Search

In the study of safety enhancements on advanced sodium-cooled fast reactors (SFRs) by the Japan Atomic Energy Agency (JAEA), it has been essential to clarify the thermal hydraulics under various operating conditions at high and low flow rate conditions in a fuel assembly (FA) with wire-wrapped fuel pins to assess the structural integrity of the fuel pin that achieves a high-performance core with high burnup ratio and high power density. A finite element thermal-hydraulic analysis code named SPIRAL has been developed by JAEA to analyze the detailed thermal-hydraulic phenomena in the FA of a SFR. In this study, numerical simulations of 37-pin bundle sodium experiments at different Reynolds (Re) number conditions, including a transitional condition between laminar and turbulent flows and turbulent flow conditions, were performed to validate the developed hybrid k-ε/kθ-εθ turbulence model equipped in SPIRAL to consider the low Re number effect near the wall in the flow and temperature fields. The temperature distributions predicted by SPIRAL were consistent with those measured in the sodium experiments at the Re number conditions. Through the validation study, the applicability of the hybrid turbulence model in SPIRAL to the thermal-hydraulic evaluation of sodium-cooled FAs in a wide range of Re numbers was confirmed. [ABSTRACT FROM AUTHOR]

To support the design efforts of advanced sodium-cooled fast reactors (SFRs), a series of computational fluid dynamics (CFD) simulations are performed to investigate the pressure change along various flow passages in the proposed SFR system. The simulations are carried out with the state-of-the-art spectral element flow solver, Nek5000. Two specific case studies are presented in this paper: the flow exiting the axial neutron reflector channels and the flow entering the fuel pin bundle. Due to the high Reynolds numbers expected, a Reynolds-averaged Navier-Stokes (RANS) approach is necessary to model the turbulence. A newly developed regularized k − ω RANS model is adopted in the related CFD calculations. The first case study explores the effect of Reynolds number on the pressure change when flow exits the reflector channels. The pressure change in this case has two major contributors: the change due to wall friction and the Bernoulli effect. It is noted that the nondimensional pressure loss follows a log-linear trend up to Re = 105, and then the trend is flattened. In the second case study, the advanced NekNek coupling capability is tested where an integral domain can be divided into multiple subdomains with coupling interfaces, which would greatly ease the meshing process of complex engineering geometries and potentially save computational resources. The preliminary results obtained so far confirm the consistency between the NekNek results and those produced by regular Nek5000 simulation. The presented work demonstrates the readiness and flexibility of the related CFD techniques, which is part of the broader effort to leverage cutting-edge CFD to inform the advanced nuclear reactor designs. [ABSTRACT FROM AUTHOR]

A one-dimensional (1-D) thermal stratification (TS) model was recently developed in our research group to predict the TS phenomenon in pool-type sodium-cooled fast reactors. This paper performs uncertainty quantification (UQ) of the 1-D TS model to evaluate its performance by considering the aleatoric uncertainties that existed in the model parameters and to identify the plausible sources of the epistemic uncertainties. The Latin hypercube sampling-Monte Carlo method (LHS-MC), which is elaborated with an example in this paper to facilitate its understanding and implementation, is used for the UQ process. The advantages of LHS-MC, including both better stability and better accuracy than the conventional random sampling-Monte Carlo method with fewer realizations, are demonstrated in this paper. In total, 648 temperature measurements acquired from nine experimental transients performed in a university-scale Thermal Stratification Experimental Facility are used to evaluate the performance of the computational 1-D TS model. The UQ result shows that 77.5% of the experimental data can be predicted by the 1-D TS model within uncertainty ranges, which indicates the good performance of the computational model when the aleatoric uncertainties are correctly captured. The rest 22.5% of the experimental data are found located outside of the uncertainty ranges, which reveals the existence of the epistemic uncertainties caused by the lack of understanding of the TS phenomenon and defects in the 1-D model. The simple jet model currently employed by the 1-D TS model is thought to be one of the attributors to these defects. [ABSTRACT FROM AUTHOR]

