bluuebunny/crossref_metadata_2025 huggingface dataset:
Applying Computational Fluid Dynamics Methodologies to Hydro-Dynamic Loads in the Suppression Pool of a BWR
Authors: Mauritzson David , Perzon Sven , Probert Thomas
Published: October 2018

The direct and indirect hydro-dynamic loads in the suppression pool of the BWR fleet were developed well before 2000. These loads were based on scaled experiments and numerical solutions using one dimensional models. The analysis was cumbersome and conservatisms were added in multiple steps.

As these loads spread throughout the containment, as global vibrations, they are generally a part of all structural verification inside the containment. This fact has made it hard and unpractical to challenge and revise these loads, as a change could lead to significant re-work. A consequence of this is that loads in the suppression pools have seldom been revisited, regardless if they cause local or global vibrations. This is problematic when new equipment is needed. The design of this equipment suffers from the very conservative loads.

Experience has shown that these loads can be challenged and refined using updated techniques, leading to significantly lower loads. This was realized during the modernization of Swedish plant Oskarshamn Unit 2 with Mark II containment, where loads following pool swell proved to be particularly challenging and it was decided to investigate possibilities to reduce the conservative loads.

Due to the large scale of the condensation pool coupled with the transient and small scaled condensation of steam at the drywell vent pipe nozzle full CFD resolution is not feasible. Instead lumped models in GOTHIC was used to increase resolution from the 1D approach of normal containment analysis, to a resolution that can account for the features of the condensation pool. This showed that the pool swell was less uniform than initially thought, leading to fewer objects affected as well as lower loads on objects that suffered from these loads. A full CFD analysis was then used to resolve phenomena working on even shorter time scales leading to a complete rework of all local loads.

The loads addressed using updated codes and modelling techniques was pool swell impact (PSI), pool swell drag (PSD) and local drag loads due to pressure relief valve opening, LV/SRV.

The current work shows that using updated modelling techniques and aligning results with previous analysis and documentation, it is possible to reduce loads for some events in the suppression pool without violating safety for the power plant. The results from the GOTHIC model shows that the cylindrical shape of the pool will create an uneven velocity distribution radially at pool swell resulting in much smaller loads at the outer boundary.

CFD analysis of the LV/SRV event shows that the loads are reduced in comparison with previous methodology and this is mainly because of shadow effects. The calculations also verifies assumptions used in the previous methodology.
Preliminary Study on Thermal Performance of Inert Matrix Disperse Pellet Using FEA Method
Authors: Zhiwei Lu , Yun Li , Yongdong Zhang , Lei Li , Zhaodandan Ma , Tong Liu
Published: October 2018

Based on the tri-isotropic (TRISO) fuel technology of high temperature gas cooled reactor, a new fuel concept for improving the accident tolerance of light water reactor (LWRs) named inert matrix dispersion pellet (IMDP) was proposed. Through the silicon carbide matrix and embedded TRISO fuel particles, the safety of the nuclear fuel could be enhanced. Recently, dummy IMDPs were fabricated by China General Nuclear Power Corporation (CGN) and thermal conductivity was tested. According to the tested data, a FEA model using ABAQUS combined its secondary development function was developed and benchmarked. Several influence factors of the effective thermal conductivity (ETC) of the IMDP were studied by the FEA model, such as burn-up, TRISO packing fraction and temperature. The heat transfer behaviors of IMDP and UO2 under typical normal PWR operating condition were also studied.
Analysis and Improvement of Global-Local Self-Shielding Calculation Scheme for AIC Control Rods
Authors: Li Jikui , Zu Tiejun , Cao Liangzhi , Wu HongChun , He Qingming
Published: October 2018

Global-local self-shielding calculation scheme is a new high-fidelity resonance calculation model proposed by NECP laboratory of Xi'an Jiaotong University. Neutron Current Method (NCM) is utilized for resonance calculation in the global aspect to obtain Dancoff factors. Then each fuel pin is transformed into individual 1D cylindrical problems by conserving Dancoff factors. The Pseudo-Resonant-Nuclide Subgroup Method (PRNSM) is used to conduct resonance calculation in the local aspect for each 1D cylindrical pin. Global-local self-shielding calculation scheme has been successfully implemented in high-fidelity numerical nuclear reactor physics code NECP-X. Verification results of global-local self-shielding calculation scheme showed good accuracy for UO2 fuels. The maximum relative error of microscopic absorption cross sections (XSs) for 238U in resonance range was 1.5% compared with MCNP5 [1]. AIC control rods serve as strong absorbers in reactor. Strong self-shielding phenomenon occurs when AIC control rods are inserted. Analysis was performed to determine the effects of AIC control rods on the accuracy of global-local self-shielding calculation scheme and the sources of error. Evaluation results showed that the main part of error was introduced by NCM and radius searching. The relative errors were larger than 10% in several resonance groups. Therefore, a supercell model is proposed to couple with global-local self-shielding calculation scheme to treat resonance calculation for AIC control rods in this paper. Numerical results show that this model improves the accuracy of the global-local self-shielding calculation scheme. The relative errors of microscopic absorption XSs for AIC in most resonance groups were decreased to less than 2%.
Experimental Nusselt Number in Rod Bundles Cooled by Heavy-Liquid Metals
Authors: Pacio J. , Daubner M. , Wetzel T. , Di Piazza I. , Tarantino M. , Martelli D. , Angelucci M.
Published: October 2018

A detailed safety assessment of innovative Generation IV reactor designs with heavy-liquid metal coolants, such as lead and lead-bismuth eutectic (LBE), requires an evaluation of the maximum core temperature in several postulated scenarios.

