bluuebunny/crossref_metadata_2025_split huggingface dataset:
Experimental Investigation on Boiling Flow Characteristics Under Passive IVR-ERVC Conditions
Authors: Wang Fan , Kuang Bo , Liu Pengfei , He Longkun 
Published: October 2018

In vessel retention (IVR) of molten core debris via water cooling at the external surface of the reactor vessel is an important severe accident management feature of advanced passive plants. During postulated severe accidents, the heat generated due to the molten debris relocation to the lower reactor pressure vessel head needs to be removed continuously to prevent vessel failure.
Besides the local critical heat flux (CHF) of outer wall surface which is the first importance, a stable feature of natural circulation flow and an effective natural circulation capability within the external reactor vessel cooling (ERVC) channel tend to be rather crucial for the success of IVR. Under this circumstance, a full-height ERVC test infrastructure for large advanced pressurized water reactor (PWR) IVR strategy engineering validation, namely reactor pressure vessel external cooling II test facility (REPEC-II), has been designed and constructed in Shanghai Jiao Tong University (SJTU). And therefore, a brief introduction to the SJTU REPEC II facility as well as the experimental progress to date, is hereby given in the paper. During test campaign on the REPEC II facility, the one-dimensional natural circulation boiling flow characteristics during IVR-ERVC severe accident mitigation are investigated, with the experimental observation and measurement on natural circulation flow characteristics within the REPEC II test facility.
Based on the abundant results acquired in the test campaign, it is attempted, in this paper, to summarize and evaluate the ERVC performances and trends under various practical engineered conditions. The main evaluation results includes: influence on ERVC flow characteristics of various non-uniform heat load distribution cooling limits, the observed sinusoidal oscillation is suggested to be flashing-induced density wave oscillations and the oscillation period correlated well with the passing time of single-phase liquid in the riser. It is expected that these conclusions may help designers to have a reliable estimate of the impact of some related engineered factors on real IVR-ERVC performance.
Foreseen Capabilities, Bottlenecks Identification and Potential Limitations of Serpent MC Transport Code in Large-Scale Full 3-D Burnup Calculations
Authors: Ferraro Diego , Garcia Manuel , Mercatali Luigi , Sanchez Espinoza Victor Hugo , Leppänen Jaakko , Valtavirta Ville 
Published: October 2018

Continuous improvement in Nuclear Industry Safety Standards and reactor designer’s and operator’s commercial goals lead to an increasing demand of fast running and highly accurate methodologies oriented to improve the prediction capabilities of main reactor’s parameters under steady state and transient situations.
An increasing effort has been observed during past years to develop high accurate multi-physics approach for nuclear reactor analysis, based both on the availability of advanced codes and the constant increase of computational resources with massive parallel architectures. As a result, several improvements have been observed in the implementation of coupled full 3-D Monte Carlo (MC) neutronic models for nuclear reactor cores, including not only the coupling to thermal-hydraulics but also fuel behavior codes. This approach has proved at concept level to be able to develop high accurate models that would allow to predict important safety and performance parameters of nuclear reactors with less conservativism.
Under this framework, the European Research and innovation project McSAFE is a coordinated effort started in September 2017 with the objective of moving MC stand-alone and coupled solution methodologies to become valuable tools for core design, safety analysis and industry like applications for LWRs of gen II and III.
In this work the Serpent 2 code, a high performance MC code developed by VTT, is used by the aim of performing the preliminary screening of capabilities, performance and limitations of such challenging objective.
As a result, simplified analysis are developed to identify full 3-D modeling computational requirements for typical LWR configurations, including burnup aspects. Potential bottlenecks and limitations are presented and discussed, providing foreseen alternatives and solutions for further code improvements.
European Utility Requirements for Advanced LWR Issue of EUR Revision E and Ongoing Assessments
Authors: Chappell Peter , Jacquart Guillaume , Ferraro Giovanni , Kymäläinen Olli 
Published: October 2018

