bluuebunny/crossref_metadata_2025 huggingface dataset:
Results of a LES Application to LBE Turbulent Flow in a Wire-Wrapped Single Rod Channel
Authors: Pucciarelli Andrea
Published: October 2021

Innovative Liquid Metal Fast Reactors are among the most promising concepts for the upcoming Nuclear Gen IV; nevertheless, several challenges still have to be overcome before achieving a sufficiently mature understanding of all the involved phenomena. Concerning Thermal-Hydraulics aspects, the correct estimation of the turbulent heat flux contributions and their impact on the predicted heat transfer is still being investigated. Particularly, this becomes relevant when addressing complex geometries, such as the wire-wrapped rod bundles considered for the reactor core, in which simple assumptions may no more be sufficient for a suitable representation of the actual turbulent quantities’ distributions.

The present paper reports on the results of a LES application to a wire-wrapped single rod channel performed adopting the STAR-CCM+ code. The considered fluid is Lead-Bismuth Eutectic Alloy (LBE), often adopted as the working fluid in several experimental facilities built to support the development of Gen. IV LFRs while the considered Reynolds number is in the range of 8600 and the heated length is 0.1 m. The obtained results provide relevant information to be considered for the development and tuning of RANS turbulence models specialized in addressing operating conditions involving liquid-metals: the distributions of the turbulent heat fluxes (THF) is here particularly taken into account.
Codes and Methods Improvements for VVER Comprehensive Safety Assessment: The CAMIVVER H2020 Project
Authors: Verrier Denis , Vezzoni Barbara , Calgaro Barbara , Bernard Olivier , Previti Alberto , Lafaurie Clément , Hashymov Artur , Groudev Pavlin , Stefanova Antoaneta , Zaharieva Neli , Damian Frédéric , Mosca Pietro , Tomatis Daniele , Bieder Ulrich , Willien Adrien , Dos Santos Nicolas , Mercatali Luigi , Sanchez-Espinoza Victor Hugo , Forgione Nicola , Paci Sandro
Published: October 2021

CAMIVVER (Codes And Methods Improvements for VVER comprehensive safety assessment) is a three years project launched under the European Union research program HORIZON 2020. The main aim of the Project is to develop and improve computer codes and methods for VVER comprehensive safety assessment.

VVER reactors constitute a significant and dynamic part of the European energy market. Their safe Long Term Operation (maintenance, refuelling, safety-upgrade, revamping) relies on the industrial use of neutronics and thermal-hydraulics codes and methods that allow studying the behaviour of the plant in normal and accidental conditions.

Computer codes development is strongly required for the following reasons:

1) Current codes and methods used for VVER safety assessment are subjected to growing international export controls from outside EU, threatening the EU sovereignty and security in terms of energy supply.

2) A new generation of innovative codes and methods are under development within Europe. They are improving 3D-multiphysics modelling and uncertainty quantification capabilities and are worth being transferred from lab to industry as they will substantially improve the physics comprehension of PWR and VVER.

3) European codes and methods development for VVER safety assessment will open the VVER market to the European nuclear industry.

CAMIVVER will perform developments required for the new generation codes and will generalize 3D-multiphysics coupling, performing benchmark against current industrial codes used for PWR and VVER safety assessment. CAMIVVER will demonstrate CFD assets and compatibility with uncertainty propagation in the frame of a safety assessment.
Further Study on Real-Time Positioning of Fuel Failures Method in HFETR
Authors: Sun Shouhua , Li Jian
Published: October 2021

Research reactor’s fuel element integrity is essential for nuclear safety. In normal operations, radionuclides could be well contained in fuel element. Unfortunately, under severe conditions, the amount of radionuclides may increase, releasing to the primary coolant and, as a result, crack or damage the fuel element mechanically or chemically. Thus, it is imperative to help operators take measures effectively, locating damaged fuel assembly in reactor cores rapidly. In this paper, a quantitative fuel failure positioning method based on cesium (134Cs and 137Cs) detection was studied. An exact model to describe the relation between cesium radioactivity and fuel burnup based on nuclide conversion chain was put forward. Radioactivity measurement data has approved the validation of this method when it applied to HFETR. Moreover, based on the measurement of 24Na activity, we proposed an approach to identify fuel failures and cladding corrosion. The validation results further indicate the accuracy and good applicability of this method, that provides guarantee for the safe operation of HFETR.
Estimation of Long-Term Ex-Vessel Debris Cooling by Water in Fukushima Daiichi Nuclear Power Plant Unit-3
Authors: Sato Ikken , Yamaji Akifumi , Li Xin , Madokoro Hiroshi
Published: October 2021

Interpretation for the two-week long Unit 3 ex-vessel debris cooling behavior was conducted based on 1F plant and site data such as pressure, temperature, gamma ray level and live camera pictures. It was estimated that the debris relocated to the pedestal was in partial contact with liquid water for about initial two days. With the reduction of the sea water injection flowrate, the debris, both in-vessel and ex-vessel ones, became “dry”, in which the debris was only weakly cooled by vapor and this condition lasted for about four days until the increase of the sea water injection. During this dry period, in-vessel and ex-vessel debris were heat up and it took another four days to re-flood the heated up debris.
Preliminary Core Design of the Solid Moderator Reactor for Investigation of the In-Depth Europa Ice Layer
Authors: Fukizaki Shuta , Yamaji Akifumi , Fukuda Takanari
Published: October 2021

