bluuebunny/crossref_metadata_2025_split huggingface dataset:
Development of Wireless Tomography System to Measure Solid-Liquid Flow in Rotating Body
Authors: Atagi Yuya , Akimoto Yuya , Nagae Fumiya , Okawa Kazuya , Iso Yoshiyuki , Ikeda Ryosuke , Matsumi Yuuki , Takei Masahiro 
Published: October 2017

A wireless tomography system has been developed to measure the real-time behavior of solid-liquid two phase flow in the centrifuge for controlling a rotational velocity and particle supply rate.
The purpose of this study is development of real-time behavior of solid-liquid two phase flow measurement technique and image conductivity distribution in rotating body because a technique for measuring the behavior of solid-liquid two phase flow in the centrifuges has not existed yet. In this study, a device to measure wirelessly behavior of solid-liquid two phase flow in stationary body has developed as a preliminary to measure in rotating body.
A centrifugation technology for industry process should be improved to obtain more effective separation, shorten process time and save energy. These requirements are achieved by optimizing rotational velocity and particle supply rate. The real-time measurement of behavior of solid-liquid two phase flow for controlling the rotational velocity and the particle supply rate is essential. In other words, the real-time behavior of solid-liquid two phase flow measurement and the rotational velocity control become innovative technologies. Typical behavior of solid-liquid two phase measurement technics with easy handling in particles liquid two-phase flow uses process tomography because the process tomography is suitable for real-time measurement in a rotating body. Process tomography has high temporal resolution. This detector was used with the wireless because electrical cables are not available for centrifuges under high speed rotational condition. This wireless tomography system was used for a lab scale rotating machinery measurement experiment.
Consequently, we could wirelessly measure the behavior of solid-liquid two phase flow in stationary body in real time and get data to image the behavior of solid-liquid two phase flow. This result images the behavior of solid-liquid two phase flow. Observing it enable to control a rotational velocity and particle supply rate.
So this wireless tomography system can separate solid-liquid two phase flow efficiently.
Technology Development on Reactor Dismantling and Investigation of Contamination in FUGEN
Authors: Soejima Goro , Iwai Hiroki , Nakamura Yasuyuki , Hayashi Hirokazu , Kadowaki Haruhiko , Mizui Hiroyuki , Sano Kazuya 
Published: October 2017

Advanced Thermal Reactor (ATR) FUGEN is the heavy water-moderated, boiling light water-cooled, pressure tube-type reactor. The commercial operation of FUGEN started on Mar. 1978 and terminated on Mar. 2003 and the decommissioning of FUGEN has been carried out since the decommissioning plan was approved in 2008.
In order to perform the decommissioning work such as dismantling and decontamination safely and reasonably, technology development for the decommissioning has been promoted actively.
This paper describes a part of technology development as follows.
(1) Technology development on reactor dismantling
The reactor of FUGEN is made of various materials such as stainless steel, carbon steel, zirconium alloy and aluminum which have relatively high activity concentration by operation for 25 years.
With consideration of these characteristics, the reactor will be dismantled under water remotely in order to shield the radiation and prevent dust from migrating from water to air generated by the cutting considering the usage of zirconium alloy which is likely to be oxidized. In addition, laser cutting method whose features are fast cutting speed and less secondary waste in cutting will be applied for reactor dismantling.
However, laser cutting method has no experiences to be applied to dismantlement of reactor facilities. Therefore, laser cutting for actual dismantled objects in air was demonstrated in controlled area in FUGEN using laser cutting system composed of articulated robot and laser cutting head. As a result, safety and applicability of laser cutting system was confirmed. From now on, primary cutting work in air, cutting demonstration with a relatively high dose rate and mock-up test in water for dismantling the actual reactor will be carried out.
(2) Technology development on investigation of contamination
It is necessary to evaluate radioactive inventory in the facilities accurately in order to reflect the evaluated data to dismantling plan appropriately. Therefore, the investigation of the contamination for the facilities has been carried out for safe and reasonable decommissioning work.
The in-situ simple investigation method for the contamination of inner pipes which is mostly dominated by Co-60 is started to develop using the portable NaI(Tl) spectrometer. This method complements conventional investigation method to take samples from the pipes and to analyze them by radiochemical method to figure out the contamination of the whole facility.
Investigation on Gas Dispersion Inside the Large-Scale Containment by CFD Simulation With Code_Saturne for Experimental Scenario Definition
Authors: Song Yuting , Yun Xiangyu , Chen Tian , Zhang Huiyong , Min Jiesheng , Chen Guofei 
Published: October 2017

