Thermal-hydraulic Calculations Research Articles

Analysis of hypothetical channel flow blockage of a tubular fuel assembly in a 10 MWth nuclear research reactor

Abstract Heat removal from the core of nuclear reactors is a crucial safety requirement that needs much analysis and attention. Coolant channel flow blockage is one of the unintentional cases that may influence the required heat removal rate. As a result, unsustainable over heating of the fuel cladding could occur resulting eventually in affecting the fuel integrity. This paper deals with simulating the effect of total channel flow blockage in a single fuel assembly on its steady state thermal hydraulic parameters. Thermal hydraulic calculations are carried out for the much less studied IRT-4M ducted fuel assembly type in a 10 MWth WWR-S tank in pool research reactor. The simulated IRT-4M fuel assembly comprises six rounded rectangular fuel elements in which seven channels are used for cooling. A mathematical model is developed in which all steady state thermal hydraulic parameters and their distributions could be predicted for the blocked and unblocked flow channel cases. Calculations were performed for two separate hypothetical flow channel blockage cases in the innermost channel and the middle channel of the IRT-4M fuel assembly. The results showed that the maximum clad temperatures for the two-flow blockage cases are nearly 127.4 °C, and 132.2 °C, respectively. The temperature increases resulting from a single channel blockage are not high enough to cause fuel damage, although some local coolant boiling might occur. Model validation and results support the model outcomes and its applications in the safety of the research reactor during abnormal operation.

Modeling and Visualization of Coolant Flow in a Fuel Rod Bundle of a Small Modular Reactor

This article presents the results of an experimental study of the coolant flow in a fuel rod bundle of a nuclear reactor fuel assembly of a small modular reactor for a small ground-based nuclear power plant. The aim of the work is to experimentally determine the hydrodynamic characteristics of the coolant flow in a fuel rod bundle of a fuel assembly. For this purpose, experimental studies were conducted in an aerodynamic model that included simulators of fuel elements, burnable absorber rods, spacer grids, a central displacer, and stiffening corners. During the experiments, the water coolant flow was modeled using airflow based on the theory of hydrodynamic similarity. The studies were conducted using the pneumometric method and the contrast agent injection method. The flow structure was visualized by contour plots of axial and tangential velocity, as well as the distribution of the contrast agent. During the experiments, the features of the axial flow were identified, and the structure of the cross-flows of the coolant was determined. The database obtained during the experiments can be used to validate CFD programs, refine the methods of thermal-hydraulic calculation of nuclear reactor cores, and also to justify the design of fuel assemblies.

Thermo-neutronic integrated coupling effects on nuclear reactor core calculations

Small Modular Reactors (SMRs) have gained significant attention as the next generation of nuclear reactors due to their unique features, such as improved economic viability and safety. This study focuses on the SMART reactor, the first certified design of an SMR, and investigates various Thermo-Neutronics aspects using an integrated coupling scheme of nuclear codes. The Neutronics cell and core calculations are performed using the DRAGON/DONJON codes, while the COBRA code is employed for thermal-hydraulic calculations. The analysis considers parameters including axial and radial power peaking factors (PPFs), critical heat flux (CHF), minimum departure from nucleate boiling ratio (MDNBR), axial average coolant temperature in hot channel and core, and maximum fuel temperature. These parameters are evaluated using the developed coupling system. Code-to-code verification is conducted to ensure the accuracy of the results, demonstrating good agreement. The coupling approach reveals significant effects compared to stand-alone code modeling, emphasizing the importance of considering the coupling between Neutronics and thermal-hydraulics, especially during the cycle length. This study contributes to the understanding of SMRs, particularly the SMART reactor, by providing insights into key parameters and their interdependencies. The integrated coupling scheme enhances the accuracy of the analysis and highlights the significance of considering coupling effects in the modeling process during the steady state and the cycle length.

Development of multi-physics calculation method in lead cooled fast reactor code system MOSASAUR

A full-core three-dimensional multi-physics calculation method was researched based on the Lead cooled fast reactor (LFR) code system MOSASAUR, which is focused on both core steady-state and transient analyses of LFR. In the previous version of MOSASAUR, cross-sections generation module and core steady-state simulation module have been developed. In this research, the stiffness confinement method is employed to solve the time-dependent multi-group neutron transport equation to expand the capabilities of core transient analyses. And the thermal hydraulic calculation method is developed as thermal feedback module for core steady-state and transient analyses. Based on the fixed-point iteration scheme, the neutron module and thermal hydraulic module are coupled at each time step. The LMW benchmark and the problem based on ORAL 19-rod bundle are employed to verify the accuracy of transient calculation and thermal hydraulic module respectively. Finally, the multi-physics calculation method is utilized to simulate the transient process of the MicroURANUS core. The simulation results demonstrate the capability of the constructed multi-physics calculation framework to accurately simulate the transient behaviors of LMRs. Numerical results showed the good accuracy of the newly-developed multi-physics calculation module with MOSASAUR.