Considering the complex geometry of fuel assemblies (FAs), and the low Prandtl number of the coolants, this flow scenario is challenging for the models used in numerical simulations, e.g. for relating the turbulent transport of momentum and heat. Thus, reliable experimental data are needed for validation.

In recent years, a series of comprehensive heat transfer tests in fuel pin bundle simulators was performed at ENEA (Italy) and KIT (Germany) in the framework of the European collaborative projects THINS and SEARCH. Both grid and wire spacer geometries are considered, in a wide range of operating conditions (temperature, flow velocity and power density) representative of the ALFRED and MYRRHA fuel assemblies, in natural and forced-convective flow regimes. Although different experimental approaches were followed by each group (e.g. thermocouple position and average data treatment), there is a relatively good agreement on results in the overlapping regions.

These experimental studies indicate that the mean Nusselt number is in well agreement with the predictions of empirical correlations developed for sodium systems. In particular, for wire-spaced FA, heat transfer results show values close to the Kazimi-Carelli correlation both for low and high flow rates at ENEA and KIT respectively. For grid-spaced FA, results are more in agreement with Ushakov correlation.

Furthermore, large temperature differences are measured by thermo-couples installed at selected rods and sub-channels. A discussion on the influences of the spacer design and bundle size is included. This wide comparison allows an overview of the research on the HLM cooled fuel assembly in Europe.
Analyzing Droplet Size Distributions Inside a Self-Priming Venturi Scrubber for Filtered Containment Venting Systems
Authors: Papadopoulos P. , Lind T. , Prasser H.-M.
Published: October 2018

After the accident in the Fukushima Daiichi nuclear power plant, the interest of adding Filtered Containment Venting Systems (FCVS) on existing nuclear power plants to prevent radioactive releases to the environment during a severe accident has increased. Wet scrubbers are one possible design element which can be part of an FCVS system. The efficiency of this scrubber type is thereby depending, among others, on the thermal-hydraulic characteristics inside the scrubber. The flow structure is mainly established by the design of the gas inlet nozzle. The venturi geometry is one of the nozzle types that can be found in nowadays FCVS. It acts in two different steps on the removal process of the contaminants in the gas stream. Downstream the suction opening in the throat of the venturi, droplets are formed by atomization of the liquid film. The droplets are contributing to the capture of aerosols and volatile gases from the mixture coming from the containment. Studies state that the majority of the contaminants is scrubbed within this misty flow regime. At the top of the venturi, the gas stream is injected into the pool. The pressure drop at the nozzle exit leads to the formation of smaller bubbles, thus increasing the interfacial area concentration in the pool. In this work, the flow inside a full-scale venturi scrubber has been optically analyzed using shadowgraphy with a high-speed camera. The venturi nozzle was installed in the TRISTAN facility at PSI which was originally designed to investigate the flow dynamics of a tube rupture inside a full-length scale steam generator tube bundle. The data analysis was focused on evaluating the droplet size distribution and the Sauter mean diameter under different gas flow rates and operation modes. The scrubber was operated in two different ways, submerged and unsubmerged. The aim was to include the effect on the droplet sizes of using the nozzle in a submerged operation mode.
Research on Nuclear Power Plant Safety Functional Requirements Analysis and Function Allocation
Authors: Ming Jia , Huan Huang , Xuegang Zhang
Published: October 2018

This paper researched the safety functional requirements analysis and the allocation of functions between man and machine for the nuclear power plant. The safety functional requirements are identified from accident handling needs and refined from system configuration consideration. Through the analysis of design conditions, some safety features were extracted to mitigate accidents. Then, components (e.g. pumps, valves, tanks) were determined to implement each of the safety features at the system design stage. At this stage, some implicit safety features, which could not be obtained directly from the accident analysis, were added, according to the specific conditions of system configuration and operation. Finally, after further judgement on possible inconsistency, a complete list of safety functions for the nuclear power plant was formed. As an illustration, this paper provided a list of safety functions related to the safety injection function, and a list of equipment for the safety injection system. Furthermore, these identified safety functions, were appropriately allocated between man and machine, to be performed either by system components automatically, or by operators locally or remotely from the control room, or under the cooperation of operators and system components. Seven factors were considered in the allocation: a) performance requirements; b) the capability or limits of man and machine; c) existing practices; d) operating experience; e) management requirement; f) technical feasibility; g) cost. The allocation of functions for the safety injection system was validated using a simulator.
Numerical Study on the Two-Phase Flow for a Gas/Liquid Metal Magnetohydrodynamic Generator
Authors: Liao Meng-Ran , Dai Chun-Hui , Ma Can , Liu Yong , Zhao Zheng-Xing , Liu Zhou-Yang
Published: October 2018

