bluuebunny/crossref_metadata_2025 huggingface dataset:
Analysis of Pressure Field on Wire-Wrapped Pin Bundle Surface for Concerns of FSI
Authors: Brockmeyer Landon , Obabko Aleksandr , Merzari Elia , Hassan Yassin
Published: October 2018

Wire wrapped fuel pins, typical of sodium fast reactor designs, introduce a strong secondary rotational flow and enhanced mixing compared to bare pins found in many other nuclear reactor designs. The transverse flow is created by deflection off of the helical wires, exerting a force onto the wires. The resulting pressure field can be more intricate than found for bare fuel pins. There is an increased concern for fluid structure interactions such as twisting, bending, and vibration due to the fluid deflection and increased transverse flow. The present study uses LES to simulate the fluid flow in a wire-wrapped fuel pin bundle. The simulation is performed using Nek5000, a spectral element LES/DNS code. The time-dependent, fluctuating pressure field at the surface of select fuel pins are specifically captured for analysis. Statistical analysis is performed on the pressure data to find areas of persistent force or bending moment, which over large periods of time may result in deformation of the pins. Further, the pressure data is analyzed in search of areas with large fluctuations in pressure, which may result in flow induced vibrations. Frequency spectra of the force fluctuations are analyzed.
The Effects of Compressibility and Piping Geometry on Steamhammer Loads
Authors: Moody Frederick J. , Stakenborghs Robert
Published: October 2018

Nuclear power plants typically consider a turbine trip and rapid closure of the main turbine stop valves as a normal transient event. As required by ASME Code [1], the piping loads generated by the unbalanced pressures in the system resulting from the rapid valve closure are part of the analyzed spectrum of conditions in the piping and support analysis. The analysis that determines the magnitude and timing of the loads is often referred to as a “steamhammer” analysis.

Currently, there are several computerized analytical techniques to determine the steamhammer piping and support loads [2], but because of compressibility assumptions the equations become more difficult to solve than in the analogous incompressible waterhammer models, which are quite straightforward.

This paper highlights the effect of fluid compressibility by comparing results predicted by both waterhammer (slightly compressible) flow models and compressible (steamhammer) flow models. Guidelines are offered to show how parameters of a piping system (such as pipe length, valve closure time and flow characteristic, steam initial state properties, and velocity) can be interpreted to determine if compressible effects are insignificant or if they play a significant role.
Interfacial Drag Force Improvement in Two-Fluid Model
Authors: Zhu Longxiang , Shan Jianqiang
Published: October 2018

Interfacial drag force, which indicates the momentum transfer between liquid phase and vapor phase, is a key constitutive equation in the two-fluid model. Based on the “drift-velocity approach” (utilized in RELAP5/MOD3) and the “drag coefficient approach” (utilized in RELAP5/MOD2 and CTF), three improvements are proposed, which are: 1) improved drag coefficient closure, 2) improved drag coefficient formulation approach, 3) improved bubble radius closure approach. The comparison among the two original approaches and the three improved approaches has been made with the ORNL experiment data in high pressure-low flow condition and the results have been discussed. Results indicates: 1) the EPRI correlation predicts the void fraction worse than the drag coefficient approaches; 2) the drag coefficient correlations in CTF predicts the void fraction better than improved drag coefficient formulation approach, which are the original equations in Ishii’s model; 3) the improved drag coefficient formulation approach predicts similarly with the original RELAP5/MOD2 correlations, though it gets rid of the dependence on interfacial area concentration; 4) improved bubble radius closure approach over-predicts the void fraction, however more experiment tests should be calculated before a conclusion is drawn.
Study on Emergency Planning Zone Determination for CAP200 Small Modular Reactor
Authors: Xuan Wang , Fenglei Du , Dawei Sun , Te Tang
Published: October 2018

Determination of the SMR emergency planning zone (EPZ) is one of the important external constraint factor of its marketing and application, which means that it is very important to formulate appropriate classification criteria and establish proper size range.

In China, due to the requirement of “Criteria for emergency planning and preparedness for nuclear power plants: Part 1, The dividing of emergency planning zone.” (GB/T 17680.1-2008), for PWR nuclear power plant, its external plume EPZ should be within 7km–10km, and its internal plume EPZ should be within 3km∼5km. However, the scope of the standard for the emergency planning area is currently limited to conventional nuclear power plants, and for the current SMR, its emergency planning size is not included.

In this paper, we will analyze the classification method of SMR EPZ based on the traditional Nuclear Power Plants feedback experience, including selection of source term, accident cutoff probability, determination method of the plume EPZ and the ingestion EPZ. Three typical nuclear power plant sites in China are chosen as CAP200 case study sites, including two inland nuclear power plant sites and one coastal site. The three sites can represent most of the meteorological and terrain characters of China nuclear power plants.

