Fast Reactor Core Research Articles

Generating a system of group constants for neutron-physical calculations of fast reactors based on ROSFOND-2020.2 library files

The subject of constant support development is currently becoming increasingly important in connection with the national strategy of transition to a closed nuclear fuel cycle. At the same time, the importance of the tasks of refining calculation models and minimizing methodological, statistical, and constant errors is increasing. In this connection, the idea of creating a universal system of constants describing with equal accuracy both uranium loading and loading with mixed oxide uranium-plutonium fuel and the possibility to perform on its basis both precision Monte Carlo calculations (data format – ACE) and multigroup calculations (formats – ABBN and ANISN) was laid in the basis of this work. In this paper we analyze the freely available libraries of evaluated nuclear data in order to justify the choice of source files for the formation of a new system of group constants for neutron-physical calculations of fast reactor cores, describe the process of formation of a new library of reactor constants, and verify the obtained system on computational test models of BN-800 reactor and critical systems. The process of formation of the new system of group constants included selection of initial neutron data files, updating of data tables of basic neutron cross-sections, self-shielding factors and Doppler coefficients, as well as data on fission spectra for the main fuel nuclides. The methodological component of the error for test models of the BN-800 core was evaluated. Using the ABBN-RF22 system of group constants it was possible to estimate the correction in the 299 groups calculation, which amounted to 0.3%. Previously, when using the ABBN-93 library, there was no such possibility due to the lack of continuity between the files of evaluated neutron data and the group constants used.

A subchannel analysis code for advanced liquid metal fast reactor cores and study on heat transfer characteristics of core geometry parameters

The liquid metal fast reactor represents an important reactor type for the future development of nuclear energy. Accurately modeling and predicting the subchannel flow and heat transfer phenomenon in the rod bundles plays a key role in ensuring core safety. Consequently, a subchannel code suitable for the liquid metal fast reactor is analyzed based on the subchannel analysis code of the Pressurized Water Reactor (PWR). The modified code incorporates various types of liquid metals, such as lead, lead-bismuth eutectic (LBE), and sodium, along with their constitutive models, enhancing the applicability of the liquid metal reactor systems. This code has been verified and validated at several levels, including comparing the code simulation results with other similar subchannel codes results, high-dimensional Computational Fluid Dynamics (CFD) simulations, and experiment data. The sensitivity analysis of core geometric parameters and turbulent mixing coefficients is performed based on the modified version code. The influence of the core geometry, including the rod Pitch to Diameter ratio (P/D), Rod to Wall gap (RTW) and the wire wrap pitch (H) on the sensitivity of outlet temperature has been investigated. The results show that the new subchannel analysis code suitable for liquid metal cooled reactor shows reasonable agreement with the verified and validated results. The geometric parameters, such as P/D, RTW and H have noticeable effects on the outlet temperature difference of the subchannels. Additionally, the outlet temperature distribution in the subchannel is significantly affected by different turbulent mixing coefficients and the distribution of the turbulent mixing coefficient also influenced by the reactor core sizes. Overall, this study could support the rod bundles design and safety analysis in the liquid metal reactors.

Numerical investigation on the core thermal hydraulic behavior of pool-type sodium-cooled fast reactor (SFR)

A thorough understanding of the reactor core thermal hydraulic behavior is essential for the design and safety analysis of Sodium-cooled Fast Reactors (SFR). Due to the application of hexagonal subassembly, the core thermal hydraulic behavior is significantly affected by the flow field within the subassemblies, the inter-wrapper region and the hot pool. Analysis of the core thermal hydraulic behavior requires a model coupling the three regions mentioned above, which has been identified as one of the thermal hydraulic challenges in SFR. In the present study, a 3D model that covers the three regions was developed for the core of the China Experimental Fast Reactor (CEFR) with the Computational Fluid Dynamic (CFD) code, Fluent. The inter-wrapper region and the hot pool were modeled in detail, while the subassemblies were modeled with a special porous medium model. The core thermal hydraulics behavior under steady state was studied, more specifically, information for the flow field distribution at the core outlet, the inter-wrapper flow and the duct wall temperature distribution was obtained. Under steady state, liquid sodium in the inter-wrapper region is supplied by the inner region of the hot pool. And it enters the inter-wrapper region from the core outer region and returns back to the hot pool inner region from the core central region. The inter-wrapper flow is cooled by non-fuel subassemblies and heated up by fuel subassemblies. For non-fuel subassembly, the ratio of the total heat transfer rate between the inter-wrapper flow and the subassemblies to the heat generated within subassemblies could reaches 96%; for fuel subassemblies, the maximum ratio of the total heat transfer rate between the inter-wrapper flow and the subassemblies to the heat generated within subassemblies is 2.45%. Significant temperature gradients have been observed on the duct wall, with maximum values of 156.69 K/m in the vertical direction and 2,196.00 K/m in the circumferential direction. The largest temperature gradient appears on the duct of subassemblies adjacent to the transition region of fuel subassemblies and non-fuel subassemblies.

