bluuebunny/crossref_metadata_2025_split huggingface dataset:
Development, Distribution and Maintenance of Application Security Controls for Nuclear
Authors: Waedt Karl , Ding Yongjian , Ciriello Antonio , Lou Xinxin 
Published: October 2017

The generic concept of Security Controls, as initially deployed in the information security domain, is gradually used in other business domains, including industrial security for critical infrastructure and cybersecurity of nuclear safety I&C. A Security Control, or less formally, a security countermeasure can be any organizational, technical or administrative measure that helps in reducing the risk imposed by a cybersecurity threat. The new IAEA NST036 lists more than 200 such countermeasures. NIST SP800-53 Rev. 4 contains about 450 pages of security countermeasure descriptions, which are graded according to three levels of stringency. In order to facilitate and formalize the process of developing, precisely describing, distributing and maintaining more complex security controls, the Application Security Controls (ASC) concept is introduced by the new ISO/IEC 27034 multipart standard. An ASC is an extensible semi-formal representation of a security control (e.g. XML or JSON-based), which contains a set of mandatory and optional parts as well as possible links to other ASCs. A set of Application Security Controls may be developed by one company and shipped together with a product of another company.
ISO/IEC 27034-6 assumes that ASCs are developed by an organization or team specialized in security and that the ASCs are forwarded to customers for direct use or for integration into their own products or services.
The distribution of ASCs is supported and formalized by the Organization Normative Frameworks (ONF) and Application Normative Frameworks (ANF) deployed in the respective organizational units.
The maintenance and continuous improvement of ASCs is facilitated by the ONF Process and ANF Process.
This paper will explore the applicability of these industry standards based ASC lifecycle concepts for the nuclear domain in line with IEC 62645, IEC 62859 and the up-coming IEC 63096. It will include results from an ongoing bachelor thesis and master thesis, mentored by two of the authors, as well as nuclear specific deployment scenarios currently being evaluated by a team of cybersecurity PhD candidates.
Setup of the Supercritical CO2 Test-Facility “Scarlett” for Basic Experimental Investigations of a Compact Heat Exchanger for an Innovative Decay Heat Removal System
Authors: Flaig Wolfgang , Mertz Rainer , Starflinger Jörg 
Published: October 2017

Supercritical fluids show great potential as future coolants for nuclear reactors, thermal power and solar power plants. Compared to the subcritical condition, supercritical fluids show advantages in heat transfer due to thermodynamic properties near the critical point. This can lead to the development of more compact and more efficient components, e.g. heat exchanger and compressors.
A specific field of interest is a new decay heat removal system for nuclear power plants which is based on a turbine-compressor-system with supercritical CO2 as the working fluid. In case of a station blackout this system converts the decay heat into excess electricity and low-temperature waste heat, which can be emitted to the ambient air. This scenario has already been investigated by means of the thermo-hydraulic code ATHLET, numerically demonstrating the operation of this system for more than 72 h. The practical demonstration is carried out within the Project “sCO2-HeRo”, funded by the European Commission, in which a small scale demonstration unit of the turbo compressor shall be installed at the PWR glass model at GfS, Essen, Germany. To guarantee the retrofitting of this decay heat removal system into existing nuclear power plants, the heat exchanger needs to be as compact and efficient as possible. Therefore, a diffusion welded plate heat exchanger (DWHE) was developed and manufactured at IKE. It has been designed with rectangular mini-channels (0.5–3 mm hydraulic diameter) to ensure high compactness and high heat transfer coefficients. Due to uncertainties the DWHE has to be tested in regard to the actual possible transferrable heat power and to the pressure loss. According to this demand a multipurpose facility has been built at IKE for various experimental investigations on supercritical CO2, which is in operation now. It consists of a closed loop where the CO2 is compressed to supercritical state and delivered to the test section. The test section itself can be exchanged by other ones for various investigations. After the test section, the CO2 pressure is reduced and the liquid is stored in storage tanks, from where it is evaporated and compressed again. The test facility is designed to carry out experimental investigations with CO2 mass flows up to 0.111 kg/s, pressures up to 12 MPa and temperatures up to 150 °C. The first subject of interest will be the study of the thermal behavior of a DWHE using supercritical CO2 as a working fluid close to its critical point. Experiments concerning pressure loss and heat transfer will be carried out as a start for fundamental investigation of heat transfer in mini-channels. This paper contains a detailed description of the test facility, of the first test section and first results regarding heat transfer power and pressure loss.
Development of Probabilistic Risk Assessment Methodology of Decay Heat Removal Function Against Combination Hazard of Low Temperature and Snow for Sodium-Cooled Fast Reactors
Authors: Nishino Hiroyuki , Yamano Hidemasa , Kurisaka Kenichi 
Published: October 2017

