bluuebunny/crossref_metadata_2025_split huggingface dataset:
Thermal Hydraulics Analysis on a Simplified Model of Spent Fuel Pool
Authors: Liang Min , Lu Daogang , Xia Huining , Zhang Yuhao 
Published: February 2014

The spent fuel still has massive decay heat and is temporarily stored in the spent fuel pool, after unloaded from the reactor core; it is cooled by the circulating water in the spent fuel pool. The present study generally agreed that the spent fuel pool is relatively safe when the cooling water works, even if under accident conditions. It is generally considered that the thermal-hydraulic process in the spent fuel pool is very slow, and does not endanger the safety of nuclear power plant. But the boiling in the pool has been given more and more attention after the Fukushima nuclear accident; especially after the failure of core safety injection and the cooling water in the spent fuel pool is lost.
School of Nuclear Science and Engineering, North China Electric Power University intends to perform overall test bench to simulate the AP1000 spent fuel pool, the experimented data can be used to validate the results of the COSINE software calculations. In this paper, the numerical simulation is performed with CFD software to calculate the temperature, velocity, pressure field of the scaled spent fuel pool experimental bench; it can provide the guideline for the selection of the test section measuring instruments and the arrangement of the measuring points. Also, it’s essential to verify the COSINE software and guide the experimental process.
Validation of TRACE Code Against ROSA/LSTF Test for SBLOCA of Pressure Vessel Upper-Head Small Break
Authors: Xing Mian , Hu Xiao , Chen Yaodong , Li Liangxing , Ma Weimin 
Published: February 2014

OECD/NEA ROSA/LSTF project tests are performed on the Large Scale Test Facility (LSTF). LSTF is a full-height, full-pressure and 1/48 volumetrically-scaled two-loop system which aims to simulate Japanese Tsuruga-2 Westinghouse-type 4-loop PWR. ROSA-V Test 6-1 simulates a pressure vessel (PV) upper-head small break loss-of-coolant accident (SBLOCA) with a break size equivalent to 1.9% of the volumetrically scaled cross-sectional area of the reference PWR cold leg.
By building a TRACE calculation model of LSTF and PV upper-head, the paper dedicated to assess the effect of different modeling options and parameters on simulating thermal hydraulic behaviors of TRACE code. The results show that TRACE code well reproduce the physical phenomena involved in this type of SBLOCA scenarios. Almost all the events in the experiment are well predicted by the model based on TRACE code. In addition, the sensitivity of different models and parameters are investigated. For example, the code slightly overestimated the break mass flow from upper head which could affect the accuracy of the results significantly. The rising of core exit temperature (CET) is significantly influenced by the bypass flow area between downcomer and hot leg.
Development of Numerical Simulation for Jet Breakup Behavior in Complicated Structure of BWR Lower Plenum: 2 — Flow Observation With Visualized Experimental Apparatus
Authors: Saito Ryusuke , Abe Yutaka , Kaneko Akiko , Suzuki Takayuki , Yoshida Hiroyuki , Nagase Fumihisa 
Published: February 2014