Application of Lie symmetry to find an analytical solution for reactor point kinetics equation with prompt jump approximation and power feedback

The reactor Point Kinetics Equations (PKE) are simpler zero-dimensional approximation to space dependent dynamical models of nuclear reactor core, that are accurate enough to describe transients in small to medium size fast reactor cores. Even these simplified equations can be solved only by numerical methods, except in a very few restrictive cases, where they are amenable to analytical solution. Symmetry methods using Lie's point symmetry have been shown to be a systematic and powerful tool to solve any given ordinary or partial differential equation. An approximation of the PKE, known as Prompt Jump Approximation (PJA) converts the coupled system of first order ODEs with one delayed-neutron precursor group and power feedback, into a single first order nonlinear ordinary differential equation. In this study, we demonstrate an application of Lie symmetry method for solving the point kinetics equation under PJA. The analytical solution obtained is compared with benchmark numerical solution of PKE with PJA.

Thermomechanical Processing for Improved Mechanical Properties of HT9 Steels.

Thermomechanical processing (TMP) of ferritic-martensitic (FM) steels, such as HT9 (Fe-12Cr-1MoWV) steels, involves normalizing, quenching, and tempering to create a microstructure of fine ferritic/martensitic laths with carbide precipitates. HT9 steels are used in fast reactor core components due to their high-temperature strength and resistance to irradiation damage. However, traditional TMP methods for these steels often result in performance limitations under irradiation, including embrittlement at low temperatures (<~430 °C), insufficient strength and toughness at higher temperatures (>500 °C), and void swelling after high-dose irradiation (>200 dpa). This research aimed to enhance both fracture toughness and strength at high temperatures by creating a quenched and tempered martensitic structure with ultrafine laths and precipitates through rapid quenching and unconventional tempering. Mechanical testing revealed significant variations in strength and fracture toughness depending on the processing route, particularly the tempering conditions. Tailored TMP approaches, combining rapid quenching with limited tempering, elevated strength to levels comparable to nano-oxide strengthened ferritic alloys while preserving fracture toughness. For optimal properties in high-Cr steels for future reactor applications, this study recommends a modified tempering treatment, i.e., post-quench annealing at 500 °C or 600 °C for 1 h, possibly followed by a brief tempering at a slightly higher temperature.

Improving tally efficiency and accuracy of multi-group scattering matrix calculations in the Monte Carlo code NECP-MCX

Two issues arise in the calculation of the multi-group scattering matrix when employing a continuous-energy Monte Carlo code for generating homogenized multi-group cross-sections. Firstly, the analog estimator is used to evaluate group-to-group elements, which leads to large statistical uncertainty. Secondly, employing the scalar flux as the weighting function in generating the high-order scattering matrix introduces errors in fast reactor calculations. For the first issue, the repeated collision approach and pre-tabulated cross-section approach are adopted to improve the tally efficiency. For the second issue, the average scattering cosine is calculated based on the conservation of the mean square displacement of neutrons, which is then used to correct the first-order self-scattering cross-section. To evaluate the effectiveness of the above approaches, a PWR pin-cell problem and fast reactor core problems are tested. The results demonstrate that: 1) The figure of merit for multi-group scattering matrix calculations was improved by 8–12 times with the pre-tabulated cross-section approach. 2) Biases of keff were reduced from over 500 pcm to less than 300 pcm when using the corrected self-scattering cross-section. 3) The corrected self-scattering cross-section also yielded higher accuracy for the assembly power calculations, where the maximum biases are reduced from 5 % to 1 %.