According to the CAP200 source term and meteorological data of the sites, MACCS2 computer program is used to calculate the severe accidents consequence. Conclusions show that for the CAP200 SMR, the accident cutoff probability can be 1.0E−08 to 1.0E−07 per reactor per year, and its project dose exceeding probability in the three sites boundary is far below 30%, which directs that for CAP200 SMR, its plume and ingestion emergence planning zone is limited to the on-site area, and its off-site emergency response can be simplified.
Research on Source Inversion for Nuclear Accidents Based on Variational Data Assimilation With the Dispersion Model Error
Authors: Liu Yun , Liu Xinjian , Li Hong , Fang Sheng , Mao Yawei , Qu Jingyuan
Published: October 2018

In a nuclear accident, radioactive release source term is the critical factor of nuclear emergency response and accident assessment. The modelling of source inversion based on variational data assimilation (VAR) is capable of balancing the environmental radioactive monitoring data to obtain the global optimal source term. But it could be influenced by the discrepancy between predictions of the atmospheric dispersion model and observations, which is defined as the dispersion model error in this study. In order to reduce this influence, the VAR with the dispersion model error (DME-VAR) is proposed. In the DME-VAR, the dispersion model error is quantified by the error coefficients at every monitoring station. These error coefficients and the release source term are estimated at the same time. For limiting the runtime, the DME-VAR program supports parallel processing. Two sets of wind tunnel experiment data for a typical Chinese nuclear power plant site are used to validate and evaluated the performance of the DME-VAR. The results demonstrate that the DME-VAR effectively estimates the error coefficients, and outperforms the VAR in both release rate estimation and radioactive contamination predicting. Moreover, the runtimes of these verification experiments are all reasonable, even for the application in the nuclear emergency response.
Prediction of Subcooled Boiling With Uncertainties From Boundary Conditions and Interaction Models
Authors: Cong Tenglong , Peng Minjun , Zhang Xiang
Published: October 2018

Subcooled boiling has been investigated by using the RPI wall boiling models in the last two decades. High accuracy of such models has been achieved by improving the submodels for interphase actions or tuning the model parameters. However, the reliabilities of the models are still suspicious due to the limited model validation and the experimental data based model calibration. The applicability of calibrated model parameters in the new experiment data can not be assured. The effects of model parameters need to be calibrated were treated as the uncertainties of these parameters. The critical parameters that dominate the prediction of subcooled boiling were selected by using the hierarchy analysis. After that, the input samples for uncertainty analysis were obtained by an efficient Monte-Carlo sampling technology — Latin Hypercube Sampling based on the hypothetic normal distribution. Then, the samples were transferred into the FLUENT code for CFD calculations. Results from CFD code were extracted for statistical analysis. Besides, the uncertainties from boundary conditions were also analyzed to quantify the effects of experimental uncertainties. The dependency of the predicted subcooling parameters and the input parameters can be obtained. A PIRT table can be drawn from the generated correlation coefficients between inputs and outputs to quantify the importance of model parameters on subcooled boiling.
A Criticality Evaluation of Fukushima Daiichi Unit 1 Fuel Debris
Authors: Freiría López María , Buck Michael , Starflinger Jörg
Published: October 2018

After the Fukushima accident, the interest of the scientific community in severe accident research has been renewed. One of the severe accident research issues that needs to be further investigated is the potential for recriticality of the fuel debris, which is formed after the core meltdown. In this study, a conservative criticality evaluation of the Fukushima Daiichi Unit 1 debris bed has been carried out. Parameters, such as debris size, porosity, particle size, fuel burnup and the coolant conditions, including the water density and the content of boron, were considered. The effect of these parameters on the neutron multiplication factor was analysed and safety parameter ranges, i.e. zones where the recriticality can be totally excluded, have been identified. The content of boron in water required to secure the subcriticality was calculated for those zones with recriticality potential. It was found that recriticality is achievable for a wide range of fuel debris conditions. 1600 ppm B would ensure subcriticality under any conditions.
Application of Virtual Reality Technology in Nuclear Power Plant Control Room Simulator
Authors: Xiyun Li , Xi Wang , Chenchen Liang , Shaohua Wang
Published: October 2018

Main control room simulator is widely used in design verification and operator training for nuclear power plant. The simulator needs to implement the arrangement, environment, human machine interface and function of main control room, which should be the same as much as possible. For designer, each type of reactor needs an individual simulator for design verification. As the number of unit increased, the simulator will consume a lot of space and difficult to reuse for other project. In addition, design verification for control room and I&C system need to start at the early stage of a project and is usually an iterative process with the design work. Build a control room facility for simulator needs a lot of time and is difficult to modify once constructed. To make the simulator more flexible and match the project schedule, virtual reality technology can be used to replace or extend traditional control room simulator with approximately the same arrangement, environment, human machine interface and function. In the full scope engineering simulator of HPR1000 unit, virtual reality control room interface has been designed as an extension of real control room implementation. The designer or operator can control and monitor the power plant in virtual reality environment, which just feels like real control room. It also can be used for other type of reactor by connecting to other simulator server and adding corresponding control room model in virtual reality software. With this preliminary application, control room simulator can be implemented in a short time and flexible for modification, which give designer more time and space for design verification and optimization. Once it applied in training simulator of nuclear power plant in future, it may provide a low cost and flexible option for operator training.
Study on Advanced PWR NPP Safety Related Equipment Qualification Function Requirement Design Methodology
Authors: Zhao Xu , Zhuang Miao , Ke Yi
Published: October 2018

