bluuebunny/crossref_metadata_2025_split huggingface dataset:
The Preliminary Experimental Study Based on Failure Mechanism of the RPV Lower Head
Authors: Meng XianKe , Fei LiKai , Zhang Aijing , Xiong SiJiang , Cui Lei , Chen YaoDong 
Published: October 2016

In-Vessel Retention is a key severe accident management strategy for reactors such as AP/CAP series reactors. The IVR success evaluation criterion is whether the RPV is melted through or not at the final RPV state. Once the RPV lower head melt through, the liquid corium will flow into the reactor cavity and will lead to complex phenomena, such us steam explosion and the reaction between the corium and concrete. These will make temperature and pressure of the containment vessel rise quickly and is a threat to the integrity of the containment vessel. When the wall surface of RPV lower head heating condition exceed the critical heat flux, the temperature rises rapidly, it is generally assumed that the RPV lower head in this state will inevitably melt through. This is the so-called IVR failure.
In order to study the possible failure modes and mechanism of RPV lower head under the IVR measures, an experimental facility called TRECT is built. By measuring the parameters such as temperature, flow of the test section to study the influence to CHF by the parameters such as flow velocity and angle. All of these can provide reliable basis to the effectiveness appraisal and model development on the area of severe accident mitigation measures (IVR). To be specific, the test section is rectangular channel whose section is 50 × 20 mm. The upper surface is the heat surface and using a direct current heating mode to supply heat power. The heat flux can reach 1.5MW/m2. We use this upper surface heated rectangular channel to simulate RPV ERVC channel. By adjust the angle of test section to simulate the different circum ferential location of RPV lower head. And the Adjusting range can be 0° to 90°. The experimental results show that flow rate was reduced by 11% in the experiments, the critical heat flux density increased by 4.5%. Inclined angle increased from 16° to 29°, CHF increased by 7.9%.
Effects of External Grid Disturbances to Nuclear Power Plants
Authors: Arians Robert , Arnold Simone , Mueller Christian , Quester Claudia , Sommer Dagmar 
Published: October 2016

The reliability of the auxiliary power supply of a nuclear power plant (NPP) is of high importance for safe operation. The loss of the electrical power supply is one of the major contributions to the calculated core damage frequency in probabilistic safety assessments. Among others, the events in Forsmark in 2006 [1] and 2012 [2] as well as in Byron in 2012 [3] illustrate that disturbances in the external power grid can propagate into the NPP and have an impact on the safety important electrical equipment. Therefore, the grid reliability contributes considerably to the reliability of the auxiliary power supply.
In the research work presented in this paper the international operating experience has been evaluated concerning events which include disturbance in the external grid to discover those types of grid disturbances which may have influence on the safe operation of the NPPs. The identified events have then been categorized within a developed classification scheme to determine those with the highest relevance. Based on this scheme representative scenarios of grid disturbances have been developed.
The investigation of the impact of the developed scenarios on the electrical equipment of NPPs will be performed using a grid analysis, planning and optimization tool which also allows executing dynamic simulations of electrical grids [4]. Therefore, a generalized auxiliary power supply of a pressurized water reactor was modeled according to German NPPs of the type Konvoi.
In this paper, an overview of the developed scenarios of grid disturbances and the actual status of the simulation of the auxiliary power supply of NPPs is presented.
Development of Prediction Technology of Two-Phase Flow Dynamics Under Earthquake Acceleration: (17) Effect of Structure Vibration on a Rising Single Bubble
Authors: Miyazaki Akifumi , Kato Yuki , Kanagawa Tetsuya , Kaneko Akiko , Monji Hideaki , Yoshida Hiroyuki , Abe Yutaka 
Published: October 2016

