bluuebunny/crossref_metadata_2025 huggingface dataset:
A Neutron Transport Calculation Method for Deep Penetration and its Preliminary Verification
Authors: Yang Wankui , Yuan Baoxin , Zhang Songbao , Guo Haibing , Liu Yaoguang , Deng Li
Published: October 2018

Deep penetration problems exist widely in reactor applications, such as SPRR300 (Swimming Pool Research Reactor 300), a light water moderated, enriched uranium fueled research reactor in China. Deterministic transport theory is intrinsically suitable for deep penetration. But there exist some problems when it’s applied in SPRR-300research reactors. First, the reactor core is complicated for geometry description in deterministic theory codes. Monte Carlo method has advantages in complex geometry modeling. And it uses continuous energy cross sections which are independent with specific reactor types and research objections. But usually it’s difficult to converge well enough to deal with deep penetration problems, even though there are a number of variance reduction techniques. Based on the advantages and disadvantages of Monte Carlo and Deterministic method, we proposed a coupled neutron transport calculation method for deep penetration. It combines advantages of these two methods. Firstly, we use Monte Carlo code to finish fine modeling and do the whole reactor core calculation. Domestically developed Monte Carlo code JMCT is used to do the neutron transport calculation. Then homogenized group constants in each mesh are calculated from JMCT output by a self-developed script. Afterwards, we do the whole reactor calculation with deterministic theory code TORT. It directly uses group constants generated by Monte Carlo code. Finally, we can get the deep penetration calculation results from TORT output. Verification is carried out by comparing the group constants of benchmark problem, and by comparing keff calculated by this method with continuous energy Monte Carlo method. Benchmark calculation is conducted with OECD/NEA slab benchmark problem. The comparison shows that group constants generated by this study are in good agreement with results from published references. Then above group constants are applied to 3-dimensional discrete ordinates deterministic theory transport code TORT. But keff calculated by TORT is a little lower than that calculated by Monte Carlo code JMCT. To minimize other influence factors, different Sn/Pn order, and different mesh size in TORT has been tried. Unfortunately the keff difference between these two methods remains. Even though the keff results in this benchmark are less than keff calculated by continuous energy MC method, Benchmark results show that all the group constants generated by this method are in good agreement with existing references. So it can be expected that after further verification and validation, this coupled method can be effectively applied to the deep penetration problem in such kind of research reactors.
Numerical Research on Fuel Rod Progression During Core Degradation Process Using MELCOR
Authors: Feng Tangtao , Tian Wenxi , Song Ping , Wang Jun , Wang Mingjun , Su G. H. , Qiu Suizheng
Published: October 2018

Core degradation experiment will be conducted in Fuel Rod Melt Progression apparatus (FROMA) to investigate the distribution of mass and energy in designed fuel rod during a core degradation process. Numerical research on the core degradation experiment is mainly to evaluate the designed parameters of FROMA. The pre-numerical study was conducted using the widely accepted severe accident analysis software MELCOR. The fuel rods used in the experiment consisted of real reactor fuel elements, and the designed fuel rods were electrically heated with internal center tungsten-rhenium rod to a very high temperature. Numerical analysis model is described, and the input parameters are in accord with experimental conditions. In the transient period of core degradation, the production hydrogen and the distribution of mass and temperature are obtained. All the predicted MELCOR results will be compared with the experimental measurements.
Transient Thermal Behaviors of SBO Accident for a 200MW OFNP Under Heaving Motion Conditions
Authors: Yan Qiqi , Luo Simin , Zhang Yapei , Liu Limin , Su Guanghui , Qiu Suizheng
Published: October 2018

For some Pressurized Water Reactors (PWR) operated on automobiles, boats or deep sea vessels, system characteristics is important for understanding their safety during severe accidents. The development of an analysis code and the transient thermal beaviors of a floating nuclear reactor under heaving motion are described in this paper. By modifying the control equations based on the mathematical models of ocean conditions, an ocean condition available system analysis code named RELAP5/GR was developed from RELAP5 MOD3.2 to simulate the transient thermal-hydraulic response of the nuclear reactor systems to the motion conditions in accidents, which is an advanced and independent node programming code. Using the code, the analysis model was established for a small 200MW offshore floating nuclear plants (OFNP). The transient thermal behaviors of the whole system were analyzed in the cases of the station blackout accident under heaving motion conditons. The analysis shows that all the results can be reasonably explained and the code development is successful at this stage.
The Influence of Pebble Placement on the Wake of Tandem Pebbles in a Free Stream
Authors: Botha Gerrit , Hassan Yassin A. , Kurwitz Richard C. , Merzari Elia
Published: October 2018