In order to improve the safety of Boiling Water Reactor (BWR), it is required to know the behavior of the plant when an accident occurred as can be seen at Fukushima Daiichi nuclear power plant accident. Especially, it is important to estimate the behavior of molten core jet in the lower part of the containment vessel at severe accident.
In the BWR lower plenum, the flow characteristics of molten core jet are affected by many complicated structures, such as control rod guide tubes (CRGT), instrument guide tubes and core support plate. However, it is difficult to evaluate these effects on molten core jet experimentally. Therefore, we considered that multi-phase computational fluid dynamics approach is the best way to estimate the effects on molten core jet by complicated structure.
The objective of this study is to develop the evaluation method for the flow characteristic of molten core jet including the effects of the complicated structures in the lower plenum. So we are developing a simulation method to estimate the behavior of molten core jet falling down through the core support plate to the lower plenum of the BWR. The method has been developed based on interface tracking method code TPFIT (Two Phase Flow simulation code with Interface Tracking). To verify and validate the applicability of the developed method in detail, it is necessary to obtain the experimental data that can be compared with detailed numerical results by the TPFIT. Thus, in this study, we are carrying out experimental works by use of multi-phase flow visualization technique.
For the experiment works, we constructed two experimental apparatuses, one is single-channel experimental apparatus and the other is multi-channel experimental apparatus. The single-channel experimental apparatus simply simulate single flow channel between four CRGTs. The multi-channel experimental apparatus is a 1/10 planar type model of a lower plenum of BWR. In the experiments, we will obtain the jet breakup behavior and surrounding velocity profiles of the jet by using LIF and PIV.
In this paper, the outline of two-experimental apparatuses are shown. And the results of the single-channel experimental apparatus with/without modelled complicated structures are also shown. In the results, it was confirmed that the complicated flow channel affects the jet injection and breakup behavior. And it was also confirmed that the complicated structures restrain diffusion of fragments of the jet.
Factors Promoting Embrittlement of T91 Steel in Liquid Lead-Bismuth Eutectic Alloy
Authors: Ye Changqing , Vogt Jean-Bernard , Proriol-Serre Ingrid 
Published: February 2014

The mechanical behavior of the T91 martensitic steel has been studied in liquid lead-bismuth eutectic (LBE) and in inert atmosphere. Several conditions were considered to point out the most sensitive embrittling factor. The Small Punch Test technique has been employed by using smooth specimens.
T91 appeared in general as a ductile material, and became brittle in the considered conditions only at least, but not necessarily, if the test was performed in LBE.
It turns out that the loading rate appeared as a critical parameter for the occurrence of liquid metal embrittlement of T91 in LBE. For standard heat treatment condition, loading the T91 very slowly instead of rapidly in low oxygen LBE resulted in brittle fracture.
Nuclear Safety Culture Construction in Nuclear Power Design Enterprise on the Perspective of Knowledge Workers in China
Authors: Li Qiu , Wu Jun 
Published: February 2014

Design plays a leading role in the whole life cycle of nuclear power production, including site selection, design, manufacture, construction, operation and decommission. The quality of the design products directly affects the intrinsic safety of the nuclear power plant. The quality of final products in design process depends on a number of factors, including not only technology capability, but also the improvement of design quality assurance system, the capability and responsibility of the designers to nuclear safety, the thinking mode and working habit of the designers, and the extent of complete implementation of the review systems at all levels. These factors consequently affect the level of safety of the nuclear power plant.
Nuclear power design industry is typical knowledge-intensive business, in which knowledge workers consist of essential assets to the enterprise. The article analyzes the asset structure and staff structure of nuclear power design industries in contrast with other industries, and discusses the contribution of knowledge workers to the development of the enterprise. This paper also tries to document the characteristics of knowledge workers in nuclear power enterprises. They are characteristic of superior income and welfare, high level education and ability, high specialization, and tacit knowledge.
Based on these analyses, the article addresses four major principles in nuclear safety culture construction for nuclear power design enterprises as follows: 1. adopt motivation factors as major incentives, 2. emphasize self management, 3. management by objectives (MBO), 4. team work.
Design Analysis of Corium Immobilization and Cooling of the Reactor Vessel in Severe Accidents at NPP With Commercial Unit V-320
Authors: Stremedlovskyi Roman A. 
Published: February 2014