Thermal stratification (TS) is a thermal-fluid phenomenon that can introduce large uncertainties to nuclear reactor safety. The stratified layers caused by TS can lead to temperature oscillations in the reactor core. They can also result in damages to both the reactor vessel and in-vessel components due to the growth of thermal fatigue cracks. More importantly, TS can impede the establishment of natural circulation, which is widely used for passive cooling and ensures the inherent safety of numerous reactor designs. A fast-running one-dimensional (1-D) model was recently developed in our research group to predict the TS phenomenon in pool-type sodium-cooled fast reactors. The efficient 1-D model provided reasonable temperature predictions for the test conditions investigated, but nonnegligible discrepancies between the 1-D predictions and the experimental temperature measurements were observed. These discrepancies are attributed to the model uncertainties (also known as model bias or errors) in the 1-D model and the parameter uncertainties in the input parameters. In this study, we first recognized through a forward uncertainty analysis that the observed discrepancies between the computational predictions and the experimental temperature measurements could not be explained solely by input uncertainty propagation. We then performed an inverse uncertainty quantification (UQ) study to reduce the model uncertainties of the 1-D model using a modular Bayesian approach based on experimental data. Inverse UQ serves as a data assimilation process to simultaneously minimize the mismatches between the predictions and experimental measurements, while quantifying the associated parameter uncertainties. The solutions of the modular Bayesian approach were in the form of posterior probability density functions, which were explored by rigorous Markov Chain Monte Carlo sampling. Results showed that the quantified parameters obtained from the inverse UQ effectively improved the predictive capability of the 1-D TS model. [ABSTRACT FROM AUTHOR]

When pressurized water or vapor leaks from a failed heat transfer tube in a steam generator (SG) of sodium-cooled fast reactors, a high-velocity, high-temperature jet with sodium-water chemical reaction may cause wastage on the adjacent tubes. For safety assessment of the SG, a computational fluid dynamics code SERAPHIM, in which a compressible multicomponent multiphase flow with sodium-water chemical reaction is computed, has been developed. The original SERAPHIM code is based on the finite difference method. In this study, an unstructured mesh-based numerical method was developed and introduced into the SERAPHIM code to advance a numerical accuracy for a complex-shaped domain including multiple heat transfer tubes. The multiphase flow under the tube failure accident is calculated by the multifluid model considering compressibility. The governing equations are solved by the Highly Simplified Marker And Cell (HSMAC) method. The original HSMAC method was modified for compressible multiphase flows in the unstructured mesh. Validity of the unstructured mesh-based SERAPHIM code was investigated through the analysis of an underexpanded jet experiment, which is a key phenomenon in the tube failure accident. The calculated pressure profile showed good agreement with the experimental data. Numerical analysis of water vapor discharging into liquid sodium was also performed. The calculated behavior of the reacting jet agreed with the previous experimental knowledge. It was demonstrated that the proposed numerical method could be applicable to evaluation of the sodium-water reaction phenomenon. [ABSTRACT FROM AUTHOR]

Fuel-coolant interaction (FCI) is an important issue for the assessment of severe accident safety for both sodium-cooled fast reactors (SFRs) and pressurized water reactors (PWRs). For the ASTRID SFR demonstrator, FCI is a key phenomenon affecting the relocation of molten fuel in engineered discharge tubes between the core region and the core catcher plenum. FCI controls jet fragmentation and debris bed formation and raises the issue of potentially energetic vapor explosions in the ASTRID lower head. In this frame, experimental data will be necessary to validate SCONE, the fuel-sodium interaction code under development at CEA. For PWRs, one of the configurations of interest lies within the residual case where in-vessel retention would fail. In this case, it is expected that a light metallic layer would be the first to interact with water, before a heavier oxide melt discharge. Here, steam explosion and debris bed formation are the two major points of interest. Based on the experimental expertise gained from the KROTOS facility and its X-ray radioscopic imaging system, new test facilities have been designed to carry out prototypic (depleted uranium-containing) corium interactions with either sodium or water in PLINIUS2, the CEA future large-mass experimental platform dealing with masses above 100 kg. Some test sections have been specially designed to ensure proper visualization of the fuel, liquid coolant, and vapor phases by an improved X-Ray imaging system. This paper presents the future PLINIUS 2 platform as well as the experimental programs foreseen to study both water-corium and sodium-corium interactions. [ABSTRACT FROM AUTHOR]