In order to guarantee NPP operation safety, it is necessary to ensure the reliability operation of safety related equipment by equipment qualification. In the past, plant equipment qualification function requirement is an envelope requirement based on engineering experience, and it is not beneficial to NPP economical efficiency. These days, representative Gen III NPP (e.g. AP1000 and EPR) adopt improved technology in equipment qualification function requirement design, and more accurate requirement is designed. In this paper, AP1000 and EPR equipment qualification function requirement design methodology is studied and analyzed as the first step. Then, a safety related equipment qualification function requirement design methodology which is applicable for China self intellectual property Gen III NPP is provided. Furthermore, an example of equipment qualification function requirement design is carried out by analyzing nuclear instrument system power range channel sensor.
Research Concept of Decommissioning Knowledge Management for the Fugen NPP
Authors: Taruta Yasuyoshi , Yanagihara Satoshi , Iguchi Yukihiro , Kitamura Koichi , Tezuka Masashi , Koda Yuya
Published: October 2018

In 2002, the International Atomic Energy Agency (IAEA) mentioned the strengthening of nuclear knowledge, technology and application. This background has that there are aging of nuclear facility and nuclear power plant staffs. In addition, it would be difficult to succession a nuclear knowledge, technology, and skills. For example, undergraduate departments of nuclear energy and science are decreasing. The IAEA discussing those situations and pointed out the importance of a nuclear knowledge management. The nuclear knowledge management (NKM) is developing a database science as management on nuclear knowledge and information. In recent years, the IAEA has also advanced knowledge taxonomies on nuclear accidents as one of a nuclear knowledge management. In Japan, this achievements of nuclear knowledge taxonomy was using in the organization of information on accidents in Fukushima. A few studies are attempts to appropriately arrange and utilize huge amounts of information.

Even in nuclear facilities in Japan, it is pointed out a veteran or expert staff retirement and loss of knowledge and skill caused by this retirement. This problem is common issue in the world. Then, we created a prototype database system to utilize past documentation of knowledge and information. The database made from semantic web technology. The semantic web is a method of preparing a frame of categorized knowledge and linking information related to it. The target is a nuclear reactor of ATR Fugen that is decommissioning from 2008. Until now, cases of decommissioning completion are 17 cases in the world. One case of JPDR in Japan. It is not enough to understand a good method of decommissioning.

In general, the decommissioning project requires many information related to dismantling and decontamination. Particular, past information is important to know a past contamination situation and so on. This study focus on an access method for past data and information. However, we need to pay attention to other side of decommissioning project. Because of history of operating reactor has different tasks that are design, construction, operation and decommissioning. It is not appropriate to use the collected information as it is. For that reason, we will continue our research on the points pointed out above.
Numerical Investigation on the Characteristic of the Reverse Flow Phenomenon in a Z-Type Parallel Compact Heat Exchanger With Large Number of Tubes
Authors: Zhou Jian , Ding Ming , Bian Haozhi , Zhang Yinxing , Sun Zhongning
Published: October 2018

The parallel compact heat exchangers have been widely applied in the various fields such as heat exchangers in chemical engineering, the solar collector, fuel cells and the passive removal heat exchanger in passive containment cooling system (PCCS), etc. The heat exchangers in the PCCS removes out the heat brought by the steam coming out from the broken reactor or primary cooling system. Therefore, the performance of the passive containment cooling system heat exchanger (PCCS HX) will greatly influence the safety and integrity of the containment. In previous investigations on the parallel compact heat exchangers, attentions are focused on the pressure distribution and flow distribution in the heat exchangers. A bad flow distribution in the heat exchanger will reduce the heat performance. More seriously, the coolant in some tubes may boils and the tubes will be overheated, resulting in explosion of tubes. Therefore, the characteristic of pressure distribution and the flow distribution should be investigated for a uniform flow distribution.

In the past studies of the compact heat exchangers, the numbers of tube are almost under 72 which is relatively small, while the number of tubes PCCS HX is usually over than 100. And the pressure distribution in compact heat exchangers is assumed that the pressure recovery plays a leading role.

However, the more numbers of tube will bring more flow maldistribution, if the geometry design is selected inappropriately. The reverse flow may occur in the heat exchanger, which means that in some tubes, the coolant flows from the tube outlet to the inlet. This phenomenon of reverse flow have never been mentioned in previous studies. The occurrence of the reverse flow will significantly decrease the performance of the heat exchanger and cause a bad influence on the safety of the containment.