The gas/liquid metal magnetohydrodynamic generator (G/LM-MHD) with the mixture of gas and liquid metal as working fluids shows a promising future due to recent development of liquid metal cooled nuclear reactors. Previous efforts on the G/LM-MHD energy conversion systems have predicted a higher efficiency than traditional thermodynamics cycle. However, most of the earlier studies focus on the conception designs, feasibility analysis and preliminary experiments, while less attention paid on some specific problems such as the bubble phenomenon in the two-phase flow. Therefore, this paper deals with numerical study on the performance characteristics of the gas/liquid metal two-phase flow in an ideal Faraday-type MHD channel, of which the geometry structure is 30 × 30 × 80 mm cuboid segmentary style. The conductive mixture fluid is composed of nitrogen as the gas phase and gallium as the liquid phase (N2/Ga). The temperature at the channel inlet is about 600 K considering the heat transfer after the mixing chamber, while the inlet velocity is around 10 m/s and gas volumetric void fraction is 50%. The external magnetic field is assumed as 4 Tesla adopting the superconducting technology, which seems essential for MHD industrial practice. Then the simulation is accomplished using a modified two-phase mixture model considering the electromagnetic influence. The simulation results show that the distribution of temperature changes much weaker than pressure and velocity, which agrees with earlier one-dimension analysis. On the other hand, the results characterizes clearly the increase of the void fraction close to the electrodes, which can explain intuitively the decrease of the power-generating capacity. Besides, the power output is predicted to reach maximum 22.5 kW while the voltage is 1.2 V and the power density can be 312.5 MW/m3 which is far beyond traditional steam turbines. This study shows a promising future of the gas/liquid metal MHD generator for the small nuclear plants and power systems.
Discussion About Sodium-Concrete Reaction in Presence of Internal Heater
Authors: Kawaguchi Munemichi , Miyahara Shinya , Uno Masayoshi
Published: October 2018

Sodium-concrete reaction (SCR) is one of the important phenomena during severe accidents in sodium-cooled fast reactors (SFRs) owing to the generation of large sources of hydrogen and aerosols in the containment vessel [1–3]. In this study, SCR experiments were performed to investigate the chemical reaction beneath the internal heater (800 °C), which was used to simulate the obstacle and heating effect for SCR. Furthermore, the effects of the internal heater on the self-termination mechanism were discussed.

The internal heater on the concrete seemed to hinder the transport of Na into the concrete. Therefore, Na could react with the concrete at the periphery of the internal heater. The concrete ablation depth at the periphery was larger than under the internal heater. However, the Na concentration around the reaction front was about 30 wt.% despite the position of the internal heater. The Na concentration was similar to that of Na2SiO3, which was almost same as that in our past study [4–5]. It was found that the Na concentration condition was one of the dominant parameters for the self-termination of SCR, even in the presence of the internal heater.
Evaluation of the Kinetics of Molten Pool Stratification in Case of In-Vessel Melt Retention Strategy
Authors: Carénini L. , Fichot F.
Published: October 2018

The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and retain the core melt in the reactor pressure vessel. This type of Severe Accident Management (SAM) strategy has already been incorporated in the SAM guidance (SAMG) of several operating small size LWR (reactors below 500MWe, like VVER440) and is part of the SAMG strategies for some Gen III+ PWRs of higher power like the AP1000.

One of the main issues for the demonstration of the success of the IVR strategy lies in the evaluation of the transient heat fluxes applied by the corium pool along the vessel wall. Indeed, these transient heat fluxes, during the corium pool stratification evolution, are expected to be higher than the steady-state ones, in particular due to the concentration of the heat flux in the top metal layer when it is thin (so called focusing effect). Another issue appears when a heavy metal is initially formed and rises later to the top (inversion of stratification): in such a situation, the metal goes through the oxide phase and accumulates a significant superheat which is likely to produce a high transient heat flux. Thus, it is of primary importance to be able to evaluate the duration of these transient peaks in order to evaluate the minimal residual vessel thickness after such fast transient ablation and draw conclusions about the vessel integrity.

This paper first presents the phenomenology associated to the transient molten pool stratification and the model implemented in the severe accident integral code ASTEC (Accident Source Term Evaluation Code) to evaluate this kinetics. Then, evaluations are presented, based on a typical PWR reactor configuration. A sensitivity study is proposed to consider the impact of the main uncertainties on parameters which govern this kinetics. A particular focus is made on the physical phenomena driving the transient stratification of material layers in the corium pool and on the identification of critical situations with possible consequences in terms of vessel failure. The characteristic times of each individual process (chemistry, stratification, natural convection) are compared. In particular, the limiting cases of very fast chemistry or very slow chemistry are evaluated.