The use of different nuclear reactor systems as power source of deep-space explorers has been studied over the past decades as solar power cannot be expected in deep-space beyond Jupiter. In the preceding study, a cylindrical solid moderator reactor concept of about 500 kg in weight, which consists of 20% enriched UN fuel, YH1.5 moderator and Be reflector without using working fluid, was developed. In this study, we propose to further extend use of the developed core concept for the in-depth ice layer’s investigation of Europa (one of the moons of Jupiter). At the end of the journey, the bare reactor is to be landed on the ice layer of Europa and sink down through the ice-layer as it melts the ice layer with its thermal power. For this purpose, this study aims to flatten the core power distribution to increase the core surface temperature for efficient ice melting while keeping the peak temperature of the core below a design limit. The core neutronics characteristics and power distributions are evaluated with neutron diffusion approximation and ablation and sinking behavior of the reactor through the ice layer is analyzed with Moving Particle Semi-implicit (MPS) method. Among different designs, the Hollow core could reduce the radial power peaking relatively well. The trial analysis by MPS method showed that modeling convective heat loss of the core surface and / or modifications to the core design may be necessary to prevent excess heat-up of the core before it sufficiently melts the ice and sinks down into the ice layer.
Influence of Inlet Turbulent Flow Generated by Periodic Computations on the Pressure Drop and Axial Velocity Distribution Predictions
Authors: Qiu Chufa , Raverdy Bruno , Faucher Vincent , Bergeron André
Published: October 2021

This study presents the CFD modeling of the adiabatic rod bundle test (MANIVEL) with a Simple Support Grid (SSG) at a Reynolds number of 100 000. Calculations are carried out on grid using a one-block fully hexahedral mesh. To reduce the computational power requirements, the computational domain is limited to one span with one SSG. The inlet is a fully developed turbulent flow obtained using periodic computations that are applied on two different domains: an inlet box and a span box. The mean axial velocity distribution at different elevations and pressure drop are compared with measurements. Small discrepancies of the mean velocity distributions and of the pressure drop per span are observed between the two types of periodic inlet conditions. The influence of SSG is not negligible and is still present downstream. The application of periodic conditions generated on span box is closer to experimental data and so is more appropriate.
Affect Analysis of Surface Liquid Film Coverage on the Safety Performance of Containment
Authors: Shi Xingwei , Shaoxin Zhuang , Xinfang Cui , Wei Song , Jiaxu Zuo
Published: October 2021

Under the severeaccident condition, the containment is the last shielding for preventing the accidentalrelease of radioactive fission products to the atmosphere. To figure out the failure margin in accident conditions, it is necessary to research and analyze the effects of surface liquid film coverage on the containment safety. The Passive Containment Cooling System’s thermal-hydraulic model has been built with passive containment analysis code in Large Power Passive Reactor. Based on the reference conditions of Double Ended Guillotine (DEG) on a cold leg and Main Steam Line Break (MSLB), the sensitivity analysis of parameters related to surface liquid film coverage has been performed, the cases of the water distribution tube failure and the spray pipe pluggage added the failure of water distribution tube, and the scram under different power level have been carried out, respectively. Results: The analysis results show that: (1) the containment would fail in 24 hours after DEG accident start as a result of the liquid film coverage lower than 45%; (2)under the MSLB condition, the containment pressure is highest when scram power level is 30% of Fission Power (FP) under keeping the design liquid film coverage, and with the decrease of liquid film coverage, the containment is not failing as a result of overpressure, although the pressure is highest on the scram power level 0% FP.
Implementation and Validation of an Improved Interfacial Area Concentration Model for Two-Phase Flow CFD Simulations
Authors: Zhang Xiang , Peng Minjun , Cong Tenglong , Lu Chuan , Wang Chenyang
Published: October 2021

The interfacial area concentration (IAC) is an important parameter in the calculation of interfacial transfers in two-fluid model, which can affect the accuracy of the boiling simulations. In this paper, an improved IAC model based on drag force and drift velocity is obtained, which can make full use of the experimental data and the models of the drag force and the drift velocity to avoid the shortage of IAC algebraic model in two-phase flow simulations theoretically. The improved model is validated by the DEBORA boiling flow experiment data. The reasonable radial distributions of void fraction, liquid temperature and phase velocity can be obtained, which indicates that the improved IAC model coupled in boiling flow model can be applied in CFD simulation of two-phase boiling flow. The improved model provides a new calculation approach for the IAC in the boiling flow with multi flow regimes.
Study on Main Radionuclides of Liquid Waste in Containment Under Severe Accident
Authors: Wang Shaowei , Chen Haiying , Li Wei
Published: October 2021