During accidental situations, large quantities of hydrogen are generated and released due to metal-water reaction. Eventual stratification formed in the top region of containment and locally high concentration threatens the integrity of nuclear power plants. The stratification of hydrogen may be eroded by several measures, such as air jet, spray, ventilation. So far, several activities have been carried out on the hydrogen stratification’s break-up study. The well-known OECD/SETH-2 project used PANDA and MISTRA test facilities to study this phenomenon with variable test conditions. These test cases are also wildly used to validate the computational fluid dynamics (CFD) codes.
The large-scale containment test facility – COntainment Thermal-hydraulics and Hydrogen Distribution (COTHYD) was designed and constructed at CGN to study the containment thermal hydraulics behavior and hydrogen risk in PWR during severe accidents. A series of tests including helium and air stratification, helium and air stratification eroded by air jet, as well as steam dispersion and condensation are planned to be performed on this test facility. The objective is to set up high quality database for code validation and physical phenomena research.
During the test cases preparation, the relevant tests carried out since 2000 in MISTRA, TOSQAN, THAI, PANDA were collected to give the reference for the cases design in COTHYD. For the tests of stratification’s break-up on COTHYD, helium is discharged from top and it accumulates at the top region of containment. On the preparation stage, Code_Saturne was used to define the test scenario and predict helium distribution. Both dead volume and open volume (with ventilation) are modeled to investigate helium stratification formation and helium-rich layer concentration evolution. These results will be used as test matrix configurations. Code_Saturne, EDF in-house open-source software, has been used in the simulation of hydrogen dispersion. In 2015, B. Hou presented their work on the simulation of the break-up phenomenon of helium stratification by air (ST1_7 and ST1_10 test cases of OECD/SETH-2 project) (Hou et al., 2015) [1]. Simulation results demonstrated that Code_Saturne can well predict and simulate this phenomenon. This software is, as a consequence, used as the case design and validation tools in the pre/post experiment steps.
Experimental Study on Critical Heat Flux of In-Vessel Retention for Large Nuclear Reactor
Authors: Sun Canhui , Chen Wei , Geng Jun 
Published: October 2017

During a rare accident involving severe core damage in a nuclear power plant, if the molten core material can be contained within the boundary of the reactor vessel, the severity of the accident is expected to be greatly reduced. Therefore, the severe accident management strategy based on in-vessel retention (IVR) of molten core debris is highly desirable, and has been adopted by advanced reactor designs such as AP1000, APR1400 and CAP1400.
Previous studies assumed a fully developed molten pool, where molten metal and molten oxides separated, and liquid metal form a layer on the top of the oxidic pool. The limiting failure mechanism of the lower head is related to occurrence of boiling crisis on the outer surface of the vessel. The boiling crisis would occur, when the local heat flux through the vessel wall exceeds the critical heat flux value. The evaluation of thermal regime is then done by comparing the heat flux distribution from the molten corium pool and the CHF variation on the outer wall in order to define the available margins to CHF. When a larger power nuclear reactor is designed, the margin of the IVR could be decreased. Therefore, the study on critical heat flux enhancement for lager nuclear reactor is necessary.
In this study, an experimental facility is conducted to analysis the critical heat flux mechanism and enhancement for lager power nuclear reactor. The facility includes two loops, one is primary loop, and the other is auxiliary loop which could import injection flow in the primary loop. The test section is an inclined downward heated rectangular channel, and the angle is variable. A series of critical heat flux experiments are conducted in this facility. The results are compared with previous experiment results and calculation correlation in the large nuclear reactor operating conditions.
The European Utility Requirements for Advanced Light Water Reactors (EUR): Recent Achievements and New Challenges
Authors: Jacquart Guillaume , Kymäläinen Olli , Vivier Jean-François , Ferraro Giovanni , Vieilletoile Emmanuel 
Published: October 2017

