bluuebunny/crossref_metadata_2025_split huggingface dataset:
Evaluation of Heat Removal During the Failure of the Core Cooling for New Critical Assembly
Authors: Eguchi Yuta , Sugawara Takanori , Nishihara Kenji , Tazawa Yujiro , Tsujimoto Kazufumi 
Published: October 2018

The Japan Atomic Energy Agency (JAEA) has been conducting the research and development (R&D) on accelerator-driven subcritical system (ADS) as a dedicated system for the transmutation of long-lived radioactive nuclides. To foster the R&D of ADS, the Transmutation Physics Experimental Facility (TEF-P) in the J-PARC project has been planned to build by JAEA [1]. The TEF-P is used minor actinide (MA) fuel which has large decay heat, so during the failure of the core cooling system, the evaluation of the core temperature increase is important.
This study aims to evaluate the natural cooling characteristics of TEF-P core and to achieve a design that does not damage the core and the fuels during an accident (the failure of the core cooling system).
The experiments using mockup device was performed to validate the heat transfer characteristics in the empty rectangular lattice tube. It was obtained that the actual heat transfer coefficient of empty rectangular lattice tube was about 2.2 times larger than the theoretical free convection model. It was also confirmed that the insertion of any block into the empty rectangular lattice tube could achieve the higher heat transfer coefficient.
Using the heat transfer coefficient obtained by experiment results, thermal analysis was performed by the three-dimensional heat transfer analysis. As a result, the calculation results showed that the maximum core temperature will be 294 °C which is less than the design criterion of temperature, 327 °C. It was presented that the design condition which the core temperature will be below the design criterion during the failure of the core cooling system through this study.
Analysis of the Condensation Model Used in Severe Accident Integral Codes
Authors: Zhu Yiming , Liu Zhuo , Yang Xiaoming 
Published: October 2018

Steam condensation plays a key role in removing heat from the containment atmosphere, therefore the calculation of the steam condensation has a significant effect on the judgment of severe accident process. This paper discussed the calculation models of condensation used in severe accident integral codes.
Firstly two kinds of condensation calculation models introduced and analyzed for application in integral codes: models based on experiment correlations — Uchida model, Dehbi model, and the models based on analogy of heat-mass transfer — Kreith model and Collier model. And then two calculation models based on Kreith model were proposed: simplified calculation model and completely iterative calculation model.
After that, a heat structure calculation code was designed and the two model was applied. The results of the code with the two model along with Uchida and Dehbi models were compared with the experiment result of Wisconsin. It was found that the simplified model had a satisfied calculation speed and the relative errors were under 13% in total situations, which was acceptable for the integral code. Besides, it was shown that the completely iterative calculation model had smaller relative errors, which was under 11%. In addition, this paper also investigated the effects of some parameters on the heat transfer coefficient for these different models, the velocity of the bulk of gas, length of surface, pressure of the gas and mole fraction of air were selected and studied separately.
Free Release of Radioactive Waste Containing Very Low Level Waste and Short Lived Radionuclides at Nucleco
Authors: Andreozzi C. , Bianchilli B. , Dodaro A. , Gagliardi F. , Mauro E. , Sisti M. 
Published: October 2018