In the PCCS, the Z-type heat exchanger is one of the choice of PCCS HX (heat exchanger) design. Therefore, the present study focus on the characteristic of reverse flow phenomenon in Z-type heat exchangers. The pressure distribution and the flow distribution have been separately investigated deeply. The conclusion of this study will provide a guide to the geometry design of the PCCS HX with large number of tubes.
On Experimental and Computational Investigation of Heat Transfer Deterioration and Hydraulic Resistance in Annular Channel and SCWR 3-Rod Bundle
Authors: Filonov Vladislav V. , Filonova Yuliia S. , Razumovskiy Victor G. , Pis'mennyi Evgeniy N.
Published: October 2018

The experiments on upward flow of supercritical water in a vertical annuli and 3-rod tight bundle simulator made of 485-mm heated-length tubes of 5.2-mm OD and 4.5-mm ID with four helical ribs of 0.6-mm height, 1-mm width, and axial 400-mm pitch are presented.

Heat transfer and pressure drop under various operating conditions (inlet pressure and temperature, flow mass rate and heat flux) were investigated. Longitudinal wall temperature profiles made it possible to determine the place and flow thermal state of heat transfer deterioration (HTD) onset. Analysis of the obtained data (about 200 regimes) proved their good enough agreement with the correlations previously derived by the authors both for the heat flux rate (q/G)b of HTD beginning and for pressure drop in round tubes and annular channels. These correlations were updated to correct the results of their prediction.

Computational fluid dynamics and its counterpart computational heat transfer were used for modeling the above-mentioned thermohydraulic processes studied in the first part of the work by finding the most adequate flow turbulence model and optimized domain meshing. The accepted model was benchmarked by some data on heat transfer and pressure drop in tubes and annular channels cooled by SCW.
The Experimental Investigation of Steam Condensation With Non-Condensable Gas Under Natural Convection
Authors: Ma Xizhen , Fu Wen , Jia Haijun , Li Peiyue , Li Jun
Published: October 2018

The non-condensable gas is used to keep the pressure stable in the steam-gas pressurizer. The processes of heat and mass transfer during steam condensation in the presence of non-condensable gas play an important role and the thermal hydraulic characteristics in the pressurizer is particularly complicated due to the non-condensable gas. The effects of non-condensable gas on the process of heat and mass transfer during steam condensation were experimental investigated. A steam condensation experimental system under high pressure and natural convection was built and nitrogen was chosen in the experiments. The steam and nitrogen were considered in thermal equilibrium and shared the same temperature in the vessel under natural convection. In the experiments, the factors, for instance, pressure, mass fraction of nitrogen, subcooling of wall and the distribution of nitrogen in the steam, had been taken into account. The rate of heat transfer of steam condensation on the vertical wall with nitrogen was obtained and the heat transfer coefficients were also calculated. The characteristics curve of heat and mass transfer during steam condensation with non-condensable gas under high pressure were obtained and an empirical correlation was introduced to calculated to heat transfer coefficient of steam condensation with nitrogen which the calculation results showed great agreement with the experimental data.
Study on Hydrogen Migration in Small Water Leak of Sodium-Cooled Fast Reactor
Authors: Deng Xinjie , Cao Xuewu
Published: October 2018

Steam generator is a key component for realizing the heat exchange between the secondary sodium circuit and the tertiary circuit water in the sodium-cooled fast reactor. Any tiny crack of the heat-transfer tube in the steam generator may lead to the contact of water and sodium, resulting in the sodium-water reaction and the generation of hydrogen, reaction heat and other corrosive products, which further corrode the break and the adjacent tubes. In order to effectively curb the further development of small leak, timely detection of the occurrence of sodium-water reactions is particularly important. With the background of the small leak sodium water reaction caused by the crack of the steam generator heat transfer tube, in this paper the migration behavior of hydrogen in the secondary sodium circuit is analyzed, based on the mass, momentum and energy conservation equations, and the concentration of hydrogen in the secondary sodium circuit is numerically calculated one-dimensionally. The results show that the hydrogen concentration at a certain point in the circuit increases with the increase of water leaking time as a whole, when the water injection is intermittent, the hydrogen concentration at a certain point in the circuit fluctuates periodically and the basal concentration becomes higher and higher, which provides a reference for the arrangement of the secondary loop small leak detection system.
Study on Dynamic Characteristics and Flow Induced Vibration of Tube Bundles Based on the Fluid Structure Coupling Method
Authors: Feng Zhipeng , Huang Qian , Liu Shuai , Cai Fengchun , Lv Xi , Jiang Xiaozhou
Published: October 2018