The application study of conservative best-estimated approach for steam line break accident

The most important characteristic of main Steam Line Break (SLB) accident is the asymmetry cooldown between loops due to secondary break spurting, which results non-uniform reactivity feedback in different reactor core sections. The fuel rod DNBR limit would be more challenged in the core section with peak power. The SLB safety analysis methodologies in Administration License Application for typical PWR in China are investigated in this paper. Based on the conservative universal methodology, the best-estimated code RELAP5/MOD3 is used to perform the SLB thermal-hydraulic calculation for a three-loop PWR. In this analysis, the critical conservative models submitted for Licensing process are implemented in RELAP5/MOD3 specified modeling, which includes the Three-Channel Core Model, core inlet mixture array, core outlet mixture, vessel upper head flow and the more realistic SG model. The “second pressurizer” phenomenon in vessel upper head flashing is studied, which affects the system pressure response and mitigation. The effect of SG spouting during the transient is also studied in this paper, which is different with lumped parameter SG model. The combined approach for Deterministic safety analysis is validated in this paper, which is composed with best-estimated code, Conservative assumptions and Conservative initial & boundary conditions. Because of the flexibility and adequate conservatism, it can be spread for design optimization and independent safety assessment from the third party.

Numerical investigation of flow boiling characteristics in narrow rectangular channels using two-fluid Eulerian CFD model covering a wide range of pressures

Narrow channel heat exchange technology is one of the key technologies for the development of integrated small modular reactors (SMR). The ‘wall heat flux partitioning’ (WHFP) model in conjunction with an Eulerian–Eulerian Multiphase Flow (EEMF) method is found to be an important tool for two-phase hydrothermal analysis. In this paper, the accuracy of the EEMF-WHFP method with various sub-models is systemically evaluated over a large database of boiling flows in narrow rectangular channels. Prediction results using a total of 9 model combinations, are compared with data from 22 experimental conditions covering a range of the mass flux of 134.2 to 2200 kgm-2s-1, the inlet fluid subcooling of 7.2 to 74.2 K, the pressure of 0.1 to 14 MPa and the heat flux of 9.54 to 1700 kWm-2 in narrow rectangular channels with gaps of 1.0 to 3.0 mm. The evolution of distributions of void fraction, wall temperature and the net vapor generation (NVG) point along the flow channel is simulated and compared with experimental data. It is found that due to the limitation of the range of applicability of sub-models, certain models produced better prediction results under different pressure ranges than others, and the recommended combinations are given according to the quantitative assessments. In addition, the ability of the model combinations to predict the NVG point, as well as the flow structure and boiling process in narrow channels is investigated. The critical local void fraction near the wall at the NVG point is quantitatively estimated based on numerical results. The research conclusions in this paper can be used in thermal–hydraulic calculations in a wide range of operating pressures covering the applications of compact heat exchangers, flat plate fuels, climbing/falling film plate evaporators and the cooling of microelectronics, among other applications.

A new method of fuel pebble temperature calculation based on neutronic and thermal coupling equivalent model

As one of the six reactor types in the fourth-generation nuclear energy systems, the Very High Temperature Gas-cooled Reactor (VHTR) possesses inherent safety features. One VHTR fuel pebble contains tens of thousands of randomly distributed TRISO fuel particles, displaying a double heterogeneity characteristic. Current research models employ the homogenized method for fuel pebble temperature calculation, which neglected the neutronic and thermal (N/T) coupling effect and double heterogeneity characteristic. This study proposes a fuel pebble coupled equivalent method (PCEM) for temperature calculation. The method is based on the effective thermal conductivity obtained by the fine structure fuel pebble N/T coupling method. The new method is evaluated on steady-state and transient VHTR multi-scale neutronic and thermal hydraulic (N/TH) coupling calculation. The results show that the traditional temperature calculation method is conservative. Under steady-state conditions, the maximum degree of conservatism of the fuel pebble average temperature and peak temperature are 0.6 K and 14.2 K respectively. For the sudden drop in coolant flow rate transient condition, the average temperature and peak temperature of the fuel pebble calculated by PCEM are smaller than the values calculated by traditional method at the same time. Due to the influence of the fuel Doppler effect, the transient response of the traditional calculation method is slower.

