bluuebunny/crossref_metadata_2025 huggingface dataset:
Hydraulic Characteristics Research on SG Under Tube Plugging Operations Using FLUENT
Authors: Zhao Xiaohan , Wang Mingjun , Tian Wenxi , Su G. H. , Qiu Suizheng
Published: October 2018

Steam Generator (SG) is a critical equipment in the nuclear power plant, it is the huge heat exchanger in reactor system which can achieve removing fission energy from the reactor system effectively to ensure safety of the whole nuclear system. It is located between the primary and the secondary loop in reactor system act as the intermediate hub of energy and the security barrier in nuclear power plant. Generally, there are numerous of U-shaped heat transfer tubes in SG it is one of the weakest structures throughout the primary loop system. So the integrity of the SG especially its heat transfer tubes is important to the safety of reactor operation. The degradation problem of heat transfer tubes together with ruptures accidents often occur under suffer environments in reactors, which include thermal stress, mechanical stress and so on, it is noteworthy that this kind of accidents is inevitable due to the limited properties of existing materials. The performance of the SG is seriously affected by the number of failure tubes. Plugging operations through various mechanical means is the most common method to solve the tubes ruptures problems which can reduce the economic losses to the utmost extent. However, plugging operations will make huge impact on the thermal hydraulic performances of both sides of SG. It’s meaningful to research the characteristics of the plugging affects under different operations. In this paper the hydraulic characteristics of primary side in AP1000 SG under a certain fraction of heat transfer tube plugging conditions is researched. Three dimensional hydraulic characteristics of primary side coolant in SG under different plugging conditions are obtained by using the thermal hydraulic software FLUENT. The typical plugging fraction in this simulation model is 10 percent, and the effect of plugging locations also be considered through changing the plugging positions using the zone marking method. The results shows that the pressure drop under the structure integrated SG is 358.01MPa which is accordance with the results from Westinghouse 343KPa. The pressure drop values varies when changing positions of the plugging tubes under the same plugging fraction condition. The flow fields in bottom head also change meanwhile and the maximum pressure drop can reach up to 388.05KPa when the plugging fraction is 10%. The growth rate become significant when tube plugging fraction larger than 5%, and differences between maximum and minimum values of total pressure drop under different plugging positions become larger gradually. Finally the local resistance coefficients and flow field distributions of primary side in SG under various plugging conditions are obtained which is meaningful for the reactor safety and it can be a good reference for the maintenance of SG.
A Critical Experimental Study of Bubble Effect in the Process of Spent Fuel Dissolving
Authors: Xiaoping Zhou , Shuhong Liang , Zhaodong Xia
Published: October 2018

The influence of gas introduction on the critical safety of the nuclear fuel system under the condition of cold condition, given reactor material and geometry structure is studied.

Refer to bubble effect test experiment on nuclear critical safety test device (YSR) and considering solid-liquid two-phase nuclear fuel system with uranyl nitrate solution - uranium dioxide fuel element as the experimental platform, the dynamic process of the real behavior of bubbles in uranyl nitrate solution has been simulated in the quasi-static way by replacing bubble generator with aluminous bubble simulation elements.

Bubble effect is the reactivity change caused by the change of volume of solution, neutron leakage and absorption property in the nuclear fuel system due to the bubbles generated in the solution. In the dissolving process of spent fuel, oxygen or nitrogen are usually added to accelerate the dissolution of fuel element shear section, and some other bubble production are also caused by the heat released during the dissolution process. Here, the bubble production caused by the heat is omitted and only artificial gas introduction is considered in my study. When there are enough bubbles in the uranium solution system, the volume of the solution will increase, which will inevitably influence the absorption and leakage property of the neutrons, and further influence the reactivity of the nuclear fuel system.

The corresponding relationship between the bubble-intake rate and the bubble equivalent diameter, arising velocity and bubble share is determined through fluid dynamics modeling to manufacture the aluminous bubble simulation elements. The theoretical calculation by MONK9A and the critical experimental measurements are also compared and analyzed in this paper.

The results showed that the reactivity caused by bubbles was negative, and the greater the intake rate, the greater the negative effect. Meanwhile the theoretical calculated value was in good agreement with the experimental value and the maximum deviation was 63.4 pcm.
A Framework and Model for Assessing the Design Point Performance, Off-Design Point Performance, Control, Economics and Risks of Brayton Helium Gas Turbine Cycles for Generation IV Nuclear Power Plants
Authors: Gad-Briggs Arnold , Pilidis Pericles , Nikolaidis Theoklis
Published: October 2018