Thermal-hydraulic analysis of a full-scale lead-bismuth cooled fast reactor core considering inter-wrapper flow

The Lead-bismuth cooled Fast Reactors (LFR) utilizes wrapped hexagonal fuel assemblies arranged in a honeycomb pattern. The inter-wrapper flow (IWF) between adjacent assemblies has a significant influence on the core thermal-hydraulics. In this paper, the LFR core assembly analysis model, inter-wrapper flow and heat transfer model, and multi-scale coupling analysis model were developed and were implemented into the previously self-developed subchannel-level three-dimensional thermal hydraulic analysis code CorTAF-LBE for whole core of LFR, which is based on the OpenFOAM computational fluid dynamics simulation platform. The inter-wrapper flow and heat transfer model was validated against the KALLA-IWF experiment. Whole core steady state operating condition and low flow rate condition simulations were performed with reference to the 19-rod bundle geometry from KALLA laboratory and typical LFR core assembly arrangement. The whole core steady state distribution laws of key thermal-hydraulic parameters were obtained, and the influence of low flow rate operation on safety parameters such as cladding and pellet temperatures was analyzed. Under low flow rate operating conditions, the IWF heat transfer was accounted for 4.87% of the core thermal power, which was 315.6 kW higher than that under normal operating conditions, playing an important role in core heat removal and safe operation. This work provides important insights into LFR core design and the assessment of the role of IWF under low flow rate conditions.

CorTAF-LBE: A full scale subchannel-level thermal-hydraulic characteristics analysis code for LBR core based on OpenFOAM

The lead-bismuth cooled fast reactor core adopts box-type design, and the fuel rod bundle is fixed by wrapping wires. The complex geometric structure of the fuel assemblies significantly affects the flow and heat transfer status of the liquid metal coolant. Building upon the open-source CFD platform, OpenFOAM, this work established conservation equations for liquid metal coolant, fuel rod thermal characteristic models, inter-wrapper coolant analysis models, and coupled heat transfer models, and developed the subchannel level thermal-hydraulic analysis code CorTAF-LBE for whole-core analysis of LBR core using the finite volume method. The wire-wrapped rod bundles internal flow and heat transfer experiments of lead-bismuth, carried out based on the THEADES and KYLIN-II circulation loops, were selected to perform verification and validation. The results match well with the experimental data and CFD fine-modeling prediction. The steady-state numerical simulation of three wire-wrapped rod bundle assemblies with inter-wrapper flow was carried out under normal operation condition. The temperature and pressure distributions of the coolant in each subchannel and inter-wrapper flow region, the temperature of cladding and other key thermal-hydraulic parameters were obtained. The role of the inter-wrapper coolant in heat removal from the fuel assemblies was revealed. The work in this paper has referential value for the optimization of lead-bismuth cooled fast reactor design and safety assessment during operation.

High-temperature creep and corrosion behavior of 316LN stainless steel in oxygen-saturated sodium

The high-temperature performance and sodium corrosion of in-core components are the main problems of liquid sodium-cooled reactors, and low-carbon nitrogen-controlled 316 stainless steel (316LN) is the main research and development direction of the main materials for the core of sodium-cooled fast reactors. In this work, the creep behavior of 316LN in oxygen-saturated liquid sodium at high temperatures (650 ℃, 700 ℃ and 750 ℃) was tested and compared with the results of creep tests in air at the same conditions. It is found that the creep life in the sodium environment is reduced by 63.44%, 67.93% and 40.07%, respectively. The specimens in oxygen-saturated sodium exhibit reduction in ductility, selective dissolution as well as oxidative corrosion. The corrosion mechanism of 316LN under external loading in oxygen-saturated sodium is illustrated.

METAL: Methodology for liquid metal fast reactor core economic design and fuel loading pattern optimization

The design of Liquid Metal-cooled Fast Reactor (LMFR) cores encompass two levels of design. The first is the physical core design, which is concerned with the design of the fuel assembly geometry in the context of a full core. The second is the fuel loading design, which is an in-core fuel management problem concerned with the design of fuel enrichment and core loading pattern. In this paper, an optimized fuel design search is developed to improve core loading patterns. As part of the implementation, the multiphysics simulation suite LUPINE has been enhanced to allow for fuel shuffles and an in-core fuel management optimization methodology has been added with the objective of reducing the fuel cost. The design methodology is referred to as Methodology for Economical opTimization of Applied LMFR (METAL), and this methodology is demonstrated using the popular Super Power Reactor Innovative Small Modular (SPRISM) sodium-cooled fast reactor (SFR) as a reference core. One challenge currently facing the deployment of LMFRs is the availability of TRansUranium (TRU) and high assay low enriched uranium (HALEU) driver fuel. The two forms of fuel are expensive and not widely available, which poses a challenge on the startup of LMFR cores. To address the availability of driver fuel, and to lower fuel costs, low enriched uranium (LEU) blankets are investigated as replacements to the natural uranium or depleted uranium blankets typically used. The advantage of this solution is the wide availability of LEU and its ability to lower the amount of TRU or HALEU driver fuel needed. The design objectives, constraints, and optimization algorithm for a SFR are identified. Then, METAL is applied to design a SFR core fueled by TRU driver fuel assemblies and LEU blanket assemblies. To demonstrate the advantage of introducing LEU blankets, METAL is also used to design another SFR core following the same objectives and constraints, but using naturally enriched blankets. By comparing the two cores, it was found that the introduction of LEU blankets results in a ∼28% reduction in the driver fuel mass requirements. Upon development of a levelized fuel cycle cost (LFCC) model, the 28% reduction in driver fuel mass corresponds to a 10% decrease in the LFCC. Additional advantages of using LEU blankets include reducing the core conversion ratio (CR), and improving the assembly radial power peaking factor (RPPF), which enhances the safety and non-proliferation performances of the SFR core.