Since 1991, the European Utility Requirements (EUR) organisation has been actively developing and promoting harmonised technical specifications for the new mid- and large-size LWR designs to be proposed by the vendors in Europe.
The EUR document consists of a comprehensive set of requirements covering the whole Nuclear Power Plant (NPP). It encompasses all aspects (safety, performance, competitiveness) and all parts of a NPP (nuclear island and conventional island). The document can be used by the utilities (guide for design assessment, technical reference for call for bids) and by the vendors, as a technical guide. The harmonisation which is sought after by the EUR aims at delivering the safest and most competitive designs based on common rules shared all over Europe. Fourteen nuclear operators across Europe are members of the organisation.
After the publication of the Revision D of the EUR document (October 2012), the EUR organisation has been extremely active. This presentation describes the main results obtained during the last four years and the new challenges for the coming years (roadmap 2016–2018) in the three following fields.
First, the revision of the EUR document, in order to maintain it at a state-of-the-art level, remains the highest priority for the organisation. Regarding the new revision E of the EUR document [1], expected to be issued in the 2nd quarter of 2017, the presentation describes the most significant updates implemented in many fields among which: revised safety requirements taking into account the most recent European and international safety standards issued by WENRA and IAEA, the lessons learnt from the Fukushima accident, including re-evaluated seismic and external natural hazards approach and the most recent international standards, for example for Instrumentation and Control (I&C).
The assessment of new designs is the second main technical activity of the EUR organisation. The MHI EU-APWR design has been assessed against the revision D between 2012 and 2014. New design assessments are in progress (namely Korean KHNP’s EU-APR and Russian AEP’s VVER TOI) and are planned to be ended in 2017–2018. A new applicant is in the file. The presentation briefly recalls the EUR design assessment objectives and process and the progress of the different assessment projects.
The third topic to be covered by the presentation is the interaction between the EUR and the other stakeholders, in particular the other international organisations (ENISS, WNA/CORDEL, WENRA, IAEA, EPRI/URD). The presentation describes how the EUR organisation is connected to these stakeholders and the corresponding cooperation results and future projects.
Flow Field Analysis of the Reactor Coolant Pump Internal Cooling Circulation
Authors: Qin Jie , Xu Qingmu , Yuan Junkai , Cai Kun 
Published: October 2017

Reactor coolant pump (RCP) is one of the most critical devices in third generation of pressurized water reactor nuclear power plant. EMD shield pump and KSB wet winding pump are two representative kinds of RCPs without complex shaft seal system. Due to cancellation of shaft seal system, the entire rotors (including the flywheel) are immersed in the coolant. The losses in RCPs take one third of the total power including rotation loss caused by rotor in the water, electromagnetic loss in the shielding sleeve,the heat transferred through high temperature coolant, and heat generated by bearing.Because of the losses listed above, bearing and winding are heated,and the losses make temperature rise. in order to ensure that the motor is working properly at low temperatures, the company EMD and KSB design the RCP internal cooling circulation which brings the heat out to ensure the normal operation of the RCPs.
The RCP internal cooling circulation includes inlet flow area, auxiliary impeller, thrust bearing, the lower flywheel, motor can, upper radial bearing, upper flywheel, outlet flow area, and external heat exchanger,etc. Flow characteristics in every flow path determine the flow distribution and heat transfer, and the flow distribution determines whether the cooling performance of RCP internal cooling circulation meets the requirements. In order to control operating temperature of motor and bearing, and to optimize heat transfer, adjusting the size of flow area and changing the flow characteristics arecritical. flow field and temperature field in RCP internal cooling circulation need overall analysis.
Flow distribution can be obtained theoretically through the calculation of an overall three-dimensional model.But on the one hand, the calculation time is long due to a complex three-dimensional model with a large quantity of grids, on the other hand, it is easier to casue errors in local processing and the errors are difficult to find or correct. For rapid analysis and optimization of flow and heat transfer in RCP internal cooling circulation, ensure the motor winding and bearing operate at an appropriate temperature, the local characteristics of RCP internal cooling circulation are studied, one-dimensionalanalysis method of RCP internal cooling circulation is developed. This one-dimensional analysis method can be used to predict the flow distribution of each part of RCP internal cooling circulation according to change of the channel geometry parameters, key dimensions, boundary conditions and rotor speed. The geometric parameters are optimized by analyzing the flow distribution, and the purpose of design guidance are achieved.
Leakage Flow Investigations on a Through-Wall Crack Related to Thermal Fatigue of Nuclear Power Plant Piping
Authors: Schmid Stefan , Kulenovic Rudi , Laurien Eckart 
Published: October 2017

For the validation of empirical models to calculate leakage flow rates in through-wall cracks of piping, reliable experimental data are essential. In this context, the Leakage Flow (LF) test rig was built up at the IKE for measurements of leakage flow rates with reduced pressure (maximum 1 MPA) and temperature (maximum 170 °C) compared to real plant conditions. The design of the test rig enables experimental investigations of through-wall cracks with different geometries and orientations by means of circular blank sheets with integrated cracks which are installed in the tubular test section of the test rig.
In the paper, the experimental LF set-up and used measurement techniques are explained in detail. Furthermore, first leakage flow measurement results for one through-wall crack geometry and different imposed fluid pressures at ambient temperature conditions are presented and discussed. As an additional aspect the experimental data are used for the determination of the flow resistance of the investigated leak channel. Finally, the experimental results are compared with numerical results of WinLeck calculations to prove specifically in WinLeck implemented numerical models.
Neutronic Analyses for CFETR With Modular Helium Cooled Lithium Ceramic Blanket
Authors: Xu Kun , Ye Minyou , Song Yuntao , Lei Mingzhun , Mao Shifeng 
Published: October 2017

