Gen-IV Sodium-cooled Fast Reactor Research Articles

A validation study of the SLTHEN code for hexagonal assemblies of wire-wrapped pins using liquid metal heating experiments

This paper presents a validation study of the subchannel analysis code SLTHEN used for the core thermal-hydraulic design of the Prototype Gen-IV sodium-cooled fast reactor (PGSFR). To assess the performance of the ENERGY model of SLTHEN, four liquid metal heating experiments conducted by ORNL, WARD, and KIT with hexagonal assemblies of wire-wrapped rod bundles were analyzed. These experiments were performed with 19- and 61-pin bundles and varying power distributions of axial and radial peaking factors up to 1.4 and 3.0, respectively. The coolant subchannel temperatures measured at different axial locations were compared with the SLTHEN predictions with the Novendstern, Chiu-Rohsenow-Todreas (CRT), and Cheng-Todreas (CT) correlations for flow split and mixing in wire-wrapped pin bundles. The results showed that the SLTHEN predicts the measured subchannel temperatures reasonably well with root-mean-square errors of ∼10 % and maximum errors of ∼20 %. It was also observed that the CRT and CT correlations consistently outperform the Novendstern correlation.

Validation of applicability of induction bending process to P91 piping of prototype Gen-IV sodium-cooled fast reactor (PGSFR)

The application of the induction bending process to pipe systems in various industrial fields is increasing. Recently, efforts have also been made to apply this bending process to nuclear power plants because it can innovatively reduce welded parts of the curved pipes, such as elbows. However, there have been no cases of the application of induction bending to the piping of nuclear power plants.In this study, the applicability of the P91 induction bending piping for the sodium-cooled fast reactor PGSFR was validated through high temperature low cycle fatigue tests and creep tests using P91 induction bending pipe specimens. The tests confirmed that the materials sufficiently satisfied the fatigue life and the creep rupture life requirements for P91 steel at 550 °C in the ASME B&PV Code, Sec. III, Div. 5. The results show that the effects of heating and bending by the induction bending process on the material properties were not significant and the induction bending process could be applicable to piping system of PGSFR well.

Numerical analysis of temperature fluctuation characteristics associated with thermal striping phenomena in the PGSFR

Thermal striping is a complex thermal–hydraulic phenomenon caused by fluid temperature fluctuations that can also cause high-cycle thermal fatigue to the structural wall of sodium-cooled fast reactors (SFRs). Numerical simulations using large-eddy simulation (LES) were performed to predict and evaluate the characteristics of the temperature fluctuations related to thermal striping in the upper internal structure (UIS) of the prototype generation-IV sodium-cooled fast reactor (PGSFR). Specific monitoring points were established for the fluid region near the control rod driving mechanism (CRDM) guide tubes, CRDM guide tube walls, and UIS support plates, and the normalized mean and fluctuating temperatures were investigated at these points. It was found that the location of the maximum amplitude of the temperature fluctuations in the UIS was the lowest end of the inner wall of the CRDM guide tube, and the maximum value of the normalized fluctuating temperatures was 17.2%. The frequency of the maximum temperature fluctuation on the CRDM guide tube walls, which is an important factor in thermal striping, was also analyzed using the fast Fourier transform analysis. These results can be used for the structural integrity evaluation of the UIS in SFR.

Assessment of subchannel flow mixing coefficients for wire-wrapped hexagonal fuel rod bundles

Flow mixing between adjacent subchannels within a wire-wrapped hexagonal fuel rod bundle affects radial heat transfer and determines the maximum cladding temperature, which is a key parameter to ensure the fuel safety margin in a sodium-cooled fast reactor (SFR). In the ENERGY model, both the effective eddy diffusivity and the edge swirl velocity ratio are employed to characterize the subchannel flow mixing. The Korea Atomic Energy Research Institute conducted subchannel flow mixing tests on the 37-pin and 61-pin bundle assemblies. The test assemblies were fabricated by reflecting thermal–hydraulic similarities of the Prototype Gen-IV SFR candidate cores. The subchannel flow mixing experiment was conducted using a wire-mesh sensing system. The subchannel flow mixing coefficients are determined by the least-square method so that the difference between the SLTHEN code prediction and experimental data is minimized numerically. The results show good agreement with the previous correlations, especially with the Cheng-Todreas model.