A framework – NuTERA (Nuclear Techno-Economic and Risk Assessment) has been developed to set out the requirements for evaluating Generation IV (Gen IV) Nuclear Power Plants (NPPs) at the design conceptual stage. The purpose of the framework is to provide guidelines for future tools that are required to support the decision-making process on the choice of Gen IV concepts and cycle configurations. In this paper, the underpinning of the framework has been demonstrated to enable the creation of an analyses tool, which evaluates the design of an NPP that utilises helium closed Brayton gas turbine cycles. The tool at the broad spectrum focuses on the component and cycle design, Design Point (DP) and Off-Design Point (ODP) performance, part power and load following operations. Specifically, the design model has been created to provide functionalities that look at the in-depth sensitivities of the design factors and operation that affect the efficiency of an NPP such as temperature and pressure ratios, inlet cycle temperatures, component efficiencies, pressure losses. The ODP performance capabilities include newly derived component maps for the reactor, intercooler and recuperator for long term Off-Design (OD) operation. With regard to short term OD, which is typically driven by changes in ambient conditions, the ability to analyse the cycle load following capabilities are possible. An economic model has also been created, which calculates the component costs and the baseline economic evaluation. An incorporated risk model quantifies the performance, operational, financial and design impact risks. However, the tool is able to optimise the NPP cycle configuration based on the best economics using the Levelised Unit Electricity Cost (LUEC) as a measure. The tool has been used to demonstrate a typical decision-making process on 2 Gen IV helium closed gas turbine cycles, which apply to the Gas-cooled Fast Reactors (GFRs) and Very-High Temperature Reactors (VHTRs). The cycles are the Simple Cycle Recuperator (SCR) and Intercooled Cycle Recuperator (ICR). The tool was able to derive the most efficient cycle configurations for the ICR (53% cycle efficiency) and SCR (50% cycle efficiency). Based on these efficiency figures, the baseline LUEC ($/MWh) for the year 2020 is $62.13 for the ICR and $61.84 for the SCR. However, the inclusion of the cost of contingencies due to risks and the subsequent economic optimisation resulted in a cost of $69.70 and $69.80 for the ICR and SCR respectively.
Liquid Film Flow on Vertical and Successive Inclined Plate Modeling Flow Channel of Molten Control Rod in Severe Accident
Authors: Hihara Yutaro , Monji Hideaki , Abe Yutaka , Yamashita Susumu , Yoshida Hiroyuki
Published: October 2018

When a severe accident of a nuclear reactor occurs, decommissioning work becomes important task. In the decommissioning work of a boiling water reactor after the severe accident, estimation of the sedimentation place of the molten debris is important. However, the technique to estimate exactly the sedimentation place has not been enough developed. Therefore, the detailed and phenomenological numerical simulation code named “JUPITER” is under development for predicting the molten core behavior in JAEA (Japan Atomic Energy Agency). The comparison between experimental and numerical results is necessary to clarify the validity of the numerical analysis code. The study provides the experimental data to examine the numerical simulation code. As a basic study to examine the numerical simulation code, a liquid film flowing in a modeling flow channel was studied by using water. The flow was visualized, and the flow data were obtained by image processing.
Heat Transfer and Fluid Flow Characteristics of One Side Heated Vertical Rectangular Channel Applied As Vessel Cooling System of VHTR
Authors: Fujikami Kenta , Takeda Tetsuaki , Funatani Shumpei
Published: October 2018

A Very High Temperature Reactor (VHTR) is one of the next generation nuclear reactor systems. From a view point of safety characteristics, a passive cooling system should be designed as the best way of a reactor vessel cooling system (VCS) in the VHTR. Therefore, the gas cooling system with natural circulation is considered as a candidate for the VCS of the VHTR. Japan Atomic Energy Agency (JAEA) is advancing the technology development of the VHTR and is now pursuing design and development of commercial systems such as the 300MWe gas turbine high temperature reactor GTHTR300C (Gas Turbine High Temperature Reactor 300 for Cogeneration). In the VCS of the GTHTR300C, many rectangular flow channels are formed around the reactor pressure vessel (RPV), and a cooling panel utilizing natural convection of air has been proposed. In order to apply the proposed panel to the VCS of the GTHTR300C, it is necessary to clarify the heat transfer and flow characteristics of the proposed channel in the cooling panel. Thus, we carried out an experiment to investigate heat transfer and fluid flow characteristics by natural convection in a vertical rectangular channel heated on one side. Experiments were also carried out to investigate the heat transfer and fluid flow characteristics by natural convection when a porous material with high porosity is inserted into the channel. An experimental apparatus is a vertical rectangular flow channel with a square cross section in which one surface is heated by a rubber heater. Dimensions of the experimental apparatus is 600 mm in height and 50 mm on one side of the square cross section. Air was used as a working fluid and fine copper wire (diameter: 0.5 mm) was used as a porous material. The temperature of the wall surface and gas in the channel were measured by K type thermocouples. The flow velocity distribution was obtained by a PIV method. In this paper, we discuss the heat transfer and fluid flow characteristics of the proposed channel. From the results obtained in the experiment, it was found that the amount of removed heat decreased with increasing of temperature of gas when a copper wire was inserted into the channel with high porosity. This is because the mass flow rate decreased with increasing of viscosity of gas. Since it is expected that the porosity of a porous material will have an optimum value, further studies will be needed.
Design of the Sampling Measurement and Radiochemistry Lab in the Nuclear Island of HTR-PM
Authors: Lou Mengqi , Li Wenqian , Xie Feng , Cao Jianzhu , Wei Liqiang , Tong Jiejuan , Kong Jiaji
Published: October 2018