Keeping the melt inside of the reactor vessel is the key strategy of the management of severe accidents for the exploited, as well as for the newly designed reactors with pressurized water. The concept has been substantiated for emergency cooling of VVER-440 reactor at Loviisa NPP in Finland [1]. However, for more powerful reactor, such as VVER-1000, up to now it was considered that this concept did not work.
This work analyzes the problem of the interaction of molten corium with the reactor vessel. The work is representing the results of studies that were carried out as part of research with a goal to validate design functions of the system of containment the melt inside the VVER vessel.
The VVER or WWER is a series of PWR designs originally developed in the Soviet Union, and nowadays in Russia. The reactor plant of VVER-1000 type of V-320 series is a reactor plant with a water-cooled and water-moderated power reactor for NPP. It has an electric power of 1000 MW and a thermal power of 3000 MW. The plant consists of four loops, each one with a horizontal SG and a MCP. In the work the design analysis of the accident with melting of the core of the reactor plant V-320 was conducted on codes RELAP/SCDAPSIM/MOD3.4 and MELCOR. The main processes that determine the state of the core melting in the reactor vessel were analyzed. The main results of the calculation of the molten corium interaction with the wall of the reactor vessel were represented. The results of the preliminary calculated estimation of the melt behavior and the technical ability to hold the core melt in the vessels of PWR reactors were represented.
Based on the carried out analysis and implementation of technological and design study, the recommendations were made with a purpose to improve the safety of NPP with a reactor plant V-320 using the systems of melt restraining and cooling of the reactor vessel.
Simulation Research on Combustion of Liquid Sodium Droplet
Authors: Wang Chao , Zhang Zhi-gang , Peng Kang-wei , Guo Ming , Sun Shu-bin 
Published: February 2014

Since sodium leak and sodium fire belongs to incidental accident in fast reactor. This paper made the simulation research on sodium droplet combustion, which was the most fundamental one of the analysis on sodium spray fire. Referencing to the sodium droplet combustion experiment, this paper made improvement in the traditional model of the sodium droplet and researched on the initial temperature and oxygen fraction on oxidation behavior of sodium droplet.
This paper used the new spherical model with cones on the sodium droplet surface and improved the spherical model used in the traditional calculation of sodium droplet. This made the simulation was closer with the actual combustion situation.
According to the characteristics of the reaction between the sodium and oxygen, this paper modeled sodium droplet combustion process, which was divided into three stages, namely the surface oxidation stage, pre-combustion stage and combustion stage, set up the mathematical model, compiled the calculation program and got the regulation of the sodium droplet temperature changed with time under different initial temperatures (200–500 °C) and different oxygen fractions (4–20%). This program was validated with the sodium droplet combustion experiment. Through using the proportion γ of the sodium mass participated in the oxidation reaction to vaporizing reaction in the unit time, under the different conditions of sodium droplet combustion, this paper used different γ in different reaction stage, got the simulation results, which were consistent with the experimental data and analyzed the changing regularity of γ.
Study on the Nonradioactive Environmental Impact Assessments for Chinese Nuclear Power Plants
Authors: Wei Xinyu , Ma Hongbin , Xiong Xiaowei , Yang Duanjie , Shang Zhaorong 
Published: February 2014

In China, the radioactive environmental impact assessment (EIA) for nuclear power plants (NPPs) had been carried out in detail due to the specific characteristic of radiation and the social universal attention. However, the nonradioactive environment impact assessment for NPP doesn’t get enough attention. This should be improved, since the operation of NPPs could cause some serious nonradioactive environment impacts.
Based on the investigation of EIA for American NPPs, the following suggestions were put forward for Chinese EIA: (1) the laws and regulations in China need to be revised for the EIA of NPPs; (2) the statistics of chemicals or biocides used in the operation of NPPs and their discharged concentration need to be tested; (3) the impact assessments of nonradioactive gas emissions as well as hazardous wastes need to be carried out to meet the national requirements; and (4) the monitoring of nonradioactive waste also needs to be carried out for NPPs to obtain the basic data.
The Use of Risk-Informed Inspection Approaches of Ukraine
Authors: Hryshchenko Bogdan , Polianskyi Mykhailo , Nosovskyy Anatoliy , Sevbo Oleksandr 
Published: February 2014

