bluuebunny/crossref_metadata_2025 huggingface dataset:
Residual Stress Measurement of Sealing Glass Based on Optical Fiber Sensing Technology
Authors: Mingze Li , Zhichun Fan , Xingzhong Diao , He Yan
Published: October 2018

Metal-to-glass electrical penetration assemblies (EPA) are highly sophisticated equipment and have been used for electrical connection in containment structures or pressure vessels in nuclear plants because of their high temperature resistance and good hermeticity. One important factor to keep hermeticity and reliability can be attributed to the initial residual stress in sealing glass of metal-to-glass EPA. If the residual stress is too high, small defects easily take place in the sealing materials. An insufficient prestress also cannot meet the requirement of high pressure application. To study the influence of residual stress on hermeticity, we developed a novel method of residual stress measurement in metal-glass sealing based on an embedded optical fiber sensor. The fiber Bragg grating (FBG) sensor was embedded in the glass material during the EPA manufacturing, and the residual stress along the grating could be retrieved via optical fiber sensing technique. Basing on our existing metal-glass sealing technique, the initial residual stress could be modulated by changing the sealing process, then the change of different initial residual stress was measured by the embedded FBG, through which the impact of residual stress on metal-glass sealing hermeticity could be finally revealed. A finite element model was established basing on linear elastic theory, then the localized stress along the FBG and the global stress distribution had been investigated theoretically. Taking the stress measuring by FBG as a breakthrough point, the effect of initial residual stress on sealing hermeticity was studied experimentally. The results showed that the residual stress should be an important assessment indicator to metal-to-glass sealing. This research also provided a new approach to optimize EPA manufacture.
Study on Supervision Mode of Floating Nuclear Power Plant With Small Modular Reactor
Authors: Wan Lei , Li Guiyong , Rui Min , Liu Yongkang , Yang Jue
Published: October 2018

A floating nuclear power plant (FNPP) with small modular reactor (SMR) is a combination of a civilian nuclear infrastructure and an offshore installation, which is defined as a floating nuclear facility. The article draws the lessons from studying of the engineer combination like Floating Production Storage and Offloading (FPSO) under the regulation of several government departments. It puts forward recommendations for license application and government regulation as follows in consideration with current license application for nuclear power plant and ship survey. A FNPP shall follow the requirements of construction, fueling and operation for civil nuclear installation combined with ship survey. Application is submitted to nuclear safety regulator for construction permit, while the design drawings shall be submitted to department of ship survey which checks the drawings whether meet the requirements of ship survey, considering some nuclear safety needs. The result of ship survey shall be represented in the safety analysis reports. The construction and important devices manufacturing shall be under the supervision of nuclear installation regulators and ship survey departments. In conclusion, National Nuclear Safety Administration (NNSA) and Maritime Safety Administration of the People’s Republic of China (MSA) shall establish united supervisory system for SMR on sea in China. It is suggested that NNSA is in charge of the overall safety of a FNPP, while MSA is responsible of the ship survey. The operator shall undertake obligation of a FNPP and evaluate the ship cooperating with experienced agency. It is suggested that government departments build the mutual recognition agreement of safety review. It is better to solve the vague questions by coordination.
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.
Development of a Spatial Domain Decomposition Scheme for Monte Carlo Neutron Transport
Authors: García Manuel , Ferraro Diego , Sanchez-Espinoza Victor Hugo , Mercatali Luigi , Leppänen Jaakko , Valtavirta Ville
Published: October 2018

The computing power available nowadays to the average Monte-Carlo-code user is sufficient to perform large-scale neutron transport simulations, such as full-core burnup or high-fidelity multiphysics. In practice however, software limitations in the majority of the available Monte Carlo codes result in a low efficiency when running in High Performance Computing (HPC) environments, the main issues being inadequate memory utilization and poor scalability. The traditional parallel processing scheme based of splitting particle histories among processes requires domain replication across nodes, and therefore the memory demand for each computing node does not scale, and a memory bottleneck appears for large-scale problems. The scalability of this approach usually limits the resources that can be used efficiently to a small number of nodes/processors. Consequently, massively parallel execution is not viable with particle-based parallelism, at least not by itself. In this work we propose a Spatial Domain Decomposition (SDD) approach to develop an efficient and scalable Monte Carlo neutron transport algorithm. Breaking down the geometry into subdomains, a distributed memory scheme can be used to reduce the in-node memory demand, allowing the simulation of large-scale memory-intensive problems. Additionally, with an efficient neutron tracking algorithm the overall speedup can be significantly improved.
Stress Monitoring of Sealing Materials in Electrical Penetration Assemblies
Authors: Zhichun Fan , Mingze Li , Feng Chen , Zhiyong Huang , He Yan
Published: October 2018