This work is performed in the frame of the European H2020 project IVMR (In-Vessel Melt Retention) coordinated by IRSN. This project has been launched in 2015 and gathers 27 organizations with, as main objective, the evaluation of feasibility of IVR strategy for LWR (PWR, VVER, BWR) of total power 1,000MWe or higher.
A Numerical Research of the Resistance Characteristics of the Bottom Nozzle in the Annular Fuel Assembly
Authors: Minghui Duan , Minfu Zhao
Published: October 2018

The upstream of the bottom nozzle in the annular fuel assembly is the coolant inlet, and the downstream is the fuel rod bundle, which includes both inner and outer coolant flow paths. Therefore, the resistance characteristics of the bottom nozzle are very complex. In order to obtain the nozzle local resistance characteristics, a numerical research has been performed to simulate the bottom nozzle in different flow rates and split ratios with a simplified geometric model by CFX. The calculation results indicated that the local resistance of the bottom nozzle in the annular fuel assembly existed self-modeling phenomenon. The critical Reynolds number wasn’t influenced by the split ratio. The local resistance coefficient of the bottom nozzle was only related to the split ratio. The research has provided an analysis method for the bottom nozzle resistance characteristics research.
A Simplified Numerical Benchmark for Pool-Type Sodium Fast Reactors
Authors: Radman Stefan , Fiorina Carlo , Mikityuk Konstantin , Pautz Andreas
Published: October 2018

The current work presents a simplified benchmark for a pool-type Sodium Fast Reactor based on a 2D (r-z) geometry for testing tightly-coupled spatial neutron transport, thermal-hydraulics and thermal-mechanics modeling. The new benchmark is motivated by development of the multi-physics OpenFOAM-based GeN-Foam code at the Laboratory for Reactor Physics and Systems Behaviour at the EPFL, Switzerland, and the FAST code system by the Advanced Nuclear Systems group at the Paul Scherrer Institut, Switzerland. Aiming at the improvement of modeling and code-to-code comparison, the benchmark could prove useful for developers of tightly-coupled multi-physics simulation tools for reactor analysis. The benchmark specification and the solutions obtained with GeN-Foam are presented and discussed in the paper.
Research on the Neutron Multiplicity Pulse Trains Computer Simulation
Authors: Li Sufen , Zhang Quanhu , Huo Yonggang , Zhou Man
Published: October 2018

Neutron multiplicity counting (NMC) is one of the most advanced Non-Destructive Assay (NDA) techniques. In this paper, the principle and method of generation of neutron pulse trains is studied. A model of 4π cube counter is established for 238Pu sample, the capture time of neutrons is gained by simulation code, then output the neutron pulse trains for post dispose from generated fissile initial-times. The result from post dispose of this neutron pulse trains show a small bias compared with input value, it’s an evidence for the validity of simulation in this paper.
Investigation on Thermal-Hydraulic Parameters and Instability Under Two Different Heating Conditions Based on RELAP5 Code
Authors: Xia Zhaoyang , Zhou Zhiwei , Peng Tianji
Published: October 2018

The subject of heat transfer behaviors of steam generator is of great interest for better heat transfer efficiency and safety in industry. On most previous study, the models of SG based on RELAP5 used heat structures as the power generation, which is different from heating by primary side. This paper investigates the distinctions of different heat transfer methods on the thermal-hydraulic parameters by RELAP5. Two different heat transfers are selected, case 1 is uniformly heated by heat structure and case 2 is heated by the primary side of SG. The results show that the onset position of critical heat flux for case 2 is ahead of that for case 1. Besides, the temperature of tube wall corresponding to position of critical heat flux increases abruptly in case 1, and the wall temperature in superheated region in case 1 increases faster than that in case 2 when the water becomes fully superheated vapor. Then instability distinctions for two heat transfer cases are studied in this investigation. It is detected the fluid flow heated by second method is more stable than the first one, and the system has a higher ultimate thermal power when density wave oscillation occurs for case 2.
NPP I&C Safety Assessment by Aggregation of Formal Techniques
Authors: Babeshko Eugene , Kharchenko Vyacheslav , Leontiiev Kostiantyn , Odarushchenko Oleg , Strjuk Oleksiy
Published: October 2018