The effect, the pebble spacing has, on the fluid flow past two spheres in a free stream, was investigated by utilizing Direct Numerical Simulation (DNS), a Spectral Element Method (SEM) and NEK5000. There was seen that for a spacing of 2 pebble diameters (D), between the centroids, the turbulence statistics were asymmetric. For this study four different pebble spacings were investigated: 1.5D, 2.0D, 2.5LD, 3.0D. For all the cases the inlet flow, (U0) was selected that the Reynolds number, (Re) equalled 1000. The flow was allowed to develop for 200 Convective Units (CU = D/U0) before tests were conducted to ensure compliance to DNS. With the Kolmogorov scales captured and adequate Y+ values, time averaging was done for 800 CU to calculate the turbulence statistics. Calculated parameters include turbulence statists for selected line profiles and turbulent kinetic energy (TKE) budget terms. The generated statistics will be compared for each spacing selection.
TREAT Transient Modeling and Impact of Graphite Thermal Scattering
Authors: Sorrell Nina C. , Hawari Ayman I.
Published: October 2018

The Transient Reactor Test Facility (TREAT) is a high enriched, graphite moderated, air cooled reactor built for experimental transient fuel testing. Recently, the reactor was returned to operation after having been shut down since 1994. Transients at TREAT are controlled largely by control/transient rod movement and temperature feedback that is attributed to the core’s graphite-fuel matrix. To date, TREAT simulations use the standard ENDF/B-VII.1 graphite thermal neutron scattering cross sections that assume an ideal crystalline form for the core’s graphite. Historically, it has been reported that the use of these cross sections may result in a −2000 pcm discrepancy when attempting to predict TREAT criticality [1]. In this work, a multi-physics simulation of a TREAT transient is performed using the standard ENDF/B-VII.1 graphite thermal scattering cross section libraries and compared with results using graphite libraries that assume a porous graphite structure and a corresponding density consistent with TREAT graphite. The transient simulation methodology couples a full-core transient Monte Carlo calculation in the Serpent code with feedback calculated from temperature estimates derived using the computational fluid dynamics code OpenFOAM. Steady state simulations show that use of the “porous” graphite libraries allows predicting TREAT criticality to within a few hundred pcm. In the current transient simulations, the reactor’s time dependent power behavior is successfully reproduced. With this model, observables such as maximum fuel temperatures and temperature-dependent flux spectra are calculated, using both the traditional ENDF/B-VII.1 and the “porous” graphite thermal scattering libraries.
The EUR Assessment Process and Highlights of the Compliance Analysis for the EU-APR Standard Design
Authors: Declercq Cedric , Ballard Andrew , Ferraro Giovanni , Touré Anicet
Published: October 2018

The EU-APR standard design is a two-loop pressurised water reactor of the range of 1400MWe, developed by Korean Hydro & Nuclear Power (KHNP) from the Korean APR1400 reference plant. It is the objective of the EU-APR to comply with the current main European regulatory and utility requirements for safe and competitive LWR NPPs. A full assessment of the EU-APR against the European Utility Requirements (EUR) can therefore be regarded as an important step in its introduction to the European market.

KHNP’s request to assess its EU-APR standard design was granted by the EUR Organisation and in autumn of 2015, a two-year assessment process started, during which the EU-APR was assessed against the twenty chapters of the EUR Document Volume 2, encompassing over four thousand individual requirements having in view principally NPP safety and performance.

The EU-APR standard design was assessed against each single requirement by experts of the appropriate discipline from within the EUR member companies. The resultant assessment reports were reviewed by the Coordination Group (working level team) and then validated by the EUR Administration Group and EUR Steering Committee. The syntheses of the requirement-by-requirement assessments per chapter were collected in the principal project output: the EU-APR dedicated subset of EUR Document Volume 3. This Volume 3 subset (Subset I) also contains a technical description of the plant, issued by KHNP and reviewed by the EUR organisation. Furthermore, it explains the assessment process and highlights the main assessment results. Volume 3 Subset I was approved by the EUR Steering Committee in September 2017.

The EU-APR standard design was found to be fully compliant for a majority (almost 90%) of all requirements applicable in this assessment. Furthermore, the EU-APR was found to be compliant with the objectives of an additional 6% of the investigated requirements. The EU-APR standard design can be considered to be in the late basic design phase. Nonetheless, the portion of requirements which were not assessable is very low (below 3%), explained by the abundant availability of reference plant (detailed design) documentation, which often helped demonstrate by means of extrapolatable examples KHNP’s adherence to design principles established for the EU-APR. In total, less than 1% of assessed requirements were found to be non-compliant. The main non-compliances relate to choices of design principles and analysis methods that are explicitly different from those stipulated by EUR, and to the relatively low level of development of reactor core and fuel studies to demonstrate operational nuclear plant manoeuvrability and fuel cycle optimisation in view of safety margins and economics.
Large Eddy Simulations of a Coolant Flow in Spacer Grid Fuel Assemblies With a Spectral Element Solver
Authors: Yuan Haomin , Makarashvili Vakhtang , Merzari Elia , Obabko Aleksandr , Yu Yiqi
Published: October 2018