After the Fukushima accident, the disposal of radioactive liquid waste has received great attention. The analysis of nuclide activity in the liquid waste of the containment after a severe accident is the prerequisite for the treatment of the liquid waste. Based on the source and reduction mechanism of radionuclides in the liquid waste of the containment after the severe accident, the composition change of nuclides in the liquid waste was studied. The results showed that during the release stage of nuclides from core to containment, the activity of each group nuclides in the liquid waste gradually increased. Then the activity of each group nuclides gradually decreased with time except for the lanthanum group nuclides. The total iodine activity decreased the fastest, and the total activity of the cesium group decreased the slowest. At the beginning of the accident, iodine was the main nuclide in the liquid waste, followed by cesium group. The percentage of cesium group in the total activity decreased first and then increased, while that of other groups first increased and then decreased. 134Cs, 137Cs, 89Sr, 90Sr, 127Tem, 106Ru, 90Y, 144Ce, 241Pu decayed very slowly, which were the important components of the radioactive liquid waste.
Research on Multi-Objective Optimal Design of Plate Heat Exchanger in Nuclear Power Plant Cold Chain System
Authors: Zhao Weiguang , Zhou Xiuan , Shi Jiangwu , Yan Changqi , Wang Jianjun
Published: October 2021

In order to improve the scientificity and rationality of cold chain system design, the plate heat exchanger of a branch of the nuclear power plant cold chain system of nuclear power plants is taken as the research object. Based on the basic principles of optimization design, appropriate optimization calculation algorithms are developed and plate type heat exchangers are developed. The design program of the heat exchanger forms an optimal design analysis tool. Taking a plate heat exchanger as an example, taking its mass, flow resistance on the hot side and flow resistance on the cold side as the optimization objectives, a new hybrid genetic algorithm was used to carry out a multi-objective optimization case analysis. The calculation results show that there is a competitive relationship between the mass of the plate heat exchanger of the nuclear island cold chain system of the nuclear power plant and the flow resistance on the hot and cold sides, and the optimal solution cannot be obtained at the same time. There is a correlation between the flow resistance on the cold side and the flow resistance on the hot side. If the heat exchanger mass is reduced by 6.39%, the flow resistance of the hot side and cold side will increase by 6.55% and 7.71% respectively; if the flow resistance on the hot side and cold side are reduced by 32.22% and respectively 32.01%, its mass will increase by 2.16%; on the premise of keeping the mass of the heat exchanger basically unchanged, the flow resistance on the hot side of the heat exchanger is reduced by 23.84% and at the same time the flow resistance on the cold side is reduced by 19.77%, and the flow resistance on the cold side is reduced by the largest 23.35% and at the same time reduce the flow resistance on the hot side by 12.54%. Therefore, there is a lot of room for optimization in the design of nuclear island cold chain system of nuclear power plant and there are many optimization schemes. The research results can provide reference for the subsequent optimization of nuclear power plant cold chain system design and improve the scientific and economic efficiency of system design.
Sensitivity Analysis of Power Related Parameters in a Reactivity-Initiated Accident of a Molten Salt Reactor
Authors: Wang Chaoqun , Yang Qun , Wang Kai , Jiao Xiaowei , Wu Yanhua , He Zhaozhong
Published: October 2021

Molten salt reactor (MSR) is one of the promising candidate Generation IV reactor technologies for future consideration. The rapid power rise during the accident is usually a critically important threat for the safety of a reactor. In a MSR, the increasing of reactor structural temperature caused by power rise may damage the integrity of reactor pressure boundary. During the accident scenario simulation of a MSR, the trend of reactor power really depends on the initial power, power peaking factors, and fuel salt and moderator temperature reactivity coefficients. This work aims to identify how these parameters influence the consequences of the accident for a MSR. In this study, a reactivity-initiated accident (RIA) is considered as the initiating event, which is caused by the withdrawal of control rod during power operating. This accident will lead to a rapid increase of reactor power and core temperature. A local sensitivity analysis was utilized to analyze the power related parameters. And a wide variety of values were selected for each parameter as the assuming initial conditions of the analysis. Multiple thermal-hydraulic simulation runs through a modified RELAP5 code were performed to acquire the accident analysis results for sensitivity analysis. According to the results of sensitivity analysis: power related parameters have a low impact on the consequence of the reactivity-initiated accident under power condition of the molten salt reactor. And the consequence of the accident is more sensitive to the changes in power peaking factors than the other parameters.
Dynamic Characteristics Analysis of NuScale in Frequency Domain
Authors: Yang Jingrui , Ma Qian , Han Lingtong , Sun Peiwei
Published: October 2021

NuScale is a small nuclear reactor that relies on natural circulation. Its modular production and inherent safety can not only be used to generate electricity in some remote areas, but also provide energy for water desalination and regional heating. However, the dynamic characteristics of the NuScale are different from those of the traditional PWRs because of its passive characteristic. Therefore, it is necessary to study and analyze the system dynamic characteristics of NuScale.

The NuScale PWR model is established based on MATLAB&Simulink. It includes point-reactor kinetics model with six groups delayed neutrons, coolant system and steam generator system. The model is established based on the conservation equations of mass, energy and momentum. And the correctness of the model is verified by the comparison between the steady-state calculation results and the design parameters. Transient calculation results are verified by mechanism analysis.