Safety assessment of nuclear power plant instrumentation and control systems (NPP I&Cs) is a complicated and resource consuming process that is required be done so as to ensure the required safety level and comply to normative regulations. A lot of work have been performed in the field of application of different assessment methods and techniques, modifying them and using their combinations so as to provide unified approach in comprehensive safety assessment. Anyway, performed research have shown there are still challenges to overcome, including rationale and choice of the safety assessment method, verification of assessment results, choosing and applying techniques that support safety assessment process, especially in the nuclear field. In our work we present developed framework that aggregates the most appropriate safety assessment methods typically used for NPP I&Cs. Key features that this framework provides are the formal descriptions of all required input information for every safety assessment method, possible data flows between methods, possible output information for every method. Such representation allows to obtain possible paths required to get necessary indicators, analyze the possibility to verify them by application of different methods that provide same indicators etc. During safety assessment of NPP I&Cs it is very important to address software due to its crucial role in I&C safety assurance. Relevant standards like IEC 60880 [1] and IEC 62138 [2] provide requirements for software related activities and supporting processes in the software safety lifecycle of computer-based I&C systems of nuclear power plants performing functions of safety category A, B and C, as defined by IEC 61226 [3]. Requirements and frameworks provided by IEC 60880 and IEC 62138 for the nuclear application sector correspond to IEC 61508, part 3 [4]. These standards define several types of safety related software and specify particular requirements for each software type. So as to verify software and confirm correspondence to required safety level, different techniques are suggested in normative documents. We share our experience obtained during software failure modes and effect analysis (software FMEA) and software fault insertion (software FIT) processes into FPGA-based platform, NPP I&C systems based on that platform, and RPCT, integrated development environment used by RPC Radiy and end users to design user application logic, specify hardware configuration etc. We apply software FIT to outputs of RPCT, considering source code, configuration files and firmware files. Finally, we provide a case study of application the developed safety assessment framework and software FMEA/FIT practices during practical assessment of FPGA-based NPP I&C system.
Evaluation of Electromagnetic Fields From Wireless Technologies in a Nuclear Plant
Authors: Cappelli Mauro , Lopresto Vanni , Cecchi Riccardo , Marrocco Gaetano
Published: October 2018

The aim of this work is to show a preliminary investigation on the propagation of electromagnetic fields generated by wireless technologies inside a nuclear facility or power plant. First, a survey of currently proposed wireless technologies for nuclear facilities and plants has been carried out. Then, for selected scenarios, the electromagnetic field propagation has been studied by means of electromagnetic simulation tools, and the presence of the nuclear environment has been simulated by properly modeling environmental parameters and engineered barriers. The choice of the proper simulation techniques and tools is mandatory in order to simulate the effect of the realistic environment on the propagation. Accordingly, the feasibility of wireless technologies application at nuclear facilities can be assessed on the basis of results achieved from simulated scenarios. The goal is to analyze, for selected scenarios, possible issues due to the propagation of an electromagnetic field in presence of simplified barriers mimicking the real nuclear environment. This approach can provide indications on how to deploy potential benefits of wireless technologies in a nuclear environment, evaluating pros and cons of the investigated technologies.
Experimental Investigation and Model Analysis on Upper Plenum Entrainment in AP1000
Authors: Xiang Y. , Wu Y. W. , Wang Mingjun , Tian W. X. , Qiu S. Z. , Su G. H.
Published: October 2018

Upper plenum entrainment plays an important role for the coolant inventory in small break LOCA. When the mixture level in the pressure vessel is lower than the height of the hot leg, the coolant will still be entrained to the hot leg in the form of upper plenum entrainment with further decrease of the liquid level in the pressure vessel, posing threat to core uncovering and melting. ADETEL (Automatic Depressurization and Entrainment Test Loop) was scaled after the automatic depressurization lines and the pressure vessel in AP1000 to investigate the upper plenum entrainment. The main droplet entrainment mechanisms from the two-phase interface to the hot leg has been obtain, including soaring, splashing, fragmentation, hitting and fusion. The dimensionless analysis was applied obtain the significant dimensionless number based on the mass conservation and energy conservation. It can be found that the predict data was different from the experiment data with an error of ±50%.
The Effect of Initial Condition of Melt Jet on the Jet Breakup Phenomenon in the Subcooled Water Pool
Authors: Jung Woo Hyun , Park Hyun Sun , Moriyama Kiyofumi , Kim Moo Hwan
Published: October 2018

The melt jet breakup phenomenon in a pre-flooded reactor cavity during a severe accident is related to the debris bed coolability. It is important to predict the jet breakup length for the evaluation of the debris bed coolability.

A large volume of works on the jet breakup length have been performed. However, the consistency between experiments and correlations was difficult to achieve. Some data follow the Saito correlation (include Froude number in the correlation) and others follow the Epstein correlation (doesn’t include Froude number).

The separation of the jet breakup length correlation along the water subcooling was reported based on the experimental data using the low melting temperature materials in our previous works. Since the previous experiments show an unclear jet shape before entering the water pool which could be an uncertainty factor, a slide gate system for a clear jet shape was additionally installed. Experiments were conducted with the similar condition of previous work and different initial condition of melt jet. With a clear jet shape, the jet breakup length in the subcooled water show different tendency following the Saito correlation.

To figure out the effect of the entry condition of the melt jet, the jet diameter and the method of estimating the jet breakup length are revisited. Our previo0us experiments show large uncertainties on the jet diameter, leading to the large discrepancy of the dimensionless jet breakup length. Also, early broken jet core is reported in subcooled water cases.