Radiation monitoring system is very important to the safe operation of the nuclear power plant. The radiation monitor, including the on-line type and the off-line type, can provide the radioactive level for a given system, certain areas, or unique substance promptly. However, as a supplementary method, the sampling measurement can supply more accurate information about the source term. In this paper, we present the design of the sampling measurement in the nuclear island of HTR-PM. The sampling measurement contains the helium sampling from the primary coolant, the radioactive dust sampling from the primary loop, the liquid sampling from the tritiated water and some process systems, the gas sampling from certain areas, etc. The frequency of the sampling measurement depends on regulatory requirements and actual operational demands. The control values of the samples are settled on the basis of the source term analysis and regulatory requirements. The radiochemistry lab has been designed to meet the requirement of the sampling measurement, in which the main instruments include a NaI γ spectrometer, two high-purity germanium γ spectrometers, a four-channel ultralow background α/β analyzer, and a low background liquid scintillation counter. The characteristics of these instruments will be described in details in the paper.
Numerical Investigation of Mass Mixing in the Reactor Core of a Pebble Bed HTGR
Authors: Lu Wei , Li Xiaowei , Wu Xinxin
Published: October 2018

In the reactor core of a pebble bed HTGR, fuel pebbles are in random arrangement. The coolant flow passages are complicated. Under the air and water ingress accident conditions, the partially enriched oxidizing gas corrodes the fuel pebbles by oxidation reaction. How the oxidizing gases (oxygen and water vapor) diffuse in the pebble bed will greatly influence the oxidation process. Analysis of the mass mixing is important for analyzing the fuel element graphite corrosion and safety of HTGR. In this paper, a three-dimensional simple cubic structure packed cells of pebble bed core model with surface contact method was established. The flow, and mass mixing processes were numerically and theoretically studied using numerical simulation and porous media methods. The effects of molecular diffusion, turbulent diffusion and mass dispersion on mass mixing effect were discussed respectively.
Tritium Transport Characteristics Analysis in Molten Salt Reactor Under Transient Conditions
Authors: Qin Hao , Wang Chenglong , Qiu Suizheng , Zhang Dalin , Tian Wenxi , Su Guanghui
Published: October 2018

Tritium control is one of the most critical issues for the development of Molten Salt Reactor (MSR). The point kinetic model has been added into the TAPAS (Tritium trAnsPort chAracteristicS code) to research the dynamic responses of tritium transport characteristics during transients. Two typical transient accident conditions for Transportable Fluoride-salt-cooled High-temperature Reactor (TFHR) are simulated, including unprotected reactivity insertion (URIA) and unprotected overcooling (UOC). The distributions of the key variables in the primary loop and variations with time are obtained including TF, T2 and Cr2+. Numerical results show that during transients, tritium has quite different transport characteristics compared with steady state, ranging from tritium production and speciation, absorption by graphite, Cr corrosion and deposition to tritium permeation. The results are discussed in comparison with earlier findings in detail. This study may provide a valuable reference for the tritium control in MSR.
Experimental Study on Flow Instability During Gravity-Driven Reflooding
Authors: Tong Lili , Zhang Dandi , Hou Liqiang , Wang Xiaoji
Published: October 2018

Due to the inherent safety characteristics of passive core injecting system, it is one of the most important mitigation strategies under loss of coolant accident (LOCA). However, flow instability during passive gravity-driven reflooding may occur, which can obstruct core reflooding process and reduce the heat removal rate from the reactor core. Therefore, exploring the characteristic and influencing mechanisms of flow instability during gravity-driven reflooding is necessary to analyze the effectiveness of passive core injecting system and mitigate the consequences of LOCA.

In this work, small scale experiments are performed to investigate the flow instability during gravity-driven reflooding when the open ratio of steam outlet are in the ranges of 3.33–50%, gravity driving head are 16–20kPa and system pressure are 0.1–0.3MPa. The test facility consists of a cooling water storage tank, a test section and a condensate tank. The results show that flow instability could occur under specific conditions and have a strong regularity, which is divided into three stages: cooling water initial injection, cooling water expulsion and cooling water re-injection. In addition, the effects of open ratio of steam outlet, gravity driving head and system pressure on flow instability phenomenon are investigated. Increasing the open ratio of steam outlet accelerates the discharge rate of steam and prevents the accumulation of steam at the inner cylinder, which can contribute to cooling the heated rods and restrict the flow instability. And under low gravity driving head of 16kPa and low system pressure of 0.1MPa the temperature and pressure acutely change and flow instability is more likely to occur.
Computational Fluid Dynamics As a Tool for Deriving Subchannel Model Parameters: The PSBT Case Study
Authors: Puragliesi R. , Mukin R. , Clifford I. , Ferroukhi H. , Seidl M.
Published: October 2018

The development of subchannel models for fuel assemblies and reactor cores requires accurate information on flow distribution, wall friction and loss coefficients in order to accurately predict the pressure, temperature and flow distribution on a subchannel level. This paper discusses the use of Computational Fluid Dynamics (CFD) simulations as a practical tool for characterising inlet velocity boundary conditions, an approximation of wall friction factor and spacer grid pressure loss coefficient, which are of fundamental importance to correctly generate a consistent subchannel model of a given assembly system. The geometry of the simplified PWR assembly presented here is based on the NUPEC PWR subchannel and bundle tests. Comparison of the derived friction factors and grid pressure loss coefficient with published and recommended values are reported. Discrepancies are also explained using additional calculations. A comparison of the overall system pressure drop, by comparing_numerical and analytical solutions, and the local axial pressure distribution at subchannel level are presented. To make a one-to-one comparison between CFD and subchannel solutions, volume-averaging is applied to the CFD results according to the chosen subchannel nodalization.

