bluuebunny/crossref_metadata_2025_split huggingface dataset:
Structure Innovation and Thermal Hydraulics Analysis for Cold Trap of CEFR
Authors: Zu-guo Chen , Yi-jun Xu , Wen-jing Qi 
Published: October 2018

As one of the significant equipment in CEFR operation system, cold trap plays an important role in purifying the coolant sodium and ensuring the safe and steady operation of the reactor. After several years’ of operation, the cold trap of CEFR can’t work very well as it has been designed previously, which results in the low impurity aggradation efficiency, being easily blocked and frequent replacement. Analyzing the structure of the current cold trap, there are two possible reasons causing the problems mentioned above. The first one is that the sodium flow path is improper in the cold trap resulting in the blockage in specific area. The second one is that the density and layout of wire mesh didn’t take the zoning scheme, which would make the precipitation of impurities become easier in exterior area and the flow path is thus blocked.
Based on the two reasons, a modified scheme is proposed from the view of structure and hydraulics to improve the impurities precipitation mode so that the lifetime of the cold trap can be prolonged. The modified scheme is to block about 1/3 orifices on interior cylinder wall in wire mesh zone and to open two rows of diversion holes on the upper support plate of wire mesh so that the velocity distribution can be improved. Meanwhile, in order to make impurities precipitate from the bottom of the cold trap, the wire mesh is arranged to be in upper, middle and lower areas using different wire mesh density. By 3-D modeling and numerical analysis and then comparing the results to that of current cold trap of CEFR, it can be concluded that the velocity distribution in modified scheme is more reasonable and the new layout of the wire mesh can meet the impurities precipitation requirements better. Further comparative analysis with the results under the actual operational condition will be carried on after the completion of the manufacture for modified cold trap.
A Methodology to Determine SMR Build Schedule and the Impact of Modularisation
Authors: Lloyd Clara A. , Roulstone Anthony 
Published: October 2018

Light-water cooled Small Modular Reactors (SMRs) are a potential game-changing technology for energy supply. The potential benefits of SMRs are however conditional on solving the key standardisation and construction issues that have troubled large reactor (LR) projects, which have in turn led to high build costs and long project durations.
Initiatives to determine the build schedule of SMRs are hindered by a lack of SMR construction experience and related data. The methodology used in this paper, to deal with the lack of SMR-specific data, draws conclusions about SMRs based on data from actual large pressurised water reactor (PWR) construction experience.
It is expected that SMR build schedules can be greatly reduced because of the smaller physical size of structures, fewer components, and other size-related features. However, the construction work space will be more constrained, which could negatively impact build durations. As a result, simple geometric scaling and reduction arguments cannot necessarily be applied to SMR schedules. This paper defines the key areas in which SMR construction differs from LRs, such as smaller geometries as well as modularised and standardised build processes, and describes how these differences might be expected to impact build duration quantitatively.
The model developed in this paper presents an approach to determining SMR build schedule durations for a range of reactor sizes. It starts with an LR build schedule based on real data from the UK’s only PWR, Sizewell B. The available data are used to establish a reference point for a non-modular, stick-built SMR schedule. This scheduling approach assumes that, for each major element, part of the time spent on fabrication and installation tasks will vary with reactor size while the remaining fraction will remain constant regardless of reactor size (e.g. due to quality assurance and commissioning tasks). The accuracy of the model generated here is assessed against available construction data and models from a range of actual reactor build projects.
The objective of this work is to consider how modularisation can reduce build schedule of SMRs of varying size, by employing modular design and construction principles to both remove tasks that are of long duration from the critical path and to improve construction productivity. Mechanisms by which modularisation reduces build schedule are investigated. Build reduction scenarios are presented based on analysis and subsequent modularisation of the SMR critical path and are compared with other related analyses.
Thermal Hydraulic Analysis of Pressurized Thermal Shock for Loviisa NPP Using Apros Simulation Code
Authors: Žemulis Gintaras , Junninen Pasi , Kytömäki Petri 
Published: October 2018