Thus, the uncertain characteristics of the jet breakup length analysis is discussed in this paper including the jet diameter and the method to estimate the jet breakup length.
Development of a Standard for Fusion Needs: Example of Introduction of Eurofer in RCC-MRx
Authors: Muñoz Garcia J. E. , Pétesch C. , Lebarbé T. , Lamagnère P. , Lejeail Y.
Published: October 2018

The 2018 edition of the RCC-MRx Code [1] will be issued by the end of the 2018, in French and English versions by AFCEN (Association Française pour les règles de Conception et de Construction des Matériels des Chaudières Electro-nucléaires). This Code set up design and construction rules of research reactor components (coming from the RCC-MX 2008 code developed within the context of the Jules Horowitz Reactor project), and to components operating at high temperature and to the Vacuum Vessel of ITER (coming from the RCC-MR 2007).

The extension of the scope of the code to innovative systems such as fusion reactors leads to revisit the background of the code to define the requirements to introduce a new process or a new material.

The developed methodology has been applied to the introduction of the Fe–9%Cr–1%W–TaV steel (Eurofer), today in the Probationary Phase Rules of RCC-MRx. It was the first time to introduce a “new” material into the code, new in the sense of non-existing in any current standardization. This process, still in progress, highlights the need to have a minimum of information on the expectation of the code regarding the material data.

This paper describes the different steps of the introduction of the Eurofer in the RCC-MRx code as well as the tools developed to facilitate the process.
Influence of Xe-135 Dynamic Behavior on Core Operation Safety for a Molten Salt Reactor
Authors: Jian-hui Wu , Xiao-xiao Li , Ji-feng Hu , Jin-gen Chen , Cheng-gang Yu , Chun-yan Zou , Xiang-zhou Cai
Published: October 2018

The isotope Xe-135 has a large thermal neutron absorption cross section and is considered to be the most important fission product. A very small amount of such neutron poison may significantly affect the reactor reactivity since they will absorb the neutrons from chain reaction, therefore monitoring their atomic density variation during reactor operation is extremely important. In a molten reactor system, Xe-135 is entrained in the liquid fuel and continuously circulates through the core where the neutron irradiation functions and the external core where only nuclei decay occurs, at the same time, an off-gas removal system operates for online removing Xe-135 through helium bubbling. These unique features of MSR complicate the Xe-135 dynamic behaviors, and the calculation method applied in the solid fuel reactor system is not suitable. From this point, we firstly analytically deduce the nuclei evolution law of Xe-135 in the flowing salt with an off-gas removal system functioning. A study of Xe-135 dynamic behavior with the core power change, shutdown, helium bubbling failure and startup then is conducted, and several valuable conclusions are obtained for MSR design.
Thermal Stratification Modeling for Sodium-Cooled Fast Reactors: A Status Update
Authors: Morgan Sarah , Bilbao y Leon Sama , Bucknor Matthew , Anderson Mark , Baglietto Emilio , Schneider James , Weathered Matthew , Xu Liangyu
Published: October 2018

Thermal hydraulic behavior in the upper plenum of pool-type sodium-cooled fast reactors (SFRs) is a major concern, as many design challenges are concentrated in this region. As SFR designs aim for licensing and commercialization, it is important to accurately analyze and predict the thermal-hydraulic behavior in this region during accident scenarios, specifically thermal stratification.

Thermal stratification models are currently a major source of uncertainty in most system codes for all types of power plants. Most system codes, including SAS4A/SASSYS-1, a system level code developed by Argonne National Laboratory (Argonne), use very coarse meshes that cannot capture the complexities of the stratification phenomena. While the commonly employed lumped-volume based models for thermal stratification are able to run in a matter of seconds, they result in approximate results and can only handle simple cases. Other 2-D and 3-D methods, such as computational fluid dynamics (CFD) models, can analyze simple configurations with higher fidelity, but come with a relatively large computational expense. Finding a modeling solution that is both accurate and computationally efficient has proven difficult.

This paper provides details of a review and gap analysis of the various modeling approaches proposed to date and explores a path forward for future thermal stratification modeling efforts, with a focus on developing new models for the SAS4A/SASSYS-1 system code.
Partially Averaged Navier-Stokes Turbulence Modeling of Flow in 5x5 PWR Fuel Assembly With Spacer Grid
Authors: Busco Giacomo , Hassan Yassin A.
Published: October 2018

The highly turbulent flow inside a pressurized water reactor makes unpractical the use of scale resolving simulations, due to the large number of space and time turbulent structures. The high computational cost associated with typical large eddies simulations or direct numerical simulations techniques is unsuitable due to the large spatiotemporal resolution required. Partially averaged Navier-Stokes turbulence model is presented as bridging model between Reynolds averaged Navier-Stokes equations and direct numerical simulations. As filtered representation of the Navier-Stokes equations, the model is able to continuously shift its energy-based filter, inside the turbulence spectrum, being able to resolve the turbulent scales of interest. The choice of energy based cut-off filters gives the chance to directly impose the degree of needed resolution, where the most important large scales unsteadiness are resolved at minimal computational expenses.