The obtained results show a perfect agreement between the two codes. This outcome reflects the correct approach employed to build two consistent numerical models by properly carrying important information from the high-resolution models (CFD) to the low-resolution models (subchannel code). Furthermore, it has been found that the large discrepancies recorded in the CFD prediction of the grid pressure loss coefficients suggested in the benchmark specifications are mainly because the suggested benchmark values do not take into account the presence of the bounding channel that was present in the experimental facility.
Numerical Investigation on Turbulent Heat Transfer of Supercritical CO2 in a Helically Coiled Tube
Authors: Huang Xiao-Rui , Zhang Zhen , Yang Xing-Tuan , Jiang Sheng-Yao , Tu Ji-Yuan
Published: October 2018

Helically coiled tubes are widely used in many industrial applications such as the steam generator in the high-temperature gas-cooled reactor which is recognized as one of the new generation advanced reactors. The thermophysical properties of fluids exhibit drastic and fast changes in the pseudocritical region so that the flow and heat transfer characteristics of supercritical pressure fluids are greatly different from those at the subcritical pressure. The paper presents results of numerical investigation on turbulent heat transfer of supercritical CO2 in a helically coiled tube with a tube diameter of 9 mm, a coil diameter of 283 mm and a coil pitch of 32 mm under the constant wall heat flux. Both the RNG k-ε model with enhanced wall function and the SST k-ω model were applied in the simulations, and the results showed that the SST k-ω model agreed better with the experimental results in the literature. Effects of buoyancy and flow acceleration were evaluated. Details of developing heat transfer characteristics at three specific cross sections were analyzed. The heat transfer regularity and mechanism presented in this work can be useful for the design and development of more economic and safer design of the supercritical steam generator.
Pressurized Thermal Shock (PTS) Transient Scenarios Screening Analysis With TRACE
Authors: Mukin Roman , Clifford Ivor , Ferroukhi Hakim , Niffenegger Markus
Published: October 2018

A number of postulated accidents that can lead a nuclear power plant (NPP) to a rapid cooling condition known as Pressurized Thermal Shock (PTS) are analysed. Such PTS, which induce high thermal stresses in the RPV wall, can occur at high or low internal pressure conditions and thus can challenge the integrity of the reactor pressure vessel. It is getting more crucial during operating of a NPP due to radiation embrittlement of the reactor pressure vessels (RPV). In this work, an overview of different transients that could contribute to the risk of a vessel failure for a pressurized water reactor (PWR) is presented. In total, 25 different scenarios were analyzed including large, intermediate and small break Loss-of-Coolant Accidents (LOCA), steam generator tube rupture (SGTR), main steam line break and stuck-open pressurizer (PZR) relief valve scenarios, with different assumptions of emergency core cooling system (ECCS) availability, temperature and operator response time. Screening of all these transients is done using a TRACE model of the reference NPP. Analysis of all transients is performed by calculating key parameters that are characterizing the cooling rate of the internal reactor vessel surface and downcomer (DC), i.e. the minimum fluid temperature in the vessel, primary system pressure and heat flux on the inside of the vessel wall. Comparisons between one-dimensional (1D) and three-dimensional (3D) cylindrical spatial representations of the downcomer and lower plenum are considered. Results of the screening analysis showed that the 3D vessel model is especially crucial during an asymmetric injection of the emergency cooling water by the safety system, for instance, in case of LOCA in a cold leg (CL). The main outcome of the LOCA screening analysis is that a large break and some intermediate break LOCAs in either the CL or hot leg (HL) lead to the most severe PTS conditions. Other analyzed transients, e.g. small break LOCA, stuck-open PZR relief valve (RV) and SGTR, are not severe enough to contribute significantly to the PTS risk, with the possible exception of a stuck-open PZR-RV valve since this has a higher probability of occurrence.
Modeling of Cerenkov-Based Fiber-Optic Gamma-Ray Radiation Sensor Using Monte Carlo Simulation
Authors: Han Hwa Jeong , Park Byung Gi , Kim Beom Kyu , Park Ji Hye , Kim Won Ki
Published: October 2018

In this study, a Monte Carlo model has been developed for a Cerenkov-based fiber-optic gamma-ray sensor (CFOGRS) using the GEANT4 simulation toolkit. The detection material for gamma rays in CFOGRS is the transparent silica core of the optical fiber, which is also used for optical signal propagation. The model implemented with the GEANT4 includes the transport process of gamma rays, as well as the physical processes of Compton scattering, the Cerenkov effect, and optical photon propagation within the optical fiber. The model also simulated the applicability of the CFOGRS in a radiation environment by using the Monte Carlo code of GEANT4.
Design for Plant Modularisation: Nuclear and SMR
Authors: Wrigley Paul , Wood Paul , Stewart Paul , Hall Richard , Robertson Dan
Published: October 2018

The UK Small Modular Reactor (UKSMR) programme has been established to develop an SMR for the UK energy market. Developing an SMR is a multi-disciplinary technical challenge, involving nuclear physics, electrical, mechanical, design, management, safety, testing to name but a few.