To evaluate the dynamic characteristics of NuScale, the sine function changes with different frequencies in reactor reactivity and feedwater mass flow rate are introduced. The amplitude and phase responses of reactor power, secondary loop steam pressure, secondary loop mass flow rate, secondary loop steam temperature, and coolant average temperature are recorded. Then the Bode plot can be drawn with amplitude and phase responses in different frequencies. To evaluate the NuScale dynamic characteristics, frequency domain analysis is performed.
Experimental and Analytical Investigation on Local Damage To Reinforced Concrete Panels Subjected to Projectile Impact Part 1: Penetration Damage Mode due to Normal Impact
Authors: Kang Zuoyi , Okuda Yukihiko , Nishida Akemi , Tsubota Haruji , Li Yinsheng
Published: October 2021

Most empirical formulas have been proposed to quantitatively evaluate local damage to reinforced concrete (RC) structures caused by a rigid projectile impact. These formulas have been derived from impact tests performed to the target structure with a normal angle, while only a few impact tests involving soft projectile to the target structure have been studied. Recently, we conducted a series of impact tests to evaluate local damage to RC panels subjected to normal and oblique impact due to rigid and soft projectiles. The final goal of our study is to establish a new formula for evaluating local damage to RC structures caused by oblique impact based on experimental and analytical investigation. This paper summarizes the results of experimental and analytical investigation on penetration damage mode to RC panels subjected to normal projectile impact. Through the comparison between experimental and analytical results, the analytical method is validated.
Experimental Study of Flow Characteristics in Round Jet Flow Using Particle Image Velocimetry (PIV)
Authors: Wu Lei , Wang Jianjun
Published: October 2021

High-resolution particle image velocimetry (PIV) was used to study the flow characteristics in the nearfield of a turbulent round water jet issuing from a circular nozzle. Mean velocity, velocity gradient, turbulence intensity, Reynolds stress and other turbulence statistics are investigated over a range of jet Reynolds number Re (based on jet exit velocity and equivalent nozzle diameter) varying from 531 to 13500. The distribution of parameters of different Reynolds numbers and different height sections was studied. The velocity measurements reveal that all exit jet profiles in the near-field closely approximate a single peak shape, and they show strict self-similarity. Compare turbulence intensity, Reynolds stress with different Reynolds number jet, the self-similarity is shown along jet path when Reynolds number is low, and the self-similarity is shown after the turbulence is fully developed when Reynolds number is high. The shapes of Reynolds stress of high Reynolds number jet change from double peak to single peak along jet path, and the shapes are always single peak at low Reynolds numbers. This critical Reynolds number is about 2000. In addition, the turbulence statistics can be used to correct the coefficients of the turbulence equation and large eddy simulation.
Load Match-Oriented Coordinated Control for Modular High Temperature Gas-Cooled Reactor Based on Dynamic Matrix Control
Authors: Jiang Di , Dong Zhe , Li Bowen , Huang Xiaojin
Published: October 2021

To cope with the flexibility from both load side and supply side, nuclear power generation should provide flexible operation services to improve its economic competitiveness. The prerequisite of flexible operation is the real-time realization of the operation point which is usually achieved by unit coordinated control finding the setpoints of nuclear power, coolant flow and feedwater flow to meet various load demand and keep key parameters within reasonable limits. Modular high temperature gas-cooled reactor (MHTGR) is typically a small reactor and adopt adjustable helium flow, graphite and once through steam generator (OTSG) as coolant, moderator, heat exchanger, respectively. The thermal hydraulic characteristics of MHTGR are of significant difference compared with that of pressurized water reactor (PWR). As a result, the coordinated control design for MHTGR plant is quite different from the PWR. In this paper, the feedwater flow and control rods are solely used to regulate the steam temperature and nuclear power, respectively. Moreover, the so-called dynamic matrix control (DMC) then is utilized to realize the load match for MHTGR thermal power, where the setpoint of helium flow is regarded as manipulated variables and the modular thermal power is regarded as controlled variables to be optimized. The effectiveness of the proposed method is then tested and verified by a hardware-in-loop simulation through a commercial distributed control system (DCS).
PSA Analysis of Switch Port Disabled on DCS Layer 1
Authors: Luo Yang , Liu Deyi , Zou Shengjia , Cao Yong
Published: October 2021

This paper first introduces the background, purpose and implementation process of SDP, and then analyzes the SDP based on the disabled fault of the switch port of DCS layer 1 in nuclear power plant. This DCS failure caused 14 ports between the room level switch and the unit level switch on the first floor of DCS to be disabled, resulting in partial interruption of communication between the room level switch and the unit level switch, resulting in abnormal connection between KIC/BUP and the first floor, and all KIC control functions and BUP in order to prevent the fluctuation of working conditions from introducing new disturbance to the control system, the unit has been maintaining stable operation. During the whole event, the reactor was in a safe state, the reactor protection system was available, the three barriers were complete, and no radioactive material was released. The influence of DCS failure on core safety is analyzed by SDP, and the conclusion is given.
Migration Characteristics of Sub-Source Items in Breach Occurrences
Authors: Peng Xu , Ruifeng Tian
Published: October 2021

In the loss of coolant accident in a PWR nuclear power plant, the loss of coolant will cause the core to melt, resulting in the release of large amounts of radioactive fission products from the core[1]. Therefore, nuclear power plants must have effective protective measures to prevent the release of radioactive fission products to protect workers, the surrounding population, and the environment from radiological hazards. Most of the current scholars have conducted detailed studies using modular software such as MELCOR[2], but the three-dimensional distribution of the source material in the breach space cannot be visualized using this method.