Earthquake is one of the most serious phenomena for safety of a nuclear reactor in Japan. Structural safety of nuclear reactors has been studied and nuclear reactors were contracted with structural safety for a big earthquake. However, it is not enough for safe operation of nuclear reactors because thermal-fluid safety is not confirmed under the earthquake. For instance, behavior of gas-liquid two-phase flow is unknown under the earthquake conditions. Especially, fluctuation of void fraction is an important factor for the safe operation of the nuclear reactor. In the previous work, fluctuation of void fraction in bubbly flow was studied experimentally and numerically. In case of the earthquake, the fluctuation is not only the flow rate, but also body force on the two-phase flow and shear stress through the pipe wall. Interactions of gas and liquid through their interface also act on the behavior of the two-phase flow. The fluctuation of the void fraction is not clear for such complicated situation under the earthquake.
Our study has investigated the behavior of gas-liquid two-phase flow experimentally and numerically. In this paper effects of vibration on bubbly flow in the components and construct experimental database for validation and performs visualization experiments of a rising single bubble in a rectangular water tank on which sinusoidal vibration was applied. In this paper, results of visualized experiment evaluated by the visualization techniques, including positions of a bubble, a shape of the bubble and the bubble tilt angle were shown. In the results, bubble behavior were affected by the table oscillation. The bubble tilt angle is also almost same value of the bubble movement angle. It is implied that higher table oscillation frequency than 20 Hz quite weakly affects on fluctuation of bubble tilt angle frequency.
Parametric Study of the Burnup Credit Criticality Analysis in BWR Fuels
Authors: Wu Shang-Chien , Chao Der-Sheng , Liang Jenq-Horng 
Published: October 2016

The purpose of this study is to investigate the effects of operation parameters on BWR fuel reactivity and nuclide inventory. The three high-priority parameters that are suggested by NRC/ORNL [1], that is, axial burnup profile, axial moderator density profile, and control blade insertion during operation, were examined in order to evaluate their impact on BWR burnup credit. All of the calculations were performed using SCALE6.1 based on the ENDF/B-VII nuclear data library and involved 238 energy groups.
The results indicate that the axial burnup profile is the most critical factor in reactivity when the burnup value is greater than 10 GWd/MTU. The insertion of control blades during operation is also more important for the reactivity of spent fuel as compared to the influence of the axial moderator density profile. At a burnup of 60 GWd/MTU, these three operation parameters (i.e., axial burnup profile, axial moderator density profile, and control blade insertion during operation) in turn lead to multiplication factor differences of 18850, 7600, and 1980 pcm, respectively, for set2 nuclides. The influence between different parameters was also investigated in this study. The results reveal that multiple operation parameters affect the reactivity reduction tendency. In addition to the three parameters investigated in this study, other important factors related to burnup credit analysis are also currently under investigation.
Jet Breakup and Droplet Formation in Immiscible Liquid-Liquid System
Authors: Saito Shimpei , Iwasawa Yuzuru , Abe Yutaka , Kaneko Akiko , Kanagawa Tetsuya , Nariai Hideki , Koyama Kazuya 
Published: October 2016

Mitigative measures against the event of a core disruptive accident (CDA) are of the importance from the viewpoint of safety of a sodium-cooled fast reactor (SFR). If the CDA occurs, the so-called post-accident heat removal must be surely achieved. The present study focuses on the scenario that the molten materials are injected into the lower plenum as jets. The jet breakup behavior during the CDA will be very complicated. Therefore, a specialized study on the fundamental process during the jet breakup is believed to be an effective approach.
The aim of this paper is to understand the fundamental process of hydrodynamic interaction of jet breakup and droplet formation Using the immiscible liquid-liquid system, water and silicon oil as the test fluids, visualization via high-speed videography was performed. From the visualization results, the breakup length and droplet diameter were measured by image processing. The experimental data were scaled with ambient Weber number. When the Weber number was smaller than 1, the droplet diameter was close to the nozzle diameter, and distribution of droplet size was not observed. When the Weber number exceeded 1, the breakup length became longer and the generated droplet diameter possessed a distribution with two peaks due to satellite droplet formation. In both cases, the droplet formed at the leading edge of jet. In case that Weber number is around 100, the droplets were formed by entrainment of interfacial wave at jet side. From the mass median diameter data, we can see that the increase of the Weber number caused the decrease of median diameter and the increase of the width of the distribution.
Design and Construction of a Test Rig to Analyze Thermal-Hydraulic Phenomena in Two-Phase Flow
Authors: Lanz Thomas , Alobaid Falah , Epple Bernd 
Published: October 2016