In 2016 Upadhyay & Jain performed a literature review on modularity in Nuclear Power. They concluded that although modularisation has been utilised in nuclear to reduce costs, more work needs to be done to “create effective modules”. Hohmann et al also concluded the same for defining modules in the chemical process plant industry.

The aim of this paper is to further define modules with a particular focus on an SMR for the UK market, the UKSMR. The methods highlighted may be relevant and applied to other international SMR designs or other types of plant.

An overview and examination of modularisation work in nuclear to date is provided. The different configurations are defined for the Nuclear Steam Supply System (NSSS) in primary circuits and then for Balance of Plant (BOP) modules. A top level design process has been defined to aid in the understanding of design choices for current reactors and to further assist designing balance of plant modules.

The paper then highlights areas for additional research that may further support module design and definition.
Research on Heat Transfer Characteristics of the Thermometric Sphere in HTR-10
Authors: Sun Shiyan , Zhang Youjie , Zheng Yanhua , Fang Xiang , Yang Xiaoyong
Published: October 2018

During the operation of the High Temperature Gas-cooled Reactor (HTGR), the hot-spot temperature in the reactor core must be lower than the maximum permissible temperature of the fuel elements and the materials of construction, so that the reactor kept safe. However, no fixed temperature-measuring devices can be set in a pebble-bed core. A special spherical temperature-measuring device is adopted to make sure it brings as small impact to the reactor operation as possible. There are several metal wires with different melting points inside. The graphite thermometric balls will be put onto the top of HTR-10 reactor core, and they record and reflect the highest temperature in different positions in the core when flowing in the pebble bed. Before the reactor core temperature-measuring experiment of HTR-10, we must study the heat transfer characteristics of the graphite thermometric sphere to find out the relationship of the melting conditions and the temperature in the reactor core. A 3-D model of the graphite thermometric ball is established, and CFD method is adopted to research and figure out the thermal equilibrium time and temperature difference between the metal wires in the ball and the hot fluid outside the balls. Multiple situations are simulated, and the heat transfer process of the thermometric sphere is comprehensively studied. The heat convection is certified the most important aspect. Thermal equilibrium can be achieved within 19 minutes, far shorter than the period while the spheres flowing through the core. The simulation results can also applied to derive the thermal fluid temperature backward.
Numerical Study on Streamwise Vorticity and Entrainment Enhancement of a Round Jet
Authors: Li Bangming , Xiao Qi , Li Yong , Hu Xu , Wang Wei
Published: October 2018

The jet pump has many outstanding advantages such as simple structure, high reliability, and low operation and maintenance costs and so on. Therefore, it has been widely used in nuclear industry fields. The entrainment ability of the jet pump is the core performance. It is generally recognized that, although the vorticity value of streamwise vortex is smaller than that of spanwise, it plays an important role in the entrainment process. Thus, a numerical simulation of streamwise vortex influence on the entrainment ability of a round jet is carried out. To enhance the entrainment of the round jet, tabs (streamwise vortex generator) and vibrator (perturbation) are set at the exit of nozzle. Tabs are set of number as 0, 2, 4 and 6, and the frequency of vibrator is set as natural frequency and its half. Based on the above configuration, numerical simulations of different nozzle types were carried out. Based on above results, firstly the influence of tabs on the entrainment ability was analyzed. The vortex intensity of the jet flow is obviously enhanced under introducing tabs. Similarly, the intensity of streamwise vortex downstream also increase significantly, thus enhance the entrainment ability. From the figure of local entrainment flow rate, the jet of nozzle with 6 tabs has the strongest entrainment ability; its local entrainment flow rate is 1.5 times of the round jet. Then, the influence of perturbation on the entrainment ability is analyzed. After introducing velocity perturbation, the evolution of the mixing layer evidently accelerated. Under natural frequency perturbation, mixing layer evolution is the most rapid. The mixing layer roll into a vortex ring in less than one diameter distance downstream allowing to 1.75 times the amount of fluid entrained in comparison with the round jet. Finally, integrating the influence of tabs and perturbation, the variation of the jet entrainment ability is analyzed. Due to the combined effect of tabs enhancing vortex intensity and natural frequency accelerating the evolution process, local entrainment ability of the jet is further improved which is more than three times that of the round jet. Taking the nozzle with 6 tabs and perturbation of natural frequency as an example, the streamwise vorticity increases rapidly to a maximum. Local entrainment flow rate also has a corresponding maximum value, which is one-to-one correspondence with the intensity distribution of streamwise vortex, indicating that streamwise vortex plays a predominant role in the entrainment ability of the jet.
Reliability Prediction for the Squib Valve of Advanced Passive PWR by Hardened Test Method
Authors: Ge Shao , Qinfang Zhang
Published: October 2018