A Probabilistic Risk Assessment (PRA) should be performed not only for earthquake and tsunami which are major natural events in Japan, but also for other natural external hazards. However, PRA methodologies for other external hazards and their combination have not been sufficiently developed. This study is intended to develop PRA methodology for a combination of low temperature and snow for a Sodium-cooled Fast Reactor (SFR) that uses the ambient air as its ultimate heat sink for decay heat removal under accident conditions.
Annual excess probabilities of low temperature and of snow are statistically estimated based on the meteorological records of low temperature, snow depth and daily snowfall depth.
To identify core damage sequence, an event tree was developed by considering the impact of low temperature and snow on decay heat removal systems (DHRSs), e.g., plugged intake and/or outtake for the DHRS and for the emergency diesel generator (EDG), unopenable door on the access routes due to accumulated snow, failure of the intake filters due to accumulated snow, possibility of freezing of the water in cooling circuits. Recovery actions (i.e., snow removal and filter replacement) to prevent loss of DHRS function were also considered in developing the event tree. Furthermore, considering that a dominant contributor to snow risk can be failure of snow removal around the intake and outtake induced by loss of the access routes, this study has investigated effects of electric heaters installed around the intake and outtake as an additional countermeasure.
By using the annual excess probabilities and failure probabilities, the event tree was quantified. The result showed that a dominant core damage sequence is failure of the electric heaters and loss of the access routes for snow removal against the combination hazard at daily snowfall depth of 2 m/day, duration time (snow and low temperature) of 1 day.
Assessment on Direct Debris Interaction at High Pressure Melt Ejection Scenario in Level 2 PSA
Authors: Hirokawa Naoki , Chitose Hiromasa , Hisamochi Kohei 
Published: October 2017

Advanced Boiling Water Reactor (ABWR) has multiple safety features to prevent core damage and mitigate the accident progression. Failures of all injection systems to the reactor (ECCS, alternative injection systems, mobile injection systems) could cause a core melt accident and the Reactor Pressure Vessel (RPV) failure. If large amount of debris particles are deposited on the structures in the containment such as access tunnel, hatches, and Vacuum Breakers (V/Bs), it may result in containment failure or Suppression Pool (S/P) bypass due to the direct heating from the corium. This is defined as Direct Debris Interaction (DDI) in this paper.
The phenomenon is introduced in the containment event trees of internal events at power Level 2 Probabilistic Safety Assessment (PSA) for an ABWR and sensitivity analyses and source term analyses are performed in this paper. As a result, the contribution of DDI to total Containment Failure Frequency (CFF) is not significant for the ABWR even if the conservative DDI probability is assumed.
Double-Wall Natural Draft Heat Exchanger Design for Tritium Control in FHRs
Authors: Zhang Sheng , Shi Shanbin , Wu Xiao , Sun Xiaodong , Christensen Richard 
Published: October 2017