Development of an inspection technique in opaque liquid-metal coolant is one of the important issues to ensure the safety of the liquid-metal fast breeder reactor (LMFBR). Performance tests of an under sodium viewer (USV), which was developed to detect an obstacle in the reactor vessel (RV) of the LMFBR Monju, have been carried out. Ultrasonic sensors and reflectors are located across the core inside Monju's RV. The USV can detect an obstacle existing between the core top and the upper core structure bottom by identifying differences of echo signals. This paper describes the USV performance tests. In the tests, the reference echo signals under various conditions were accumulated, and the signal-to-noise ratio successfully exceeded the target value. Measured signals clearly differed with and without an obstacle. These experimental results show the performance of the USV for detecting an obstacle in a specified place. [ABSTRACT FROM AUTHOR]

Severe external hazards on the Japan Sodium-cooled Fast Reactor (JSFR) have been analyzed and evaluated. For seismic design, safety components are confirmed to maintain their functions even against recent strong earthquakes. Integrity of the major components has been confirmed covering recent earthquake conditions. In the case of a tsunami, the seawater pumps for the component cooling water system (CCWS) could be damaged by the tsunami, since they are located at sea level. In the JSFR design with full natural convection decay heat removal systems (DHRSs) and an air-cooling emergency gas turbine, safety-grade components are independent of CCWS, and loss of CCWS does not affect reactor cooling. As a consei-vative case, hypothetical station blackout (SBO) has been evaluated. In the case ofSBO, decay heat is removed by natural convection DHRS, but control of the air cooler (AC) damper is lost after the battery power is out. The analysis has revealed that freezing at one of three ACs could happen due to loss of automatic control of AC dampers. However, the time margin to protected loss of heat sink is evaluated to be >10 days. Manual control of the AC damper is also investigated. Transient analyses show that the AC dampers can be controlled manually adopting a simple operation procedure with sufficient operation time. Decay heat cooling in the case of collapse in all air stacks of AC has been evaluated. The result shows that decay heat could be removed maintaining air paths in two of three ACs by accident management. In conclusion, JSFR in the 2010 design version has enough external hazard toughness mainly thanks to passive safety features and a seismic isolation system. [ABSTRACT FROM AUTHOR]

In response to the accident at the Fukushima Daiichi (IF) nuclear power plant, designers of the 2010 version of the Japan sodium-cooled fast reactor (JSFR) have been studying the robustness of JSFR's fuel handling system (FHS) against an earthquake and a tsunami. In the earthquake evaluation, the margin of seismic resistance and sloshing in the ex-vessel fuel storage tank (EVST) against an earthquake of the IF-envelop condition were estimated. In the tsunami evaluation, for the case of loss of emergency power supply, a scenario is studied where fuel subassemblies are led to a stable cooling state, and potentialities for the cooling system are examined. As a result, it is shown that the original design of the JSFR FHS already had the potential to prevent the release of radioactive material. Additionally, some design improvements of the EVST cooling system are introduced. [ABSTRACT FROM AUTHOR]

As part of the Fast Reactor Cycle Technology Development (FACT) Project, JSFR (Japan Sodium-Cooled Fast Reactor) core design efforts have been made to cope with the transuranic (TRU) fuel compositions expected during the light water reactor (LWR)-to-fast breeder reactor transition period, during which various kinds of TRU fuel compositions are available depending on the characteristics of the LWR spent fuels and their recycling method. The sodium void reactivity, which is one of the major core safety parameters, is considerably influenced by TRU fuel compositions. The criteria assigned to the JSFR core include a void reactivity effect limited to ~ 6 $; therefore, designing a core with reduced sodium void reactivity will offer a greater margin for the core to host TRU fuel. To this end, a new core concept called BUMPY is proposed. This homogeneous core exhibits a low sodium void reactivity, due to partial-length fuels with an upper sodium plenum interspersed within the core, among other standard fuel assemblies. This core configuration enhances the upward and lateral neutron leakage from the core fuel region toward the sodium plenum when voiding to reduce void reactivity. The BUMPY core is applied to the 750-MW(electric) JSFR core design. The core can meet the design target by adjusting the loading fraction of the partial-length fuels and the height of the step in fuel lengths. The calculated void reactivity of the selected BUMPY core is 2.5 $ (25% loading fraction, 30-cm step height), which is considerably reduced from the 5.3 $ value of the reference core. This allows the BUMPY core to accommodate 5% to 9% more minor actinides in the core compared to the reference core. [ABSTRACT FROM AUTHOR]