In this study we used Nek5000, an open-source, high-order spectral element CFD code developed at Argonne National Laboratory (ANL), to model the coolant flow in spacer grids. Two fuel assembly configurations were studied: 2 × 2 and 5 × 5 fuel rod arrangements. The simulations for the 2 × 2 case were based on previous studies, simulating one span of the 2 × 2 fuel rod configuration including a surrogate spacer grid and mixing vane design with typical features of spacers for energy production. Dual periodic boundary conditions were applied in the spanwise direction to take the crossflow into consideration. The study of the 5 × 5 fuel assembly was performed as part of the ANL–Framatome collaboration for advancing computational fluid dynamics (CFD) tools. An advanced numerical model was developed to simulate the experimental setup provided by Framatome. For the 5 × 5 fuel assembly study, two cases of flow geometry were simulated with Nek5000: balanced and unbalanced configurations. In the balanced flow the coolant was entering the fuel rod assembly through 121 uniformly spaced inlet holes arranged in an 11 × 11 matrix. The unbalanced case, on the other hand, featured 14 larger holes placed on only one side of the horizontal plane. Nek5000 accepts only hexahedral meshes, which bring a great challenge to the meshing process for a spacer grid fuel assembly. A tet-to-hex meshing strategy was applied to handle the complex geometric features. A tetrahedral mesh was created first, and then each tetrahedral element was converted into four hexahedral elements. Boundary layers were extruded to fit to the exact geometry. In order to account for transient flow characteristics, the large eddy simulation approach was applied in this study. The employed subgrid-scale model relies on explicit filtering, which has been proven valid for many engineering-scale simulations. We present here the simulation results obtained for both the 2 × 2 and 5 × 5 fuel assemblies.
Experimental Studies on the Thermal-Hydraulics of Dowtherm A Through the Pebble Bed With Internal Heat Generation
Authors: Liu Limin , Zhang Dalin , Li Linfeng , Yang Yichen , Wang Chenglong , Qiu Suizheng
Published: October 2018

The Fluoride-salt-cooled High temperature Reactors (FHRs) are an advanced concept using a novel combination of high-temperature coated-particle fuel, low-pressure fluoride-salt coolant and air-Brayton power conversion system. Prismatic fuel or pebble fuel are adopted for the conceptual core designs of FHRs like TMSR-SF, MK1 PB-FHR and SM-AHTR. The high-Prandtl-number FLiBe is mainly adopted as the primary coolant, which specifies in high melting and boiling point and high volumetric capacity. The experimental results obtained from the air, water or inert gas prove reliable for the Prandtl number vary from 0.7 to 7. Little experimental research has been conducted to prove applicability of the above results to the high-Prandtl fluid, fluoride salts in the packed pebble bed. In this paper, a pebble bed experimental facility has been designed and constructed for the FHRs to explore the thermal-hydraulic characteristics of fluoride salts in the reactor pebble bed core. Dowtherm A is adopted as a simulant fluid for the fluoride salts. The cylindrical test section is packed with steel pebbles. The electromagnetic induction heating system is used to provide internal heat source for the pebble beds. The forced flow and convective heat transfer of high-Prandtl-number fluid in the pebble bed with internal heat generation are investigated in the experiment. The fluid inlet temperature and mass flow rate are studied on the thermal-hydraulic characteristics.
Super Critical Carbon Dioxide Brayton Cycle Based Heat Removal System in Nuclear Power Plants
Authors: Shuhong Wei , Hua Zheng
Published: October 2018

Heat removal from the core and spent fuel is one of the fundamental safety functions. Mobile equipment for heat removal from the core and spent fuel is required after Fukushima accident, but there are various constraints for modification of current operating nuclear power plants, such as layout, especially when new equipment are needed inside the containment. New reactor designs emphasize passive safety systems, but most passive safety systems rely on large pool and the heat removal duration depends on water volume. Super critical carbon dioxide brayton cycle can work as a heat engine by itself without external power supply or water supply, and supply surplus electricity due to the difference between expansion work and compression work. Also, super critical carbon dioxide brayton cycle is small, can be designed as a modular, mobile system and has little effect to system configuration or layout of current operating nuclear power plants. Super critical carbon dioxide brayton cycle is a good choice for self-propelling or passive heat removal for nuclear power plant modifications or new reactor designs without difficult modification of system configuration or layout. Super critical carbon dioxide Brayton cycle based heat removal system in nuclear power plants is designed and its technical feasibility is analyzed.
Preliminary Research on the Oxidation Effect of the Carbon Steel Plate of Downward Facing Pool Boiling by Two-Dimensional Image
Authors: Wang Kai , Erkan Nejdet , Okamoto Koji
Published: October 2018

When considering the safety of the reactor after the core melted, the external reactor vessel cooling (IVR-ERVC) is regarded as one of the most prominent method and is now widely being studied. And in order to apply this method more efficiently, CHF is the utmost part because it limits the upper threshold of the cooling effect. There has already been a large number of investigation on the CHF both by experiment and simulation. But for the experiment, most of them used the copper for its high thermal conductivity. However, the lower plenum is actually composed of the carbon steel instead of copper since the reactor pressure vessel and some core catchers in nuclear power plants are made of this material.