Since August 2015 a new classification of radioactive waste was issued by Italian Ministry of Economic Development, in order to adapt Italian historical classification to European standards. This new classification provides 6 categories, from exempt to high level waste, and it is based on the waste final destination: from free release to final disposal or interim storage (high level waste and intermediate level waste with α-content higher than 400 Bq/g) [1].
Nucleco is a State owned Company acting as Waste Management Organization for radioactive waste coming from hospitals, industries and research and development activities not related to electricity production by nuclear plants. Nucleco collects, safely manages and temporarily stores waste that will be sent to the National Repository (site definition phase is still ongoing), while for the Short Lived Radionuclides and Very Low Activity waste Nucleco performs all necessary operations to be compliant with the conditions of release prescribed by the Italian Control Authority.
Short Lived Radionuclides are those whose half-life is shorter than 100 days or reach the condition of non-radiological relevance in 5 years: they are mainly produced by bio-medical applications of radioactive materials. Very Low Activity waste are characterized by activity concentrations lower than 100 Bq/g (of which less than 10 Bq/g of α-emitting radionuclides) and reach the condition of non-radiological relevance in 10 years: these waste usually came from research institutions and industrial activities.
This work presents the authorized operating procedures, the radiological measurements criteria and the technical know-how put in place by Nucleco to fulfil the provisions of Italian regulations for unconditioned release of radioactive waste.
A case study of ISO 20’ containers is discussed in the current paper.
Main emphasis will be addressed to:
• gathering of historical information about the state of the material to be released and definition of the reference radiological spectrum;
• sampling procedures to ensure representativeness of the samples from homogeneous waste batch to be released and then subjected to radiological characterization;
• characterization phase consisting of the integration of several state-of-art techniques aiming to collect the most complete set of radiological data;
• data processing protocols needed for the calculation of the activity concentrations for each radionuclide of the reference spectrum (or other radionuclides eventually detected);
• evaluation of the main sources of uncertainty affecting the results;
• comparison of the activity concentration (including the uncertainty) of each radionuclide with the corresponding authorized concentration limits.
Study on Thermal-Hydraulics Characteristics of the Flat and High-Thermal-Conductivity Core-Catcher
Authors: Takeyama Daiki , Iwaki Chikako , Onitsuka Yoichi , Tahara Mika 
Published: October 2018

For mitigation of severe accident with core melting, core-catcher has been developed to catch and cool molten core. The core-catcher developed by Toshiba for Advanced Boiling Water Reactor was designed to be installed under Reactor Pressure Vessel (RPV) and catch molten core in basin with thermal-resistant material. It also has structure including risers and downcomers to generate natural circulation flow of cooling water. On the other hand, there is not enough space to install it in the existing plants due to the height of the inclined cooling channels.
Then, we have been developing flat core-catcher with flat cooling channels for the existing plants to reduce total height of the structure. Finned channels will be adopted to increase heat transfer rate by increasing heat transfer area. However, the thermal-hydraulics characteristics of such core-catcher has not been clarified due to the specific configuration, that is, the horizontal rectangular finned channel with the heated surface from upper side.
This present study investigated the natural circulation characteristics and heat transfer behavior in the horizontal rectangular finned channel by experiment. Pressure drop, natural circulation flow and temperature were measured by changing heat fluxes. The flow was visualized to obtain flow pattern in the finned channel by a high-speed camera. The maximum value of test range of heat flux was 250 kW/m2, which is the value when the total amount of the molten core would be dropped to the core-catcher. The partial simulated test section of finned channel with 3.5 m length was heated from upper side by heaters to simulate the heat flux from the molten core. This length is the same as the inner diameter of RPV pedestal of the existing plants.
The natural circulation mass flux increased as the heat flux increased and then two-phase flow pressure loss also increased. Consequently, circulation flow turned to decrease. As a result of the test, when the heat flux was 50 kW/m2, the circulation mass flux got to the maximum value, 230 kg/m2s. Under all the conditions except for the maximum heat flux of 250 kW/m2, the fin surface temperature was around the saturated temperature. At maximum heat flux 250 kW/m2, the temperature got 540 K. However, the structural soundness was maintained because it is lower than melting temperature of the fin. It can be concluded that the flat and high-thermal-conductivity core-catcher has enough cooling performance to catch and stabilize the molten core.
FMEDA and FIT-Based Safety Assessment of NPP I&C Systems Considering Expert Uncertainty
Authors: Yasko Alexander , Babeshko Eugene , Kharchenko Vyacheslav 
Published: October 2018