The partially averaged Navier-Stokes modelling approach has been tested for a Reynolds number of 14,000, inside a 5 × 5 fuel bundle, with a single spacer grid and split-type mixing vanes. Four different filters have been tested, whose resolution ranged from Reynolds averaged Navier-Stokes and large eddy simulation. A comparison with large eddy simulation will be presented. The results show that the partially averaged Navier-Stokes modeling produces results comparable to those of large eddy simulation when the appropriate cut-off energy filter is chosen. The turbulence models results will be compared with the available particle image velocimetry experimental data.
Online Adversarial Learning of Reactor State
Authors: Li Yeni , Abdel-Khalik Hany S. , Bertino Elisa
Published: October 2018

This paper is in support of our recent efforts to designing intelligent defenses against false data injection attacks, where false data are injected in the raw data used to control the reactor. Adopting a game-model between the attacker and the defender, we focus here on how the attacker may estimate reactor state in order to inject an attack that can bypass normal reactor anomaly and outlier detection checks. This approach is essential to designing defensive strategies that can anticipate the attackers moves. More importantly, it is to alert the community that defensive methods based on approximate physics models could be bypassed by the attacker who can approximate the models in an online mode during a lie-in-wait period. For illustration, we employ a simplified point kinetics model and show how an attacker, once gaining access to the reactor raw data, i.e., instrumentation readings, can inject small perturbations to learn the reactor dynamic behavior. In our context, this equates to estimating the reactivity feedback coefficients, e.g., Doppler, Xenon poisoning, etc. We employ a non-parametric learning approach that employs alternating conditional estimation in conjunction with discrete Fourier transform and curve fitting techniques to estimate reactivity coefficients. An Iranian model of the Bushehr reactor is employed for demonstration. Results indicate that very accurate estimation of reactor state could be achieved using the proposed learning method.
Multi-Fluid Gas Turbine Components Scaling for a Generation IV Nuclear Power Plant Performance Simulation
Authors: Osigwe Emmanuel O. , Gad-Briggs Arnold , Nikolaidis Theoklis , Pilidis Pericles , Sampath Suresh
Published: October 2018

One major challenge to the accurate development of performance simulation tool for component-based nuclear power plant engine models is the difficulty in accessing component performance maps; hence, researchers or engineers often rely on estimation approach using various scaling techniques. This paper describes a multi-fluid scaling approach used to determine the component characteristics of a closed-cycle gas turbine plant from an existing component map with their design data, which can be applied for different working fluids as may be required in closed-cycle gas turbine operations to adapt data from one component map into a new component map. Each component operation is defined by an appropriate change of state equations which describes its thermodynamic behavior, thus, a consideration of the working fluid properties is of high relevance to the scaling approach. The multi-fluid scaling technique described in this paper was used to develop a computer simulation tool called GT-ACYSS, which can be valuable for analyzing the performance of closed-cycle gas turbine operations with different working fluids. This approach makes it easy to theoretically scale existing map using similar or different working fluids without carrying out a full experimental test or repeating the whole design and development process. The results of selected case studies show a reasonable agreement with available data.
Multi-Diversity for FPGA Platform Based NPP I&C Systems: New Possibilities and Assessment Technique
Authors: Kharchenko Vyacheslav , Kovalenko Andriy , Leontiiev Kostiantyn , Panarin Artem , Duzhy Vyacheslav
Published: October 2018

Diversity approach is used to decrease risk of common cause failure (CCF) of Nuclear Power Plant (NPP) Instrumentation and Control systems (I&Cs). Application of a multi-diversity, i.e. a few different types of version redundancy allows minimizing CCF risk. On the other side, implementation of diversity increases cost and complicates maintenance of multi-version I&Cs. Hence, it is important to find optimal solution according with criteria “required level of diversity (safety) / minimal cost and maintenance complexity. Modern FPGA technology creates additional possibilities to meet requirements of the standards (such as NUREG/CR-7007, IEEE Std 7-4.3.2-2016, IAEA SSR-2/1:2016, IAEA NP-T-3.17:2016 and others) by developing main and diverse subsystems on the basis of the same FPGA platform. Existing diversity normative base should be enhanced in three directions — scope, depth and rigor to provide more detailed description of possible applied techniques and tools for quantitative assessment.

The goals of the paper which overviews practical issues of diversity application are the following:

- present extended classification of diversity considering additional types of version redundancy for FPGA platform based I&Cs (logical processing equipment, life cycle, logic/algorithm etc.) in comparing to NUREG7007;

- describe the modified technique of diversity assessment taking into account three and more levels of diversity classification;

- illustrate and discuss variants of assurance of the required degree of diversity by use of the RadICS FPGA platform to develop main and diverse subsystems.

The classification is specified considering diversity of hardware and FPGA designs. In particular, diversity of hard logic and soft processors, interfaces and buses, self-diagnostics means and others are described and embedded into NUREG/CR-7007 classification.