The major roles of a prototype sodium-cooled fast reactor (SFR) planned to be developed at the Korea Atomic Energy Research Institute are (a) to provide an irradiation test capability for fuel and structural materials and (b) to obtain operational experience on the systems and components. The power level of the prototype SFR should be large enough to provide an appropriate irradiation test environment. Trade-off studies were therefore performed from a neutronics viewpoint to determine the power level. Specifically, core designs were performed for power levels of 125, 250, 400, and 500 MW(thermal). The selected core performance and economic efficiency indices became insensitive to the power at ~400 to 500 MW (thermal) and sharply deteriorated at ~125 to 250 MW (thermal) with decreasing core sizes. For the fuel management scheme, the transuranic (TRU) core performance compared with that of the uranium core, and the sodium void reactivity, were also evaluated with increasing power levels. It was found that increasing the number of batches shows a higher-burnup performance and economic efficiency. However, increasing the cycle length resulted in a lower economic efficiency. The irradiation performance of TRU and enriched TRU cores was improved by ~20% and 50%, respectively. A maximum sodium void reactivity of 5.2 $ was confirmed as less than the design limit of 7.5 $. As a conclusion of our entire study, the power capacity of the prototype SFR should not be <250 MW (thermal), and would be appropriate at ~500 MW (thermal) considering the performance and economic efficiency. [ABSTRACT FROM AUTHOR]

A conceptual design study of Japan Sodium-cooled Fast Reactor (JSFR) is in progress in the Fast Reactor Cycle Technology Development (FACT) project in Japan. hz the design of JSFR, a selector valve mechanism is adopted for its failed-fuel detection and location (FFDL) system. Since JSFR has only two FFDL units for ~600 fuel subassemblies due to its compact design of the reactor vessel, one FFDL unit must handle a much larger number of subassemblies than in previous designs. In addition, during the long plant life of 60 years, the wear length of the selector valve will become longer than those of past reactors. Therefore, the endurance of the selector valve becomes important. To demonstrate the manufacturability and endurance of the selector valve, a full-size mockup valve including coating to protect the sliding mechanism was manufactured, and an endurance experiment of the mockup model under high-temperature sodium was conducted. Dimensional inspections and seal performance showed manufacturability of the selector valve. The cross-section observation, hardness measurement, and chemical assay results after the endurance experiment showed that the coating layer on the sliding surface still remains. Thus, the endurance of the JSFR selector valve was demonstrated. [ABSTRACT FROM AUTHOR]

The Commissariat it L'Energie Atomique et aux Energies Alternatives (CEA) and the Japan Atomic Energy Agency (JAEA) intend to develop prototype or demonstration sodium-cooled fast reactors (SFRs) within two decades. The common final goal of their respective programs is SFR commercialization. The target of a commercial SFR for CEA and JAEA is basically consistent with Generation IV goals. Based on their industrial backgrounds and feedback from past and existing reactor experiences, CEA and JAEA have selected pool and loop configurations for the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) and the Japanese Sodium-Cooled Fast Reactor (JSFR), respectively. CEA and JAEA have cross-analyzed both pool and loop concepts (ASTRID and demonstration JSFR, respectively). The analysis results show that both concepts are technologically feasible and meet design goals. From the viewpoint of collaboration, the analysis identifies a wide range of collaborative areas in several fields: design principles (e.g., design target and design standard); development of components and systems; development of component parts, devices, or subsystems; design methods; simulation tools; etc. [ABSTRACT FROM AUTHOR]

Key technologies for Japan Sodium-cooled Fast Reactor (JSFR) have been evaluated. The ten technologies--high-burnup core, safety enhancement, compact reactor vessel, two-loop cooling system using high-chromium steel, integrated intermediate heat exchanger/pump component, reliable steam generator, natural-circulation decay heat removal system, simplified fuel handling system, containment vessel made of steel plate-reinforced concrete, and advanced seismic isolation system--have been confirmed to be ready for installation to a conceptual design of demonstration JSFR and large-scale demonstration experiments. [ABSTRACT FROM AUTHOR]