Loviisa Nuclear Power Plant consists of two VVER-440 type pressurized water reactor units and it is owned by a Finnish energy company Fortum Power and Heat Oy. Only some years after Loviisa Unit 1 start-up, during early 1980’s, some reactor pressure vessel embrittlement by fast neutron irradiation was recognized. Embrittlement of the reactor pressure vessel materials increases the risk of a fast fracture under emergency core cooling conditions. This led to an extensive study of pressurized thermal shock (PTS) scenarios, which included experimental research, large set of probabilistic and deterministic safety analyses, code development and plant modifications to both lower overcooling transient probability and to mitigate it’s consequences. Current operating licenses for Loviisa 1 and 2 Units reactor pressure vessels were applied by Fortum in 2012 to cover operation up to the years 2027 and 2030, respectively. The operating licenses were granted as applied by the Finnish Radiation and Nuclear Safety Authority (STUK) according to application. However a know-how of PTS scenarios shall be maintained. This becomes even more important in a light of any big power plant automation modernization projects targeting also safety systems. Currently Fortum is involved in an automation renewal project also known under its ELSA-project name. The project is scheduled to be ready in 2018.
The project targets at replacing parts of old analogue automation with new digital systems and also extending their diversity by introducing new safety functions for emergency management scenarios. As a result, changes to some emergency operating procedures are inevitable. Due to the changes to emergency operating procedures PTS-scenarios for Loviisa NPP have been also updated.
In 2016 Fortum launched PTS-project in connection with ELSA-project. Previous PTS-scenario related thermal hydraulic analyses were based on the original 80’s decade analyses, which were subsequently revised many times following other modernization projects that took place at Loviisa NPP. In this new round of PTS-analyses the most important cases were recalculated using Apros (Advanced Process Simulator). ELSA-project introduced changes are not only emergency procedures specific but also involve safety system related process and automation logic modifications. The changes affected therefore boundary conditions needed in the analyses. In total probabilistic safety analyses included over 160 different PTS-sequences. The most limiting sequences were selected for deterministic analyses. PTS-analyses were performed in co-operation with Platom Oy and plant owner Fortum Power and Heat Oy. The main results of PTS-analyses are presented in the paper.
Development of the LMFBR Subchannel Analysis Code ATHAS-LMR and Analysis of Blockage Accident
Authors: Du Peng , Shan Jianqiang , Zhang Bo 
Published: October 2018

ATHAS is the subchannel analysis code, independently developed by Nuclear Safety and Operation Laboratory (NUSOL) of Xi'an Jiao Tong University (XJTU), the original version of the ATHAS is only suitable for pressurized water reactor. This paper is aimed at the characteristics of sodium cooled fast reactor core mentioned above, developing the analysis program AYHAS-LMR, which is applicable for LMFBR.
In the design of sodium cold fast reactor, flow blockage must be considered to meet the design criteria. In view of this, a new model is proposed in this paper and added in ATHAS-LMR, considering the flow blockage which changes the flow area and the wetted perimeter. This model is based on the original grid model, we use the grid to simulate the presence of block, increase the resistance of the blockage area artificially. Then, ATHAS-LMR is able to simulate the flow blockage in LMFBR.
In this paper, simulations of FFM5B blockage test were done for the high flow case with ATHAS-LMR, and we compare the simulation results with COMMIX-1. The result indicates that ATHAS-LMR is suitable for the analysis of the flow blockage.
Visualization Study on Droplet-Entrainment in a High-Speed Gas Jet Into a Liquid Pool
Authors: Sugimoto Taro , Saito Shimpei , Kaneko Akiko , Abe Yutaka , Uchibori Akihiro , Ohshima Hiroyuki 
Published: October 2018