In order to study the dynamic characteristics and fluid structure interactions of tubular structures under the action of fluid in reactor, such as fuel rod bundles and heat transfer tube bundle of steam generator, the dynamic equations and the acoustic wave equations of structures are discretized by finite element method. The acoustic wave equations are simplified from continuity equation and momentum equation of fluid field. Based on the fluid structure coupling method, the dynamic characteristics of the tube under the internal flow, external flow and combined action of internal and external flow are calculated respectively. The influence of flow field domain, element type and grid number on the dynamic characteristics of the tube is also analyzed. Secondly, based on the computational fluid dynamics and computational structural dynamics, the interaction between the two physical fields of fluid and structure is considered simultaneously. The finite volume method is used to discretize the fluid control equations and the turbulent flow is investigated using the large eddy simulation method (LES). The Newmark algorithm is used to solve the structural dynamic equations. Combined with the dynamic mesh control technique, a numerical model for flow induced vibration of three-dimensional flexible tube is established. Finally, the flow induced vibration of a three dimensional flexible single tube and a square arrangement tube bundle is calculated using the numerical model. By comparing with the existing research results, it is found that the numerical simulation results are in good agreement with the experimental results. Thus, the correctness of the model is verified. It is also shown that the numerical model established in this paper can be used to simulate the dynamic characteristics and flow induced vibration of tubular structures.
Boiling Visualization and Critical Heat Flux (CHF) Phenomena on PCB in a Saturated Pool at Various Surface Orientations
Authors: Tanjung Elvira F. , Jo Daeseong
Published: October 2018

The onset of nucleate boiling (ONB) and critical heat flux (CHF) of saturated water on printed circuit board (PCB) that is immersed in a stainless steel pool were investigated at various surface orientations. Additionally, photographic analyses and image processing were conducted to observe the pool boiling phenomena and determine the bubble dynamics from upward to downward surface orientations (0°, 45°, 90°, 135°, 150°, and 180°). Results revealed that boiling commenced from downward to upward surface orientations (180° to 0°). When the heater was placed facing upwards (0° and 45°), the generated vapor bubble grew and coalesced together to form an elongated bubble which eventually departed from the heated surface due to the buoyancy in the vertical direction of the heated surface. In contrast to those orientations, at 90°, 135°, and 150°, the generated vapor bubble exhibited similar behavior except that it drifted along the heated surface before it departed at the most upper part of the heater. On the other hand, when the heater placed horizontally facing downward (180°), the generated vapor bubble grew and coalesced together until it ultimately covered the entire heated surface. Early ONB occurred when the heater was placed facing downwards (135°, 150°, and 180°). Moreover, it was observed that the CHF began when the heater was placed horizontally facing downward (180°) as the boiling commenced at this surface orientation. Furthermore, based on the photographic and image processing, the results revealed that the approximate bubble departure diameter with the heater facing upwards (0° and 45°) including at 90° was bigger compared to that of the heater facing downwards (135° and 150°). On the other hand, bubble frequency decreased with increasing surface orientation (0° to 90°) and increased as the surface orientation went beyond 90°. Moreover, it was observed that the bubble with the biggest departure diameter had the smallest departure frequency, and needs longer time to generate another bubble. Based on these results, it is apparent that the surface orientation significantly affects the ONB, CHF, and bubble dynamics on PCB in a saturated water pool.
Injectable Sacrificial Material System to Contain Ex-Vessel Molten Corium in Nuclear Accidents
Authors: Louie David L. Y. , Wang Yifeng , Rao Rekha , Kucala Alec , Kruichak Jessica
Published: October 2018

An ongoing Sandia National Laboratories’ (SNL) research study is evaluating a potential design of an injectable sacrificial material (SM) system that could contain and cool corium ejected from a reactor vessel lower head failure during a potential severe accident involving melting fuel at a commercial light water nuclear reactor (LWR). An injectable system could be installed at any existing LWR, without significant modification to the cavity or to the drywell pedestal region of the plant. The conceptual design under consideration is a passive system. The SM is being optimized to quickly cool the corium mixture while creating gas to form porosity in the solid, such that subsequent water flooding can penetrate the structure and provide additional cooling. The SM would form a barrier and limit corium-concrete interactions.

This three-year project takes a joint experimental and computational approach. In this paper, we will first discuss the success of our small-scale experiments conducted on the interactions between the surrogate corium material (SCM) and SM, used to evaluate the injectable concept. A larger experimental study, currently underway, will further validate the injectable concept, with a focus on accurately measuring interactions. This paper details the modeling study and its progress, including modeling the experiments on a surrogate system and extending the model to bench-scale corium flow from validation experiments. The project’s modeling studies will use the SNL engineering code suite SIERRA Mechanics to understand the interaction of injectable SM and molten corium and predict corium spreading. Spreading is modeled using a level set method to track the front in conjunction with a pressure-stabilized finite element method on the fully three-dimensional mass, momentum, and energy conservation equations. Using this diffuse-interface method, the corium spreading front can be tracked and an appropriate pseudo-solidification viscosity models can be implemented to accurately model the corium spreading physics. Finally, an injectable SM delivery system is discussed along with its deployment to the six-common commercial LWR designs currently operating in the United States.