Currently in Ukraine inspection activity is based on deterministic conservative principles, operation experience and expert appraisal of the inspector. Possibility and benefits of PSA are used with low efficiency.
Results of the study conducted by IFC (International Financial Corporation) indicate the absence of a risk-informed approach in the practice of inspections in Ukraine. Also, according to the World Bank appraisal of the investment climate in Ukraine it should be concluded that until regulatory authorities begin implementation of a risk-informed approach in planning inspections in Ukraine random unscheduled events will dominate and won’t provide the goal of State Inspection.
Information which obtained from the PSA helps to direct human and financial resources to the problems research that are the most important for safety, and to eliminate or reduce the requirements, which will reduce expenses of solution for significant issues.
Inspection in the planning, preparation, implementation and evaluation of the results of which, in addition to deterministic estimates, operating experience and expertise evaluating risk are used is called the risk-informed inspection. The use of risk-informed approaches allows to: focus inspections on design and operational aspects, which have dominant influence on the safety of nuclear power plants (NPP); improve the schedule of inspections (recording of risk assessments in determining the scope, frequency, and type of inspection); and use an additional source of information on the systems and the components of power units, personnel availability.
Applying a risk-informed approach to inspection, inspectors can focus primarily on systems that make the largest contribution to core damage frequency, failure of which leads to significant increase of CDF. Risk-informed approach allows to select the most important elements to test systems that will increase the efficiency and quality of inspections.
Based on the above it can be concluded that the experience of inspector, his knowledge of the power unit design, the process, the mechanisms of failure of equipment and of accident running, the use of information on the importance of components and systems for the safety of nuclear power plants, obtained from PSA — an effective way to achieve the best results in improving safety.
Analysis and Evaluation of the Setting Value Drift of SEBIM Relief Valve Used in Nuclear Power Plant
Authors: Yang Xiaochuan , Luo Chunlei , Qu Meng , Shen Rongfa , Che Yinhui 
Published: February 2014

SEBIM relief valve is mainly used in safety important system of the nuclear power plant to provide the overpressure protection for the pressure vessel and pressure pipeline which related to the nuclear safety. It is probable that the system will lose the overpressure protection function and lead to a major security hidden danger in the nuclear power plant if the relief valve setting value happens drift. So the problem of the SEBIM relief valve setting value drift has becoming the focus of attention of the whole nuclear industry.
It is of great significance to study the setting value drift condition of SEBIM relief valve in the practical application. This paper analyzes the structure and operating principle of the SEBIM valve, and studies the main failure mode of the setting value drift of SEBIM relief valve combining with the operating environment of the equipment. And this paper collects the setting value check data of the SEBIM valve used in two domestic nuclear power plants (These two power plants are named NPP A and NPP B in this paper) during its operation period, and carries on the analysis and evaluation.
Research and Development of Self-Priming Venturi Scrubber for Filter Venting: Preliminary Analysis and Observation of Hydraulic Behavior in Venturi Scrubber
Authors: Horiguchi Naoki , Yoshida Hiroyuki , Uesawa Shin-ichiro , Kaneko Akiko , Abe Yutaka 
Published: February 2014