The Squib valve is never used in the Pressurized Water Reactor (PWR) before, and it plays an important role in the advanced passive PWR accident mitigation. Thus the reliability of squib valve has received wide attention and high consideration. As a pyrotechnic mechanical device, the currently accepted approach for predicting the reliability of the squib valve requires hundreds or thousands of consecutive, successful tests of very similar components. The hardened test method for predicting the function reliability of the squib valve is presented in this paper. The squib valves are tested under hardened conditions. If they can successfully be actuated under hardened conditions, then the functional reliability under normal condition can be predicted based on the stress-strength model and one-shot product reliability assessment theory. In this way, high reliability of the squib valve can be assessed with only a limited number of tests. The hardened test method can provide considerable cost savings in testing pyrotechnical devices.
The Influence of Centrifugal Pump Characteristics on Dynamic Loadings on Pipelines After Power Failure
Authors: Marcinkiewicz Jerzy , Karaskiewicz Krzysztof , Joheman Claes
Published: October 2018

The paper presents a Relap5 study of the influence of the centrifugal pump characteristics on the dynamic loads on piping system after power failure. Interpolated and experimental pump characteristics are used.

The differences between the interpolated and measured pump curves, the general description of Relap5 model and results of calculations in form of selected time curves for rotational speed, volume flow, pressures and dynamic forces are presented and discussed.

The analysis of the results shows that the maximal dynamic force on pipe section calculated with experimental pump curves can be up to 6 % higher than respective calculated using interpolated curves. However, it is not possible to determine to what extent the differences are caused by the interpolation itself or caused by the differences in the design of the centrifugal pumps. The latter since it differs more than 50 years between the pumps whose characteristics are used for interpolation and the pumps with corresponding experimental characteristics.
Transient Reactor Test Facility Restart 23 Years Later
Authors: Heath Bradley K. , Race Cody C. , Nelson Lee O.
Published: October 2018

The Transient Reactor Test (TREAT) Facility, located at the Idaho National Laboratory (INL), is a versatile test facility able to subject experimental specimens to various transient nuclear conditions. TREAT was placed in standby after operating from February 1959 through April 1994, resulting in the loss of nearly all transient testing capability in the United States. Recently, the US Department of Energy (DOE) determined this capability was again needed. After DOE completed National Environmental Policy Act actions in February 2014, INL established the Resumption of Transient Testing Program (RTTP). RTTP was a multi-year effort to restart TREAT to reestablish a domestic transient testing capability. After 23 years of standby operations, the RTTP completed restart activities on August 31, 2017, 13 months ahead of schedule and nearly $20 million under budget. RTTP activities included an Environmental Assessment that resulted in a Finding of “No Significant Impact” associated with restarting TREAT, establishment of a compliant Safety Analysis Report (SAR), refurbishment and/or replacement of key reactor systems and components, key system knowledge recovery, reestablishment of configuration management, procedure updates, personnel training and qualification, and demonstration of operational readiness for reactor operations. Several noteworthy factors that contributed to the restart of TREAT include:

• Funding to acquire personnel and material resources provided in a timely fashion.

• Close coordination with the regulator’s (DOE) nuclear safety program during updates, interactive review, and approval of safety documentation provided for timely update of the TREAT SAR and implementing documents.

• Effective management control enabled by utilization of standard outage management techniques with a focus on age-related degradation and updated standards and requirements.

• DOE program management ensured efficient implementation of program management tools. These tools focused on clear high-level milestones and spend plans allowing flexibility for the contractor to respond to evolving facility conditions and information in a near-real time manner and with minimal program overhead. This approach enabled efficient execution of work in an environment where determination of required work scope was dependent on performance of inspection, testing, analysis, and evaluation activities.

• Implementation of the Contractor Assurance System, with frequent internal and externally-led assessments that facilitated process improvements and corrective actions to ensure the operational readiness for required contractor and DOE readiness assessments and safe nuclear operations.

• The RTTP benefited from archived plant documentation and maintenance performed while the plant was in a safe-standby status.

• Unique methods of reactivity control allowed for individual and integrated reactor system functional testing, procedure vetting, and personnel training while maintaining the reactor in a safe state.
Numerical Investigation of Corium Coolability in Core Catcher: Sensitivity to Modeling Parameters
Authors: Guo Liancheng , Rineiski Andrei
Published: October 2018