Tritium control is potentially a critical issue for Fluoride salt-cooled High-temperature Reactors (FHRs) and Molten Salt Reactors (MSRs). Tritium production rate in these reactors can be significantly higher compared to that in Light Water Reactors (LWRs). Tritium is highly permeable at high temperatures through reactor structures, especially. Therefore, heat exchangers with large heat transfer areas in FHRs and MSRs provide practical paths for the tritium generated in the primary salt migrating into the surroundings, such as Natural Draft Heat Exchangers (NDHXs) in the direct reactor auxiliary cooling system (DRACS), which are proposed as a passive decay heat removal system for these reactors.
A double-wall heat exchanger design was proposed in the literature to significantly minimize the tritium release rate to the environment in FHRs. This unique shell and tube heat exchanger design adopts a three-fluid design concept and each of the heat exchanger tube consists of an inner tube and an outer tube. Each of these tube units forms three flow passages, i.e., the inner channel, annular channel, and outer channel. While this type of heat exchangers was proposed, few such heat exchangers have been designed in the literature, taking into account both heat and tritium mass transfer performance.
In this study, a one-dimensional heat and mass transfer model was developed to assist the design of a double-wall NDHX for FHRs. In this model, the molten salt and air flow through the inner and outer channels, respectively. A selected sweep gas acting as a tritium removal medium flows in the annular channel and takes tritium away to minimize tritium leakage to the air flowing in the outer channel. The heat transfer model was benchmarked against a Computational Fluid Dynamics (CFD) code, i.e., ANSYS Fluent. Good agreement was obtained between the model simulation and Fluent analysis. In addition, the heat and mass transfer models combined with non-dominated sorting in generic algorithms (NSGA) were applied to investigate a potential NDHX design in Advanced High-Temperature Reactor (AHTR), a pre-conceptual FHR design developed by the Oak Ridge National Laboratory. A double-wall NDHX design using inner and outer fluted tubes was therefore optimized and compared with a single-wall design in terms of performance and economics.
Source Term Analysis Considering B4C/Steel Interaction and Oxidation During Severe Accidents
Authors: Ishikawa Jun , Sugiyama Tomoyuki , Maruyama Yu 
Published: October 2017

The Japan Atomic Energy Agency (JAEA) is pursuing the development and application of the methodologies on fission product (FP) chemistry for source term analysis by using the integrated severe accident analysis code THALES2.
In the present study, models for the eutectic interaction of boron carbide (B4C) with steel and the B4C oxidation were incorporated into THALES2 code and applied to the source term analyses for a boiling water reactor (BWR) with Mark-I containment vessel (CV). Two severe accident sequences with drywell (D/W) failure by overpressure initiated by loss of core coolant injection (TQUV sequence) and long-term station blackout (TB sequence) were selected as representative sequences. The analyses indicated that a much larger amount of species from the B4C oxidation was produced in TB sequence than TQUV sequence. More than a half of carbon dioxide (CO2) produced by the B4C oxidation was predicted to dissolve into the water pool of the suppression chamber (S/C), which could largely influence pH of the water pool and consequent formation and release of volatile iodine species.
Theoretical and Experimental Research on the Single-Phase and Incipient Boiling Heat Transfer of Liquid Metal Sodium
Authors: Zhang Dalin , Chen Jing , Song Ping , Wang Shibao , Wang Xin'an , Li Linfeng , Qiu Suizheng , Su G. H. 
Published: October 2017

Sodium-cooled Fast Reactor (SFR) is the most promising reactor among the six Gen-IV nuclear reactor systems. The coolant of SFR is liquid metal which has very different heat transfer characteristics from conventional fluids. The study on heat transfer characteristics of the liquid metal sodium at single-phase and two-phase has not only great academic but also application meanings for SFRs design and safety analysis. Xi’an Jiaotong University (XJTU) performed comprehensive studies from the theoretical and experimental aspects.
In the theoretical studies, the main focuses are paid on the sodium single-phase heat transfer flowing in rod bundles, incipient boiling, two-phase thermal-hydraulic and critical heat flux (CHF). The experimental research on the flow and heat transfer characteristics of sodium in annuli is performed on XJTU single-phase sodium loop and two-phase boiling sodium loop, in which the single-phase friction and Nu correlations, the incipient boiling superheat (IBS) correlations, the two-phase friction multiplier factor and the heat transfer coefficient of sodium boiling are obtained. In this paper, the established theoretical models and the experimental studies for the single-phase heat transfer and the IBS in the annuli or the pin bundles are presented. The theoretical models are validated by the experimental data, and the comparison shows good agreement.
The Application of Task Analysis and Task Support Verification in HPR1000 Main Control Room Human System Interface Design
Authors: Shiguang Deng , Xiaoyang Ming , Yuan Gao 
Published: October 2017