Numerical analysis of the temperature distribution of the EM pump for the sodium thermo-hydraulic test loop of the GenIV PGSFR

The temperature distribution of an electromagnetic pump was analyzed with a flow rate of 1380 L/min and a pressure of 4 bar designed for the sodium thermo-hydraulic test in the Sodium Test Loop for Safety Simulation and Assessment-Phase 1 (STELLA-1). The electromagnetic pump was used for the circulation of the liquid sodium coolant in the Intermediate Heat Transport System (IHTS) of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) with an electric power of 150 MWe. The temperature distribution of the components of the electromagnetic pump was numerically analyzed to prevent functional degradation in the high temperature environment during pump operation. The heat transfer was numerically calculated using ANSYS Fluent for prediction of the temperature distribution in the excited coils, the electromagnet core, and the liquid sodium flow channel of the electromagnetic pump. The temperature distribution of operating electromagnetic pump was compared with cooling of natural and forced air circulation.The temperature in the coil, the core and the flow gap in the two conditions, natural circulation and forced circulation, were compared. The electromagnetic pump with cooling of forced circulation had better efficiency than natural circulation even considering consumption of the input power for the air blower. Accordingly, this study judged that forced cooling is good for both maintenance and efficiency of the electromagnetic pump.

Design and evaluation of reactor vault cooling system in PGSFR

In the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR), a Reactor Vault Cooling System (RVCS) is prepared to maintain the structural integrity of the reactor vessel and the concrete during normal operation as well as to mitigate accidents during severe accident conditions. In this research a series of analysis was described during the RVCS design refinement process. The one-dimensional design code analysis was used to examine various design parameters that are able to enhance RVCS cooling capability without deteriorating RVCS performance in severe accident condition. As a results RVCS gap distance was decreased by 10 mm. Multi-dimensional thermal–hydraulic CFD analysis were utilized to analyze the real geometry effects on flow field. Based on the analysis, RVCS design refinement has been done within an acceptable range. In order to evaluate the RVCS design refinement to design requirement under more realistic condition, a Head Access Area (HAA)-RVCS integrated reactor system model was built and an integrated simulation was performed at normal operating conditions. It was found that proposed RVCS fulfills the design requirements of normal operation condition.

Design of A scale-down experimental model for SFR reactor vault cooling system performance analyses

We propose a scaled-down experimental model of vertical air-natural convection channels by applying the modified Ishii–Kataoka scaling method with the assistance of numerical analyses to the Reactor Vault Cooling System (RVCS) of the Proto-type Gen-IV Sodium-cooled fast reactor (PGSFR) being developed in Korea. Two major non-dimensional numbers (modified Richardson and Friction number) from the momentum equation and Stanton number from the energy balance equation were identified to design the scaled-down experimental model to assimilate thermal-hydraulic behaviors of the natural convective air-cooling channel of RVCS. The ratios of the design parameters in the PGSFR RVCS between the prototype and the scaled-down model were determined by setting Richardson and Stanton number to be unity. The friction number which cannot be determined by the Ishii-Kataoka method was estimated by numerical analyses using the MARS-KS system code. The numerical analyses showed that the friction number with the form loss coefficient of 2.0 in the scale-down model would result in an acceptable prediction of the thermal-hydraulic behavior in RVCS. We also performed experimental benchmarking using the scaled-down model with the MARS-KS simulations to verify the appropriateness of the scale-down model, which demonstrated that the temperature rises and the average air flow velocity measured in the scale-down model.