This CHF experiment here used carbon steel plate on a downward face. The carbon steel plate was attached to the copper base by solder. The results of the carbon steel plate experiment were then analyzed. After polishing the surface by sandpapers, the heat flux is being increased until CHF was reached. We found that the surface is somehow changed during the heating process. This process was repeated several times, and the growing tendency of the CHF was found. Detailed images of the heating surface acquired by high speed camera under different heat fluxes were also obtained and analyzed. It was found that the more the oxidization was, the fewer bubbles were generated and the bigger the CHF was. Finally some theories of the CHF mechanism were also analyzed. The authors hope that study could shed some light on surface effect on causing different CHF.
Modification of RELAP/SCDAPSIM/MOD4.0 for Liquid Metal in Contact With Noncondensable Gas
Authors: Sun Qian , Peng Tianji , Zhou Zhiwei , Chen Zhibin , Wang Shisheng
Published: October 2018

The use of liquid metal alloy as a coolant in nuclear systems receives more and more attention in recent years, such as Dual-functional Lithium Lead Test Blanket Module (DFLL-TBM) for ITER, Lead-bismuth spallation target for accelerator driven sub-critical system (ADS) and Lead-alloy-cooled fast reactor (LFR). A system safety analysis code is an important tool for the liquid metal nuclear system safety analysis. In order to analyze some of the basis accidents, there is the need to simulate the mixing of liquid metal and noncondensable gas. While the current system safety code RELAP/SCDAPSIM/MOD4.0 which was initially designed to predict the behavior of light water reactor systems is incapable of modeling the mixture of liquid metal fluids and noncondensable gas. This paper first briefly introduce the two-fluid model in RELAP5/MOD4.0 and the reason for its incapability of modeling liquid metal in contact with a non-condensable gas. Then, a solution to solve the problem and the modification of the RELAP5/MOD4.0 code is proposed. Last, several typical problems in DFLL-TBM system were simulated and the results demonstrate the feasibility and validity of the modified RELAP5/MOD4.0 in modeling the mixing of liquid metal and non-condensable gas. The modified code provides a powerful tool for liquid metal nuclear system designs and safety analysis.
Application of Level 2 PSA in the Design of Cavity Injection System for Nuclear Power Plant
Authors: Zhu Wentao , Li Wenjing
Published: October 2018

After Fukushima nuclear power plant accident, severe accident is getting more and more concerns all over the world. In order to mitigate severe accident and improve the safety of nuclear power plant, two different strategies are applied in different plants. One is in-vessel melt retention strategy, and the other is ex-vessel melt retention strategy. Tianwan nuclear power plant is an improved Gen II nuclear power plant and in-vessel melt retention strategy is adopted in the plant. In order to achieve this strategy, cavity injection system is designed for the plant. Probabilistic Safety Analysis is the most commonly used quantitative risk assessment tool for decision-making in selecting the optimal design among alternative options. For this plant, in order to optimize the design of cavity injection system, improve the safety level of nuclear power plant, and meanwhile, improve the engineering implementation and economization, Level 2 PSA was used for this decision-making process. In this paper, the Level 2 PSA for this plant and the application for the design of cavity injection system are introduced.
Discussion on Safety Classification for Equipment of Nuclear Power Plants
Authors: Dan He , Yue Zhang
Published: October 2018

Nuclear power plant is composed of many structures, systems and components. In the design and development of nuclear power plants, in order to improve their safety, more and more designers realized that the standards and methods of design, manufacture and supervision of nuclear facilities and equipment have changed a lot. In the new period, facing the deteriorating environment, green, safety and environmental protection are all the things we have to pursue in our real life. Therefore, it is urgent to develop clean and efficient nuclear power sources and focus on improving the safety, reliability and economy of nuclear power plants. For extreme accidents that are extremely unlikely, the equipment of nuclear power plant should be kept safe and worthy of further exploration under the design extension conditions. In order to ensure nuclear safety, according to the practical experience at home and abroad and the practice of new nuclear power plant, it is proposed to improve the safety classification theory and method of nuclear power plant equipment. At the same time, combined with the in service inspection, equipment identification, quality assurance requirements, maintainability and technical maturity of nuclear power plant equipment, we comprehensively and deeply study the safety classification of nuclear power plant equipment.
Discussion of Issues for Small Modular Reactor PSA
Authors: Tao Liu
Published: October 2018