A sodium-cooled fast reactor (SFR) is now under development in Japan. A shell-and-tube type once-through heat exchanger is to be installed to generate steam in the design. Low-pressure hot sodium flows in the shell side and high-pressure water, which heated to become steam, flows in the tube side. It has been anticipated that a pin hole is formed on the tube wall and high-pressure steam blows out from the hole. When a high-pressure steam flows out from the tube hole, a high-speed steam jet is formed in the sodium coolant. Fine sodium droplets are torn off from the sodium surface and entrained into the steam jet. Sodium-water chemical reaction causes an increase of entrained droplet temperature. The hot and high-speed sodium entrained droplets attack the wall of a neighboring tube and cause a wastage on the tube wall, which may lead to a failure propagation.
In Japan Atomic Energy Agency (JAEA), an analysis code for the sodium-water reaction phenomenon, called SERAPHIM, has already been developed. Visualization data is required to validate the liquid entrainment model in this code. Since the flow velocity at the gas leakage is a sonic speed, it is extremely difficult to visualize the inside of the gas jet. Experiments have been carried out to visualize this phenomenon in the past; however, experimental data for model validation has not been entirely obtained due to the above-mentioned difficulty. Thus, the motivation of this study is to examine the possibility of visualization method and to clarify flow structure.
To this end, we first performed the preliminary experiments using simple test facilities. Two types of test sections were used in the experiments: three-dimensional one and two-dimensional one. In the experiment using the three-dimensional one, we tried to visualize a more realistic phenomenon. Through this experiment, the whole gas-jet behavior was clearly captured. However, we found that the detailed droplet-entrainment behavior in a gas jet could not be obtained in this setup, especially at high-velocity conditions. Then, we carried out the experiments using the two-dimensional one. In these experiments, the flow structure of a gas jet was simplified. However, it was difficult to distinguish the liquid film formed on the wall surface of the test section from the entrained droplets. We considered that the liquid film is formed due to the nozzle outlet shape and improved the test section. By experiments with new test section, we succeeded in visualizing entrained droplets of relatively large diameter and calculated droplet diameter distribution. Then, we discussed the mechanism of entrained droplet behavior.
Mechanism of Flashing Phenomena Induced by Microwave Heating
Authors: Fujita Shunya , Abe Yutaka , Kaneko Akiko , Yuasa Tomohisa , Segawa Tomoomi , Yamada Yoshikazu , Kato Yoshiyuki , Ishii Katsunori 
Published: October 2018

In a de-nitration conversion processes of the nuclear fuel cycle, mixed oxides (MOX) are produced from the reprocessing solution (plutonium and uranium mixed nitrate solution) of used nuclear fuel. The microwave heating (MH) method has various advantages as one of de-nitration conversion techniques, i.e., this can complete rapidly processes, reduce the waste liquid, and operated remotely and easily by the waveguide. Fine crystal powders are thereby generated, and the manufacture of high-density and high-quality pellets is significant advantage.
The MH method is accompanied with transient boiling phenomena such as overflow and flashing. From the viewpoint of enhancing mass productivity and cost efficiency, in the future, scaling up for the size of de-nitration vessel and shortening processing time are desired. In addition, the safe design of device and the appropriate conditions in microwave irradiation process are required. Hence the extensive understanding of transient boiling phenomena induced by microwave heating is important, and the detailed mechanism of flushing and overflow should be clarified. The aim of this study is to clarify the transient boiling phenomena and detailed mechanism of flashing and overflow. Flashing are affected by physical factors such as solution properties, vessel characteristics, and input energy. Distilled water and different dielectric constant solution as working fluid are used. Dielectric constant was adjusted by the concentration of potassium chloride aqueous solution. A cylindrical vessel kept enough clean by ultrasonic washing machine before experiments was used. The influence of vessel diameter and initial water depth on flashing is examined. As input energy, amount of power supplied and position of the object to be heated in an oven are varied. A high-speed video camera was installed to observe boiling phenomena, and thermography and fiber optic thermometer were to measure the temperature on vessel walls and in a solution, respectively.
As a result, the higher the dielectric constant solution, the less risk of flashing phenomena. In no flashing case, the surface temperature became over 100 °C and the precipitation of potassium chloride was existed. When the high dielectric constant solution was heated by microwave irradiation, the potassium chloride which was precipitated on the inner surface of the vessel was heated to high temperature. On the other hand the temperature of water was lower than that of the deposition of the potassium chloride. As described above, the influence of the high dielectric constant solution on the flashing phenomenon by microwave heating was evaluated. And the mechanism of flashing phenomena was assumed by detailed observation of bubble nucleation. Eventually the mechanism of flashing phenomena was examined.
Investigation on Flow and Breakdown Characteristics Under Horizontal Shear of Water Film Falling Down Vertical Corrugated Plate Dryer
Authors: Wang Bo , Tian Rui-feng , Bowen Chen , Feng Mao 
Published: October 2018