Heat transfer, pressure drop, and instability phenomena in water-steam two-phase flows are important research areas with regard to thermal power plants. Here, demineralized water is vaporized to operate steam turbines that drive generators, converting the energy from fuel that is burned into electricity. To design the evaporation process, it is important to know the heat transfer coefficients of water-steam-mixtures and the pressure drop due to friction, elevation and momentum in terms of steam quality. Moreover, occurring instabilities need to be understood to prevent unstable and unsafe operation conditions. The aim of this research work is to design and construct a test rig to analyze these thermal-hydraulic phenomena. Furthermore, the obtained results will be applied to validate the heterogeneous flow model of the dynamic simulation software APROS (Advanced Process Simulation Software) from VTT (Technical Research Center of Finland).
In contrast to previous studies in this field, the fluid used for the experiments is demineralized water — instead of usually used refrigerants — to eliminate possible side effects related to different fluid properties. Furthermore, the heated test section of the test rig is a thermal-hydraulic scale down from a thermal power plant’s low-pressure evaporator tube. The focus lies on a huge parameter variability in terms of operation conditions (sub-cooling, inlet pressure, mass flow, heating power, upstream compressible volume) and test rig layouts (vertical and horizontal test section, different lengths and inner diameters, single and parallel tubes) to analyze their impact on the different instability phenomena (density-wave oscillations, pressure-drop oscillations). Literature dealing with experimental investigation of two-phase flow in horizontal evaporator tubes exposed to natural circulation at low pressure is rare. This work aims to contribute especially to this field of research.
Currently, the designed test rig is under construction. Once the experiments are performed, the results will be analyzed by means of stability maps and fast Fourier transformation (FFT) amongst others. The results will then be used to validate the heterogeneous six-equation model of the dynamic simulation software APROS with regard to heat transfer, adiabatic pressure drop and flow instabilities. The six-equation flow model (also known as Euler-Euler approach) implies a considerably higher accuracy than the usually regarded homogeneous three-equation model.
Preliminary Benchmark Validation of a Lattice Physics Code COSLATC
Authors: Yang Tongrui , Liu Zhanquan , Yan Yuhang , Li Shuo , Yu Hui , Chen Yixue 
Published: October 2016

COre and System INtegrated Engine for design and analysis (COSINE) is developed by State Nuclear Power Software Development Center (SNPSDC), which is an integrated nuclear engineering code package. A lattice physics code named COSLATC is an essential part included in COSINE. COSLATC is a multi-group two-dimensional transport code. Pin cells and assemblies of Pressurized Water Reactor (PWR) can be calculated by COSLATC. It is used to calculate few group constants and nuclei densities for core simulator.
In order to make sure the quality of the COSLATC, a strategy of verification & validation (V&V) are discussed and applied to the COSLATC. Firstly a V&V requirement analysis is performed. A test matrix considering the variety of fuel enrichments, materials, geometric and working conditions is generated. Then the corresponding benchmark is collected and classified. Finally, the numerical results of COSLATC and the reference values are compared and analyzed. According to the validation strategy discussed above, the preliminary validation is carried out.
The benchmark provides consistent and comprehensive tests for high burnup (approx. 70GWd/t) fuels of PWR. By comparison with k-infinity and isotopic composition in the UO2 and MOX fuels pin cell problems or assembly problems benchmark, the result shows that the calculated results from COSLATC code agree well with reference results. COSLATC will be an important lattice calculation code in the scientific calculation and engineering application.
Advanced Cooling Tube Mobile Heat Exchanger in the Outer Shape of a Fuel Assembly for Spent Fuel Pool Cooling
Authors: Reuter Matthias , Ornot Leo , Maier Torsten , Fuchs Thomas , Gauter Kerstin , Reck Markus , Redmond Joseph 
Published: October 2016