Small module reactors (SMRs) are in ascendant in the world due to their shorter design & construction duration and higher safety level. Probabilistic safety assessment (PSA) will play an irreplaceable role in small module reactor design demonstration and other application stages. A mature PSA methodology framework is mature and has successfully used in commercial nuclear power plant application. While several issues are worthy of discussion when PSA performed for SMRs project. Based on traditional PSA framework and its fundamental elements, potential technical issues put forward for discussion for SMRs project PSA, which will provide references for the combination of PSA and SMRs project.
Modelling the Neutronics of a Molten Salt Fast Reactor Using DYN3D-MG for the Investigation of the Application of Frozen Wall Technology
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. New results of the application of the nodal diffusion code, DYN3D-MG, to model the steady state operation of molten salt fast reactors based on circulating fluoride salt, are presented here. The cross-section set was prepared using the Monte Carlo particle transport code, SERPENT. Full core comparisons between SERPENT and DYN3D-MG demonstrate that the molten salt fast reactor steady state operation can be modelled by DYN3D-MG within the required accuracy. The work is an important initial step for the development of a coupled CFD/neutron physics code system. This code system will finally be applied for the investigation of an innovative material protection scheme to avoid the contact of the liquid salt with the structural material.
One Step Method for Multigroup Adjoint Neutron Flux Through Continuous Energy Monte Carlo Calculation
Authors: Shang Xiaotong , Shi Guanlin , Wang Kan
Published: October 2018

The adjoint neutron flux is vital in the analysis of reactor kinetics parameters and reactor transient events. Both deterministic and Monte Carlo methods have been developed for the adjoint neutron flux calculation on the basis of multi-group cross sections which may vary significantly among different types of reactors. The iterated fission probability (IFP) method is introduced to calculate the neutron importance which is able to represent the adjoint neutron flux in continuous energy problem and have been applied to the calculation of kinetics parameters. However, the adjoint neutron flux can’t be obtained directly and applied to both Monte Carlo transient event analysis and deterministic methods. In this study, a method based on IFP is studied and implemented in Monte Carlo code RMC. The multi-group adjont neutron flux can be obtained directly through the discretization of energy and space with the modification of fission neutrons through continuous energy Monte Carlo calculations. The obtained multi-group adjoint neutron flux can be used in both Monte Carlo transient analysis and deterministic methods.
Assessment of Turbulence Models Against Supercritical Hydrogen Flows in a Straight Tube
Authors: Wang Zhipeng , Ji Yu , Sun Jun , Shi Lei
Published: October 2018

Convective heat transfer to supercritical hydrogen under high heat flux is a complex phenomenon in nuclear thermal propulsion (NTP), which has been regarded as a promising technology for human space exploration in the following decades. In this paper, concentration is mainly focused on exploring convective heat transfer to supercritical hydrogen in a strongly heated straight tube numerically based on several turbulence models. Differences between the standard k-ε model and the sst k-ω model in terms of the simulation results like the wall temperature profile, the bulk parameters, the velocity magnitude distributions and the turbulence kinetic energy profile are discussed in the first part. And the approach is validated through comparisons with experimental data. In the second part, the effects of heat flux is under investigation, combined with four turbulence models where the RSM model and the V2F model are newly included. Generally, this work will contribute to the design and analysis of nuclear thermal propulsion system.
Evaluation of the Radiation Effects of Residents Living Around the NSRI Under the External Hazards
Authors: Motome Yuiko , Akiyama Yoshiya , Hiroyuki Murao
Published: October 2018

The NSRR (Nuclear Safety Research Reactor) is a research reactor of TRIGA-ACPR (Annular Core Pulse Reactor) type, located in the Nuclear Science Research Institute (NSRI). The NSRR facility has been utilized for fuel irradiation experiments to study the behaviors of nuclear fuels under reactivity initiated accident (RIA) conditions.

Under the new regulation standards after the Fukushima Daiichi accident, the research reactors are being regulated according to the risk of the facility. Graded approach is introduced in the regulation. In order to apply the graded approach, the radiation effects of residents living around the NSRI under the external hazards were evaluated and the level of the risk of the NSRR facility was investigating. This report is summarized for the result of the evaluation in case the safety functions were lost by the tornado, earthquake and following tsunami.

As the result, the risk is confirmed to be low, since the effective dose of the residents has been below 5 mSv per event due to the loss of the safety functions by the tornado, earthquake and following tsunami.
Large Eddy Simulation of 5-Tube Bundle Helical Coil Steam Generator Test Section
Authors: Yildiz Mustafa Alper , Merzari Elia , Hassan Yassin A.
Published: October 2018

Helical Coil Steam Generator (HCSG) is a specific type of shell and tube heat exchanger known for having a higher heat transfer coefficient than many other designs. For this reason, they are considered in Small Modular Reactor (SMR) and High Temperature Reactor (HTR) designs. Investigation of flow behavior in HTRs is important for better design. To investigate the flow on the shell side, five rod HCSG test section was built in Texas A&M University. Experiments were conducted at Reynolds number of approximately 9000.