Transient analysis of temperature field and fission gas about the arc plate fuel element

Obtaining the temperature field and fission gas release of fuel elements are crucial to the study on fuel behaviors and ensure the safety of elements. In order to enhance fuel efficiency, providing a design idea to make the fast reactor with high flux and multi-function, we have designed a core of the lead-based fast reactor with arc plate fuel elements. Based on above, a transient performance analysis program of arc plate fuel elements (TPAPAFE) was developed in this paper, for analyzing the thermal and fission gas module of arc plate fuel element by calculating transient variation in the maximum pellet temperature, internal and external coolant outlet temperature, and fission gas release share. The modular verification was done by comparing the results of TPAPAFE with others. Besides, the result of the arc plate fuel analyzed by TPAPAFE shows that the code is feasible and rational, and the single arc plate is so thin that coolant outlet temperatures on both sides of the plate are almost the same. Meanwhile, the error in temperature data between TPAPAFE and the steady-state program is only 5%. Furthermore, fission gas release fraction increased linearly with pellet temperature and bubble growth.

Adaptive energy group division in the few-group cross-section generation for full spectrum reactor modeling with deterministic method

Advanced nuclear reactors, especially the newly developed small and micro-reactors have complex neutron spectrum, which makes the deterministic reactor core calculations sensitive to the energy group structure of few-group cross-sections. To avoid significantly increasing the cost of energy discretization in the core calculation, two energy group structures with 31 groups and 33 groups were adopted for typical thermal and fast reactor cores, respectively. Then, an adaptive scheme of group division for reactor cores with a medium neutron spectrum was proposed. The works were based on the full spectrum nuclear reactor analysis code SARAX/TULIP. An equivalent one-dimensional model of the core was proposed to capture the key neutron spectrum features of the reactor core. Such features were used to adaptively determine a few-group structure for the following reactor core calculations. Then, the neutron spectrum in different zones with more details was calculated. With this spectrum, the cross-sections were condensed into the determined energy groups. Three tests based on different neutron spectrum were calculated to verify the schemes. The results show that using the adaptive energy group division scheme, the following core calculation can meet the accuracy requirement of different reactors with different neutron spectra.

Compacting the core of a commercial high-power lead-cooled fast reactor

Two options for the layout of a fast reactor with lead coolant are being considered. The design of the fuel assemblies and core fuel rods is identical to the BREST-OD-300 reactor. One of the layouts is traditional with a flattened core. The other layout is the most compact. It has an elongated core and is characterized by an optimal ratio of core diameter to height in terms of minimal neutron leakage. The electric capacity of the first of them is about 2.9 GW, the second - 13.3 GW. The second layout is a high-temperature reactor. When operating at rated power, the maximum coolant temperature reaches 1631 K. The use of cermets fuel based on UN-PuN mononitride micrograins and uranium metal nanopowder, coolant based on lead thorium ores and tungsten fuel rod coatings made it possible to ensure the safety of reactors in emergency modes such as "anticipated transients without scram". A distinctive feature of the layouts is to reduce the danger of emergency modes when they are applied simultaneously. In all emergency modes, large reserves are maintained up to the maximum permissible temperatures of fuel, coolant and structural materials.