The corrugated plate dryer is a very important equipment for the steam-water separation in the steam generator. Its separation efficiency determines the economic indicators of nuclear power plants. The study of the complicated steam-water separation mechanism of the corrugated plate dryer is very helpful to improve its separation efficiency.
Flow and breakdown characteristics under horizontal shear of water film falling down vertical corrugated plate dryer is investigated. Air and water are used as the two phases in the research. This paper also derives a model calculating the critical airflow velocity of water film breakdown and completes the comparison of the experimental result and the result calculated by the model. The relationship between the water film thickness and critical airflow velocity of water film breakdown is investigated through the experiment. In addition, the water film thickness is measured according to CCD high-speed camera acquisition system and the Planar Laser Induced Fluorescence (PLIF) method is also used for the measurement.
The experimental result reveals that the critical airflow velocity of water film breakdown is related to the corrugated plate structural parameters, the properties of the two phases and the water film thickness. The critical airflow velocity of water film breakdown and the water film thickness are negatively correlated. The result calculated by the model is in good agreement with the experimental result as the water film thickness is large in a certain range.
Development of Hydrogen Treatment System in Severe Accident: Part 4 — Study of Fission Products and Steam Effect on Hydrogen Treatment Characteristics
Authors: Yamada Akira , Yagyu Motoshige , Iwaki Chikako , Sugita Tsukasa , Haruguchi Yoshiko , Tanabe Masashi 
Published: October 2018

A large amount of hydrogen is generated by the metal water-reaction in the Primary containment vessel (PCV) of light water reactors in the severe accident (SA). In the present accident management for boiling water reactor (BWR), vent of mixing gas with filtered vent is regarded as the most likely method that prevents the PCV overpressure. However, it is difficult to vent in early stage of SA because of high radioactive dose. Then we have been developing the hydrogen treatment system to prevent excessive pressure without PCV vent. In focusing on the oxidation-reduction reaction of metal oxides (MOs) with high reaction rate, we have been studying hydrogen treatment system using MOs as effective device under oxygen deficit conditions like PCV of BWR. In the previous studies, we evaluated the hydrogen treatment rate using a couple of MOs, and confirmed that CuO, Co3O4, and MnO2 were effective for the hydrogen oxidation under the oxygen-free condition. We also found that granules of these three MOs could achieve the goal of hydrogen treatment rate with reactor of hydrogen treatment system. We predicted that the performance of MOs decreased with exposure to steam and fission products (FPs) in the PCV during the hydrogen treatment, and investigated their influence. The objective of the present research is to investigate how the steam and FPs, which is supposed to be a reaction-inhibiting-factor, influence hydrogen treatment rate. Then, we conducted hydrogen treatment experiments using a fixed bed reactor with MOs layer. As the results, we confirmed that the hydrogen treatment rate of MnO2 decrease from 70 g/s/m3 to 15 g/s/m3 when partial pressure of vapor went above 0.1 MPa-abs, though, that of CuO didn’t depend on the partial pressure of vapor and sustain the same rate about 40 g/s/m3. We also confirmed that the hydrogen treatment rate was decreased with the consumption of granulated MOs faster than our expectation estimated with unreacted-core model*. We also estimated that CsI selected as typical FPs could not affect the hydrogen treatment rate of CuO. From these results, we have evaluated the reaction rate equation including the steam influence in CuO, which could estimate the hydrogen treatment rate of reactor unit.
*Gas reacts only on the surface of solid and generates shell of products around reactants core. The core shrinks with reaction.
Vaporisation and Condensation in the Feed-Water System in the Turbine Building: How the Phenomenon Arises and How it Can Be Avoided
Authors: Probert Thomas Arne 
Published: October 2018