As revealed by Fukushima Daiichi nuclear disaster, countermeasures against severe accident in nuclear power plants are an urgent need. In particular, from the viewpoint of protecting containment and suppressing the diffusion of radioactive materials, it is important to develop the device which allows filtered venting of contaminated high pressure gas. In the filtered venting system that used in European reactors, so called Venturi scrubbers are used to realize filtered venting without any power supply. The Venturi scrubber operates without a power supply of high pressure gas filled into the containment. In this apparatus, scrubbing of contaminated gas is promoted by both gas releases through a submerged Venturi tube which is one of the major components of Venturi scrubber and liquid splay flow formed by liquid suction through a hole for suction provided at the throat part of the Venturi tube. This type of Venturi scrubbers is called self-priming ones. However, the mechanism of a self-priming Venturi scrubber including effects of gas flow rate and shape of Venturi tube is understood insufficiently in the previous studies.
In this study, to provide knowledge about the detailed mechanism of filtering and evaluation method for operating condition of the filtered venting system with Venturi scrubber, both experimental and numerical research works have been performed. In this paper, as a previous step of designing and making a self-priming Venturi tube, hydraulic behavior in a rectangular Venturi tube is taken by high speed camera, by the air-water experiment under atmospheric condition to check the operation in a self-priming mode and the validity of the experimental method and the visualization technique. And numerical analysis of a circular Venturi tube is conducted as a preliminary analysis, by TPFIT (Two-Phase Flow simulation code with Interface Tracking) to predict its operation. As results, the Venturi scrubbers used in experiment and simulated in numerical analysis were operated in a self-priming mode. By observed results of the hydraulic behaviors in the Venturi tube, the validity of the experimental method and the visualization technique were confirmed. And it is expected that the simulated circular Venturi scrubber in this numerical simulation was operated in a self-priming mode in a real situation.
Calculation of External Radiation in Dry Spent Nuclear Fuel Storage Container
Authors: Zaiets V. 
Published: February 2014

Zaporozhye Nuclear Power Plant (NPP) commissioned a Spent-Fuel Dry-Storage Facility (SFDSF) with a 50-year service life. The SFDSF design is based on the proven technology of the US Duke Engineering & Services Company and meets international standards, rules and regulations on nuclear and radiation safety [1].
The objectives of this paper are calculation of the external neutron and γ-radiation of a single container, and prediction of radiation characteristics during a storage period up to 50 years.
The calculations were made for VSC-24 special ventilated containers manufactured by Sierra Nuclear Corporation (USA) [2].
The analysis shows that the change in gamma energy spectrum results in a faster decrease in dose with time which is in agreement with observations in the SFDSF. The γ-radiation dose rate is mainly composed of 0.5 to 2.5 MeV quanta.
Selection of Chemical Hazard Prediction Models for Plant Safety and Control Room Habitability Evaluations
Authors: Richmond Mary C. , Wan Ping K. , Dague Brady P. , Davis Kyra W. 
Published: February 2014

Assuring the protection of people and the environment from unnecessary exposure to radiation is of great concern to nuclear electric power generators and regulators. In order to secure and maintain a license for operation of a nuclear power plant in the United States, the applicant is required to assess postulated scenarios to determine if an accidental chemical release either onsite or offsite will result in a design-basis event. When determining design-basis events, accident categories such as fire, explosion and/or toxic vapor cloud formation are considered.
Evaluations must consider whether the postulated accidental chemical release will result in operator impairment or damage to safety related structures, systems or components (SSCs), either of which may prevent safe shutdown of the nuclear plant. A critical step in performing such safety evaluations is selection of an appropriate model which will most closely reflect the behavior of a chemical release in the scenario under consideration. Many chemical dispersion models are available for use in safety evaluations for accidental chemical releases; however, it is imperative that the model selected be appropriate for the postulated release conditions. Model selection should be based on careful evaluation of factors such as release location, meteorological conditions, terrain, chemical inventory and storage conditions, as well as the physical and chemical properties of the chemical under consideration in the postulated release.
Failure to select a model suitable for the conditions under which the chemical is released or to appropriately evaluate the physical properties of the chemical and the corresponding limitations of the model may result in underpredicting or overpredicting the impact of a fire, explosion or toxic chemical release. Underpredicting could leave the facility susceptible to damage of safety related SSCs or lead to operator impairment both of which may affect the ability of the plant to safely operate following an accident. While overpredicting the impacts could lead to unnecessary and costly overdesign. This paper will address the considerations that must be evaluated when selecting a chemical dispersion model and will illustrate the importance of model selection in performing nuclear safety evaluations through examples and case studies.