Multiphysics and multi-scale design and analysis of nuclear thermal propulsion pellet bed reactor based on spherical cermet fuel element

Due to the advantages of high thrust and high specific impulse, nuclear thermal propulsion (NTP) is a preferred near-term rocket engine technology for future Mars missions. This article introduces the reactor design based on spherical cermet fuel element, with the aim of studying the feasibility of applying this new fuel in NTP system. First, design requirements are determined based on the unique features of NTP. In terms of neutronics, the effect of rocket launch accidents and fuel loss caused by high-temperature hydrogen are considered. Thermal hydraulics refer to the requirements of NASA's Mars Design Reference Architecture 5.0 (DRA 5.0) for impulse, thrust, and thrust to weight ratio. Second, based on the requirements above, the optimal reactor configuration parameters are obtained with the goal of maximizing the thrust to weight ratio: reactor radius 51.1 cm, height 106 cm, aspect ratio 1.04, reflector 13 cm, and reactor mass 3940 kg. Finally, the performances of the optimal reactor design are analyzed. The neutron spectrum shows a characteristic of fast neutron spectrum, and fuel loss analysis shows that the core can withstand a fuel mass loss of about 4 wt%. In addition, the core axial and radial power distribution are calculated, with an overall power peak factor of 1.26. The thermal hydraulic calculation provides a range of core operating conditions that satisfy the design requirements of DRA 5.0 and temperature, with power of 620 ∼ 950 MW and flow rate of 16 ∼ 24 kg/s. Compared with the ANL200/2000 reactor using prismatic cermet, it has the advantage of higher specific impulse. The main disadvantage is the large pressure drop, which prevents it from being applied to the scenarios of high power and high thrust. Therefore, for NTP reactor using spherical cermet elements, it should be applied to low to medium power scenarios.

Nuclear safety operating margin using new power peaking factors for NUR nuclear research reactor

Among a series of important safety parameters necessary for a complete analysis of a nuclear reactor core, the nuclear safety operating margin of a nuclear reactor is of great importance. The methodology generally used is to combine thermalhydraulic calculations with neutronic calculations. In this work, new Power Peaking Factors (PPFs) are calculated for the NUR nuclear research reactor core X-1 configuration by two ways; a deterministic way using WIMSD-CITVAP neutronic calculation in a fine calculation model, and a second way by a probabilistic method using MCNP5 model. The obtained results give the PPFs for each nuclear reactor core channel and the hot channel is then determined. Once this channel known, an asymptotic thermalhydraulic model is applied. This model gives the coolant and the cladding temperature evolutions in the hot channel for the calculated PPFs in the two presented models. All these results are compared with those obtained using the old PPF commonly used in such calculations. Results show that PPFs obtained by deterministic method are the most conservative. Thus, this last result is retained for a nuclear safety evaluation. An equation set-up is made for calculating the critical heat flux ratio. This ratio is verified to be greater than the limit of nuclear safety operating margin for NUR nuclear reactor.

Analytical model development for investigation on hydrogen management studies in the containment of a nuclear reactor

Containment is the ultimate physical barrier to prevent the spread of active air-borne fission products as well as radiation shielding in normal operation as well as accident condition and is designed to enclose whole reactor systems. The design of containment system for Fleet mode of 700 MWe Indian Pressurized Heavy Water Reactors (IPHWRs) incorporates a Primary Containment (PC) of pre-stressed concrete enveloped by a Secondary Containment (SC) of reinforced concrete. For the specification of the design pressure of the containment system, analysis is performed for various postulated accidents such as Loss of Coolant Accident (LOCA) and Main Steam Line Break Accident (MLSB). Further Studies on hydrogen generation and distribution for various accident scenarios in reactor containment of Indian Pressurized Heavy Water Reactors (IPHWR) indicate that during severe accident with core damage, the possibility of global deflagration in absence of mitigating features is high and may challenge the integrity of containment. Therefore, it is essential to incorporate mitigating provisions inside the containment to prevent the hydrogen reaching local detonation and global deflagration during such a scenario.In order to arrive at the design basis of containment system and simulation of large numbers of accident sequence a computer code is required, where mitigating features such as Passive Auto-Catalytic Recombiners, Containment Spray System & Containment Filtered Venting System can be modeled. A system thermal hydraulic computer code has been developed for containment response calculation during normal operation as well as accident conditions including, severe accident to include the new models for mitigating features. This article presents the details of various models of computer code PACSR-SI-2.0, its validation with data observed for hydrogen distribution & removal in Recombiner Test Facility,Battelle Spray Test for containment system & Spray test at IIT Mumbai.