In this paper, a typical pressurized water reactor nuclear power plant is taken as the research object. CFD method is used to analyze the condition of coolant loss accident, study the force of radioactive aerosols matter from different heights away from the wall under the accident condition, and consider the influence of ventilation system on the distribution and migration characteristics of the source matter in space.

The results show that the migration behavior of particles is mainly affected by gravity and drag force. For particles with small size, the combined action of gravity and thermophilic force causes them to deposit at the wall surface, while for particles with large size, gravity causes them to deposit at the wall surface. The particle deposition on the wall surface can be effectively reduced by increasing the flow velocity and decreasing the fluid temperature at the breach.
Application of Finite Difference Jacobian Based Newton-Krylov Method for Coupled Neutronics/Thermal Conduction Problems of Nuclear Reactor
Authors: Liu Baokun , Wu Yingjie , Zhang Han , Guo Jiong , Li Fu
Published: October 2021

The design and safety of nuclear reactor are strongly dependent on the neutronics and thermal conduction. The interaction between these coupled issues forms multi-physics nonlinear system, which has limited the economy of the nuclear power plant because of the complexity. And the accurate prediction of the reactor behavior still remains a challenge for the nuclear engineering. Traditionally, the neutronics and thermal conduction are solved separately by operator-splitting, which may be unstable and introduce significant errors. In recent years, fully coupled Newton-Krylov methods are prefer to be adopted due to these methods are more robust, accurate and efficient. However, each Newton-Krylov method has merits and demerits. In this study, finite difference Jacobian based Newton-Krylov (DJNK) and widely used Jacobian-free Newton-Krylov (JFNK) method are employed to solve the coupled neutronics/thermal conduction problems of a high-temperature gas-cooled reactor (HTGR). A steady state at thermal power of 250 MW and a transient state of supercritical accident are simulated. The results show that, DJNK is more efficient compared with JFNK, because DJNK performs better in preconditioning, which is the key to successful application of Newton-Krylov methods. And different preconditioning methods are utilized to reduce the number of Krylov iterations. A speedup ratio of DJNK over JFNK is observed to be 3. Synthesis results indicate that compared with JFNK, DJNK is a more potential method, contributing to a deeper understanding of the reactor behavior in a more efficient way.
General Discussion on Terminal Velocity for Rising Single Bubble
Authors: Wang Qinghua , Kawara Zensaku , Yokomine Takehiko , Kunugi Tomoaki
Published: October 2021

Distinct from solid particle, bubble is soft matter, which tends to adjust its shape according to motion, and according to the work of Tomiyama, for bubble with diameter larger than 1.3mm, the rising path would change from rectilinear to zig-zag or helical, which make the determination of the terminal velocity more difficult. Even for the bubble with diameter larger than 0.075 mm reported by Pawliszak, terminal velocity scatter started.

The existence of surfactants or contaminants and different initial deformations of bubbles introduced by varying capillary or nozzle diameters were generally considered to be the two main reason account for the terminal velocity scatter.

In this paper, the rising characteristics of small bubbles with 1 < Re < 100 were investigated. Especially the motion characteristic right after the departure. Comparisons were made between three bubbles with almost the same diameter (0.58 ± 0.008mm), the same aspect ratio (0.98 ± 0.01), it was found that the only differences lied on oscillation frequencies of aspect ratio. And the higher oscillation frequency of aspect ratio was, the higher steady-state velocity the bubble would achieve. Therefore, it was thought that the deformation is one of the ways for bubble to accelerate itself, i.e., bubbles accelerated itself through the oscillations of aspect ratio. The oscillation of aspect ratio is of high possibility to be one reason for velocity scatter.
Anomaly Detection of Electric Gate Valve Based on Multi-Kernel Support Vector Machine
Authors: Luo Jing , Wang Hang , Peng Minjun
Published: October 2021

Valve is an indispensable fluid control component in nuclear power system. Nuclear power station has a large number of gate valve equipment, which works under high temperature, high pressure, high radioactivity and other harsh conditions. In nuclear power plant accidents and economic losses, a considerable part of them are caused by valve failure. Aiming at the fault of electric gate valve, this paper proposes an anomaly detection method based on multi-kernel support vector machine. Firstly, the acoustic emission instrument is used to measure the fault state data and extract the fault features. Secondly, on the basis of classical support vector machine, multiple kernel function combinations are selected to decompose the model into convex optimization problems to realize the abnormal state detection of internal leakage fault of electric gate valve in nuclear power plant. The results show that, compared with the classical support vector machine method, the constructed support vector machine method based on multikernel learning has better effect and higher accuracy in anomaly detection of electric gate valve.
Comparative Study on Thermal Stress Analysis and Fatigue Curve in Stress and Fatigue Calculation of Nuclear Equipment
Authors: Shao Xuejiao , Xie Hai , Zhang Yixiong , Zhang Liping , Fu Xiaolong , Mi Xue , Li Hui
Published: October 2021

