacement and temperature under transient conditions are also compared. The thorium-based fuel was found to have a higher safety margin and accident resistance than conventional UO2 fuel under both normal operating conditions and accident conditions.
Stochastic Simulation Method for Reasoning of Dynamic Uncertain Causality Graph (DUCG)
Authors: Nie Hao , Zhang Qin 
Published: October 2020

Abstract
Dynamic Uncertain Causality Graph (DUCG) is an innovative model developed recently on the basis of dynamic causality diagram (DCD) model, which has been proved to be reliable for fault diagnosis of nuclear power plants. DUCG can represent complex uncertain causal relationship graphically, with both high efficient inference and support of incomplete expression. Therefore, DUCG is often built much larger than Bayesian Network (BN). However, as the scale of real problem is so large, DUCG still has the problem of combination explosion. Stochastic Simulation is a common solution for it. However, it is almost impossible to use traditional sampling algorithms for DUCG because the joint probability of evidences could be less than 10−20. In this paper, the algorithm based on conditional stochastic simulation for the inference of DUCG was proposed. It obtains the probability of evidences by calculating the expectation of the conditional probability in sampling process instead of using the sampling frequency, which overcomes the difficulty. What’s more, this algorithm uses recursive reasoning method of DUCG to calculate conditional probability distributions of node for sampling, which means this process only depends on its parent nodes’ states. As a result, the algorithm features in lower time complexity. In addition, it has the potential of parallelization like other sampling algorithms. In conclusion, this algorithm is promising to provide a new solution to the inference of the DUCG in large-scale and complex state situations.
CFD-DEM Simulations of Graphite Particle Collisions in Opposed Jet Mill
Authors: Peng Sifan , Liu Yujia , Gui Nan , Yang Xingtuan , Tu Jiyuan , Jiang Shengyao 
Published: October 2020

Abstract
Graphite is widely used in nuclear reactors as moderator and structural material. Among present graphite preparation methods, air flow mill is considered to be qualified in the control of particle size and purity, and promising for future mass production. In this work, an opposed jet mill is designed to crush large graphite particles. The opposed jet mill accelerates the particles through two supersonic jet flows in opposite directions, and finally the particles collide in the crushing cavity. In order to estimate the performance of opposed jet mill, it is necessary to solve the coupling calculation of the compressible flow and the collision process of discrete particles. However, the research on calculating the compressible gas solid coupling problems is scarcely rare. In this paper, coupled CFD-DEM model is used to simulate the particle movement process with jet flows and accompanying jet in opposed jet mill. By comparing with experimental results, it is proved that these simulation results of the acceleration process of compressible gas through these nozzles and the collision process of the final two supersonic jet flows in the opposed-jet mill are accurate, with the accuracy model of the coupled CFD-DEM provided. The practice has proved that the contrastive flow mill has a broad application prospect in the production of graphite particles.
Natural-Convective Transport Model of Radon From Building From Building Materials Into Indoor Atmosphere
Authors: Feng Shengyang , Chen Puxin , Cui Yu , Wu Yurong , Li Ce 
Published: October 2020

Abstract
Building material is one of the main sources of indoor radon in China. Radon exhaled from building materials enters the indoor atmosphere and transports in the indoor air space driven by buoyancy-driven airflow. This paper established a numerical model to reveal natural-convective effects on radon transport from building materials into the indoor air space. The building wall is approximated by a rectangular porous medium with uniform porosity and isotropic permeability. The buoyancy-driven flow is assumed to be turbulent and incompressible, ignoring viscous dissipation. The model can predict the dynamic coupling of radon in building materials and indoor atmosphere at different temperature gradient. This paper focuses on the analysis of radon transport with average Nusselt number and Sherwood number of interface between the building wall and the indoor atmosphere, which reveal the mechanism of radon exhalation from building materials into indoor atmosphere.
Study on Natural Circulation Flow Instabilities in Rod Bundle Channel Under Rolling Condition
Authors: Cheng Kun , Deng Jian , Cai Rong , Qian Libo , Qi Peiyao , Ke Bingzheng 
Published: October 2020

Abstract
The effects of rolling condition on the flow instability characteristics of natural circulation (NC) in rod bundle channel were experimentally studied. A 3 × 3 rod bundle channel is used as the testing section. The experimental system pressure range is 0.2 to 0.6 MPa, and the range of inlet subcooling is 10 to 70 °C. The ranges of rolling motion amplitude and period are 10 ∼ 20° and 10 ∼ 30s, respectively. Two typical two-phase flow instabilities in rod bundle channel under rolling condition were found in experiments: (a) the trough-type oscillation caused by the vapor generation at the minimum point of flow fluctuation and (b) the compound oscillation formed by the superposition of the trough-type oscillation and DWOI. Experimental results show that the rolling motion can reduce the threshold heating power of trough-type oscillation and cause the occurrence of NC flow instability in advance. But the rolling motion cannot affect the dimensionless boundary of DWOI in rod bundle channel.
New Control System of VR-1 Training Reactor
Authors: Kropík Martin , Rataj Jan , Reisner Ladislav , Matousek Jiri 
Published: October 2020

Abstract
The VR-1 training reactor with 500W maximum power operates at the Czech Technical University in Prague. Its control system was approaching obsolescence. It was therefore decided to upgrade the control system using up-to-date hardware and software using a new industrial PC with Windows 10 and the RTX real time system. The control system provides control and safety functions. Regarding the control function, the control system involves reactor I&C tests, reactor startup, reactor operation, and automatic power control. The control system receives operational data from the reactor safety (protection) system, computes average reactor power and power rate safety variables, and calculates a third safety variable that provides a percentage deviation of the reactor real power from the given (wanted) power. The control system also provides safety functions; it can shut down the reactor when safety limits are exceeded. The development of the control system software was accompanied by a quality assurance process, and verifications of the software development phases were carried out. After the control system installation, thorough non-active tests of the control system with removal of two fuel elements (because of nuclear safety) from the reactor core were carried out. Input signals for the reactor I&C (neutron chamber pulses and current) were simulated by computer controlled pulse and current generators. All safety functions, reactor start up, and operation were carefully tested. After successful non-active tests, active tests with a standard reactor core configuration were successfully conducted.