Failure of sealing materials is a principal cause to the leakage of electrical penetration assemblies (EPA) in nuclear plants, and the essence can be attributed to small deformations or defects taking place in the sealing materials (glass, epoxy, etc.) as a result of harsh environment influence (high temperature, pressure and ionizing radiation), which leads to leakage when the stress/strain exceeds the threshold value. Metal-to-glass sealing EPA has unique advantages of higher temperature and pressure endurance than organic material sealing EPA, and has been applied in the pressure vessel of High-Temperature Reactor Pebble-bed Modules (HTR-PM) at the Shidao Bay Nuclear Power Plant in China. To achieve on-line state monitoring, we proposed a new method to monitor the stress in the sealing glass by optical fiber sensing technique. Our research showed that the stress in sealing glass could be measured via embedding an optical fiber Bragg grating (FBG) sensor in glass. Optical fiber sensing technique has been widely used for stress measurement in many fields, however applications in metal-to-glass sealing EPA have not been reported in the literature yet. Taking advantage of the small size of a fiber sensor, the embedding of fiber will not affect the sealing structure. And taking advantage of the similar chemical content, fiber and glass can be fused together easily without affecting insulation. In this paper, a brief review on applications of FBG in nuclear facilities was present. The model of FBG embedded EPA was built based on finite element method. Sensitivity analysis about the impact of environment parameters including temperature and pressure on stress had been studied numerically. And the theoretical Bragg wavelength shift of the embedded sensor was derived from the strain/stress distribution. Experiments had been carried out in some main aspects, including pressure and thermal test, from which the relationship between environment parameters and Bragg wavelength shifts was obtained. This research makes an initial attempt for realizing an on-line real-time long-term state monitoring and sets a base for the life cycle diagnostics of EPA in nuclear 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.
Numerical Simulation on Primary Side of AP1000 Steam Generator by Porous Media Model
Authors: Li-qiang Hu
Published: October 2018

This paper employed porous media model to simplify the U-tube bundles on primary side of AP1000 steam generator. The simplified model kept the geometry of the U type tube bundles that have 150mm in height in support plate, and simplified the other part in porous media. Flow resistance was computed by Darcy’s law. Heat sources were modeled in two different ways: an average heat source function and a distributive heat source function that couples the heat transfer between the primary and the second sides. It is concluded that the geometry simplification and porous model are appropriate and accurate to describe and predict the characteristics of flow and heat transfer of the primary side well of the U-tube bundles in the steam generator.
Analysis of Unsteady Flow in a Supercritical Carbon Dioxide Radial Compressor Stage
Authors: Ma Can , Wang Wei , Wu Jun , Dai Lu
Published: October 2018

For nuclear power system, the supercritical CO2-based Brayton cycle is very promising for its potentially higher efficiency and compactness compared to steam-based Rankine cycle. Compressor is the critical component in the supercritical CO2-based cycle, which typically operates at an inlet fluid state close to the fluid critical point for optimal cycle efficiency. As the fluid parameters vary significantly near the critical point, the compressor is more vulnerable to flow instabilities and care must be taken in designing the compressor. The supercritical CO2 radial compressor features a compact design and the unsteady interactions between the impeller and the vaned diffuser are typically strong. A comprehensive understanding of the unsteady flow effects in the compressor is very helpful in guiding the aerodynamic design. However, little work has been performed on the flow analysis of the unsteady impeller-diffuser interactions in the supercritical CO2 radial compressor. In this work, the unsteady flow simulation of a supercritical CO2 radial compressor stage is carried out. Strong flow unsteadiness is observed and the isentropic efficiency shows a variation of over 20% within one revolution.
A Choking Model With Thermal Non-Equilibrium for Initially Subcooled Water
Authors: Yufeng Lv , Minfu Zhao , Weiqing Li
Published: October 2018

Mechanical non-homogeneous and thermal non-equilibrium phenomenon exists in two-phase critical flow compared with single phase flow. A one-dimensional two-fluid critical flow model is developed for initially subcooled water flowing in pipe or orifices. The model accounts for thermal nonequilibrium between the liquid and vapor bubbles and for interphase relative motion. In this model, an improved correlation to calculate flashing inception location and surperheat is proposed. The model consists of six conservation equations as well as a seventh equation representing bubble growth in bubbly flow. Closure of the set of governing equations is performed with constitutive relationships which determine the interfacial momentum terms due to mass exchange, wall to liquid and wall to vapour frictional forces, liquid to gas interfacial force and interfacial heat transfer rate. The model considers the development of three flow regimes, namely, bubbly, churn and annular flow regimes. Model predictions compare favorably with experimental data over a wide range of pressures and pipe diameters and lengths.
Modelling Frozen Salt Films in a Molten Salt Fast Reactor
Authors: Cartland-Glover Gregory M. , Rolfo Stefano , Skillen Alex , Emerson David R. , Moulinec Charles , Litskevich Dzianis , Merk Bruno
Published: October 2018