In the stress and fatigue evaluation of primary nuclear equipment, thermal stress calculation and fatigue curve have a great impact on the calculation results. There is a problem of stability in the step integral when calculating the thermal stress. The transient heat conduction changes rapidly near the initial temperature time point. It is critical to choose the right sub steps. In the nuclear code, the design fatigue curve in the air environment is given by applying a factor of 2 on stress or a factor of 20 on cycles. The factors are used to consider various influences such as data scatter, structure size, surface roughness and load. The factors of fatigue curves in ASME BPVC 2007 edition + 2009 addendum and French AFCEN’s RCC-M 2017 edition RPP 2 have changed. Due to the changes of the original test data and modificative factors, the new fatigue curve and the original fatigue curve are crossed. In code case N-761 of ASME, fatigue curves considering the influence of primary coolant environment are given under different strain rates.

In the calculation of thermal transient using commercial ANSYS software, the setting of time step has a great influence on the thermal stress. This paper compares and analyzes the influence of different calculation methods on the thermal stress at different positions of the model. The effect of different fatigue curves on fatigue life is also discussed considering different air design curve and different strain rate fatigue curve in water.

The results show that the fatigue results obtained by using the modified ASME or RCC-M air fatigue curve are smaller than those before the change when the usage factor is large to a certain extent. After considering the environmental fatigue curve, the fatigue results increase significantly. The accuracy of thermal stress calculation can be significantly improved by adding time points and refining time steps near the time points after the transient sharply changes, otherwise the influence of thermal stress on stress and fatigue results will be underestimated obviously.
Extending a Fluid-to-Fluid Similarity Rationale for Heat Transfer at Supercritical Pressure to R134a
Authors: Kassem Sara , Pucciarelli Andrea , Ambrosini Walter
Published: October 2021

In the frame of the development of Gen-IV supercritical water reactors (SCWRs), making use of the experimental data collected so far by different fluids in view of simulating the behavior of water is an important and very useful step. However, conducting experiments with simulant fluids poses the problem of establishing a clear rationale for inferring the expected water behavior from the information gathered with other fluids.

In literature, several attempts have been made to suggest dimensionless groups to be preserved in establishing a similarity. However, difficulties were encountered in getting appreciable success from these attempts and relevant reviews declared a generally unsatisfactory behavior of available theories.

Recently, some of the authors of this paper were able to demonstrate the good success of their proposal for a fluid-to-fluid similarity theory addressing mainly four relevant fluids being water, carbon dioxide, ammonia and R23. In the frame of the EU ECC-Smart Project, the refrigerant R134a will be used as a simulant fluid for performing experimental analyses in support to the achievement of the project objectives. Hence, the present paper is aimed to show the capabilities of the similarity theory when applied to R134a, proposing a rationale to be used in selecting the boundary conditions of experiments in order to make them really representative of the addressed water plant conditions.
Evaluations of TiO2 Deposition on Structure Surfaces and Water Radiolysis for the Corrosive Environment in a Reactor Pressure Vessel
Authors: Mawatari Takashi , Yamamoto Yasushi , Shibasaki Osamu , Hara Takahiro , Horayama Yusuke , Takagi Junichi
Published: October 2021

It is important to avoid a risk of stress corrosion cracking (SCC) of structure materials in nuclear power plants. One of the methods for preventing SCC is titanium-dioxide (TiO2) injection. TiO2 is a photo-catalyst, and injected TiO2 deposits on the surfaces of structure materials, and the deposited TiO2 can decrease the electrochemical corrosion potential (ECP) by irradiation of Cherenkov-light in the reactor. Generally, the ECP is an important indicator about susceptibility to SCC, and ECP-lowering makes a risk of SCC lower. Therefore, initiation and propagation of SCC will be mitigated by the TiO2 injection.

The evaluation model for corrosive environment had been already developed. That “conventional” model was consisted of some analysis such as water radiolysis, ECP and others. However, the model did not consider an effect of TiO2 deposition and was one-dimensional simple model. Therefore, to evaluate ECP distribution in a reactor after TiO2 injection, the “newly developed” model was build: analyses integration, highresolution and three-dimensional model.

In this paper, the TiO2 deposition and water radiolysis analyses integrated into flow analysis were contained. Both the deposition and the radiolysis depend on flow in the RPV, so that these analyses were built into the computational fluid dynamics (CFD) analysis model for CFD code. The target plant type for the model was 1100 MWe-class BWR-5.