Optimization of outer core to reduce end effect of annular linear induction electromagnetic pump in prototype Generation-IV sodium-cooled fast reactor

An annular linear induction electromagnetic pump (ALIP) which has a developed pressure of 0.76 bar and a flow rate of 100 L/min is designed to analysis end effect which is main problem to use ALIP in thermohydraulic system of the prototype generation-IV sodium-cooled fast reactor (PGSFR). Because there is no moving part which is directly in contact with the liquid, such as the impeller of a mechanical pump, an ALIP is one of the best options for transporting sodium, considering the high temperature and reactivity of liquid sodium. For the analysis of an ALIP, some of the most important characteristics are the electromagnetic properties such as the magnetic field, current density, and the Lorentz force. These electromagnetic properties not only affect the performance of an ALIP, but they additionally influence the end effect. The end effect is caused by distortion to the electromagnetic field at both ends of an ALIP, influencing both the flow stability and developed pressure. The electromagnetic field distribution in an ALIP is analyzed in this study by solving Maxwell's equations and using numerical analysis.

Design and validation of a fuel assembly simulator for PGSFR reactor flow distribution test facility

A test facility to investigate the flow characteristics inside reactor vessel for the Prototype Gen-IV Sodium-cooled Fast Reactor was constructed. In the test facility, reactor vessel and the main in-vessel components are linearly reduced at a scaling ratio of 1/5 and water is used as the working fluid. In the reactor vessel of the test facility, the main components such as core, UIS, PHTS pump and IHX are installed. The exteriors of the main components are conserved following the scaling ratio of 1/5, but the internal flow paths inside the fuel assemblies and IHXs are uniquely designed for precise measurement of flow rate and conserving the pressure drop characteristics. In order to determine the configuration and specific dimension of the internal flow path, separate analysis and experiment for validation are required. In the present paper, the basic design concept of internal flow path for fuel assembly simulator used for the reactor flow distribution test was established, and the detailed size of main design factors were estimated using commercial CFD code. The internal flow path of the simulator is composed of a receptacle, variable-resistance rotating orifice spool, venturi tube, and connection lines. The flow rate through the simulator could be estimated by measuring the differential pressure between inlet section and the throat of the venturi tube. The orifice spool is installed at the downstream of the receptacle, by which entire pressure drop through the simulator is controlled. A series of CFD analysis was conducted to estimate the throat diameter of the venturi tube and the size of the holes at the orifices. The geometry of the orifice is determined to obtain target pressure drop when the angle between two orifice plates is from 0° to 45°. The design specifications were applied to the fabrication of the fuel assembly simulators, and the performance of them was verified experimentally. The pressure drop of 112 fuel assembly simulators was successfully adjusted to be within ±1% of the target pressure drop. The relationship between the mass flow rate and differential pressure of the venturi tube was also obtained, and empirical correlation was suggested.

Overview of the SAS4A Metallic Fuel Models and Extended Analysis of a Postulated Severe Accident in the Prototype Gen-IV Sodium-Cooled Fast Reactor

The SAS4A safety analysis code, originally developed for the analysis of postulated severe accidents in oxide fuel sodium-cooled fast reactors (SFRs), has been significantly extended to allow the mechanistic analysis of severe accidents in metallic fuel SFRs. The SAS4A metallic fuel models simulate the metallic fuel thermomechanical and chemical behavior and track the evolution and relocation of multiple fuel and cladding components during the pretransient irradiation and during the postulated accident, allowing an accurate description of the changes in the local fuel composition. The local fuel composition determines the fuel thermophysical properties, such as freezing and melting temperatures, which in turn affect the fuel relocation behavior and ultimately the core reactivity and power history during the postulated accidents. Models describing the fuel-cladding interaction and eutectic formation, the effects of the in-pin sodium on the in-pin fuel relocation, and the postfailure reentry of the molten fuel and fission gas from the pin plenum have also been added. This paper provides an overview of the SAS4A key metallic fuel models emphasizing the postfailure metallic fuel relocation models included in the LEVITATE-M module of SAS4A. The capabilities of the SAS4A metallic fuel models are illustrated through an extended SAS4A analysis of a postulated unprotected loss-of-flow and transient-overpower accident in the metallic fuel prototype Gen-IV sodium fast reactor. The results show that the maximum relative power reached during the postulated accident is 1.19 P0. The favorable characteristics of the metallic fuel cause a significant decrease in net reactivity and relative power due to prefailure in-pin fuel relocation. Negative net reactivity values persist after cladding failure, and the postfailure fuel relocation events occur at low and decreasing power levels.