As the brain of Nuclear Power Plant (NPP), Main Control Room (MCR) is an integrated system which comprises varieties of Human System Interfaces (HSIs), procedures, operators and environment. HSI is the medium that the operators monitor and control the Nuclear Power Plant. To improve the operating performance and avoid human errors, Human Factor Engineering (HFE) is required to be implemented in the new plant HSI design for plant safety.
HPR1000 has followed the NUREG-0711 as the guidelines of HFE program, in which 12 elements are included. Task Analysis (TA) and Human Factor Verification & Validation (HF V&V) are two important elements for the HSI design. For task analysis, the design engineers could analyze the human actions (HAs) to identify the tasks that the operators must perform to accomplish the plant functions. Task analysis identifies not only the specific tasks needed to accomplish HAs but also the information, and controls required to complete those tasks which are considered as important input for the MCR HSI design.
HF V&V is one of the final stages of the Human Factors Services as included in NUREG-0711. The HF V&V objective is to evaluate that the HSI design conforms to HFE design principles and that the adequacy of the interfaces enable plant personnel to successfully perform their tasks and other operational goals, assuring plant safety. The V&V element consists of four major activities: Sampling of Operational Conditions, Design Verification, Integrated System Validation, and HED Resolution. For different targets, Design Verification could be separated into two kinds of activities: HSI Task Support Verification (TSV) and HFE Design Verification (DV). TSV is an evaluation to verify that the HSI supports the requirements of personnel tasks, as defined by task analyses. The HAs, controls and information identified in task analysis could be used as the basic input for HSI TSV.
To provide adequate input for the HSI design and TSV, the operational conditions of task analysis should be sampled as many as possible to cover all the typical conditions, which include normal conditions, transients, Design Based Accidents (DBA), Beyond Design Based accidents (BDBA), even severe accidents. The sampled conditions for task analysis would be available for the task support verification of the related HSIs.
In conclusion, the task analysis could be used as the design input for HSI, while the task support verification could review the designed HSI with the output information from TA. This paper addresses the research and engineering implementation methodologies for task analysis and HSI task support verification. As the first reactor of HPR1000, HFE program have been carried out in Fuqing 5&6 MCR from preliminary design to final test. The Fuqing 5&6 TA and TSV achievements for MCR HSIs design are summarized in the paper.
Loading Scheme Research of Small Modular Sodium Cooled Reactor in a Factory
Authors: Xiao Hu , Xiao Liang Chen 
Published: October 2017

An innovative small modular sodium cooled fast reactor called S1 is designed by China Institute of Atomic Energy (CIAE). As an encapsulated nuclear source with thermal power of 3MW, S1 is characterized by small volume, light weight, high safety and sound reliability. The S1 systems adopt modularization, by which the core will be loaded in a factory and filled with sodium, then shipped to the assembled onsite, thus the construction time for S1 can be substantially reduced.
In this paper, the Monte Carlo code MCNP is used to calculate the loading scheme of S1. Considering factory’s special conditions for loading and active chemical property of sodium, a special loading pattern is adopted for S1 — loading fuel first, then sodium filling.
Ensure that the neutron source and detectors are well matched during loading, and detector counting rate is no less than 2cps when only neutron source but no fuel exists in the core. Three positions where the 252Cf neutron source is placed are studied in this paper: (1) at the center of the core; (2) at the middle of outside core plane; (3) at the bottom of outside the reactor vessel. Through MCNP simulation calculations and comparison of large resulting data, it finds the neutron source should be reasonably placed at the bottom of the reactor vessel where 252Cf strength is 105 s−1 neutrons, and the ex-core detectors are distributed symmetrically at the center of outside core plane; the most befitting moderator material of detector surface is methacrylate-C5H8O3. In this paper, 1/N extrapolation method is used during loading and kinds of loading schemes have been studied with reference to the principles of China experimental fast reactor (CEFR) and regulations of relevant research reactors, and 5-batch loading scheme is finally chosen as the optimal loading scheme. S1 is prepared for sodium filling at 250 °C. It shows that neutron flux variation of core can be more reliably monitored when the ex-core detectors are placed about 120cm away from the center core through MCNP simulation calculation. Such arrangement can also meet the monitoring requirements for loading and sodium filling.
Sensitivity Analysis on the Detailed Plant Operational State for Shutdown PSA
Authors: Shiota Daichi , Hirokawa Naoki , Hisamochi Kohei 
Published: October 2017