Generating multi-group cross-sections using continuous-energy Monte Carlo method for fast reactor analysis

The deterministic two-step method, comprising multigroup cross-section generation and core calculation, is widely applied in fast reactor design and analysis. Monte Carlo (MC) methods with continuous energy and fine geometry provide high-precision cross-sections essential for advanced fast reactor neutronics analysis. This paper presents an analysis of integrating MC-generated homogenized cross-sections with various core solvers, demonstrating their effectiveness and potential improvements in fast reactor simulations. For diffusion core calculations, the superhomogénéisation (SPH) technique reduces control rod worth overestimation from 13.5% to 0.35% in the MET-1000 benchmark, improving power distribution predictions. In transport core calculations, the flux-moment homogenization technique (MHT) addresses reactivity overestimation by incorporating cross-section anisotropy, reducing error by 698 pcm. For Method of Characteristics (MOC) core calculations, transport-corrected multigroup cross-sections yield high precision in pin-by-pin power distribution for a 100 MWe lead-based fast reactor benchmark. While MC methods require significant computational resources, such as 62 CPU-hours for the MET-1000 core and 85.5 CPU-hours for the 100 MWe lead-based fast reactor core, they offer flexibility in geometry modeling. This work highlights MC multigroup cross-section generation methods applicable to diffusion, MOC, and transport core calculations for fast reactor analysis, suggesting further exploration into their performance in various reactor parameters and computational efficiency.

Thermal-hydraulic analysis of LBE flow in rod bundles with wire spacers based on heat transfer framework of OpenFOAM

The wire spacer allows for a compact layout in the core of lead-cooled fast reactor (LFR). However, it also introduces complexity to the flow and heat transfer of the coolant, lead-bismuth eutectic (LBE). This paper aims to analyze the effect of wire spacer in rod bundle and the applicability of heat transfer framework to LBE flow. Three Reynolds-Averaged Navier–Stokes (RANS) turbulence models with two representative turbulent Prandtl number (Prt) 1.5 and 2.0 are employed within the framework of OpenFOAM. The sensitivity of the models is examined based on turbulent heat transfer in tube flow. The simulations considered with an extensive mesh sensitivity study were analyzed. The buoyantSimpleFoam solver is compared with direct numerical simulation (DNS) data. Thermal-hydraulic analysis are conducted, considering temperature, velocity distribution, and local Nusselt (Nu) number, and pressure drop based on 7-rod and 19-rod bundle with wire spacers. The influence of wire initial starting angles from 0°, 10°, 20°, 30°, 40°, 50° on heat transfer is carried out. The study reveals that an optimum angle range from 20° to 40° exists in the sub-channels. This work validates the basic heat transfer solver for LBE using OpenFOAM.

Supercritical CO2-cooled fast reactor and cold shutdown system for ship propulsion

A neutronics study of a supercritical CO2-cooled fast reactor core for nuclear propulsion has been performed in this work. The thermal power of the reactor core is 30 MWth and a ceramic UO2 fuel can be used to achieve a 20-year lifetime without refueling. In order to make a compact core with inherent safety features, the drum-type reactivity control system and folding-type shutdown system are adopted. In addition, we suggest a cold shutdown system using gadolinium as a spectral shift absorber (SSA) against flooding. Although there is a penalty of U-235 enrichment for the core embedded with the cold shutdown system, it effectively mitigates the increment of reactivity at the flooding of seawater. In this study, the neutronics analyses have been performed by using the continuous energy Monte Carlo Serpent 2 code with the evaluated nuclear data file ENDF/B-VII.1 Library. The supercritical CO2-cooled fast reactor core is characterized in view of important safety parameters such as the reactivity worth of reactivity control systems, fuel temperature coefficient (FTC), coolant temperature coefficient (CTC), and coolant temperature-density coefficient (CTDC). We can say that the suggested core has inherent safety features and enough flexibility for load-following operation.

Research of PM-ADC scheme for optimized computational CFD analysis in fast reactor cores

Several challenges exist in the analysis of the sodium-cooled fast reactor (SFR) core using computational fluid dynamics (CFD)codes. These include the intricate task of geometric modeling and mesh partitioning and the substantial computational resources required by CFD calculations. These challenges have been shown to create significant impediments to the practical analysis of the SFR core. This paper proposes a CFD scheme called the Parameterized Modeling and Automated Distributed Computing Approach (PM-ADC) to address the challenges. This approach includes a fuel assembly model tailored for the fast reactor core and incorporates parameterized geometric modeling coupled with a system to automate the distribution of parallel computing. The PM-ADC scheme is divided into two critical components: rapid preprocessing design and the generation of geometric models and meshes for core assemblies, enabled by Parameterized Modeling (PM). In contrast, the Automatic Distributed Computing (ADC) component is responsible for the enhancement of computing efficiency and resource utilization. This is achieved by the automatic allocation and distribution of CFD computing resources within a partitioned domain, resulting in the completion of CFD analyses for all fuel assemblies at full height in a significantly reduced timeframe. The performance of the PM-ADC scheme is analyzed on CFD calculations on representative areas of 81 fuel assemblies within the core of the Chinese Experimental Fast Reactor (CEFR). The results show that the PM-ADC scheme can accurately and efficiently simulate the flow and heat transfer characteristics of the CEFR fuel assembly, demonstrating its potential for high-fidelity analysis of large engineering systems. The study's findings also provide new insights and tools for the ongoing development and analysis of SFRs, contributing to the evolution of nuclear reactor design and safety.