Experimental identification of pressure drop characteristics at the shell side of an intermediate heat exchanger inside a prototype generation-IV sodium-cooled fast reactor

The flow characteristics at the shell side of a prototype intermediate heat exchanger (p-IHX) in a prototype generation-IV sodium-cooled fast reactor (PGSFR) were investigated experimentally in this study. The shell side of the p-IHX consisted of an inlet window, tube bundles, grid plates, and an exit flow channel. Two different test facilities were newly constructed to identify the pressure drop characteristics due to the complicated configuration with minimized scaling distortion. An intermediate heat exchanger test loop for PGSFR (iHELP) was constructed to characterize the flow resistance of the tube bundle regions with the grid plates and inlet window. The iHELP test section with a slab-shaped cross section was manufactured with a 1/29.6 vol scale ratio by preserving the height of the p-IHX. The hydraulic diameter at the tube bundle regions, the porosity of the grid plates, and the configuration of the tube bundles were conserved. In addition, the Reynolds number was preserved, which determined the flow rate conditions for the iHELP experiments. To investigate the pressure drop characteristics of the exit flow channel (EC), an intermediate heat exchanger test loop for the exit flow channel (IEC) was built with a 1/5 linearly reduced length scale by preserving the geometry of the flow path based on the EC of the p-IHX. The Euler number was conserved with a 1/8 Reynolds number ratio corresponding to a 1/1 flow velocity ratio to the p-IHX for dynamic similarity. By using each test facility the experimental pressure drop data satisfying the experimental uncertainty requirements were obtained separately for the bundle and exit channel regions. The pressure drop at the shell side of the p-IHX under a wide range of flow rate conditions was determined by using the pressure drop results from the iHELP and the IEC test facilities, which would be useful in improving and validating the pressure drop correlations in the IHX design computational code (SHXSA).

Description and thermal simulation of the DIAMINO irradiation experiment of transmutation fuel in the OSIRIS reactor

French GENIV sodium-cooled fast reactor (SFR) system could include Americium (Am) recycling in the radial fertile blankets located in the outer core regions. An issue related to this option is the lack of knowledge regarding the in-pile behaviour of the Am-bearing blanket (AmBB). The DIAMINO separate-effect experiment was irradiated in the OSIRIS Material Test Reactor (MTR) with the goal of studying AmBB gaseous release and swelling as a function of temperature, AmBB microstructure and gas production rate.A brief description of the experiment is first provided, in terms of objectives and planning. The innovative design developed for the DIAMINO experiment is described in details. The DIAMINO irradiation time history is then summarized and the first results of non-destructive post-irradiation examinations are given.The LICOS fuel design code of the PLEIADES platform was used for the thermo-mechanical modelling of the DIAMINO experiment. The Pleiades platform, co-developed by CEA and its industrial partners EDF and FRAMATOME, gathers several fuel performance codes, each generally dedicated to the fuel element of a particular nuclear system. For its part, the LICOS code allows performing studies on innovative fuel elements or on the design of irradiation devices for instance. The thermo-mechanical modelling of the DIAMINO experiment device using the LICOS code is detailed in the paper together with the validation of the computations on the basis of some experimental temperature measurements performed on an in-pile mock-up of the experiment. The use of the simulation tool for the monitoring of the AmBB sample temperature under irradiation and for the post-irradiation calculations is explained and illustrated through some examples.

Validation of the MCNP6 code for SFR shielding design analysis

In this paper, the uncertainty of the Monte Carlo code MCNP6 for the sodium-cooled fast reactor (SFR) shielding design is studied. Shielding analysis, which ensures the radiation safety of the core design, is challenging for the Monte Carlo modeling because it is associated with large uncertainties. In order to evaluate the performance of the MCNP6 relative to the shielding design of the SFR, four SFR shielding benchmarks from the Shielding Integral Benchmark Archive Database benchmark suite, i.e. JANUS Phase VIII, SDT12, EURAC_Na and HARMO_Na were selected and analysed. In this research, the weight window variance-reduction technique and the neutron data library ENDF/B-VII.1 were used in the modeling of the benchmark problems. The results and the validation of the MCNP6 models with the available measurement data are presented in this paper. These results contribute to the assessment of radiological protection and shielding design of the Korean Prototype Gen-IV Sodium-cooled Fast Reactor.