An axially heterogeneous sodium-cooled fast reactor design is developed for converting minor actinide waste isotopes into plutonium fuel. The reactor design incorporates zirconium hydride moderating rods in an axial blanket above the active core. The blanket design traps the active core's axial leakage for the purpose of transmuting 241Am into 238pu. This 238pu is then co-recycled with the spent driver fuel to make new driver fuel. Because 238pu is significantly more fissionable than 241Am in a fast neutron spectrum, the fissile worth of the initial minor actinide material is upgraded by its preconditioning via transmutation in the axial targets. Because the 241Am neutron capture worth is significantly greater in a moderated epithermal spectrum than the fast spectrum, the axial targets serve as a neutron trap that recovers some of the axial leakage lost by the active core. A low transuranic conversion ratio is achieved by a degree of core flattening that increases axial leakage. Unlike a traditional "pancake" design, neutron leakage is recovered by the axial target/blanket system. This heterogeneous core design is constrained to have sodium void and Doppler reactivity worth similar to that of an equivalent homogeneous design. Contrary to a homogeneous design, concentrating minor actinides (MAs) in an axial blanket mitigates the problem of above-threshold multiplication during sodium voiding. Because minor actinides are irradiated only once in the axial target region, elemental partitioning of the minor actinides from plutonium is not required. This fact enables the use of metal targets with pyroprocessing. After reprocessing, the target's newly bred 238pu and remaining unburned MAs become the feedstock for the next batch of driver fuel. [ABSTRACT FROM AUTHOR]

In the Fast Reactor Cycle Technology Development (FaCT) project, a sodium-cooled fast reactor (SFR) with mixed-oxide (MOX) fuel and an SFR with metal fuel were selected as the primary and the secondary candidates, respectively, for the Japan Sodium-Cooled Fast Reactor (JSFR). The present study focuses on the effects of transuranium (TRU) composition in the design for the JSFR core with MOX fuel. In the transitional stage from light water reactor (LWR) to fast breeder reactor (FBR), there is the possibility for FBR fuel to have high minor actinide (MA) content due to the recycling of LWR spent fuel. High MA content affects core and fuel designs as follows: the neutronic reactivity characteristic changes; the linear power limit is reduced because of decreases of the melting point and thermal conductivity in the fuel; the gas plenum length is extended because of an increase in He gas generation. Thus, to evaluate the effects quantitatively, design studies for cores with two TRU compositions were conducted: an FBR multirecycle composition with ∼1 wt% (in heavy metal) of MA content and an LWR recycle composition for which 3 wt% of MA content was assumed as a tentative target. The results show that the change from the FBR multirecycle composition to the LWR recycle composition leads to a sodium void reactivity increase of 10%, a linear power limit decrease of 1 to 2%, and a gas plenum length increase of 5%. As a result, the effects of TRU composition on the core and fuel designs were revealed to be benign. [ABSTRACT FROM AUTHOR]

This paper describes the current status of the design study and the related research and development for an advanced loop-type fast reactor: the Japan sodium-cooled fast reactor (JSFR). First, the development targets and the major design requirements are established. Then, a cost-down approach in which developing innovative technologies is key to being competitive with another future energy source is discussed. Here, the development status of several innovative technologies such as a two-loop cooling system, reliable reactor system, simplified fuel-handling system, passive reactor shutdown system, mitigation measure against a core disruptive accident, and minor actinide-bearing oxide fuel core is described. Last, a review of JSFR development and the demonstration plan for the innovative technologies are discussed. [ABSTRACT FROM AUTHOR]

Annular sodium-cooled fast reactor cores [600 MW(electric)] with low sodium void worth are developed for burning transuranic nuclides discharged from light water reactors. The several core design variants are developed by changing the core configuration, the core height, the fuel assembly design and type of nonfuel assemblies in the core, and their core performance parameters including safety-related reactivity coefficients are analyzed and inter-compared. The study focuses on the core neutronic parameters without going into the detailed safety and material compatibility studies. The study shows that the several cores of the annular type can be designed to have low sodium void worth, high transmutation capability, and all the negative temperature reactivity coefficients except for the positive one related to coolant expansion that can be compensated for by the reactivity coefficients by the fuel axial expansion and the fuel Doppler effects under the off-normal events, which increase temperatures. Of the cores considered, the use of a larger central control region and fuel assemblies with high coolant flow area in the core boundaries is found to be the most effective and simple way to achieve low sodium void worth and high transmutation capability. [ABSTRACT FROM AUTHOR]