The purpose of the European Utility Requirements (EUR) Organisation is to actively develop and promote harmonised requirements for new mid- and large-size LWR NPPs that are proposed for construction in Europe. The harmonisation which is sought by the fourteen member utilities of the EUR Organisation, aims at delivering the safest and most competitive designs based on common requirements shared across Europe.
The harmonised requirements are presented in the EUR document. This consists of an extensive set of requirements covering all aspects (safety, performance, competitiveness) and all parts of a NPP (Nuclear and Conventional Islands). It can be used by Utilities (e.g. for assessment of the GEN3 designs proposed by vendors, technical reference for call for bids) and by Vendors (e.g. understanding of customer’s expectation of GEN3 NPPs, facilitating the licensing process).
The presentation will describe the main outcomes of the last 3 years of EUR Organisation activities (roadmap 2016–2018) and the challenges for the coming near future, in the following three fields.
First, completion of Revision E of the EUR Document was achieved in December 2016 and issued in July 2017, followed in October 2017 by a training course that was attended by 90 participants. The presentation will describe the most significant updates including revision of the safety requirements to align to the most recent European and International safety standards issued by WENRA and IAEA, lessons learned from the Fukushima accident, including re-evaluated Seismic and External Natural Hazards approach and updated international standards (e.g. for I&C and for European Grid code). Revision E also includes feedback from previous design assessments of NPPs. Future possible development of the EUR Document (and of the assessment process) will be considered within the EUR Organisation ‘Roadmap’.
Assessment of new designs is the second main technical activity. The assessment of the KHNP EU-APR (European version of APR1400) was completed in 2017 and an assessment of the Russian AEP’s VVER-TOI is planned to complete in 2018. These assessments are against EUR Revision D. The first assessment against Revision E (of the CGN HPR1000 “Hualong” design) is planned between 2018 and 2020. The presentation will recall the EUR design assessment objectives and process and the outcomes and progress of the different assessments.
The third topic is the interaction between the EUR and other stakeholders, in particular other international organisations (ENISS, WNA/CORDEL, WENRA, EC, IAEA, EPRI/URD) with the aim of promoting Industry Requirements and influencing prospective regulation where appropriate. The presentation will describe how the EUR Organisation is connected to these stakeholders and how it presents Utility requirements to the wider nuclear industry.
Non-Condensable Gas Plugging and Mixing Behavior in PWR Steam Generator Tubes During Reflux Condensation
Authors: Janasz Filip , Prasser Horst-Michael , Suckow Detlef , Szogradi Marton 
Published: October 2018

In this paper we investigated reflux condensation in a single pressurized water reactor steam generator tube in the presence of non-condensable gases. We observed the gas plug formation, behavior and impact on heat removal under thermo-hydraulic conditions typical for the LOCA stage during which the reflux condenser regime is characteristic. We used nitrogen and helium to observe differences of the plug behavior for heavy and light gases.
Measurements reveal that the non-condensable gas type plays a crucial role in the plug formation. Light gas (He) forms sharp and narrow interface and limits condensation efficiently while heavy gas (N2) develops wide, diffuse interface with gradual temperature drop.
CFD Modeling of Annular Flow With a Three-Field Approach
Authors: Raddino Salvatore , Strömgren Tobias , Le Moigne Yann , Le Corre Jean-Marie 
Published: October 2018

Prediction of Critical Heat Flux and dryout is crucial for BWRs. Today 1-D models based on empirical correlations are used in order to model these phenomena. However, in order to get a better physical understanding and be able to model more complex geometries CFD modeling of annular flow is developed. The complexity of annular flows makes modeling of the different interactions involved challenging.
In the present study a three-field approach is used in order to model annular flow; steam, liquid droplets and liquid film represent the different fields which are coupled to each other, i.e. both mass and momentum are exchanged between the modeled fields. The steam and droplets are modeled using Euler-Lagrangian multiphase CFD model and the liquid film is modeled through empirical models implemented at the wall boundary.
The implemented model is validated against experimental data with good agreement. In order to enlarge the model validity range further improvements are needed especially concerning the entrainment modeling.
Analysis of Vibrations due to the Steam Condensation at Sub-Atmospheric Condition
Authors: Giambartolomei G. , Lo Frano R. , Mazed D. , Del Serra D. , Aquaro D. 
Published: October 2018

The condensation of steam in water may cause pressure oscillations, vibrations, and damages to piping and, in some extreme conditions to the internals of suppression tank. Significant R&D activity, especially focusing on the emergency cooling system in BWRs, has been conducted in the past decades to clarify the mechanism of the condensation oscillation evolving/related to the Direct Contact Condensation (DCC).
The present paper deals with the pressure oscillations due to the condensation of steam jet in water, at sub-atmospheric condition; phenomena that have not been fully investigated yet. Vibrations, arisen as flow patterns transformed from stable to unstable, are investigated with particular attention to the dynamic behaviour of the whole suppression system, behavior of the interface, flow patterns etc.
To the purpose, a numerical study is performed by means of suitable FEM in order to identify the dominant frequency associated to the steam jet condensation oscillation in water flow, and determine the resulting values of pressure and acceleration. The obtained results allowed to formulate a correlation between the dominant frequency and the condensation driving potential and steam mass flux.
CFD Simulations of Aerosol Dispersion and Agglomeration During the Laser Cutting of Fukushima Fuel Debris Simulants
Authors: Gelain Thomas , Porcheron Emmanuel , Chagnot Christophe , Roulet Damien 
Published: October 2018