After the Fukushima accident the development and implementation of additional safety features has become more and more important.
In March 2014 AREVA filed a patent for the invention of a mobile heat exchanger module in the outer shape similar to the outer geometry of a fuel assembly called Advanced Cooling Tube (ACT) [1]. Since January 2015 there is a full scale test facility in the German headquarters of AREVA in Erlangen to test the performance of such a mobile heat exchanger. In spring 2015 there were investigations running to implement the ACT into the spent fuel pool of a nuclear power plant in Switzerland. In the meantime, these investigations were finished and a contract has been signed.
This paper gives an overview about the main facts of the Advanced Cooling Tube as a mobile heat exchanger in the shape of a fuel assembly for spent fuel pools, how it can be implemented in any nuclear power station and what the advantages of this product are. Further information about the performance and requirements of this special heat exchanger will be given.
Study on Differences in Floor Response Spectra due to the Seismic Input Approach
Authors: Bochert Jana Sue , Schau Henry , Schmitt Timo 
Published: October 2016

In this paper the differences of floor response spectra (FRS) resulting from different ground response spectra are discussed. These spectra include the site effects which are quantified via site response analysis. This response is generated by wave propagation from the base rock through the overlying soil layers to the surface. The influences of the different layers and the corresponding dynamic soil properties are considered by using wave propagation analysis.
The paper then discusses the results obtained from seismic input at different depths conditions. Similar results might be expected, because the depth of the input spectra is adjusted for each layer. However, in comparing the floor response spectra of these calculations, significant differences are observed and therefore interpreted.
The paper is completed with the explanation of these significant differences and also with comparable floor response spectra.
Enhancing the Plant’s Capability for Design Extension Conditions Based on Best Estimate Context Evaluation of Human Performance in ATWS Events
Authors: Petkov Gueorgui 
Published: October 2016

The experience of severe accidents shows that reliable determination of technological process parameters is necessary but not always sufficient to avoid catastrophic consequences. The accident measures should be considered in a broader context that includes the human factor, organization of the nuclear technology, external influences and safety culture.
The anticipated transient without scram (ATWS) events were not considered in the original WWER (Russian PWR) design basis accidents. The design extension conditions scenarios progress in a context which is very uncertain and highly stressful for the operators. If a specific scenario requires some operators’ actions as measures to mitigate, delay or distribute the accident consequences then the dynamics of accident context seem of primary importance for ‘best estimate’ evaluations and enhancing the plant’s capability.
The paper presents the capacities of the Performance Evaluation of Teamwork (PET) procedure for enhancing plant’s capability for design extension conditions based on ’‘best estimate’ context evaluation of human performance in ATWS events. The PET procedure is based on a thorough description of symptoms of various timelines and their context quantification. It is exemplified for different ATWS scenarios of the NPP with WWER-1000 based on thermal-hydraulic simulations with RELAP5/MOD3.2 code and models.
Assessment of SIMMER-III Code Based on Steam Generator Tube Rupture Experiments in LIFUS5/Mod2 Facility
Authors: Pesetti Alessio , Del Nevo Alessandro , Forgione Nicola 
Published: October 2016

An experimental campaign investigating the postulated Steam Generator Tube Rupture (SGTR) event, in relevant configurations for Heavy Liquid Metal Reactors (HLMRs), was carried out in the separate-effect facility LIFU5/Mod2, at ENEA CR Brasimone. Ten tests were performed injecting pressurized subcooled water into the reaction tank partially filled by Lead-Bismuth Eutectic alloy (LBE) at 400°C with a cover gas of argon at about 2 bar. Fast pressure transducers, thermocouples and strain gages provided high-quality measurement data for improving the phenomena understanding and supporting the development and validation phase of computer codes for SGTR numerical simulation.
The experimental campaign is composed by two series of tests, characterized by different water pressure: 40 and 16 bar. The first two tests belonging to the low pressure experiments are presented, highlighting the pressurization time trends of the water injection tank, injection line and reaction vessel. The injected water mass flow rate and temperature trends in the reaction vessel were measured. The former test is the reference one and the latter was carried out for investigating the injection of water with higher sub-cooling.
A post-test analysis of the two mentioned tests was carried out by SIMMER-III code. The pressure profile in the water injection tank was set as boundary condition of the calculation. The numerical analysis provided injection line and reaction tank pressurization in agreement with the experimental data. The lower water temperature test provided a better accordance with the measured data, due to the lower evaporation along the injection line. The SIMMER-III analysis also studied the water-LBE interaction from the volume fraction point of view and the energy released in the total reaction tank and in its cover gas.
Effect of Soil Structure Interaction on the Seismic Fragility of a Nuclear Reactor Building
Authors: El-Bahey Samer , Alzeni Yasser 
Published: October 2016