To avoid settling of molten materials directly on the vessel wall in severe accident sequences, the implementation of a ‘core catcher’ device in the lower plenum of sodium fast reactor designs is considered. The device is to collect, retain and cool the debris, created when the corium falls down and accumulates in the core catcher, while interacting with surrounding coolant. This Fuel-Coolant Interaction (FCI) leads to a potentially energetic heat and mass transfer process which may threaten the vessel integrity. For simulations of severe accidents, including FCI, the SIMMER code family is employed at KIT. SIMMER-III and SIMMER-IV are advanced tools for the core disruptive accidents (CDA) analysis of liquid-metal fast reactors (LMFRs) and other GEN-IV systems. They are 2D/3D multi-velocity-field, multiphase, multicomponent, Eulerian, fluid dynamics codes coupled with a fuel-pin model and a space- and energy-dependent neutron kinetics model. However, the experience of SIMMER application to simulation of corium relocation and related FCI is limited. It should be mentioned that the SIMMER code was not firstly developed for the FCI simulation. However, the related models show its basic capability in such complicate multiphase phenomena. The objective of the study was to preliminarily apply this code in a large-scale simulation. An in-vessel model based on European Sodium Fast Reactor (ESFR) was established and calculated by the SIMMER code. In addition, a sensitivity analysis on some modeling parameters is also conducted to examine their impacts. The characteristics of the debris in the core catcher region, such as debris mass and composition are compared. Besides that, the pressure history in this region, the mass of generated sodium vapor and average temperature of liquid sodium, which can be considered as FCI quantitative parameters, are also discussed. It is expected that the present study can provide some numerical experience of the SIMMER code in plant-scale corium relocation and related FCI simulation.
Feasibility Assessment of Air-Cooling System As an Ultimate Heat Sink of the ATOM System
Authors: Shin Doyoung , Seo Gwang Hyeok , Na Min Wook , Kim Sung Joong , Kim Yonghee , Lee Jeongik
Published: October 2018

Nowadays Small Modular Reactors (SMRs) have been receiving considerable attentions worldwide for potential advantages of an excellent flexibility for siting, low capital investment, and advanced safety. In Korea, a new research project has launched for the development of a conceptual design of a further advanced SMR which aims for a naturally-safe and autonomous operation, so called Autonomous Transportable On-demand reactor Module (ATOM). Major design objectives of the ATOM system are focused on the soluble boron-free (SBF) primary coolant system which enables the SMR to operate automatically in a load following mode. For the secondary system, the SCO2 power conversion cycle with air-cooling system as a final heat sink is being considered. The air-cooling system is expected to show flexible response even to extreme environmental conditions, such as a desert where utilization of cooling water is limited.

The objective of this study is a feasibility assessment for applying the air-cooling system as a final heat sink of the ATOM by means of experimental work. As a 1st phase of the ATOM development, we first conducted the experiments using a typically considered primary coolant, water-steam, to verify that air flow has enough cooling capability to remove developed heat which the coolant carries. An Integrated Condensation Loop with Air-cooling System (ICLASS) experimental facility with three pressure boundaries (Steam, coolant, and air) was established. The cooling capability of the air-cooling system was evaluated by varying steam mass flow rate, coolant flow rate, and air environment temperature as experiment variables. Overall heat transfer rate by condensation was compared with numerical simulations of a 1D thermal-hydraulics analysis code, using the MARS model of the ICLASS facility.
Low-Cycle Strength of Elements of Constructions
Authors: Zvorykin Alexander , Popov Roman , Bobyr Mykola , Pioro Igor
Published: October 2018

Analysis of engineering approach to the operational life forecasting for constructional elements with respect to the low-cycle fatigue is carried out. Applicability limits for a hypothesis on existence of generalized cyclic-deforming diagram in case of complex low-cycle loading (deforming) are shown.

It is determined, that under condition of plane-stress state and piecewise-broken trajectories of cycle loading with stresses and deformation checking the cyclic deforming diagram is united in limits of deformations, which are not exceeded 10 values of deformation corresponding material yield point. Generalized kinematic equation of material damageability is described.

The method of damageability parameter utilization for increasing of accuracy calculation of structural elements low-cycle fatigue by using the effective coefficients of stresses and deformations taking into account the damageability parameter is given.
Vibration Reduction Design for Radiation Material Transport Package With Finite Element Method
Authors: Dongyuan Meng , Shutang Sun , Hongchao Sun , Guoqiang Li , Lei Chen
Published: October 2018

Foam is a material used in the design of radioactive material (RM) transport package in recent years. It can undertake vibration reduction, heat insulation and some other functions. In this paper, a typical cask model was designed. With the Finite Element (FE) simulation, the stress response and energy ratio of the inner cask in 9m free drop test were calculated by changing the thickness of the foam and the weight of the content. A method to determine the foam thickness of cask design was introduced, and the next research direction was proposed.
Reactor Core Cooling Performance of a Passive Endothermic Reaction Cooling System During Design and Non-Design Basis Accidents
Authors: Murray Nathan R. , Sailsbery Mitchell E. , Bischoff Samuel E. , Wilding Paul R. , Memmott Matthew J.
Published: October 2018

A passive endothermic reaction cooling system (PERCS) is proposed to provide reactor core cooling during a station blackout (SBO). During a SBO, a PWR in which PERCS has been installed has a peak reactor core outlet temperature remains below 640 K (692.3°F) for 30 days, which is well below the nominal accident core outlet temperature during a SBO. During a LOCA, LOFA, and LOHSA, installation of a PERCS has no significant impact on safety performance. It should be noted that the PERCS will represent a minimal heat source (unless the PERCS is very large) during DBAs as emergency systems lower the coolant temperature below the PERCS temperature.