The NUREG7007-based diversity assessment techniques supporting all stage of analyzing options are discussed, and algorithms for versions choice are described. This technique takes into account more detailed specification of diversity classification (for types, subtypes and sub-subtypes of diversity for logic diversity, logic processing equipment diversity and others) and options to evaluate weight coefficients.

Case study is based on description of two options of RadICS FPGA platform application to develop two-version NPP I&C, which meets standard requirements to diversity.
First Steps Towards the Development of a 3D Nuclear Fuel Behavior Solver With OpenFOAM
Authors: Scolaro Alessandro , Clifford Ivor , Fiorina Carlo , Pautz Andreas
Published: October 2018

A new 3D fuel behavior solver is currently under collaborative development at the Laboratory for Reactor Physics and Systems Behaviour of the École Polytechnique Fédérale de Lausanne and at the Paul Scherrer Institut. The long term objective is to enable a more accurate simulation of inherently 3D safety-relevant phenomena which affect the performance of the nuclear fuel. The current implementation is a coupled three-dimensional heat conduction and linear elastic small strain solver, which models the effects of burnup- and temperature dependent material properties, swelling, relocation and gap conductance. The near future developments will include the introduction of a smeared pellet cracking model and of material inleasticities, such as creep and plasticity. After an overview of the theoretical background, equations and models behind the solver, this work focuses on the recent preliminary verification and validation efforts. The radial temperature and stress profiles predicted by the solver for the case of an infinitely long rod are compared against their analytical solution, allowing the verification of the thermo-mechanics equations and of the gap heat transfer model. Then, an axisymmetric model is created for 4 rods belonging to the Halden assembly IFA-432. These models are used to predict the fuel centerline temperature during power ramps recorded at the beginning of life, when the fuel rod performance is still not affected by more complex high burnup effects. Finally, the predictions are compared with the experimental measurements coming from the IFPE database. This first preliminary results allow a careful validation of the temperature-dependent material properties and of the gap conductance models.
Cyclic Plasticity Behavior of 90° Back-to-Back Pipe Bends Under Cyclic Bending and Steady Pressure
Authors: Cho Nak-Kyun , Chen Haofeng
Published: October 2018

Back-to-back pipe bends are widely adopted applications in many industries including nuclear sectors. Evaluation of their load bearing capability under complex cyclic loading is very important. Recently, a couple of research reported shakedown boundary of a 90° back-to-back pipe bends by adopting a conservative approach but no comprehensive post yield structural behaviors have been dealt with. In this research the concerning pipe bends subjected to cyclic opening in-plane (IP)/out-of-plane (OP) bending and steady internal pressures are analyzed to construct shakedown and ratchet limit boundary by means of the Linear Matching Method. Analyzed results present that the concerning pipe bends under out-of-plane bending has higher resistance to cyclic bending than under in-plane bending. In additions, the out-of-plane bending causes very small alternating plasticity areas, unlike the in-plane bending. Full cyclic incremental analyses known as step-by-step analysis are performed to verify the structural responses either side of each boundary and confirm correct responses. Parametric studies are carried out with respect to changes in geometry of the concerning pipe bends subjected to the same loading, and semi-empirical equations are derived from relationships of the reverse plasticity limit and the limit pressure with the bend characteristic. This paper offers broad understandings of structural responses of the 90° back-to-back pipe bends under the complex cyclic loading as well as providing key points to be considered for the life assessment of the piping system.
Experimental Activities on Thermal Hydraulics of Heavy Liquid Metal Flow in Typical Elements of Nuclear Power Stations
Authors: Kashinsky O. , Kurdyumov A. , Lobanov P. , Svetonosov A. , Vorobyev M. , Pribaturin N.
Published: October 2018

The report presents the results of an experimental study of the thermal hydraulics of a heavy liquid-metal coolant in the elements of a reactor installation. Hydrodynamics and heat transfer are considered in the coolant flow in a model assembly of vertical rods and in a T-junction. These tasks are important both from the point of view of justifying the design solutions of a reactor plant with a heavy liquid-metal coolant, and from the point of view of verification of calculation codes. The results of an experimental study of the temperature distribution in the 7-rod experimental model of a fuel assembly with spacer grids are presented. To create a thermal heterogeneity, a heating zone was created in one of the rods. The evolution of the temperature field is studied with the variation of the mutual location of the heated, measuring rods and the spacer grid. Validation measurements of fluid velocity profiles and its pulsations in a similar flow in a similar geometry of a water coolant have also been carried out. Results of experimental study of mixing of liquid metal with different temperatures in a T-junction also are presented. The liquid with higher temperature was injected through the branch of the T-junction. Data on wall temperature evolution for the mixing of hot and cold liquids were obtained for different ratios of flow rates of main and side flows. Experimental data on fluctuational and spectral structure of temperature field in the mixing zone were obtained for different flow conditions. Results obtained may serve as a basis for construction of experimental data base for verification of CFD.