Molten salt reactors are a very promising option for the development of highly innovative solutions for the nuclear energy production of the future. The techniques used to model thermal hydraulics of a molten salt fast reactor when frozen salt wall technology is applied to the core vessel wall are presented here for 2D numerical models of a hyperboloid reactor core region with a heat exchanger was applied in Code_Saturne. A 3D simulation of the fluid flow and heat transfer with 16 recirculation loops containing the heat exchangers is also presented. It was found that there is strong cooling in separated flow regions in the external heat exchanger, which freezes where the porous model is applied.
Effect of Subcooling and Nozzle Diameter on Heat Transfer Characteristics of Downward Facing Hot Surfaces Using Mist Jet
Authors: Sharma Avadhesh Kumar , Meena Monika , Soni Anirudh , Sahu Santosh K.
Published: October 2018

The jet impingement cooling is always preferred over the other cooling methods due to its high heat removal capability. However, rapid quenching may lead to the formation of cracks and poor ductility to the quenched surface. Mist jet impingement cooling offers an alternative method to uncontrolled rapid cooling, particularly in steel and electronic industries. In mist cooling, the droplets are atomized by compressed air. Experiments are performed under transient conditions using two full-cone spray nozzles (Lechler Pneumatic atomizing nozzle 136.115.xx.A2 and 136.134.xx.A2) to study the effect of subcooling and nozzle diameter on surface heat flux. The hot surface used for the experiment is a stainless steel foil (AISI-304) of thickness 0.15mm. The initial surface temperature of the plate is maintained at 500±10°C with the help of an AC transformer. Infrared thermal image camera (A655sc, FLIR System) is used for data estimation. The IR camera and the nozzle are positioned on either side of the plate. The variation in surface temperature has been acquired at 8 different spatial points. It has been observed that that as we move away from the stagnation point then irrespective of air and water flow rates surface heat flux decreases. The maximum surface heat flux obtained at the stagnation point. With the increase in diameter surface heat flux increases irrespective of pressure values. The correlation between qm/qstag experimental and predicted values has been shown.
Review and Security Assessment of Red Oil Explosions in Evaporator
Authors: Yiren Lian , Hongchao Sun , Lei Chen , Dongyuan Meng , Guoqiang Li , Dajie Zhuang , Shutang Sun , Jiangang Zhang
Published: October 2018

Purex progress is widely applied in the nuclear fuel reprocessing plants all over the world. However, various security problems occur in reprocessing facility involving the intense attention of red oil explosion. The exothermic reactions among TBP, nitric salts and nitric acid were responsible for the red oil explosion. In this paper, explosion events at nuclear fuel reprocessing plants initiated by red oil phenomena were reviewed. The formation and reaction mechanisms of red oil causing explosions were analyzed. Moreover, the evaluation and analysis model was built up to evaluate the security assessment of red oil explosion from the associated calculations of data in the typical red oil explosive accidents and the results of subsequent laboratory studies on the chemical reaction.
The Simulation Research of Noncondensable Gas to Condensation in Secondary Side Condenser of Floating Nuclear Power Plant Based on RELAP5
Authors: Yan Si-wei , Li Chun-mei , Liang Tie-bo , Zhao Jing , Hao Cheng-ming , Wang Yu
Published: October 2018

Similar to conventional nuclear power plant, condensate water subcooling is a common problem in secondary coolant of floating nuclear power plant, which is caused by many reasons. In this article, RELAP5 is used to simulate the phenomenon of condensate water subcooling caused by noncondensable gas. The influence of noncondensable gas to condenser pressure, subcooling temperature, heat transfer rate, terminal temperature difference, cooling water temperature rise is presented. The results obtained through this study have shown that the model with non-condensable gas in steam can simulate condensate water subcooling, and reveal the discipline of condenser heat transfer characteristics as a function of noncondensable gas content.
A Multi-Objective Optimization of the Reactor Power Plant
Authors: Lei Chen , Zhen Jia , Cong Wang , Zili Gong , Yi Liao , Chen Hu
Published: October 2018