Assessment of ex-core neutron detectors in prototype Gen-IV sodium-cooled fast reactor (PGSFR)

This paper presents the assessment of the ex-core detectors in the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) in response to the movement of single primary control assembly, to evaluate the detectors’ locations and to determine a proper number of detectors. The detectors are located below the containment vessel and distributed in different ranges from the center and different azimuthal angles. Another position considered is in the radial direction of the core, at the sodium coolant gap between the outermost assembly and core shroud. To calculate the detection rate, the contribution of each fuel region to the detector and the core power distribution are required. Initially, the aforementioned contribution is generated by performing a continuous-energy forward fixed-source Monte Carlo calculation using MCNPX. The reactor kinetics code FREK simulates the power change due to an inadvertent withdrawal of a primary control assembly up to 10% from its critical position followed by full insertion of secondary control assemblies. The response from each detector location was evaluated, and consequently, the detectors located below the containment vessel, which can “see” all the fuel assemblies, responded better to the core power increase with a minor error of less than ∼0.6%. In addition, the impact of locating the detector below the containment vessel at different locations from the center or at different azimuthal angles was insignificant but several detectors should be distributed at even azimuthal angles for effective core monitoring and redundancy purposes.

Component design and accident analysis of fuel assembly for prototype GEN-IV sodium-cooled fast reactor

Abstract The Korea 150 MWe Prototype GEN-IV Sodium cooled Fast Reactor (PGSFR) project has reached the middle of its preliminary safety analysis report (SAR) design phase. The core design studies are being conducted by KAERI. Some design features for the core have been made to comply with the GEN-IV reactor objectives. The main objective to improve the safety of the core design. A negative sodium void worth was achieved thanks to a new fuel assembly design including U-Zr and U-TRU-Zr axially heterogeneous fuel rods, a large cladding/small spacer wire bundle, a sodium bonding between metal slug and cladding, and upper gas plenum above the active core region, and upper/lower reflectors. Finite element calculations were performed to increase the stiffness of the above core load pad and top load pad in order to analyze their effect on the core mechanical behavior during hypothetical radial core flowering and compaction events. More generally, all design choices for the PGSFR have been made with the permanent objective of minimizing the fuel assembly height to decrease the overall costs of the reactor and the fuel cycle. This paper describes the fuel assembly design for the GEN-IV core at the end of the preliminary SAR design phase and studies performed to optimize the performances. The basic design and fuel assembly analysis results were described in this work. In specially, the design analysis results under the accident condition is focused in this paper. As the results of that, the structural integrity of a fuel assembly was satisfied for the PGSFR fuel assembly mechanical design criteria.

Design optimization analysis of a large electromagnetic pump for sodium coolant transportation in PGSFR

An annular linear induction electromagnetic pump (ALIP) with a flow rate of 0.86 m3/s and a developed pressure of 3.6 bar is designed by conducting an optimization analysis of its geometrical, electromagnetic, and hydrodynamic variables. The purpose of the ALIP involves circulating liquid sodium in the intermediate heat transport system of the Prototype Generation-IV Sodium-cooled Fast Reactor (electric output: 150 MWe) that is currently under development in Korea. The efficiency and developed pressure of the ALIP are derived as a function of its design variables and include the core length, flow gap, number of poles, flow velocity, pole pitch, input frequency, and input current. In terms of electromagnetic variables that can be controlled during its operation, hydraulic efficiency is maximized at an input frequency of a few tens of Hertz. The P–Q characteristics of the optimized ALIP are analyzed and verified via a theoretical comparison with an extremely large ALIP.