The flow analysis was carried out at steady-state, assuming the TiO2 injection during the rated operation of the plant. Then the TiO2 deposition and the radiolysis analyses were conducted by transient calculations using the result of flow, and the three-dimensional and high resolution results were obtained.
Features of a BWR Neutron Absorber Melt Relocation in an Oxidative Environment During the CLADS-MADE-02 Test
Authors: Pshenichnikov Anton , Nagae Yuji , Kurata Masaki
Published: October 2021

The work on the understanding of the accident progression at the units of the Fukushima Dai-Ichi Nuclear Power Plant (1F) is ongoing. This contribution gives a part of detailed investigations of the control blade melt propagation downwards through the prototypic BWR bundle assembly during the CLADS-MADE-02 test, where the conditions of the 1F Unit 3 was simulated. Interesting features emerged in an oxidative environment.
Influence of Fuel Pellets’ Thermal Expansion on Temperature Feedback Regulation for Megawatt-Class Space Gas-Cooled Fast Reactor
Authors: He Yuhao , Yuan Dongdong , Ning Kewei , Wang Xiaoyu , Zhao Fulong , Tan Sichao
Published: October 2021

As human beings explore the outer space, the demand for the nuclear sources in space missions is increasing. Almost all the existing space nuclear reactors have adopted the design of gas-cooled fast reactor, and the temperature feedback effect of the fuel of the small fast neutron reactor is different from that of the traditional pressurized water reactor. For the space fast reactors, they are designed to be compact and with a long operating lifetime to accomplish the long-term space missions. Due to the high-level enrichment of U-235 in the fuel, the Doppler effect of U-238 has little impact on the reactor reactivity. In the present paper, based on the design of the megawatt-class space gas-cooled fast reactor, the overall temperature feedback adjustment capability of the core, and the overall temperature coefficient of the core including fuel, structural materials, and coolant are investigated. The calculation results show that the temperature of the coolant has little effect on the reactivity. The temperature coefficient of the fuel is positive, and the structural material with a large negative temperature feedback coefficient. At the same time, the deformation of the fuel pellets due to thermal stress at high temperatures and the reactivity changes caused by the deformation are analyzed.
Modeling of Irradiation-Induced Thermo-Mechanical Coupling Behavior in Metal Matrix Microencapsulated Fuel
Authors: Wei Hongyang , Zhu Fawen , Ru Jun , Wang Haoyu , Zhang Jing , Li Hua , Li Yun , Huang Chunlan , Li Yuanming , Ding Shurong
Published: October 2021

The application of TRISO-based dispersion fuel to advanced light water reactors, small modular reactors and micro-reactors has attracted increasing attention globally. With combination of superior fission products retention capability of tri-structural isotropic (TRISO) particle and stability of zircaloy matrix, metal matrix microencapsulated (M3) fuel becomes a promising candidate. To guarantee the safety and obtain the maximal fissile loading, it is necessary to predict inpile thermo-mechanical coupling behavior of M3 fuel. In this study, a three-dimensional finite element analysis for thermo-mechanical coupling behavior of M3 fuel at light water reactors (LWRs) operation condition is performed based on ABAQUS platform. Irradiation effects are considered comprehensively, especially kernel-buffer-IPyC interaction is depicted based on the newly developed model. The actual irradiation boundary conditions in light water reactors are adopted based on existing studies. Distribution and evolution of temperature, displacement and stress-strain are obtained and analyzed. Numerical simulation shows that (1) irradiation-induced swelling of fuel kernel can be well accommodated by porous buffer layer and the particle volume increases only 0.039% at 19% FIMA; (2) the maximal first principal stress for the coating layers will not exceed the corresponding fracture strengths, so that the safety of TRISO particle and zircaloy matrix can be guaranteed. This study serves as a basis for advanced fabrication of M3 fuel in the future.
Automated Eddy Current Array Sensor Delivery Tool for Nondestructive Examination of Spent Fuel Pool Liner
Authors: Smith Michael , Abbate Emily , Phillips Joey , Yoon Byungsik
Published: October 2021

Inspection of a spent fuel pool (SFP) liner (e.g., detection of defects in the pool liner) is important to validate the integrity of this component to perform its intended function. Eddy current array (ECA) testing offers a reliable and sensitive nondestructive examination (NDE) technique to inspect SFP liners, but requires a delivery system to transport the ECA sensor to these hard-to-reach and harsh environment locations. This work presents an ECA sensor delivery tool for use with inspection of an SFP. The design for a payload delivery tool proof-of-concept model is proposed that can transport an ECA sensor with positional feedback. The tool allows the sensor to traverse the wall of a nuclear SFP liner, with both horizontal and vertical motion capabilities to support positioning and inspection. The tool provides both manual and automatic operating modes with variable speeds. Additionally, the delivery tool can be adapted to accommodate other NDE sensors (e.g., ultrasound testing probes). Computational based simulations and analysis validate that the performance requirements for the ECA sensor delivery tool were satisfied to support inspection of an SFP liner upper region.
Neutron Physics Characterization & Optimization Analysis of the ACPR100 Small Modular Reactor
Authors: Liu Songyang , Wang Xiang
Published: October 2021

The ACPR100 is a small modular pressurized water reactor design proposed by China General Nuclear Power Corporation which integrates most important components of the reactor into one pressure vessel. This paper aims to model the reactor core by Monte Carlo code Serpent.