A typical PWR with an installed PERCS is modeled using RELAP5-3D. The results of the model demonstrate the high level of passive safety afforded by the PERCS which contributes to the mitigation of SBO consequences without adversely affecting nuclear plant safety during a LOCA, LOHSA, or LOFA. Future work in validating the PERCS as a method of passive safety for existing light water reactors is underway, including the refining the physical design, determining better kinetic and thermodynamic properties for MgCO3, updating the PERCS model, and using a more robust PWR plant model.
What Is a “Known and Established” Standard?
Authors: Miranda Samuel
Published: October 2018

The NRC staff, and licensees will sometimes cite a Known and Established Standard, especially when there is evidence that the staff’s position, with respect to a Known and Established Standard, may have changed. Licensees will often flag such changes as backfits, and demand that they meet the requirements of the Backfit Rule in the Code of Federal Regulations (10 CFR §50.109). However, the term, Known and Established Standard, is not clearly defined. If the term is not defined by the NRC staff, then it’s possible that it will be defined by others, most notably by licensees with technical issues or license amendment requests (LARs) that are undergoing the staff’s review.

What follows is a discussion of the origin and application of the NRC’s Known and Established Standard, and how it might be defined. The Known and Established Standard is first deconstructed into its two components, Known and Established, and then each is defined separately. When this is done, it becomes apparent that some Known and Established Standards can be more important than others. The two definitions can be used, together, to distinguish between unequal, even incompatible Known and Established Standards. This could help the NRC staff identify the most relevant Known and Established Standard to apply in its review or evaluation of a particular technical issue or license amendment request (LAR). Simply choosing the most recent of Known and Established Standards might not be adequate.
Advances in Modeling Critical Heat Flux in LWR Boiling Flows With the NEK-2P CFD Code
Authors: Tentner Adrian , Vegendla Prasad , Tomboulides Ananias , Obabko Aleks , Merzari Elia , Shaver Dillon
Published: October 2018

The paper focuses on the extension of the NEK-2P Wall Heat Transfer model, which was initially developed for the analysis of Critical Heat Flux (CHF) under Dryout (DO) conditions to the simulation of CHF under Departure from Nucleate Boiling (DNB) conditions. The paper presents results of recent NEK-2P analyses of several CHF experiments including both DO and DNB conditions. The CHF experiments analyzed have measured the axial distribution of wall temperatures in two-phase boiling flow in a vertical channel with a heated wall. The axial distribution of the calculated wall temperatures is compared with the corresponding experimental data. Reasonably good agreement with measured data is obtained in predicting the CHF location and post CHF wall temperature magnitudes illustrating the ability of the NEK-2P code and Extended Boiling Framework (EBF) models to simulate the CHF phenomena for a wide range of thermal-hydraulic conditions.
QUENCH-06 Experiment Post-Test Calculations and Integrated Uncertainty Analysis With RELAP/SCDAPSIM/MOD3.4 and MOD3.5
Authors: Allison C. , Le B. T. , Gerova G. , Spasov I. , Perez-Ferragut M. , Hohorst J.
Published: October 2018

The Quench-06 experiment (ISP-45) has been used as a benchmark and training aid for Innovative Systems Software (ISS) and our users/students since it was completed in the early 2000s. The experiment was first analyzed by several international organizations using RELAP/SCDAPSIM/MOD3.2. These results were submitted to the “blind” and “open” phases of the ISP. The experiment was subsequently used for basic user training for experimental analysis by our RELAP/SCDAPSIM/MOD3.4 and MOD3.5 users. It is also used extensively in our university support and training internships.

This paper describes an integrated uncertainty analysis of the QUENCH-06 electrically heated experiment, looking at the influence of uncertainties in experimental conditions and important models/correlations. The QUENCH calculations demonstrated the use of the new IUA, “Integrated Uncertainty Analysis”, option introduced into RELAP/SCDAPSIM/MOD3.4 in the summer of 2017 and MOD3.5 in the fall of 2017. The input models and results from both versions are discussed in the paper. The MOD3.4 results are based on the original input model developed for MOD3.2 and refined in the open phase of the ISP. The MOD3.5 results are based upon two base input models. The first was developed specifically to test the impact of MOD3.5 modeling improvements for the Quench electrically heated fuel rod simulator and was used in an early paper presented at this meeting in 2014. The second input model has been refined as part of the university support and training internship program and was used originally in 2016 to look at the influence of different approaches in modeling the insulated shroud used to minimize radial heat losses.

The uncertainty analysis provided in this paper looks at the influence of uncertainties in (a) the parabolic equations for Zircaloy oxidation, (b) the tungsten heater element resistances, (c) the convective heat transfer coefficients, (d) the contact resistance of the heater elements, and (e) the thermal conductivity of the porous zirconia used in the shroud. The uncertainty analysis demonstrated very clearly that a bias was introduced into the 2014 MOD3.5 input model. This bias was subsequently determined to be primarily associated to the modeling of the shroud and associated radial heat losses. This bias was reduced in the 2016 version of the MOD3.5 input model and will be further refined as our training activities on the modeling of experiments continue.