A study on reliability assessment of a decay heat removal system for a sodium-cooled fast reactor

Probabilistic risk assessment is expected to play a major role in the design and regulation framework for future reactors. Since the safety of a sodium-cooled fast reactor largely depends on the reliability of a decay heat removal system, which adopts a passive safety feature, it is necessary to assess the reliability of this system in order to conduct probabilistic risk assessment. Currently existing reliability assessment methodology is based on the probability of component failures and human errors. However, this design does not work as a passive safety system consisting of active components that do not need any external input to operate. In this study, a methodology for evaluating the reliability of a passive safety system that includes functional failure is proposed. The Prototype Gen-IV Sodium-cooled Fast Reactor that is currently under development in Korea was selected as the reference reactor, and the functional reliability of the decay heat removal system in the reactor was calculated by uncertainty propagation of the system parameters that can affect system failure. Sensitivity analysis was also performed to determine which parameters have crucial effects on system failure.

Assessment study about the risk of Wilks’ formula for uncertainty quantification of design extension condition scenarios in prototype Gen-IV sodium fast reactor

ABSTRACT Sensitivity analysis and uncertainty quantification using Wilks’ formula and Monte Carlo for Unprotected Loss of Flow (ULOF) and Unprotected Transient OverPower (UTOP) accidents of prototype Gen-IV sodium-cooled fast reactor were performed. Multi-dimensional analysis for reactor safety for liquid metal reactors code calculations were conducted while simultaneously varying the values of all uncertain parameters according to their distribution using parallel computing platform integrated for uncertainty and sensitivity analysis to obtain uncertainty bands for Figures of Merit (FOM) – coolant, fuel centerline, and cladding temperature at the hottest fuel rod. To specify the uncertainty range of the parameters for each accident scenario, literature survey and expert judgments were consulted. By the sensitivity analysis, the importance ranking of 25 parameters in model identification and ranking table based on phenomena identification and ranking table was identified. Through Monte Carlo calculation, 95% upper limit and 95% confidence level were obtained, and about 2% and 5% under-prediction (risk) of FOM of ULOF and UTOP accidents using Wilks’ formula were confirmed, respectively.

Four-bar linkage modeling and kinematics analysis of an in-vessel transfer system for Prototype Gen-IV Sodium-cooled Fast Reactor

A Sodium-cooled Fast Reactor (SFR) as a Generation IV (Gen-IV) advanced type of nuclear fission energy requires a fuel transfer machine (or a refueling system). It is employed to exchange spent fuel-assemblies during plant operation and other failed core assemblies (CAs). The refueling system provides a means of transporting, storing and handling reactor CAs. For the refueling system of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR), developed by the Korea Atomic Energy Research Institute (KAERI), the PGSFR refueling mechanism was modeled in this paper for the refueling planning, and its kinematics analysis (mathematical steps for forward and inverse kinematics) revealed interesting results by a simple and efficient 2D four-bar linkage system as a kinematics model. This paper argues that restricting coupler link movements is far better not only for more suitable solution forms but also for dynamic safety guarding against collisions with neighboring structures than setting an orientation of the gripper. Computer simulation results in all the angular position information based on the proposed approach report that the total rotation angles of the gripper link for all the core-to-core locations is less than 90° by restricting the coupler link. It was concluded that refueling with the middle component motion restricted is a safe operation strategy with other computational benefits. By taking advantages of the proposed approach, the refueling system can be ready for more precise and safe operations confirmed by kinematics solutions for all of the target core-assembly locations.

Sensitivity and uncertainty analysis for ULOF of PGSFR using PAPIRUS

In this research, sensitivity and uncertainty analyses for 23 parameters were performed for unprotected loss of flow (ULOF) for the prototype Gen-IV sodium-cooled fast reactor (PGSFR) by using the parallel computing platform integrated for uncertainty and sensitivity analysis (PAPIRUS). Based on the development of the phenomena and model identification and ranking table (PIRT), the relative importance of the parameters was confirmed through the sensitivity analysis. The objective of the global uncertainty analysis is to evaluate all safety parameters of the system in the combined phase space formed by the parameters and dependent variables. The uncertainty propagation was performed by mapping the uncertainty bands of the model parameters through the MARS-LMR to determine the distributions for the fuel centerline, cladding, and coolant temperatures. The results show that the uncertainty bands of the temperatures are below the melting point.