China Fusion Engineering Test Reactor (CFETR) is a superconducting tokamak proposed by national integration design group for magnetic confinement fusion reactor of China to bridge the R&D gaps between ITER and DEMO. Since the launch of CFETR conceptual design, a modular helium cooled lithium ceramic blanket concept had been under development by the blanket integration design team of the Institute of Plasma Physics of the Chinese Academy of Sciences, to complete CFETR in demonstrating its fusion energy production ability, tritium self-sufficiency and the remote maintenance strategy. To validate the feasibility, the neutronic analyses for CFETR with this modular helium cooled lithium ceramic blanket were performed. The 1-D neutronic study for CFETR was done in the first place to give a preliminary and quick demonstration of the overall neutronic performance. Meanwhile, the neutronic analyses for a single standard helium cooled lithium ceramic blanket module were done in several times to give more insight for the material and geometry parameters of intra-module structures. Therefore, the principles for neutronic design and the module level optimized parameters were produced, based on which the design of practical blanket modules planted in tokamak vacuum vessel was completed. In the end, the 3-D neutronic analysis for CFETR was done utilizing the MCNP code, in which the 11.25 degree sector model (consist of blanket modules, manifold, support plate, shield, divertor, vacuum vessel, thermal shield and TF coils) was generated with the McCad automated conversion tool from the reference CAD model for analysis, the bi-dimensional (radial and poloidal) neutron source map was plugged via general source definition card to stimulate the D-T fusion neutrons. The concerned neutronics parameters of CFETR, mainly including the tritium breeding ratio to characterize tritium self-sufficiency, the energy multiplication factor to characterize power generation, as well as, the inboard mid-plane radial profiles of neutron flux densities, helium production rate, displacement damage rate and the energy deposition to characterize the shielding performance, were produced.
In principle, the neutronics performance of CFETR with modular helium cooled lithium ceramic blanket is promising. The tritium breeding capability meets the design target and, by referring to that for ITER and the EU DEMO fusion power plant, the inboard mid-plane shielding is effective to fulfill the radiation design requirement of the superconducting TF-coils, resulting in a compulsory warm-up time interval of ∼2 FPY for TF-coils. The nuclear heating loads to other CFETR components were generated. As an outcome of this work, the applicability of McCad on CFETR neutronic modeling is demonstrated.
Long-Term Cooling Strategy for the Primary Containment Vessel of the Kashiwazaki-Kariwa Nuclear Power Station in a Severe Accident
Authors: Kawaguchi Satoshi , Mizuno Satoshi , Oyama Yoshihiro 
Published: October 2017

This paper explains the strategy of our company (Tokyo Electric Power, TEPCO) regarding means of long-term heat removal from the primary containment vessel (PCV) of Units 6 and 7 (ABWR) of the Kashiwazaki-Kariwa Nuclear Power Station in a severe accident.
If the PCV continues in a high-temperature state for a long time, the strength of the PCV concrete will decline, and the risk of being affected by an earthquake will increase. Therefore, it is crucial for safety to cool the PCV and reduce its temperature to the maximum working temperature or lower.
TEPCO provides a means of cooling the reactor pressure vessel (RPV) and PCV called the alternative coolant circulation system (ACCS). This system uses the heat exchanger of the residual heat removal (RHR) system, the make up water condensate (MUWC) pump, and alternative heat exchanger vehicles. By using these measures, it is possible reduce temperature in the PCV over the long term to the maximum working temperature (design value) or less, even in severe accident scenarios such as a large LOCA + ECCS function failures + SBO (station blackout).
This function has quite high reliability, but in a scenario where these measures cannot be used, expectations are placed on the filtered vent (FV). However, due to FV characteristics, it is impossible to reduce to below the saturation temperature of 100°C at atmospheric pressure using FV alone, and it will be necessary in the medium/long-term to cool the PCV while also restoring the cooling equipment.
Therefore, the following restoration operation of PCV cooling and its dose evaluation were studied.
(1) RPV heat removal by restoring the RHR system
(2) RPV and PCV heat removal using a portable pump employing a portable heat exchanger
(3) RPV and PCV heat removal using the suppression pool water clean up system (SPCU) employing portable heat exchangers
(4) RPV heat removal using the clean up water system (CUW)
By clarifying beforehand issues such as feasibility of these systems, the on-site environment for restoration measures, and the necessary gear/systems, the authors were able to secure means of long-term cooling of the PCV, and further enhance PCV reliability.