Thermal hydraulic preliminary calculations for small modular reactor based on ACPR50S technology

This paper presents preliminary thermal-hydraulics calculation results for a small modular reactor (SMR) with a thermal capacity of 200 MW based on the ACPR50S reactor technology (Advanced customer-friendly practicable reliable 50 MWe compact SMR). The basic parameters and thermal-hydraulics safety were assessed, including the heat coolant flow through the core, fuel, and cladding temperatures to ensure the reactor’s capability to operate within safe limits. The SMR configuration was simulated using the RELAP5 program and performed the calculation in about 500 seconds in the steady state. The coolant temperature of the heat carrier along the height of the core at the inner and outer was found to be different (1.7%) from the reference value. Regarding safety, the departure from nucleate boiling ratio (DNBR) was also estimated along the average channel and the hottest channel of the core in the steady state. The results from this study will be further used in transient computations and incident analysis for SMRs.

Neutronic and thermal-hydraulic calculations under steady state and transient conditions for a generic pebble bed fluoride salt cooled high-temperature reactor

Here, we develop multiphysics coupled neutronics/thermal-hydraulics models for a recently available generic Fluoride-salt-cooled High-temperature Reactor (FHR) at equilibrium core burnup conditions with reactivity control mechanisms. We utilized the pebble bed reactor transient tool KP-AGREE with multigroup homogenized cross-sections generated by the continuous energy Monte Carlo neutron transport code Serpent to simulate transient events for an FHR. Comparing global reactor parameters and safety characteristics between Serpent and published results to standalone neutronic calculations in KP-AGREE verifies our initial conditions for our transient models. This work explores three separate transients: a loss of flow accident, an overcooling accident, and a total control rod withdrawal accident. The boundary conditions of these scenarios come from either transient benchmark definitions of a gas-cooled pebble bed reactor or transient analysis of a similar FHR design. For the loss of flow accident, our predicted peak fuel and coolant outlet temperatures show a similar response to published values, albeit slightly higher, with differences between values attributed to our modeling assumptions. The overcooling transient simulation predicts a 52 K temperature increase in the maximum fuel kernel temperature and an approximately 32.6% increase in total core power due to the negative coolant temperature feedback of the FHR design. Our control rod removal transients predicted a 235 K increase in maximum fuel kernel temperature and a 107.5% increase in total core power. For all transient cases, the fuel temperature remained significantly lower than the temperature limit, but coolant outlet temperatures for the reactivity-initiated accident exceeded external core material temperature limits.

Research on prediction and analysis of supercritical water heat transfer coefficient based on support vector machine

In order to better perform thermal hydraulic calculation and analysis of supercritical water reactor, based on the experimental data of supercritical water, the model training and predictive analysis of the heat transfer coefficient of supercritical water were carried out by using the support vector machine (SVM) algorithm. The changes in the prediction accuracy of the supercritical water heat transfer coefficient are analyzed by the changes of the regularization penalty parameter C, the slack variable epsilon and the Gaussian kernel function parameter gamma. The predicted value of the SVM model obtained after parameter optimization and the actual experimental test data are analyzed for data verification. The research results show that: the normalization of the data has a great influence on the prediction results. The slack variable has a relatively small influence on the accuracy change range of the predicted heat transfer coefficient. The change of gamma has the greatest impact on the accuracy of the heat transfer coefficient. Compared with the calculation results of traditional empirical formula methods, the trained algorithm model using SVM has smaller average error and standard deviations. Using the SVM trained algorithm model, the heat transfer coefficient of supercritical water can be effectively predicted and analyzed.