The spectral element code NEK5000 was used to do high fidelity large-eddy simulation (LES) of the test section to validate against experimental results. In this paper preliminary simulation results were presented. Pressure drop in preliminary results with seventh and ninth order polynomial expansion was consistent with experimental data.
Design Optimization of PERCS in RELAP5 Using Parallel Processing and a Multi-Objective Non-Dominated Sorting Genetic Algorithm
Authors: Wilding Paul R. , Murray Nathan R. , Memmott Matthew J.
Published: October 2018

Multi-objective optimization is a powerful tool that has been successfully applied to many fields but has seen minimal use in the design and development of nuclear power plant systems. When applied to design, multi-objective optimization involves the manipulation of key design parameters in order to develop optimal designs. These design parameters include continuous and/or discrete variables and represent the physical design specifications. They are modified across a specific design space to accomplish a number of set objective functions, representing the goals for both system design and performance, which conflict and cannot be combined into a single objective function. In this paper, a non-dominated sorting genetic algorithm (NSGA) and parallel processing in Python 3 were used to optimize the design of the passive endothermic reaction cooling system (PERCS) model developed in RELAP5/MOD 3.3. This system has been proposed as a retrofit to currently-operating light water reactors (LWR) and is designed to remove decay heat from the reactor core via the endothermic decomposition of magnesium carbonate (MgCO3) and natural circulation of the reactor coolant. The PERCS design is currently a shell-and-tube heat exchanger, with the coolant flowing through the tube side and MgCO3 on the shell side. During a station blackout (SBO), the PERCS initially keeps the reactor core outlet temperature from exceeding 635 K and then reduces it to below 620 K for 30 days. The optimization of the PERCS was performed with three different objectives: (1) minimization of equipment costs, (2) minimization of deviation of the core outlet temperature during a SBO from its normal operation steady-state value, and (3) minimization of fractional consumption of MgCO3, a metric that is measurable and directly related to the operating time of the PERCS. The manipulated parameters of the optimization include the radius of the PERCS shell, the pitch, hydraulic diameter, thickness and length of the PERCS tubes, and the elevation of the PERCS with respect to the reactor core. The NSGA methodology works by creating a population of PERCS options with varying design parameters. Using the evolutionary concepts of selection, reproduction, mutation, and survival of the fittest, the NSGA method repeatedly generates new PERCS options and gets rid of less fit ones. In the end, the result was a Pareto front of PERCS designs, each thermodynamically viable and optimal with respect to the three objectives. The Pareto front of options as a whole represents the optimized trade-off between the objectives.
Modelling of X-Ray Radioscopy for Phase Topology Estimation During Corium Sodium Interaction
Authors: Singh Shifali , Cassiaut-Louis Nathalie , Journeau Christophe , Zabiégo Magali , Estre Nicolas , Tamagno Léonie
Published: October 2018

In case of a severe accident scenario in a Sodium cooled Fast Reactor (SFR) such as the ASTRID demonstrator, the fuel might melt and interact with the coolant i.e. liquid sodium. This molten Fuel Coolant Interaction (FCI) can generate an energetic vapor explosion that can jeopardize the reactor structures. The yield of the vapor explosion is strongly dependent on the local distribution of the fragmented melt with respect to the local vapor fractions. The medium is composed of three phases, i.e. corium, liquid sodium and vapor sodium. Thus, a study of the three phase distribution within the system is a key to understand the extent of the explosion.

PLINIUS-2, the future large mass experimental platform of CEA Cadarache will be dedicated to conducting experiments to understand the behavior of prototypic corium in case of severe accidents. In order to study these interactions, a high energy X-ray imaging system is being developed. This system consists of a 15 MeV Linear accelerator producing high energy X-rays with significantly high flux, which are attenuated as it passes through the highly dense test section. The transmitted radiation is detected and re-emitted as visible light by the GADOX screen coupled to the CMOS camera. Using this system to study the interaction between corium and sodium is particularly challenging due to the small corium particulates of the size of the order of 1 mm. The qualification of the foreseen radioscopy system on the visualization of such an interaction requires the development of physical phantoms. This paper presents the preliminary simulations of expected images of corium fragments in sodium, vapor bubbles and vapor film around the fragments. The simulations are carried out using a CEA Cadarache in-house tool MODHERATO, which produces radiographic images in satisfactory agreement with the real time imaging. The simulation of particles is based on the knowledge of interaction phenomenology gained from past experiments and on the statistical analysis of the size of corium particles formed. The models which, according to MODHERATO results, qualify to be detected and resolved, help manufacturing physical phantoms to conduct the experiments.
Preliminary Validation of the Detached Eddy Simulation Model in CFD Code GASFLOW-MPI
Authors: Han Zhang , Yabing Li , Jianjun Xiao , Jordan Thomas
Published: October 2018

With the development of the computational capability, Computational Fluid Dynamics (CFD) is used more and more for design and safety issues related to reactor thermal hydraulics. A series of CFD codes have been developed and utilized to analyze the reactor thermal hydraulics behavior. GASFLOW-MPI is a well-validated parallel scalable CFD software and has been widely utilized to predict the complex physical phenomena in nuclear power plant containment. Furthermore, the simulation results are well accepted by the nuclear authorities in several countries all over the world. Turbulent mixing is a basic but important issue for thermal hydraulics safety analysis. In order to capture more unsteady turbulent information under the acceptable computational cost, the LES and RANS hybrid turbulent model Detached Eddy Simulation (DES) model has been developed in the CFD code GASFLOW-MPI. In this paper, two typical turbulent cases are simulated and compared with the experimental data to validate the DES model. The first case is the turbulent backward-facing step flow which is a well-known turbulent benchmark. The second case is a hydrogen turbulent jet which is a typical turbulent flow for hydrogen combustion risk evaluation in nuclear containment. The numerical results show that the averaged flow profiles agree well with the experimental data.
Cyber Threat Scenarios for Electrical Systems of Nuclear Power Plants
Authors: Gupta Deeksha , Bajramovic Edita , Parekh Mithil , Waedt Karl
Published: October 2018