The general context of this article is related to the development of the laser cutting technique for the fuel debris retrieval on the damaged reactors of Fukushima Daiichi. IRSN is involved in a project led by ONET with CEA, to bring relevant elements to analyze the risk occurred by the dispersion of aerosols emitted by the dismantling operations.
In this context, CFD (Computational Fluid Dynamics) simulations of dispersion of aerosols (including the transport, the deposition on the walls and the agglomeration), emitted during laser cutting of non-radioactive fuel debris simulants undertaken on the DELIA cutting laser platform from CEA, have been conducted. Indeed, evaluating the amount of aerosols likely to deposit on the walls and those to be released into the environment is one of the key safety issues in the dismantling actions of reactors of Fukushima Daiichi.
The CFD simulations have been carried out with the commercial code ANSYS CFX in which models of aerosol transport and deposition, developed and validated by IRSN, have been implemented. In a first time, airflow simulations have allowed to visualize the flow in the vessel of the DELIA laser cutting facility from CEA by the way of streamlines before injecting aerosols. Aerosol input data (size and morphology) have been acquired by IRSN team on the DELIA facility during the laser cutting of non-radioactive simulants of fuel debris for air and underwater configurations.
Those simulants of fuel debris are representative of the fuel debris present in the Fukushima reactors [1].
The simulation results allow to obtain the cartography of the aerosol deposition in the vessel for different aerosol sizes and to evaluate the amount of dispersed aerosols. These simulations also show the time evolution of the aerosols properties with the agglomeration phenomena that can influence their size evolution and therefore their dispersion and deposition.
Similar simulations have also been carried out on the geometry of a Fukushima Daiichi reactor pedestal in which laser cutting could be performed.
This paper presents the aerosol dispersion simulations, the model implemented for aerosol transport and deposition as well as for agglomeration, and finally a comparison with experimental data presented in a companion paper (ICONE26-81531).
Non-LWR Model Development for the MELCOR Code
Authors: Humphries Larry , Beeny Brad , Louie David , Esmaili Hossein , Salay Michael 
Published: October 2018

MELCOR is a fully-integrated, system-level computer code developed by Sandia National Laboratories (SNL) for the Nuclear Regulatory Commission (NRC) with the primary objective of modeling the progression of severe accidents in light water nuclear power plants [1,2,3]. Since the project began in 1982, MELCOR has undergone continuous development to address emerging issues, process new experimental information, and create a repository of knowledge on severe accident phenomena. This paper summarizes model development specifically developed for non-LWR applications such as high temperature gas reactors (HTGR), sodium fast reactors (SFR) and molten salt reactors (MSR).
Beginning in 2008, active development work began on HTGR modeling in MELCOR. Models were developed for helium gas thermodynamics, oxidation of graphite, thermal hydraulics and heat transfer for both prismatic and pebble bed designs, cavity cooling systems, fuel failure and fission product release, graphite dust generation, and aerosol transport, deposition, and resuspension.
In 2013, work commenced on the development of modeling capabilities for sodium fast reactors. This development included the addition of sodium as a working fluid as well as the addition of models for simulating containment fires (both spray and pool) as well as sodium atmospheric chemistry. Validation of these new models has been completed and code-to-code comparisons with the CONTAIN/LMR code has been performed. Work continues as development of sodium concrete interaction models is now underway.
In 2017, work began on adding capabilities for molten salt reactors. A new equation of state for FLIBE coolant has been successfully tested in MELCOR and is now undergoing validation against experiments. The alternate working fluid model has also been extended to permit both water and one alternate condensable working fluid in the same input model.
Systems Engineering Approach for Pre-Conceptual Design of DEMO Divertor
Authors: Marzullo Domenico , Dongiovanni Danilo N. , You Jeong Ha 
Published: October 2018