Comparison of thermal-hydraulic calculation in 100 MWt thorium-based HTGR using SiC and ZrC TRISO coated fuel particle

The High-Temperature Gas-cooled Reactor (HTGR) uses Tri-structural Isotropic (TRISO) coated fuel, the first barrier to fission products with unique ruggedness. One of the layers of TRISO is a SiC layer, which maintains the fuel's dimensional stability and mechanical integrity and acts as a fission product barrier. At extremely high temperatures, the SiC layer in the TRISO coating system tends to decompose, vaporizing silicon elements and forming a porous carbon structure. The weakness of the SiC layer leads to more expensive and difficult maintenance processes. ZrC is considered the primary alternative to replace the SiC layer due to its superior stability at high temperatures and more resistance to fission products especially palladium. Previous research has shown that from a neutronic perspective, the performance of ZrC and SiC layers is equally good with insignificant differences. The following research step analyzes the temperature distribution in the hot spot channel of the fuel and the overall core using both types of layers. The temperature analysis is essential to ensure that the maximum fuel temperature during normal operation does not exceed the thermal design target of 1495 °C. This research also simulates the temperature of both layers above the fuel design limit of 1600 °C. In addition to thermal–hydraulic analysis, calculations were also conducted to determine the coolant pressure drop caused by friction, elevation and form. The calculation showed that the peak temperature value during normal operation for TRISO with SiC and ZrC layer is below the thermal design target value. Both temperatures are not significantly different, with the TRISO (ZrC) temperature being higher during normal operation and the TRISO (SiC) temperature being higher during accident/high-temperature conditions. The pressure drop profile indicates that both coated fuel particles do not exhibit significant differences.

Development of a Steam Generator Mock-Up for EU DEMO Fusion Reactor: Conceptual Design and Code Assessment

Recent R&D activities in nuclear fusion have identified the DEMO reactor as the ITER successor, aiming at demonstrating the technical feasibility of fusion plants, along with their commercial exploitation. However, the pulsed operation of the machine causes an “unconventional” operation of the system, posing unique challenges to the functional feasibility of the steam generator, for which it is necessary to define and qualify a reference configuration for DEMO. In order to facilitate the transitions between different operational regimes, the Once Through Steam Generator (OTSG) is considered to be a suitable choice for the DEMO primary heat transfer systems, being characterized by lower thermal inertia with respect to the most common U-tube steam generators. In this framework, the ENEA has undertaken construction of the STEAM facility at Brasimone R.C., aiming at characterizing the behavior of the DEMO OTSG and related water coolant systems in steady-state and transient conditions. A dedicated OTSG mock-up has been conceived and designed, adopting a scaling procedure, keeping the height 1:1 of the DEMO OTSGs. The conceptual design has been supported by RELAP5/Mod3.3 thermal-hydraulic calculations. CFD and FEM codes have been used for fluid-dynamic analyses and mechanical stress analyses, respectively, in specific parts of the component.

Comparison of local and global sensitivity analysis methods and application to thermal hydraulic phenomena

The best estimate plus uncertainty analysis have been widely applied for the safety assessment of nuclear systems, and sensitivity analysis is one of the essential steps to identify and rank the important phenomena and parameters. The conventionally used local sensitivity methods are based on linear hypothesis and have limitations in the application process of nonlinear systems, while the more complicated global sensitivity methods may have problem in time costs. To compare and propose the proper method for evaluation of thermal-hydraulic calculation in a reactor system, local methods such as Pearson correlation coefficient, Spearman rank correlation coefficient, and Kendall rank correlation coefficient, global methods such as the variance-based method, moment-independent method, and elementary effects method are reviewed and applied for numerical examples and specific reactor thermal-hydraulic problem in this paper. Results show that the rank of parameter importance measured by various local analysis methods is the same, but away from the global methods. Different types of global sensitivity analysis methods have similar results for parameters with higher importance, and among these methods, the elementary effects method has the lowest computational cost with the same number of samples.

Development and validation of multiphysics PWR core simulator KANT

KANT (KAIST Advanced Nuclear Tachygraphy) is a PWR core simulator recently developed at Korea Advance Institute of Science and Technology, which solves three-dimensional steady-state and transient multigroup neutron diffusion equations under Cartesian geometries alongside the incorporation of thermal-hydraulics feedback effect for multi-physics calculation. It utilizes the standard Nodal Expansion Method (NEM) accelerated with various Coarse Mesh Finite Difference (CMFD) methods for neutronics calculation. For thermal-hydraulics (TH) calculation, a single-phase flow model and a one-dimensional cylindrical fuel rod heat conduction model are employed. The time-dependent neutronics and TH calculations are numerically solved through an implicit Euler scheme, where a detailed coupling strategy is presented in this paper alongside a description of nodal equivalence, macroscopic depletion, and pin power reconstruction. For validation of the steady, transient, and depletion calculation with pin power reconstruction capacity of KANT, solutions for various benchmark problems are presented. The IAEA 3-D PWR and 4-group KOEBERG problems were considered for the steady-state reactor benchmark problem. For transient calculations, LMW (Lagenbuch, Maurer and Werner) LWR and NEACRP 3-D PWR benchmarks were solved, where the latter problem includes thermal-hydraulics feedback. For macroscopic depletion with pin power reconstruction, a small PWR problem modified with KAIST benchmark model was solved. For validation of the multi-physics analysis capability of KANT concerning large-sized PWRs, the BEAVRS Cycle1 benchmark has been considered. It was found that KANT solutions are accurate and consistent compared to other published works.