From the view of practical engineering application, a compacter nuclear power plant is expected. The weight and the volume of a nuclear power plant can be reduced by optimal selection of the operational parameters. In this work, a thermal-hydraulic model of the reactor, mathematical models of the reactor vessel, the main pipe, the pressurizer, the steam generator, the turbine and the condenser were established for the Qinshan-I nuclear power plant based on the related technical materials. The responses of the optimal targets to the changes of the design variables were studied by the sensitivity analyses. The non-dominated solution front of the nuclear power plant was obtained by means of the immune memory clone constrained multi-objective optimization algorithm. The study shows that the component mathematical models are reliable for the optimization process, the distribution of the non-dominated solution is decided by the steam generator secondary pressure. The volume and the weight of the system could be at least reduced by 23.0% and 9.5%, respectively.
Preliminary Development of a TRISO Fuel Performance Analysis Code: FFAT
Authors: Li Jian , She Ding , Shi Lei , Zhao Jing
Published: October 2018

Tristructural isotropic (TRISO) fuel particles are chosen as the major fuel type of High temperature gas cooled reactor (HTGR). The TRISO coated particle also acts as the first barrier for radioactivity retention. The performance of the TRISO coated particle has a significant influence on the safety of HTGR. A set of fuel performance analysis codes have been developed during the past decades. The main functions of these codes are conducting stress calculation and failure probability prediction. PANAMA is a widely used German version fuel performance analysis code, which simulates the mechanical performance of TRISO coated particle under normal and accident conditions. In this code, only a simple pressure vessel model is considered, which is insufficient in stress analysis and fuel failure rate prediction. Nowadays, efforts have been done to update the fuel performance model utilized in PANAMA code, and a new TRISO fuel performance analysis code, FFAT, is under developed. This paper describes the newly updated TRISO fuel performance model and presents some first results based on the updated model.
Criticality Safety Issues in Nuclear Design of HP-SMTCs Space Reactor
Authors: Li Songyang , Wang Dingqu , Jiang Yueyuan , Li Weihua , Guo Wenli
Published: October 2018

Heat pipe-segmented thermoelectric module converters space reactor power system (HP-SMTCs SRPS) is a promising candidate for space nuclear power system. An examination was taken to discuss the criticality safety of HP-SMTCs reactor core in several accident conditions. In the original nuclear design, the case that reactor core is submerged in wet sand while the voids inside the core are filled with sea water and the BeO reflector are dismantled is defined as the worst status with the highest risk of supercritical. However, recent Monte Carlo transport calculation result shows that reactivity in the case that the core is submerged in water while the voids inside are empty is even higher than those cases with the voids full of sea water, which means that the reactor may encounter high risk of supercritical when some particular accidents occur. Detailed analysis about the neutron energy spectrum and absorption reaction rate is made in order to find out the potential reason of these unexpected results. According to the discussion about the criticality safety issues in some accidents, further evaluations may be necessary for the neutronics design of HP-SMTCs space reactors.
JSNT-S: A Parallel 3D Discrete Ordinates Radiation Transport Code on Structured Mesh
Authors: Cheng Tangpei , Mo Zeyao , Yang Chao , Wen Lili , Deng Li
Published: October 2018

To address the performance and programming challenges from the constantly evolving computer architecture in parallel discrete ordinates method, this paper presents our continue effort on developing and improving the parallel 3D discrete ordinates radiation transport code JSNT-S. A hybrid acceleration method for source iteration is presented, which combines the PCR algorithm and the Anderson algorithm for accelerating different energy groups. A patch-abstraction and hybrid MPI+threads implementation is employed for the domain partition algorithm and the sweeping algorithm. Several optimizations on the data-driven sweeping algorithm are presented, such as exploiting parallelism in angular directions, developing optimal priority strategy for scheduling and reducing DAG construction and scheduling overhead by vertex clustering. The correctness is verified using the AMBER procedure and the result of VENUS-3 benchmark is demonstrated. The parallel performance is measured using the model modified from VENUS-2 and the Kobayashi benchmark, the result of which shows reasonable scaling performance up to 76,800 cores.
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.
ACRS’ Enduring Legacy Contributing to Reactor Safety
Authors: Nourbakhsh Hossein
Published: October 2018

For over 60 years the Advisory Committee on Reactor Safeguards (ACRS) has had a continuing statutory responsibility for providing independent reviews of, and advising on, the safety of proposed or existing reactor facilities and the adequacy of proposed reactor safety standards in the United States. This paper discusses the role of the Committee as it has evolved during its more than 60 years of history, noting some of its significant contributions to reactor safety.
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.
Segmented γ Scanning Device and its Experimental Research
Authors: Hou Su-xia , Chen Chen , Zhang Quan-hu , Su Xiang-hua , Zuo Wen-ming
Published: October 2018