At the end of this project, a simplified model based on SIERRA simulations will be developed for implementation into MELCOR, a severe reactor analysis code, developed at SNL for the U.S. Nuclear Regulatory Commission. This will allow us to demonstrate the ability of the injectable SM system to mitigate the ex-vessel corium spreading, provide containment and negate the release of radionuclides.
Study on NPP Reactivity Accident Operating Strategy Design Based on Function Analysis and Task Analysis Technology
Authors: Aimin Yu , Zhao Xu , Yu Du , Qian Sun
Published: October 2018

Nuclear Power Plants (NPP) have multiple levels of defense in depth hierarchy. The NPP accident condition operation strategy belongs to the 3rd level. It is used to supervise the operator to handle the NPP under accident operating condition. NPP accident condition operation strategy is an essential and difficult work in NPP design field, hence only few organizations are able to develop the accident operating strategies independently all over the world. In this paper, a systematic NPP accident operating condition strategy design methodology is raised based on function analysis and task analysis technology. Based on the methodology, a reactivity accident operation strategy is designed and proved to be reasonable through preliminary verification and validation work.
The Choose of Decay Heat Removal Systems of SFR
Authors: Yue Nina , Cai Rong , Wang Yun , Qiu Suizheng , Zhang Dalin
Published: October 2018

A sodium-cooled fast reactor is a significant candidate for future power reactor systems. Decay heat removal is an essential function of reactor safety systems The decay heat removal system should have the capacity to remove the decay heat with natural circulation in any accident. There are three types of decay heat removal systems, namely direct reactor auxiliary cooling system, primary reactor auxiliary cooling system, and intermediate reactor auxiliary cooling system. The one dimensional systems analysis code THACS was applied to conduct transient analyses of a sodium-cooled fast reactor, and the capabilities of three types of decay heat removal systems against a station blackout accident were compared. The results indicate that these three types of decay heat removal systems can remove the residual heat effectively. For large-scale sodium-cooled fast reactor, the capabilities of primary reactor auxiliary cooling system and intermediate reactor auxiliary cooling system were better, because the cold sodium from the penetrating heat exchanger in these two auxiliary cooling systems could directly flow into the core assemblies.
Numerical Simulation of the Performance of an Axial Compressor Operating With Supercritical Carbon Dioxide
Authors: Gou Jinlan , Wang Wei , Ma Can , Li Yong , Lin Yuansheng , Li Huafeng
Published: October 2018

Using supercritical carbon dioxide (SCO2) as the working fluid of a closed Brayton cycle gas turbine is widely recognized nowadays, because of its compact layout and high efficiency for modest turbine inlet temperature. It is an attractive option for geothermal, nuclear and solar energy conversion. Compressor is one of the key components for the supercritical carbon dioxide Brayton cycle. With established or developing small power supercritical carbon dioxide test loop, centrifugal compressor with small mass flow rate is mainly investigated and manufactured in the literature; however, nuclear energy conversion contains more power, and axial compressor is preferred to provide SCO2 compression with larger mass flow rate which is less studied in the literature. The performance of the axial supercritical carbon dioxide compressor is investigated in the current work. An axial supercritical carbon dioxide compressor with mass flow rate of 1000kg/s is designed. The thermodynamic region of the carbon dioxide is slightly above the vapor-liquid critical point with inlet total temperature 310K and total pressure 9MPa. Numerical simulation is then conducted to assess this axial compressor with look-up table adopted to handle the nonlinear variation property of supercritical carbon dioxide near the critical point. The results show that the performance of the design point of the designed axial compressor matches the primary target. Small corner separation occurs near the hub, and the flow motion of the tip leakage fluid is similar with the well-studied air compressor. Violent property variation near the critical point creates troubles for convergence near the stall condition, and the stall mechanism predictions are more difficult for the axial supercritical carbon dioxide compressor.
A Theoretical Model of the Stress Intensity Factor Threshold of DHC for Fuel Cladding Tube
Authors: Chen Liang , Liu Lili , Song Xiaoming , Pang Hua
Published: October 2018

Delayed hydride cracking (DHC) is the result of a mechanism of crack initiation and slow propagation. In DHC, hydrogen diffusion in the metal is required. Gradients of concentration, temperature, and stress are all important factors in controlling diffusivity. The classic theory of DHC still has potential to be modified. In this study, a calculation formula for DHC SIF threshold is established with consideration of the temperature history, the temperature field and the stress field induced by the temperature gradient and the external mechanical loading. Moreover, the temperature gradient on crack surfaces has been considered in the model.
An Experimental Study on Freeze Valve Performance in a Molten Salt Reactor
Authors: Aji Indarta Kuncoro , Tatsuya Tokushima , Kinoshita Motoyasu , Okawa Tomio
Published: October 2018