The function of Nuclear Power Plants (NPPs) is to produce electricity. However, in order to do that electrical power is also required for the house load for operation and to keep the safety of the plant intact. The Electrical Power Systems (EPSs) are specially designed both for normal plant operation and for particular conditions other than normal operation. In this way, NPP safety is maintained and continuousness of electrical power supplies is available despite temporary interruptions or flows through regular operation and post-shutdown. The power for a plant comes from different and reliable power sources that are physically and electrically isolated. The overall EPSs design intends to assure that any single failure or malfunction will affect only a single source of power supply and the other alternative sources should not be affected. Despite the separation between electrical power sources, malicious cyberattacks [IAEA NSS 8] might disrupt the power flow and interrupt the normal operation of a NPP [IAEA NSS 17]. Therefore, to intact the nuclear safety and security of a nuclear power plant, it is very important to incorporate cybersecurity considerations into the EPS design and implementation. Particularly, threat scenarios for EPS should be developed and evaluated before a cyberattack occurs. This will help to identify possible vulnerabilities, their impact and consequences that we want to prevent. The classification of attacks and the modelling of attacks on critical infrastructure are already described in [1] as part of our SMARTEST R&D project. Here, in this paper, we are describing examples of cyber threat scenarios for the EPSs and EPS interfaces. These are necessary with regard to specifying the anticipated security impact to a physical process controlled by the Instrumentation and Control (I&C). This serves as an input for the security analysis and the closed loop virtual validation via simulation and Fault Tree Analysis (FTA).
Axial Flux Wire Measurements at the McMaster Nuclear Reactor
Authors: MacConnachie E. L. , Novog D. R. , Day S. E.
Published: October 2018

In a 2008 report on safety analysis for research reactors, the International Atomic Energy Agency (IAEA) identified experimentation as the preferred method of code validation [1]. However, many experiments currently used for code validation are performed under conditions that are not representative of real nuclear systems. Furthermore, the predominant uncertainties reported for reactor systems parameters are typically those associated with evaluated nuclear data libraries however, the significance of spatial uncertainties remains generally unknown. The magnitude of local flux measurement experimental uncertainties have not be investigated at length in the McMaster Nuclear Reactor (MNR). Such results can be used for validation of MNR models with both Monte Carlo N Particle (MCNP) and Serpent code packages. Flux wire measurements have previously been conducted at the center of an irradiation site (via the technique of neutron activation analysis), where a locally uniform flux distribution has been assumed. Early stage results show good agreement with three-dimensional neutron diffusion theory and demonstrate the viability of such measurements for continued analysis. However, the magnitude of the effects of Xenon buildup, control rod positions, and spatial sample positioning on the data remain unknown, and so a series of experiments is ongoing to address these areas of experimental variability. Full length flux wire irradiations at several high-power levels (500 kW, 800 kW, and 1 MW) are being conducted to quantify these effects. At each operating power level, several NiCr wires are irradiated, and the decay of 51Cr examined to determine the total neutron flux in the irradiation site. The use of multiple wires per irradiation provides insight into the spatial gradient of the neutron flux across one reactor site (approximately 8 × 8 cm).
Evaluation of the Mixing Vanes Effect on the Hydrogen Diffusion and Hydride Formation in the Fuel Cladding
Authors: Aly Ahmed , Petrov Victor , Avramova Maria , Manera Annalisa , Ivanov Kostadin
Published: October 2018

The fuel cladding is an important barrier to the release of fission products to the environment. Its integrity must be conserved during the in-reactor lifetime and during the spent fuel pool and dry cask storage. The corrosive interaction between the cladding and the water coolant in light water reactors leads to the oxidation of the zirconium-based cladding. A fraction of the hydrogen released due to those corrosive interactions or the radiolysis of the water coolant is picked-up by the fuel cladding. It diffuses inside the cladding driven by the concentration and temperature gradients. Eventually, its concentration can increase beyond a certain limit above which hydrogen precipitates as hydrides. The formation of hydrides can embrittle the cladding and leads to micro-cracks that can compromise the cladding integrity.