Segmented Gamma Scanner (SGS) is a commonly used nondestructive testing (Non-Destructive Assay NDA) method. SGS uses radial rotation, axial segmentation, segmented scanning of the non-uniform sample of the uniform treatment, making it possible to accurately measure the radioactivity on each segment and is currently widely used in the field of nuclear material management. This paper introduces a self-designed SGS measuring device and uses this device to perform a large number of scanning experiments on different measurement objects in the laboratory. It also studied the detection performance, stability and SGS method of different objects the accuracy of the measurement results.
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.
On the Use of Robust Command Shaping for Vibration Reduction During Remote Handling of Large Components in Tokamak Devices
Authors: Grazioso Stanislao , Di Gironimo Giuseppe
Published: October 2018

This paper proposes to use robust command shaping methods for reducing the vibrations during remote handling of in-vessel components. The need of deriving efficient vibration control strategies for a safe transportation of large and heavy pay-loads during maintenance procedures in nuclear fusion reactors is the main motivation behind this work. The approach shapes the reference motion command to the component such that the vibratory modes of the system are canceled. We perform the dynamic simulations of a large in-vessel component of the DEMOnstrating fusion power reactor during a remote handling operation. The simulations shows that the method is a possible solution to reduce the vibrations induced by the motion, in both the transient and residual phases. The benefits introduced by command shaping make the method promising towards building control framework for remote handling of in-vessel components in various tokamak devices.
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.
Development of an Optimized Transport Solver in SARAX for Fast Reactor Analysis
Authors: Xu Zhitao , Wu Hongchun , Zheng Youqi , He Mingtao
Published: October 2018

Although the neutron mean free path in typical fast reactors is long compared with that of typical thermal reactors, the heterogeneous flux distributions in angular domain and spatial domain are notable due to the strong elastic-scattering resonance and common core-level heterogeneous layout. Full-core multigroup transport analysis becomes inevitable for advanced fast reactor R&D. A fast reactor neutronics analysis system, SARAX is under development at XJTU of China. An existing in-house SN-nodal solver in triangular-Z geometry, DNTR was primarily chosen for its geometric adaptability. However, insufferable problems of computing time and storage were encountered when the solver was on active service. In this paper, the problems are firstly analysed and resolved from the theory and code levels. Then, widely used CMFD acceleration method with some stabilization techniques is implemented for the SN-nodal method in triangular-Z geometry, which can largely reduce the computing time. What’s more, a new acceleration method TCD is proposed and can obtain superior speedups with reasonable accuracy sacrifice. The updated solver, DNTR 1.1 has been developed based on these improvements, which can obtain dozens of speedup with reduced storage compared with DNTR for typical fast reactor simulation on a desktop computer.
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.
Phénix Transient Analysis for the Assessment of RELAP5-3D Based on Dissymmetric Test Benchmark
Authors: Giannetti F. , Narcisi V. , Subioli A. , Del Nevo A.
Published: October 2018

Before the definitive shutdown of Phénix fast reactor, a pool-type sodium-cooled prototype, occurred in the year 2009, the French Commissariat a l’Energie Atomique et aux Energies Alternatives (CEA) decided to carry out a final set of experimental tests, in order to gather data and additional knowledge on relevant Sodium Fast Reactors (SFR) operation and safety aspects. One of these experiments was the dissymmetrical configuration test, which has been selected as a benchmark transient on H2020 SESAME project. ENEA and Sapienza University of Rome are participating, in collaboration, in the benchmark using RELAP5-3D© code, developed by Idaho National Laboratory (INL). The analysis of the thermal hydraulic characteristics and the plant behavior focuses on the accurate core cooling prediction in natural circulation, by assessing the flow and the temperatures in the various reactor components. Aiming to investigate the asymmetric thermal hydraulic behavior inside the pool, the nodalization scheme has been developed using a detailed parallel pipe with cross-junctions for the hot and cold pool and the core has been simulated assembly per assembly, including blanket and reflector, in the first seven rings, ready for a proficient coupling with a three-dimensional neutron kinetic analysis code. The scope of the present study is to verify if a detailed thermal–hydraulics pool model could add useful data to show the differences which occur in each loop and into the pools, caused by the azimuthal asymmetry of the boundary conditions. Blind calculations results are here presented and discussed, highlighting the details added with the multidimensional nodalization scheme, compared to a mono-dimensional model. The paper will be a first step on the comparison of a Phénix mono-dimensional and three-dimensional thermal-hydraulic nodalization and on the RELAP5-3D© code assessment against the dissymmetrical experimental results.