The complexity of modern safety critical systems is becoming higher with technology level growth. Nowadays the most important and vital systems of automotive, aerospace, nuclear industries count millions of lines of software code and tens of thousands of hardware components and sensors. All of these constituents operate in integrated environment interacting with each other — this leads to enormous calculation task when testing and safety assessment are performed.
There are several formal methods that are used to assess reliability and safety of NPP I&C (Nuclear Power Plant Instrumentation and Control) systems. Most of them require significant involvement of experts and confidence in their experience which vastly affects trustworthiness of assessment results.
The goal of our research is to improve the quality of safety and reliability assessment as result of experts involvement mitigation by process automation. We propose usage of automated FMEDA (Failure Modes, Effects and Diagnostic Analysis) and FIT (Fault Insertion Testing) combination extended whith multiple faults approach as well as special methods for quantitative assessment of experts involvement level and their decisions uncertainty. These methods allow to perform safety and reliability assessment without specifying the degree of confidence in experts.
Traditional FMEDA approach has several bottlenecks like the need of manual processing of huge number of technical documents (system specification, datasheets etc.), manual assignment of failure modes and effects based on personal experience. Human factor is another source of uncertainty. Such things like tiredness, emotional disorders, distraction or lack of experience could be the reasons of under- and over-estimation.
Basing on our research in field of expert-related errors we propose expert involvement degree (EID) metric that indicates the level of technique automation and expert uncertainty degree (EUD) metric which is complex measure of experts decisions uncertainty within assessment. We propose usage of total expert trustworthiness degree (ETD) indicator as function of EID and EUD.
Expert uncertainty assessment and Multi-FIT as FMEDA verification are implemented in AXMEA (Automated X-Modes and Effects Analysis) software tool. Proposed Multi-FIT technique in combination with FMEDA was used during internal activities of SIL3 certification of FPGA-based (Field Programmable Gate Array) RadICS platform for NPP I&C systems. The proposed expert trustworthiness degree calculation is going to be used during production activities of RPC Radiy (Research and Production Corporation).
Our future work is related to research in expert uncertainty field and extension of AXMEA tool with new failure data sources as well as software optimization and further automation.
Study on the Turbulent Mixed Convection Phenomena Inside the Air-Cooled RCCS Riser
Authors: Shin Dong-Ho , Kim Sin-Yeob , Kim Chan Soo , Park Goon-Cherl , Cho Hyoung Kyu 
Published: October 2018

High Temperature Gas cooled Reactor (HTGR) that is one of GEN-IV reactor types enhances its safety adopting a passive cooling system, Reactor Cavity Cooling System (RCCS). When the active cooling system for the reactor core in a HTGR doesn’t work, the decay heat from the reactor core is transferred to the reactor vessel and the concrete wall of the reactor cavity, which are cooled by the RCCS. The RCCS consists of the vertical rectangular ducts, called risers, surrounding reactor vessel at a certain distance and chimneys that are connected to risers. Risers receiving the decay heat from the reactor vessel, the air inside the riser is heated and flows up to the chimney which accelerates exhalation of the air to the external atmosphere. The RCCS performance depends on the heat transfer rate inside the riser ducts, but the turbulent mixed convection that may occur in the riser ducts can complicate the verification. In this study, an experimental facility had been constructed to investigate heat transfer phenomena inside a riser duct and experiments with various heat flux and flow rate conditions were carried out. The experimental results showed that the turbulent mixed convection occurred for certain experimental conditions in the riser duct, which leads to heat transfer deterioration. Therefore, a correlation was suggested to predict how much the heat transfer could decrease compared to the forced convection.
Large eddy simulation was carried out for CFD analysis. The heat transfer coefficients from LES showed consistent results with the correlation. The unique velocity profile and secondary flow was observed when the heat transfer was deteriorated. This study will contribute to the evaluation of the RCCS performance.
Study on Current Status and Future Developments in Nuclear-Power Industry of the World
Authors: Pioro R. , Zvorykin A. , Machrafi R. , Pioro I. 
Published: October 2018