The DEMOnstration Fusion power Plant (DEMO) will be a key step towards Fusion Power Plant technology. It represents the single step to a commercial fusion power plant, in charge of demonstrating the viability of relevant technologies. Indeed, the development of tokamak sub-systems has to take into account interface, structural and functional requirements and multi-physics issues that can be completely known only during the development of the design process. This leads to difficulties to be faced during the conceptual design, mainly related to the identification of the main requirements, the change management and the sub-system integration. The Systems Engineering approach aims to support the design and management of complex systems over their life-cycles, providing a systematic approach for the definition of customer needs and required functionality from the early stage of the design, as well as for the design synthesis and the system validation and verification.
Among the tokamak sub-systems, the divertor is the one devoted to power exhaust management and represents, at the same time, one of the most challenging components, in terms of materials, technologies and manufacturing. Current design activities, conducted in the in the framework of EUROfusion Consortium are in a pre-conceptual phase. Despite the early design stage, a systems engineering approach is being applied as an integrated, interdisciplinary R&D effort.
The paper therefore presents the modeling effort to the conceptual design of DEMO divertor aimed at identifying both system main functions and expected behavior, given the constraints imposed from either project requirement or from current viability of technological solutions. To allow for flexibility in design needed to explore the feasibility of different solutions at this pre-conceptual stage, the impact of possible changes in high level requirement or interfaces is also investigated. This is also achieved through the allocation of the requirements to the affected components and providing efficient traceability. Therefore, the proposed modelling approach is intended to support the whole divertor conceptual design stage, allowing for requirements identification, traceability and change management.
Dose Minimization Game for Smartphones
Authors: Stelter Nolan , Das Arnav , Hanifah Zahra , Uddin Rizwan 
Published: October 2018

Due to misconceptions surrounding radiation and nuclear energy, educating the general public about basic radiation concepts has become increasingly important. The Virtual Education and Research Laboratory (VERL) at the University of Illinois at Urbana-Champaign (UIUC) has developed a 3D, virtual, interactive game that conveys the physics of radiation and principles of radiation protection to the player. The model is a scavenger hunt style game that takes place in a virtual model of a TRIGA research reactor. Several virtual radiation sources are placed in the 3D virtual model of the TRIGA facility. Radiation drops away from the radiation source. The effect of shielding can also be incorporated in modeling the radiation transport, leading to realistic radiation fields. The user’s goal is to find and collect (virtual) objects placed in this facility while minimizing the dose received in doing so. The player is meant to learn about time, distance, and shielding — key concepts in radiation protection. The start screen displays the radiation field in the form of a colored coded floor, as well as the location of the desired objects. With the given information, the player is encouraged to plan the route to collect the items and minimize exposure. By repeatedly playing the game, the player becomes familiar with the layout of the facility, and of the location of the radiation sources. This educational game is a useful teaching tool. Those unfamiliar with radiation protection concepts are able to understand how important time, distance, and shielding are in minimizing dosage. Additionally, this game proves to be a useful engagement and outreach tool. Upon completion of the game, the user is shown the score, the dose received, as well as a list of dose received in well-known instances such as eating a banana or in getting an x-ray at the dentist’s office.
The dose minimization game developed earlier for computers has now been developed for use as a game-app for cell phones. These recent developments allow for wider outreach, further increasing the use of the model as an outreach and educational tool.
Real-Time Online Risk Monitoring and Management Method for Maintenance Optimization in Nuclear Power Plant
Authors: Xu Anqi , Zhang Zhijian , Zhang HuaZhi , Zhang Min , Wang He , Ma Yingfei , Chen Sijuan , Wang Yan , Zheng Gangyang 
Published: October 2018

As a nuclear power plant system configuration risk assessment tool, Risk Monitor (RM) can periodically update Living-PSA risk monitoring model to calculate risk information. It not only provides operating personnel with the real-time risk information, but also reduces human error in test and maintenance work, in order to ensure that the high-risk configurations are identified and controlled.
Traditionally, the high-risk identification of planning activities is mainly based on technical specification (TS) and operating experience, etc. It is lack of quantitative risk assessment to support decision-making, especially when multiple systems or components are out of service or being restored.
Planning risk assessment in existing risk monitors is time-consuming and error-prone, because they typically depend on manually screening and confirming risk-related planned events. Besides, online time-dependent characteristics of NPP systems are ignored. The reliability parameters are taken as constant values providing that components and systems always work under the predefined conditions. Thus, the risk level during planning period is not affected by the current configuration, or the operation history of equipment.
With the widespread application of digital instrument control systems (I&C) in nuclear power plants, the technology of Real-time Online Risk Monitoring (RORM) was put forward. It improves the acquisition technique of current equipment status, modeling and updating technique of Living-PSA Model and provides more accurate, realistic prediction of risk level for planning configurations.
This paper briefly introduces the design of the system structure, database, interfaces and functions of Real-time Online risk monitoring (RORM). It is characterized by the following features: online acquisition of the initial configuration before planning period, time-dependent risk monitoring modeling and updating in time. And it focuses how the real-time online risk monitoring technology of nuclear power plants help the nuclear power plant minimize the risk level of maintenance plan and optimize the maintenance schedule. Also, the calculation method of risk monitoring measures is improved based on the concept of “online, time-dependent”. Finally, the risk management method for optimization of planning activities is proposed.