The multigroup library processing method for coupled neutron and photon heating calculation of fast reactor

To accurately calculate the heating distribution of the fast reactor, a neutron-photon library in MATXS format named Knight-B7.1-1968n × 94γ was processed based on the ENDF/B-VII.1 library for ultrafine groups. The neutron cross-section processing code MGGC2.0 was used to generate few-group neutron cross sections in ISOTXS format. Additionally, the self-developed photon cross-section processing code NGAMMA was utilized to generate photon libraries for neutron-photon coupled heating calculations, including photo-atom cross sections for the ISOTXS format, prompt photon production cross sections, and kinetic energy release in materials (KERMA) factors for neutrons and photons, and the self-shielding effect from the capture and fission cross sections of neutron to photon have been taken into account when the photon source generated by neutron is calculated. The interface code GSORCAL was developed to generate the photon source distribution and interface with the DIF3D code to calculate the neutron-photon coupling heating distribution of the fast reactor core. The neutron-photon coupled heating calculation route was verified using the ZPPR-9 benchmark and the RBEC-M benchmark, and the results of the coupled heating calculations were analyzed in comparison with those obtained from the Monte Carlo code MCNP. The calculations show that the library was accurately processed, and the results of the fast reactor neutron-photon coupled heating calculations agree well with those obtained from MCNP.

A revised analysis towards accurate estimation of isothermal temperature coefficients in fast reactors having different power levels and core sizes

The accurate prediction of safety coefficient such as isothermal temperature coefficient of reactivity is an important requisite in the safe operation of any reactor and also in optimizing the reactor theoretically. At IGCAR, a multi-purpose safety analysis code PREDIS is being used for this purpose by employing the spatial distribution of first order perturbation worth as input. Estimated isothermal temperature coefficient is validated against the measured value of a small 40 MWt carbide core reactor FBTR and 400 MWt FFTF. Though the results are conservative, PREDIS is found to be under predicting the safety coefficients. A detailed parametric study showed that isothermal temperature coefficient is under predicted through PREDIS analysis because of neglecting the contributions of removal worth from the non-fuelled regions surrounding the core. Relative contributions of non-fuelled regions to isothermal temperature coefficient have been systematically quantified in different fast reactor cores such as 40 MWt carbide core of FBTR, 400 MWt FFTF, 1250 MWt oxide core of PFBR and 1500 MWt oxide core of FBR 1&2. From the study, the leakage contribution to the surrounding non-fuelled region is found to be significant in a smaller core, hence the removal worth of these regions should not be ignored while estimating the safety coefficients. Conservative assumptions of considering only the core region for the prediction of safety coefficient is assumed to be good, only for the medium sized reactors, where the leakage contribution to the isothermal temperature coefficient is not that prominent.

SIMMER-IV application to safety assessment of severe accident in a small SFR

A sodium-cooled fast reactor (SFR) core has a potential of prompt criticality due to a change of core material distribution during a severe accident, and the resultant energy release has been one of the safety issues of SFRs. In this study, the safety assessment of an unprotected loss-of-flow (ULOF) in a small SFR (SSFR) has been performed using the SIMMER-IV computer code, which couples the models of space- and time-dependent neutronics and multi-component, multi-field thermal hydraulics in three dimensions. The code, therefore, is applicable to the simulations of transient behaviors of extended disrupted core material motion and its reactivity effects during the transition phase (TP) of ULOF, including a potential of prompt-criticality power excursions driven by fuel compaction. Several conservative assumptions are used in the TP analysis by SIMMER-IV. It was found out that one of the important mechanisms that drives the reactivity-inserting fuel motion was sodium vapor pressure resulted from a fuel-coolant interaction (FCI), which itself was non-energetic local phenomenon. The uncertainties relating to FCI is also evaluated in much conservative way in the sensitivity analysis. From this study, the ULOF characteristics in an SSFR have been understood. Occurrence of recriticality events under conservative assumptions are plausible, but their energy releases are limited.