In recent years, the nuclear industry and the Nuclear Regulatory Commission (NRC) have made a tremendous effort to assess the safety of nuclear power plants as advances in seismology have led to the perception that the potential earthquake hazard in the United States may be higher than originally assumed. The Seismic Probabilistic Risk Assessment (S-PRA) is a systematic approach used in the nuclear power plants in the U.S. to realistically quantify the seismic risk as by performing an S-PRA, the dominant contributors to seismic risk and core damage can be identified. The assessment of component fragility is a crucial task in the S-PRA and because of the conservatism in the design process imposed by stringent codes and regulations for safety related structures, structures and safety related items are capable of withstanding earthquakes larger than the Safe Shutdown Earthquake (SSE). One major aspect of conservatism in the design is neglecting the effect of Soil-Structure-Interaction (SSI), from which conservative estimates of In-Structure Response Spectra (ISRS) are calculated resulting in conservative seismic demands for plant equipment.
In this paper, a typical Reactor Building is chosen for a case study by discretizing the building into a lumped mass stick model (LMSM) taking into account model eccentricities and concrete cracking for higher demand. The model is first analyzed for a fixed base condition using the free field ground motion imposed at the foundation level from which ISRS are calculated at different elevations. Computations taking into account the SSI effects are then performed using the subtraction method accounting for inertial interactions by using frequency dependent foundation impedance functions depicting the flexibility of the foundation as well as the damping associated with foundation-soil interaction. Kinematic interactions are also taken into account in the SSI analysis by using frequency dependent transfer functions relating the free-field motion to the motion that would occur at the foundation level as the presence of foundation elements in soil causes foundation motions to deviate from free-field motions as a result of ground motion incoherence and foundation embedment.
Comparing the results of the seismic response analyses, the effects of the SSI is quantified on the overall seismic risk and the SSI margin is calculated. A family of realistic seismic fragility curves of the structure are then developed using common industry safety factors (capacity, ductility, response, and strength factors), and also variability estimates for randomness and uncertainty. Realistic fragility estimates for structures directly enhances the component fragilities from which enhanced values of Core Damage Frequency (CDF) and Large Energy Release Frequency (LERF) are quantified as a final S-PRA deliverable.
ESS Target Helium Experiments at Lund University
Authors: Nilsson Per , Schönborn Alessandro , Klingmann Jens 
Published: October 2016

In ESS, a 5 MW proton beam will hit a helium cooled tungsten target to generate neutrons by spallation. For the purpose of investigating various aspects of cooling with helium, an experimental system has been constructed at Lund University, Department of Energy Sciences. A helium flow of 3 g/s is circulated in a closed loop with pre-heater, cooler, filters and a test vessel. The vessel has windows for measurements and is designed for 10 bar and above 400 °C. In the test vessel, a tube forms a helium jet of up to 150 m/s onto a coin shaped material sample.
One purpose of the system is to investigate erosion of sample surfaces. Drilling and attaching equipment to the surface had to be avoided, so the setup has been benchmarked and the comparison was used to tune a CFD model. Then the model is used to assess the conditions in the experiments.
The setup, as well as the tests and calculations, are described and it is concluded that the temperature and velocity estimates are sufficient for the experiments to be performed.