Firstly, the steady-state characterization and optimization analysis within different temperature, pin-pitch and other design parameters are performed. Secondly, the loading pattern of fuel assemblies with Gd-doped fuel rods was assessed. Thirdly, based on the reference research, the temperature coefficients of fuel and coolant are calculated. The influence of control rod insertion depth was simulated additionally. Besides, to show the life cycle and the change of inventory directly, we performed burnup calculation based on pre-defined time step and discussed the fission products including radioactive minor actinides and radiopharmaceutical isotopes during the life cycle.

The results show that the ACPR100 achieved more stable performance using high abundance boric acid, and the negative temperature reactivity coefficient is sufficient to maintain the stability of the reactor operation, but the ACPR100 is not suitable for massive production of radiopharmaceutical isotopes.
A Sodium-Cooled Thermal-Spectrum Fission Battery
Authors: McDaniel Patrick J. , Forsberg Charles
Published: October 2021

We propose a low-enriched-uranium, thermal-neutron-spectrum sodium-cooled reactor with a peak sodium temperature of 700°C coupled to an air-Brayton power cycle for electricity and heat. Three low-enriched-uranium, thermal spectrum sodium-cooled reactors were built in the 1960s and 1970s; but there has been no examination of such systems for many decades. We develop a pre-conceptual design based on “new” enabling technologies since the 1970s including yttrium hydride as the high-temperature neutron moderator, commercial gas turbines and secure decay heat removal systems. We define the reactor as a sodium hydride reactor (SHR).

The initial application is as a fission battery. The concept of the fission battery (FB) is a “plug and play” nuclear reactor defined by multiple characteristics: economics enabled by factory fabrication of large numbers of identical units, easy installation and removal, unattended operation and highly reliable operations. FBs are designed to be a low-carbon replacement for fossil fuels by industrial and commercial companies that require energy to produce some product (manufactured goods, chemicals, education, data centers, ship transportation, etc.). The reactors may be owned or leased by the company. The SHR is at an early stage of development.
A Study on Radiation Imaging Mechanism and Characteristics in Different Inspection Systems
Authors: Xu Yuting , Wu Zhifang , Wang Qiang
Published: October 2021

Radiation imaging, as a key issue in nuclear technology, has received considerable attention in the industry. It is widely used in nuclear medicine, Customs supervision, and many other areas. The objective of this investigation is to get insight into the principle, operation characteristics and image characteristics of radiation imaging.

In this paper, an investigation on radiation imaging is conducted on three main inspection systems for Customs supervision, including small X-ray inspection machine, CT baggage inspection system, and large container inspection system. The principle, operation characteristics, evaluation indexes, pseudo-color processing and image characteristics are discussed in detail. The results indicate that the spatial resolution of small X-ray inspection machine is much higher than that of CT baggage/goods inspection system and large container/vehicle inspection system. It is a challenge to identify substances and specific shapes in the case of overlapping for small X-ray inspection system.

Moreover, the mechanism of X-ray images is discussed as well. The radiation images are divided into three types, including two-dimensional, pseudo-color, high spatial resolution; two-dimensional, gray, high spatial resolution; three-dimensional, pseudo-color, high density resolution. The further investigation on machine inspection images is suggested to focus on the application environment. For some objects with specific characteristics, such as amorphous, explosive, the CT baggage inspection has much better performance than other systems.

The research in this paper reveals the mechanism, parametric effect and imaging characteristics. It could provide a necessary foundation for the follow-up intelligent processing, detection, identification and annotation for radiation imaging in nuclear area. The research on inspection devices could lend strong experience to medical treatment, industry and many other fields.
Assessment Of The Interfacial Drag Models In RELAP5 With Mixture Level Swell Experiment
Authors: Zhu Wen , He Qingche , Zhang Luteng , Ma Zaiyong , Sun Wan , Xu Wangtao , Pan Liangming , Huang Tao , Ding Shuhua
Published: October 2021

The two-phase flow phenomenon can be encountered in whether nuclear reactor normal running condition or emerged accidents, and imposes significant influences on reactor safety. One of the essential characteristics of a two-phase flow is the interfacial drag, which can influence the two-phase interface’s deformation, the distribution of the two phases, the flow pressure drop, etc. In this paper, the interfacial drag models in RELAP5 are assessed using GE mixture level swell experiments and FRIGG tests. GE Small Level Swell experiment and GE Large Level Swell experiment were respectively based on small and large experimental sections to study critical flow, void fraction distribution, two-phase level, and other characteristics under reactor blowdown conditions. FRIGG tests, using steam and water as the two-phase working fluid, studied the steady-state distribution characteristics of axial void fraction in the rod bundle channel under different conditions. They are separate effects tests and used to assess the interfacial drag models for many system analysis codes. For the GE Small Level Swell experiment, the results indicate that the RELAP5’s prediction of the void fraction at the middle and top part of the test section are in good agreement. However, at the bottom of the test section, the prediction is much higher than the experimental data and beyond the experimental data’s error bar. For FRIGG tests, the most simulation results of RELAP5 are similar to the simulation results of TRACE. For high flow rate, high-level power, and large inlet subcooling operation conditions, the prediction of the low void fraction region (below 0.3) and the high void fraction region (beyond 0.6) will have some differences comparing with the experimental data. All the differences between simulation results and experimental results can be attributed to the adopted interfacial drag models’ applicability in RELAP5.