Freeze valve technology is the main feature of safety system in the molten salt reactor. Freeze valve made from frozen salt located between reactor core and drain tank. The freeze valve will automatically melt and open on the accident condition, and respectively molten salt fuel will drain out from reactor core to the drain tanks. Melting time of frozen salt is important issues on this study, where draining process of the liquid fuel must be carried out immediately after the accident. Many factors affect to the opening time of freeze valve. On this experiment, describe a melting process of frozen salt which is affected by wall effect. HTS (high transfer salt) utilized as salt material, and a metal stick planted in the frozen HTS with a certain depth. The experimental process begins when the liquid HTS poured on the top of frozen HTS and ends when the metal stick detached from the frozen HTS. This experiments focus to analyze melting time of freeze valve which impacted by several parameters; material diameter which represent about material thickness in real situation, liquid temperature which represent about molten salt fuel, length of material which propose about the freeze valve thickness, and material difference which propose about thermal diffusivity effect. Results from the experiments will be utilized as a basic to developed mathematics and numerical analysis.
Basic Experimental Study on Effectiveness of Nuclear Waste Long-Term Storage Containers With PAR for Reducing Concentration of Hydrogen Gas: Part 2 — Hydrogen Behavior in a Small-Scale Modeled Container
Authors: Hiraki Y. , Takase G. , Suzuki Y. , Tanaka Y. , Takase K.
Published: October 2018

As for the decommissioning of Fukushima Dai-ichi nuclear power plant (1F), a long-term waste storage container with high safety is requested to store radioactive materials such as fuel debris for a long period of time. Since hydrogen is generated by radioactive decomposition of water, it is important to keep the concentration of hydrogen gas below the explosion limit in order to ensure the safety of the container. Then, use of passive autocatalytic recombiner (PAR) was proposed to reduce the hydrogen concentration. PAR is installed in the container. In order to experimentally confirm the reduction of hydrogen concentration by PAR and hydrogen behavior in the container, an experimental apparatus consisting of a small-scale modeled container and a hydrogen supply system was provided. Preliminary experiments were begun for confirming fundamental performance of the experimental apparatus under the conditions that PAR and simulated fuel debris are not installed in the container. Moreover, the hydrogen behavior in the container was analyzed numerically. In addition, the steam behavior generated by the chemical reaction of hydrogen and oxygen by PAR was also predicted. This paper describes both results of the preliminary experiments and numerical simulations. The experimental results showed that the hydrogen behavior can be predicted using the temperature distributions in the container. The analysis results clarified the controlling factors on the hydrogen behavior and the steam distribution in the container by PAR.
A Concept of Intermediate Heat Exchanger for High-Temperature Gas Reactor Hydrogen and Power Cogeneration System
Authors: Hirota Noriaki , Terada Atsuhiko , Yan Xing L. , Tanaka Kohei , Otani Akihito
Published: October 2018

A new conceptual design of intermediate heat exchanger (IHX) is proposed for application to the gas turbine high temperature reactor system (GTHTR300C) which is being developed by Japan Atomic Energy Agency (JAEA). The GTHTR300C cogenerates hydrogen using the iodine-sulfur (IS) hydrogen production process and electric power using gas turbine. The IHX is used to transport high temperature heat from the nuclear reactor to the hydrogen plant. The IHX proposed in this paper is a horizontal design as opposed to conventional vertical design. Therefore, JAEA investigated the advantage of the horizontal IHX and the economic evaluation when scaling up from conceptual design of high temperature engineering test reactor (HTTR) / IHX to GTHTR300C. To meet the performance requirement, both thermal and structural designs were performed to select heat transfer tube length, tube bundle diameter, insulation thickness, and the length of shell support in a horizontal pressure vessel. It is found that the length of the heat exchanger tube can be shortened and the superalloy-made center pipe structure can be eliminated, which results in reducing the quantity of construction steel by about 30%. Furthermore, the maximum stress concentration in the tubes is found to be significantly reduced such that the creep strength to withstand continuous operation is extended to 40 years, equaling the nuclear reactor life time, without replacement.
Study on Thermal-Hydraulic Characteristics of Vertical Narrow Rectangular Channel With Large Aspect Ratio Under Transverse Uneven Heating
Authors: Sun Rulei , Yang Yichen , Zhang Dalin , Bian Jiawei , Qiu Suizheng , Su G. H.
Published: October 2018

The plate type fuel element is widely used in advanced research reactors due to its excellent heat transfer performance, low fuel core temperature and large core power-to-volume ratio, etc. The fuel plate gap is generally 1 to 3 mm, and the aspect ratio is generally greater than 20, thus forming a typical large aspect ratio rectangular channel between adjacent fuel plates. In addition, in the actual research reactor, the distribution of power is uneven along the width of the channel due to the influence of the radiation target and fuel arrangement, which inevitably causes lateral migration and cross-mixing of the fluid. As a result, the flow and heat transfer characteristics may be significantly different from those in conventional narrow channels. At present, the experimental and theoretical research on the flow heat transfer characteristics in the rectangular channel are basically carried out under the condition of uniform heating. It is rarely reported that the flow and heat transfer characteristics in narrow rectangular channel with large aspect ratio under transverse non-uniform heating. Therefore, the numerical study of thermal-hydraulic characteristics in vertical narrow rectangular channel with large aspect ratio under transverse uneven heating is carried out in this study. The numerical calculation shows that the lateral power distribution has a significant effect on the temperature field distribution of the core coolant and the fuel plate, which suggests that the distribution of the lateral power of the plate fuel should be considered in the design and safety analysis of reactor cores.
Effect of Rolling Motion on Flow Instability of Parallel Rectangular Channels of Natural Circulation
Authors: Wang Xiaoyan , Huang Siyang , Tian Wenxi , Chen Lie , Qiu Suizheng , Su G. H.
Published: October 2018