At the spacer grids locations, the mixing vanes will create swirl flow and mixing of the coolant leading to a high temperature gradient on the fuel rod cladding. This temperature gradient is a strong driving force for hydrogen to diffuse from high to low temperature locations. Therefore, the hydrogen behavior around the spacer grids with mixing vanes is important to model. In this work, the computational fluid dynamics code START-CCM+ is used to model the effect of the mixing vanes on the temperature profile on the cladding outer surface. It ws coupled with the transport code MPACT and the fuel performance code BISON. The computational model consisted of a 5 × 5 fuel rods subassembly with a guide tube in the central location. The obtained cladding temperature profile on a fuel rod of interest was applied as a boundary condition to BISON to model the hydrogen behavior around the spacer grids in a three-dimensional manner. Three spacer grids were modeled at elevations of 217.9 cm, 270.14 cm and 322.35 cm. The hydrogen behavior at each of those locations is evaluated and compared to assess the importance order of those locations.
Assessment of Control Room Radiological Habitability of High-Temperature Reactor Pebble-Bed Module in Shidao Bay Multi-Reactor Nuclear Power Site
Authors: Li Xinpeng , Fang Sheng
Published: October 2018

The control room radiological habitability (CRRH) is important for staff safety in a nuclear power plant, which is also a licensing requirement of the High-temperature Reactor Pebble-bed Module (HTR-PM) in China. Meanwhile, the complexity of the dose assessment increases for the multi-reactor site, which put forward higher requirements for building layout. The CRRH is investigated comprehensively for the multi-reactor site at Shidao Bay in this study. For a large-break loss of coolant accident of HTR-PM and CAP1000 in Shidao Bay nuclear power site, this study estimates doses of body, thyroid and skin due to three exposure pathways using NRC-recommended ARCON96 and dose calculation method in RG 1.195. To perform a realistic evaluation, the latest design and site-specific information are utilized as the input parameters, including the unique accidental source term of HTR-PM and the RG1.183-recommended source term of CAP1000, the release and ventilation parameters, the final layout and the meteorological data in a whole year. The evaluation results demonstrate that the individual dose level of staff in the control room is far below the requirement of the regulatory guide, which guarantees the CRRH of HTR-PM. The contribution of primary radionuclides suggests that tellurium and iodine are primary contributors of the inhalation dose of body and thyroid, which is worthy of paying particular attention to the CRRH design in HTR-PM.
Measuring Effects of Radiation on Precipitates in Aluminum 7075-T6 Using Differential Scanning Calorimetry
Authors: Connick Rachel C. , Hirst Charles A. , Cao Penghui , So Kangpyo , Kemp R. Scott , Short Michael P.
Published: October 2018

Radiation damage in structural materials for nuclear applications is not well-understood, especially when linking the atomic scale damage mechanisms to the macroscopic effects. On a microscopic level, particle radiation creates defects that can accumulate in the material. Defects can also interact with existing features in the material. Since both defects and features have different energies associated with them, investigation of the resulting energy spectrum in a macroscopic sample may offer insight into the connection between microscopic damage and macroscopic properties.

In alloys, changes in the size and number of precipitates will be reflected in the amount of energy required to dissolve the precipitates during thermal analysis. This can then be studied using differential scanning calorimetry (DSC). This work explores the sensitivity of the DSC measurement to detect irradiation-induced instability in metastable and secondary phase precipitates in the high-strength aluminum alloy 7075-T6 for extremely low doses of helium-ion and neutron irradiation. The precipitates in aluminum 7075-T6 are expected to grow or shrink, changing the energy spectrum measured by DSC. The magnitude of the change can then be compared to a model of irradiation-induced phase instability. This will demonstrate the ability of this thermal analysis technique to help bridge the gap between microscopic radiation effects and macroscopic properties.
Evaluation of Turbulence Modeling Approaches for Cross-Flow in a Helical Tube Bundle
Authors: Kraus Adam R. , Yuan Haomin , Merzari Elia
Published: October 2018

Helical steam generators are proposed for use in a number of advanced nuclear reactor designs. The cross-flow around the helical tubes is a complex flow-field, and accurate knowledge of this flow is necessary for estimating pressure drop, heat transfer, and risk of flow-induced vibration. However, legacy data for helical tube cross-flow are scarce, and building new large-scale experiments that investigate relevant phenomena can be costly. Thus large uncertainties must currently be taken into account in the design of these systems. Numerical modeling with CFD can provide improved insight into the flow phenomena to reduce this uncertainty, but choosing a methodology can prove difficult. LES methods provide high-fidelity data, but require immense computational time to perform even an investigatory calculation for a moderate-sized sector, let alone for many design iterations. URANS methods offer significantly lower computational time, but it can be difficult to confidently justify the accuracy of a particular model without validation, particularly given the highly three-dimensional and complex flow-field present here. To better establish a basis for URANS turbulence modeling, an LES simulation was performed using Nek5000, a massively-parallel spectral element code developed at Argonne National Laboratory, for the geometry of a legacy helical tube bundle experiment. Data from this high-fidelity LES simulation were compared with URANS simulations using a number of turbulence models with the commercial code STAR-CCM+. Turbulent kinetic energy in the flow channels as well as bundle pressure drop were compared. The values of these key parameters were found to vary significantly between different turbulence models, with some models predicting pressure drops and kinetic energies well below those seen in LES. Some models were identified that showed good potential for predicting helical tube bundle flow phenomena. Further work, at a wider range of flow velocities, will be useful to further solidify the range of applicability of these models.