Prototype 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.

Conceptual designs and characteristic of the fuel handling and transfer system for 150 MWe PGSFR and 1400 MWe SFR burner reactor

KAERI (Korea Atomic Energy Research Institute) developed the conceptual design of PGSFR (Prototype Gen-IV Sodium Cooled Fast Reactor) and Burner Reactor. Since the reactor characteristics of the PGSFR and Burner Reactor are different, the shape, size and the arrangement of the main components in the reactors must be different. Therefore, the conceptual design for the fuel handling and transfer systems needs to be performed coinciding with the structure of the reactor. Especially, because a redan structure dividing hot and cold pool is installed in the reactor vessel, the conceptual design of the fuel handling and transfer system largely changes depending on the location of the redan structure. Various elements of the conceptual design and an integral arrangement for the fuel handling and transfer system were arranged according to the characteristics, sizes and shapes of the reactors. In this paper, the conceptual designs of the fuel handling and transfer system for PGSFR and Burner Reactor are described. Especially, an A-frame method is selected as the fuel handling and transfer system for the Burner Reactor, considering the layout of the internal structure. The tilt angle, diameter and length of A-frame is determined and the strength evaluation of the A-frame is performed.

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 tritium effluent from Prototype Generation IV Sodium-cooled Fast Reactor

A tritium transport model for a pool-type sodium cooled fast reactor was developed for estimation of tritium effluents from the Prototype Generation IV Sodium-cooled Fast Reactor (PGSFR). For this purpose, the transport model developed for Experimental Breeder Reactor II and Fast Flux Test Facility was updated and validated. The tritium production rate from ternary fission and activations in PGSFR were calculated by the Argonne Reactor Computation (ARC) code suite for fast reactor analysis, and due to the potential uncertainties in source terms and variations of cold trap performance and material permeability during reactor operation, the tritium distributions and tritium effluents were estimated for several bounding cases. The upper bounding gaseous and liquid tritium release rates from PGSFR are ~ 440.7 Ci/year and 14.7 Ci/year, respectively. The gaseous tritium effluent is about few orders of magnitude smaller than the regulatory constraint of the tritium concentration in the air. The peak tritium concentration in the steam-generator feedwater is ~ 1.1x10-3 μCi/g, which is required to be diluted prior to releasing as tritiated water. The accumulated tritium in the cold trap after five cycle operations is ~ 52,000 Ci. Because of significant conservatism in these upper bounding estimations, it is expected that the tritium effluent at the nominal case with plausible source terms would be about 2 orders of magnitude smaller than that of the upper bounding case.

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.

Improvement of aseismic performance of a PGSFR PHTS pump

A design study was performed to improve the limit aseismic performance (LSP) of a primary heat transport system (PHTS) pump. This pump is part of the primary equipment of a prototype generation IV sodium-cooled fast reactor (PGSFR). The LSP is the maximum allowable seismic load that still ensures structural integrity. To calculate the LSP of the PHTS pump, a structural analysis model of the pump was developed and its dynamic characteristics were obtained by modal analysis. The floor response spectrum (FRS) initiated from a safety shutdown earthquake (SSE), 0.3 g, was applied to the support points of the PHTS pump, and then the seismic induced stresses were calculated. The structural integrity was evaluated according to the ASME code, and the LSP of the PHTS pump was calculated from the evaluation results. Based on the results of the modal analysis and LSP of the PHTS pump, design parameters affecting the LSP were selected. Then, ways to improve the LSP were proposed from sensitivity analysis of the selected design variables.

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.

Ultrasonic ranging technique for obstacle monitoring above reactor core in prototype generation IV sodium-cooled fast reactor

Abstract As the refueling of a sodium-cooled fast reactor is conducted by rotating part of the reactor head without opening it, the monitoring of existing obstacles that can disturb the rotation of the reactor head is one of the most important issues. This paper deals with the ultrasonic ranging technique that directly monitors the existence of possible obstacles located in a lateral gap between the upper internal structure and the reactor core in a prototype generation IV sodium-cooled fast reactor (PGSFR). A 10 m long plate-type ultrasonic waveguide sensor, whose feasibility has been successfully demonstrated through preliminary tests, was employed for the ultrasonic ranging technique. The design of the sensor's wave radiating section was modified to improve the radiation performance, and the radiated field was investigated through beam profile measurements. A test facility simulating the lower part of the upper internal structure and the upper part of the reactor core with the same shapes and sizes as those in the PGSFR was newly constructed. Several under-water performance tests were then carried out at room temperature to investigate the applicability of the developed ranging technique using the plate-type ultrasonic waveguide sensor with the actual geometry of the PGSFR's internal structures.

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).

Mechanical design of a sodium cooled fast reactor fuel assembly in Korea: Normal operation condition

The mechanical design of the fuel assembly of a sodium cooled fast reactor is presented in this paper. This is one of the design works of a prototype generation IV sodium cooled fast reactor (PGSFR), which is planned to be built in Korea. Among the design tasks of the fuel assembly, the normal operation conditions is primarily considered here. One of the main issues is the lack of an applicable design code because the currently available worldwide codes such as the ASME Section III Division 5 and RCC-MRx do not incorporate the material (HT-9, the ferriticmartensitic steel) of the fuel assembly’s structural components. To resolve this, the formulae of the yield and ultimate strengths of HT-9 are used here, which have been verified through comparing with the existing data in the references other than the codes. From this, we conservatively set up new design criteria and stress limits. Two fuel assemblies of different locations in the reactor core are chosen for the analysis. The results show that the stresses are far below the stress limits. In addition, a brittle fracture of the components is concerned considering the low temperature below the ductile-brittle transition temperature of the HT-9 during handling a fuel assembly in the reactor. It is found that the crack sizes that can cause a fracture failure are much larger than what can be detected using a normal non-destructive test method. Thus, it is soundly concluded that the presently designed PGSFR fuel assembly has sufficient structural integrity during normal operation and refuelling outage.

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.

Experimental study of sodium heat exchanger performance at high Pe number condition using STELLA-1 facility

Abstract In Prototype Gen IV Sodium-cooled Fast Reactor (PGSFR), there are two types of sodium-to-sodium heat exchanger, namely, Intermediate Heat Exchanger (IHX) and Decay Heat Exchanger (DHX). The role, function, and size of each heat exchanger are different, but both heat exchangers are designed by the same computer code, SHXSA, developed at KAERI. In previous work, the design code and related correlations were validated under DHX operation conditions (low Pe number range). However, the IHX operation condition is beyond this range (Pe > 150), and the heat transfer performance test was extended by modifying the facility. The experiments under large mass flowrate conditions were conducted, and the heat transfer rate of the tube-side and shell-side resulted in good agreement with the code calculation. The uncertainty analysis results were also included, and the discrepancies were −0.68 to 0.26% and −2.92 to 1.04% for the outlet temperature and heat transfer, respectively.

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.