Probabilistic Safety Assessment (PSA) for the plant shutdown state of nuclear power plants is performed based on Plant Operational States (POSs) which depend on periodic inspections and maintenance. For a typical BWR, the shutdown POSs are defined in consideration of various plant conditions: boundaries of Primary Containment Vessel (PCV)/ Reactor Pressure Vessel (RPV), reactor water level, availability of mitigation systems and temperature/pressure of primary coolant. The shutdown POSs for a BWR plant may involve conservative simplifications. One of the simplifications is treatment of PCV/RPV heads conditions during the POSs involving their opening/closing operations. Usually the PCV/RPV heads conditions are assumed to be the most onerous in terms of accident sequence modelling on both Level 1 and 2 PSA.
In this study, a sensitivity analysis is performed to understand the significance of the conservatism by refining the POSs which involve opening/closing operations of PCV/RPV heads.
Case of Geotechnical Instrumentation of Pipelines in Unstable Zones: Real Time Readings and its Development in Uncommunicated Zones
Authors: Robles Robles Dimas , Castillo Martínez Einstein 
Published: October 2017

Oil pipelines and gas pipelines usually go through geotechnically unstable areas for different reasons. These can go from situations related to the engineering stage (trace), to environmental and social aspects during the construction process. Due to these aspects, the ducts go through geotechnically undesirable areas. Usually, the geotechnical instabilities, according to the kind of movement, are low speed (cm/year), medium (m/year) and very quick processes that generate movements of tens to hundreds of meters per day. Most of Mass Removal Phenomenon (MRF) are triggered by rain and/or earthquakes and are translated into land movements which at the same time involve, occasionally, important deformations in pipelines or its breaking, depending on the movement speed and the possibility of making works before the pipeline breaking. To get to know the pipeline tensional state from the beginning of the pipeline operation, in this unstable zones, is an essential task, which depends on the early identification of the said land movements and the possibility to do measurements on the pipelines using tools such as In-line inspection running (ILI) or the installation of strain gauges. This situation makes the task of monitoring in unstable zones a vital one. The current paper is based on a breaking pipeline case due to soil movement, “monitored by inclinometers”, with the purpose to show the importance of a geotechnical and mechanical instrumentation that offers useful results. The instrumentation allows to model the interaction soil-pipeline to accomplish relevant tasks, that avoid the pipeline breaking and at the same time allow to stablish deformation thresholds of soil or pipeline, which will become early warnings to avoid breakings. Furthermore, the soil and pipeline’s deformation thresholds are documented, based on a system transport by pipelines (STP) breaking cases, to stablish threat classifications to a specific pipeline.
The called instrument reading in real time implies: detection, measurement and data broadcasting that allows the user to have daily records of the movements or required associated variables, with no need to depend on other communication systems that might be inexistent in some areas. This paper also shows the development and operation of a monitoring station that includes: inclinometers, piezometers, strain gauges and rain gauges, among others. These broadcast their data to a server that the user has access to, from any place with a Wi-Fi network, here the user will be able to display information from each one of the instruments, emphasizing the measured variables or magnitudes (displacement, water level, micro strain mm/day) into graphics. The station has a limitation over battery length of 6 months, when it’s problematic to install a recharge solar cell system.