Currently, nuclear power plays a quite visible role in the world electricity generation (∼11%). However, before the Fukushima Nuclear Power Plant (NPP) severe accident in March of 2011, NPPs generated about 14% of the world’s electricity. Accounting that after, mainly, Chernobyl NPP severe accident a number of power reactors built and put into operation in the world decreased from 120 within 1985–1990 to about 22 per 5 years (within 1995–2015), we might face a significant shortage of operating power reactors within 2030–2040.
Therefore, it is important to evaluate current status of nuclear-power industry and to make projections on near (5–10 years) and far away (10–25 years and beyond) future trends in nuclear-power industry. In the current paper statistics on all current nuclear-power reactors were analyzed and based on that future trends were estimated in terms of types of reactors to be left after 10 years, new types of reactors to be put into operation, projections of how many reactors and of which types will be build.
To make any projections an average operating term of power reactors should be estimated. In the current paper a nuclear-power-reactor operating term of 45 years was considered. Also, rates of building and putting into operation power reactors worldwide were estimated, and several scenarios of future developments in nuclear-power industry in the world and in selected countries were considered.
Development of Numerical Simulation Method to Evaluate Molten Material Behaviors in Nuclear Reactors: Estimation of Fuel Debris Distribution in the Pedestal
Authors: Yamashita Susumu , Yoshida Hiroyuki 
Published: October 2018

In the accident at the Fukushima Daiichi Nuclear Power Plant, fuel rods and reactor internals were overheated due to the radioactive decay heat and the oxidization of fuel cladding, and relocated to the lower region of the pedestal. To retrieve the fuel debris, location, distribution and characteristics of the fuel debris are quite important. However, this information still has not been clarified. To estimate the distribution of fuel debris, since experiments cannot provide sufficient information, numerical simulation will be a useful tool. There are some numerical codes for calculating conditions inside a reactor at severe accidents, however, those codes include assumptions and/or related to melt relocation, and cannot simulate relocation process directly.
In order to simulate the relocation and corium spreading behavior in PCVs without assumptions and scenario, a numerical simulation code that can phenomenologically evaluate the melting phenomena is required. Then, a phenomenologically-based numerical simulation code for predicting the melting core behavior, including solidification and relocation based on the computational fluid dynamics, JUPITER, has been developed in JAEA.
In this paper, by using JUPITER, we simulated the corium spreading process inside a pedestal. As a result, we confirmed that JUPITER has potential to estimate the debris distribution inside the pedestal region.
Study on Variation of HPGE Detector Dead Layer Thickness and its Effect on the Measurements of the Detector Response and Samples Characterization Using Monte Carlo Simulation
Authors: Abd El Gawad K. , Song Yushou 
Published: October 2018

The main objective of this work is to produce an optimal modeling for our aged Planar-HPGe detector using Monte Carlo method (MC). That optimization included the analysis of the germanium dead (inactive) layer thickness for our old detection system (planar-HPGe detector). DL is one of the important parameters needed in order to obtain the smallest discrepancy between simulated and experimental measurements of detector efficiency. Also, precise determination of 235U enrichment for UO2 samples which is necessary for purposes of nuclear materials verification in the field of nuclear safeguards.
The thickness of Germanium dead layer (DL) can be vary by time as it is not well known due to the existence of a transition zone where photons are strongly attenuated and absorbed, that cannot contribute to the total photon energy absorption which causes a significant decrease in efficiency. Therefore, using data provided by manufacturers since long years (manufacture date) in the detector simulation model is not convenient. As a result, some strong discrepancies appear between measured and simulated efficiency, in addition to that non-accurate results for 235U enrichment determination. The Monte Carlo method applied to overcome this difficulty was to vary the thickness of dead layer step by step in simulation, a good agreement (minimum deviation) between estimated and experimental efficiency was reached when a suitable germanium dead layer thickness was chosen. Calculations and measurements were performed for radioactive nuclear material samples in the form of UO2 powder with different sizes and enrichments at different locations, under different gamma-lines emitted after a-decay of the 235U nuclei. Results indicated that a good agreement between simulated and measured efficiencies is obtained using a value for the germanium dead layer thickness approximately (2.45 mm) six in comparison with (0.389 mm) provided by the detector manufacturer.
Muon Tomography for Measuring Amount of Nuclear Materials in Fuel Debris
Authors: Sugita Tsukasa , Miyadera Haruo , Yoshioka Kenichi , Kume Naoto 
Published: October 2018