Modeling of sodium-cooled fast reactor core – A sensitivity study on number of core channels and axial meshes

System thermal–hydraulic codes are used in the dynamics analysis of sodium-cooled fast reactors (SFRs). In such simulations, the accuracy of the reactor core model is very important to establish plant safety. Coupled physical phenomena, namely, heat transfer, fluid dynamics, and neutronics, need to be simulated in the core. Due to the complexity, a simplified core modeling approach has been traditionally followed in system dynamics codes. Firstly, steady-state core neutronics calculations are carried out using a 2D axisymmetric model to obtain perturbation worth and power distributions. Later, this data is mapped to several core channels (usually 10 to 20), each representing a set of fuel subassemblies (SAs) for thermal-hydraulics calculations.In this paper, a more sophisticated core modeling approach has been followed to understand the importance of such an exercise. A 3D neutronics model was used to obtain subassembly-wise perturbation worths and power distributions. For thermal-hydraulics calculations, the number of core channels was increased with the idea of one subassembly per channel. The number of axial meshes in each channel was also increased. The traditional and new modeling approaches were used to simulate the benchmark problem of FFTF LOFWOS test#13 for comparison. Steady-state parameters from both the new and the old models were comparable. Notably, there was a peaking of the axial power profile in the new model compared to the old model, which resulted in slightly higher peak fuel temperatures in the new model. However, the difference in the evolution of transient parameters between the two models was insignificant, and both models compared reasonably well with the benchmark data.

Optimization for fuel loading and reactivity initiated accident analysis for TRR-1/M1

TRR-1/M1 reactor core consists of two types of standard TRIGA fuel, 8.5% and 20% weight uranium. Periodically, the reactor core configuration is rearranged in order to optimize its performance. The objective of this work is to show that it is safe to operate TRR-1/M1 at 1 MW steady state condition for the new core configurations of which the 8.5% weight fuels in the B ring are replaced by high burn up 20% weight fuel elements. To ensure sufficient safety margins of the reactor operation, thermal–hydraulic computer codes, i.e., COOLOD-N2 and EUREKA-2/RR, were coupled with neutronic calculation based on Monte Carlo N-Particle Transport concept in order to analyse some crucial operating parameters, e.g., coolant and fuel temperature in the reactor core as well as Departure from Nucleate Boiling Ratio (DNBR) along the fuel axial direction of the hottest channel. From this study, it can be concluded that no significant effect on the reactor operation was found when replacing the 8.5% weight fuel elements in the B ring with the high burnup 20% weight fuels. Transient phenomena of Reactivity Initiated Accident (RIA) were also performed for the rapid removal of a sample from the reactor core with reactivity insertion rate of $1.00/s. It is found that the negative temperature coefficient of TRIGA fuel provided safety for the reactor even during the prescribed reactor transients. The calculation showed slightly increase in the fuel temperature during the short period of reactivity insertion which caused insignificant energy release to heat up the fuel during the accident. The maximum fuel temperature found during the accident in this study was well below TRR-1/M1 safety.

SUBSALS: a subchannel thermal-hydraulic code for IRT type fuel analysis

The LVR-15 research reactor is operated with a tube type fuel IRT-4M. Due to fuel’s unique concentric square annular shape with, coolant flow is subject to significant pressure driven crossflow. Detailed calculation of such flow patterns is beyond capabilities of standard system codes used for the thermal hydraulic safety analysis. To assess safety of core designs consisting of the IRT-4M fuel assemblies, a new subchannel code is under development at Research Centre Rež and Czech Technical University in Prague.Computer program SUBSALS is single phase, steady state, subchannel solver in development. Rigid computational mesh describes IRT-like geometries, and it is also suitable for other specific doubly connected ducts.In this contribution, general description of calculation procedure and computational capabilities is presented along with code-to-code comparison. It could be used as proof of concept for future improvements in thermal hydraulic calculations of research reactors.