In order to study the effect of rolling motion on flow instability of parallel rectangular channels of natural circulation, the natural circulation reactor simulation system is used for physical prototype. And theory analysis model of parallel rectangular channels of natural circulation system under rolling motion is established and coded by Fortran. The results of the program are verified to the experiments, and the results are in good agreement. The flow instability boundaries of different pressure under static and rolling motion are calculated respectively. The results show that: 1) under static condition, with the increase of the pressure, the instability boundary line changes, and the system becomes more stable; 2) under rolling conditions, the heating power of instability boundary decreases comparing to the stable conditions. The instability occurs earlier; 3) the stability of the system decreases with the increasing of rolling amplitude and frequency.
Assessment of RELAP/SCDAPSIM for Turbine Trip Transient in NuScale-SMR
Authors: Skolik Katarzyna , Trivedi Anuj , Perez-Ferragut Marina , Allison Chris
Published: October 2018

The NuScale Small Modular Reactor (SMR) is an integrated Pressurized Water Reactor (iPWR) with the coolant flow based on the natural circulation. The reactor core consists of 37 fuel assemblies similar to those used in typical PWRs, but only half of their length to generate 160MW thermal power (50 MWe).

Current study involves the development of a NuScale-SMR model based on its Design Certification Application (DCA) data (from NRC) using RELAP/SCDAPSIM. The turbine trip transient (TTT) was simulated and analysed. The objective was to assess this version of the code for natural circulation system modeling capabilities and also to verify the input model against the publicly available TTT results obtained using NRELAP5. This successful benchmark confirms the reliability of the thermal hydraulic model and allows authors to use it for further safety and severe accident analyses.

The reactor core channels, pressurizer, riser and downcomer pipes as well as the secondary steam generator tubes and the containment were modeled with RELAP5 components. SCDAP core and control components were used for the fuel elements in the core. The final input deck achieved the steady state with the operating conditions comparable to those reported in the DCA. RELAP/SCDAPSIM predictions are found to be satisfactory and comparable to the reference study. It confirms the code code capabilities for natural circulation system transients.
“Begging the Question” in Licensing Basis Accident Analyses
Authors: Miranda Samuel
Published: October 2018

“Begging the question” describes a situation in which the statement under examination is assumed to be true (i.e., the statement is used to support itself). Examples of this can be found in analysis reports that were prepared by analysts who are not mindful (or maybe uninformed) of the analysis criteria they’re required to fulfill.

This is generally seen in analyses of anticipated operational occurrences (AOOs). AOOs are defined in Appendix A of 10 CFR §50 [1], and in ANS-N18.2-1973 [2], where they’re also known as American Nuclear Society (ANS) Condition II events. This standard [2] also defines more serious, Condition III and IV events. Analyses of AOOs, or ANS Condition II events are required to show that:

(1) reactor coolant system (RCS) pressure will not exceed its safety limit, and

(2) no fuel damage will be incurred, and

(3) a more serious accident will not develop, unless there is a simultaneous occurrence of another, independent fault.

The three requirements are often demonstrated by three different analyses, each of which is designed to yield conservative results with respect to one of the requirements. Accident analyses that are performed to demonstrate compliance with the first two requirements are relatively straightforward. They rely mostly upon the design of safety valves and the timing of reactor trips.

“Begging the question” is seen in analyses that are designed to demonstrate compliance with the third requirement. This paper will describe how this logical fallacy has been applied in licensees’ accident analyses, and accepted by the NRC staff.
Analysis Contour Plots in RELAP/SCDAPSIM/MOD 3.4 and MOD 4.1
Authors: Sánchez-Mora Heriberto , Chávez-Mercado Carlos , Allison Chris , Hohorst Judith
Published: October 2018

RELAP/SCDAPSIM is a nuclear reactor simulator and accident analysis code that has been used in the nuclear energy industry for many years. Currently, Innovative Systems Software is developing a new tool that will show the behavior of the core components during a simulation of an accident. The addition of contour plots for the SCDAP components showing different properties: temperature, hydrogen production, etc. will allow a better understanding of core behavior during a severe accident in a nuclear reactor. The SCDAP components are fuel rods, electrically heated simulator rods, such as those used in the CORA experiments, control rods, a shroud and a BWR blade/box. This paper describes the progress in the development of the contour plot tool based on the OpenGL and FORTRAN90 libraries. The purpose of this tool is help to the user analyze the simulation of an accident and to debug an input file.