A method to measure an amount of nuclear materials in fuel debris by using muon tomography has being developed for proceeding with decommissioning of Fukushima Daiichi nuclear power plant.
As a result of the Fukushima Daiichi nuclear disaster, the molten fuels were mixed with reactor structures and accumulated as fuel debris in the reactor buildings. There is still a large amount of fuel debris remained in each reactor. Fuel debris removal is planned in the near future and the debris will be taken out in this process. The debris need to be inspected from a viewpoint of nuclear material control. Since the debris is a mixture of fuel and other structures, it is hard to quantitate nuclear materials in debris by existing measurement method.
Muons are cosmic-ray particles which have high energies, therefore, they are highly penetrative. This feature makes muon tomography sensitive to find heavy materials such as uranium or plutonium. We conducted a simulation study of applying muon tomography to measure fuel debris by using a Monte-Carlo method.
A simulation model which includes muon detectors, shielding container and fuel debris was constructed to reproduce a measurement situation at the site. In conclusion, muon tomography quantitate the nuclear materials, therefore, this method should be useful for the fuel debris removal of Fukushima Daiichi reactors.
Demonstrative HEAF (High Energy Arcing Fault) Fire Tests of High and Low Voltage Switchgears of Nuclear Power Plants
Authors: Shirai Koji , Miyagi Tsukasa , Iwata Mikimasa , Tasaka Koji , Ji Junghoon 
Published: October 2018

High Energy Arcing Faults (HEAF) have the potential to cause extensive damage to the failed electrical components and distribution systems along with adjacent equipment and cables within the zone of influence (ZOI). Furthermore, the significant energy released during HEAF event can act as an ignition source to other components within the area of the HEAF. In Japan, during the Great East Japan Earthquake occurred in 2011, the seismic induced HEAF fire event, which induced the whole damage of the multiple high voltage switchgears, was observed in Onagawa Nuclear Power Plant (NPP).
In response, in August 2017, the NRA (Nuclear Regular Authority) in Japan amended the safety requirement for the power supply to consider the influence of the successive fire due to the HEAF event (hereinafter HEAF fire event). Therefore, it is urgently necessary to establish the design criteria to prevent the HEAF fire event, and enhance the experiment data of the HEAF fire event. In order to estimate the total arc energy during the HEAF event and obtain the threshold value to prevent the HEAF fire for the existed non-arc proof electrical cabinets, several series of three-phase internal arc tests with high (6.9kV class) and low (480V class) voltage electrical cabinets were executed. We executed internal arc tests with full scale high/low voltage metal-enclosed switchgear components (non-arc proof type, copper bus conductor), and evaluated arc energy, the mechanical damage of the cabinet and the surrounding equipment due to the impulsive pressure and the possibility of successive fire occurrence. In case of high voltage switchgear, when the arcing energy exceeded 25.3MJ, successive fire was identified. Especially, in the case where the arc flash was discharged in the circuit breaker room, a 2-second arcing duration in a three-phase short-circuit current with 18.9kA (measured arcing energy over 40MJ) caused successive fire which required extinguishment. On the other hand, in case of low voltage power center, when the arcing energy exceeded 19MJ, successive fire was identified. According to these demonstrative tests, this paper presents the evaluation method to estimate total arc discharge energy during the HEAF event for high and low voltage electrical cabinets.