When restarting Oskarshamn Unit 3 after the re-fuelling outage 2016 a turbine trip occurred at low reactor power. Inspections of the feed-water system in the turbine building revealed a damaged support. A water-hammer caused by an open valve that should have been closed was suspected to be the cause of the damage. Measurement data from the event showed that the pressure in the piping near the damaged support had decreased slowly to be followed by rapid pressure peaks with high amplitudes. The damaged support was reinforced to regain its function and the plant was restarted but an investigation was initiated to analyse the event.
A study of measurement data from the plant showed that this was not a single event. Similar events had occurred earlier. The event turned out to be independent of the valve being open or closed. It was also observed that main pumps that were not in operation at the time in the feed-water system had rotated in the reverse direction during the pressure decrease (normally two out of three pumps are in operation). This event could explain a lot of the support damage in the feed-water system that had occurred since the plant was licensed and that was considered to be caused by operational displacements and vibrations.
A fruitful and intense collaboration started between Engineering, Operations and Maintenance to find the root cause of the event. This resulted in a number of observations that led to a general understanding of the nature of the event. The event was caused by depressurisation in the feed-water tank caused by a turbine trip at low power. The gradual pressure drop caused the water in the piping lower down to evaporate (flash), causing a mass-flow upwards. When the evaporation stopped the flow changed direction and fell downwards causing condensation and large pressure peaks.
The event could be simulated (by RElAP5) with results that qualitatively compared well to measurements. Simulations were then used to show how the event could be mitigated and ultimately avoided.
Hopefully future events will be avoided by changing operation procedures. This has been verified when the plant had to shut-down in November 2017 due to a fuel failure. During the shut-down process no flashing event was detected.
ALFRED Steam Generator Assessment: Design and Pre-Test Analysis of HERO Experiment
Authors: Lorusso Pierdomenico , Pesetti Alessio , Tarantino Mariano 
Published: October 2018

In the framework of the ALFRED design (Advanced Lead Fast Reactor European Demonstrator) for DEMO-LFR, a new concept of steam generator (SG) has been proposed consisting in a double wall bayonet tube bundle which improves the plant safety reducing the possibility of water-lead interaction thanks to a double physical separation between them, and allowing an easier control of eventual leakages from the coolant by pressurizing the separation region with inert gas.
In order to support the development of this innovative SG configuration, the ENEA has designed and realized the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section, a mock-up (1:1 in length) which represents the ALFRED SG. This test section, implemented in the CIRCE pool facility, aims to investigate on the thermal-hydraulic features of the system, providing a database for STH codes validation. The experimental campaign consists of high pressure tests at about 180 bar carried out in the framework of the HORIZON2020 SESAME project (Simulations and Experiments for the Safety Assessment of MEtal cooled reactors).
The secondary loop has been realized for the HERO SG feeding, consisting in an open loop circuit fed by demineralized water. The system is equipped with a volumetric pump and a heater in order to reach the water nominal working conditions of 335°C at the SG inlet and about 180 bar at the outlet.
A preliminary test analysis is carried out by RELAP5-3D© thermal-hydraulic system code. A numerical 1-D model of the HERO SG and the secondary loop has been realized in order to test the loop layout and to characterize the main components from a thermal-hydraulic point of view, defining the start-up procedures for the achievement of the working conditions of the water for the high pressure tests. Furthermore, several simulations are carried out to investigate on the secondary system behavior both for steady states and transients.