Sodium-cooled Fast Reactor Core Research Articles

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.

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.

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.

A safety enhanced sodium-cooled MOX fueled fast reactor core concept

ABSTRACT We have developed a 750MWe sodium-cooled MOX fueled fast reactor core with enhanced safety against typical ATWS. Design goals were set as follows: 1) negative void reactivity for the loss of flow (ULOF); and 2) burnup reactivity less than 1 $ for the transient over power (UTOP). To achieve the goals the following were considered: 1) adoption of the axially heterogeneous core with a sodium plenum and a gas expansion module to reduce void reactivity; and 2) addition of minor actinides (MAs) to the internal blanket to reduce burnup reactivity. Configurations of the core were optimized as follows. First, void reactivity was reduced by a parameter survey for the inner core height, B-10 content of the upper shield, and standby position of the backup control rods. Second, burnup reactivity was reduced by a parameter survey for the MA content in the internal blanket, etc. We designed the safety enhanced fast reactor core that did not undergo sodium boiling if the ULOF occurred and we maintained fuel integrity for the UTOP by the transient analysis. By installing a self-actuated shutdown system in the backup control rods, the improved operational reliability of this shutdown system and the improved safety can be expected.

Impact of Thermal-Hydraulic Feedback and Differential Thermal Expansion on European Sodium-Cooled Fast Reactor Core Power Distribution

AbstractThe objective of this paper is to quantify the coupling effect on the power distribution of sodium-cooled fast reactors (SFRs), specifically the European SFR. Calculations are performed with several state-of-the-art reactor physics and Multiphysics codes (TRACE/PARCS, DYN3D, WIMS, COUNTHER, and GeN-Foam) to build confidence in the methodologies and validity of results. Standalone neutronic calculations were generally in excellent agreement with a reference Monte Carlo-calculated power distribution (from Serpent). Next, the impact of coolant density and fuel temperature Doppler feedback was calculated. Reactivity coefficients for perturbations in the inlet temperature, coolant heat up and core power was shown to be negative with values of around −0.5 pcm/°C, −0.3 pcm/°C, and −3.5 pcm/%, respectively. Fuel temperature and coolant density feedback was found to introduce a roughly −1%/+1% in/out power tilt across the core. Calculations were then extended to axial expansion for cases where fuel is linked and unlinked to the clad. Core calculations are in good agreement with each other. The impact of differential fuel expansion is found to be larger for fuel both linked and unlinked to the clad, with the in/out power tilt increasing to around −4%/+2%. Thus, while broadly confirming the known result that standalone physics calculations give good results, the expansion coupling effect is perhaps more than anticipated a priori. These results provide a useful benchmark for the further development of Multiphysics codes and methodologies in support of advanced reactor calculations.

Safety-oriented SFR core design: Methodology and application to transient bifurcations and stability maps analyses

Within the framework of the Generation IV Sodium-cooled Fast Reactor (SFR) R&D program of CEA (French Alternative Energies and Atomic Energy Commission), a methodology is proposed to early consider safety requirement in the undergoing reactor design process. Before the use of mechanistic tools (CATHARE, SIMMER, EUROPLEXUS, etc.) whose input deck elaboration requires an advanced knowledge of the reactor design, the methodology proposed in this article involves several physical tools simulating phenomena likely to govern the choice of design parameters. These tools are mostly based on reduced order models (ROM), meaning that they mainly involve low-dimensional modelings (mostly 0D and 1D) that are validated versus experimental results. They are gathered in a platform that covers all kind of accidental phenomenology, from the initiator (pump trip, reactivity insertion, local flow blockages, etc.) until the reach of a stable and coolable state after corium relocation. Thus, enabling a large number of simulations in a reasonable computational time, this fast-running platform makes possible the characterization of some major accident transient bifurcations (such as boiling onset, boiling stabilization, primary power excursion, molten fuel vaporization, corium axial relocation in transfer tubes, etc.), in terms of probability of occurrence and of consequences on the transient evolution. It also enables to identify the main physical parameters causing the bifurcations in order to allow a straight feedback on the core design and to give some orientations for future R&D studies. In this article, a focus is firstly made on some scenario bifurcations to illustrate the platform capabilities. The boiling onset and possible reactor state stabilization are studied. The possibility of primary power excursion depending on the core design and the fuel vaporization possibilities are assessed. This platform is also of great help in order to rapidly compare several core designs when facing accidental transients. An example is given by comparing a CADOR core design to an ASTRID-like core design through flow stability maps in natural circulation conditions obtained by the fast-running tool platform. These applications demonstrate the efficiency of the presented methodology integrating safety at the very first design stages, and at facilitating the safety-oriented design of SFRs.

MOOSE Reactor Module: An Open-Source Capability for Meshing Nuclear Reactor Geometries

The U.S. Department of Energy (DOE) Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has developed numerous physics solvers utilizing the open-source Multiphysics Object-Oriented Simulation Environment (MOOSE) framework for multiphysics reactor analysis. These solvers require input finite element meshes representing the discretized spatial domain. Typically, reactor analysts turn to licensed tools for the creation of reactor geometry meshes. Recently, open-source functionality has been added to the MOOSE framework to mesh common reactor geometries and improve MOOSE-based nuclear reactor application user workflows. The new functionality is primarily contained in the new Reactor module of MOOSE and includes support for hexagonal pins, assemblies, and cores, extended Cartesian geometry support, options for modeling static and rotating control drums within a hexagonal assembly, core periphery triangulation, and automatic tagging of pin, assembly, plane, and depletion regions for easier post processing of physics results. A set of reactor geometry mesh builder objects further streamlines the construction of hexagonal and Cartesian cores and allows mapping of materials to regions during mesh generation. The meshes produced with the MOOSE Reactor module may be used directly within MOOSE-based applications or exported as Exodus II files for use in other finite element solvers. The tools have been demonstrated and verified using a variety of NEAMS physics solvers on a range of reactor applications, including a sodium-cooled fast reactor core analysis using Griffin, a fast reactor assembly thermal deformation analysis using MOOSE Tensor Mechanics, and a heat pipe–cooled microreactor coupled analysis using Griffin, Bison, and Sockeye. MOOSE’s Reactor module provides significant advantages compared to the use of external meshing tools when analyzing Cartesian and hexagonal reactor lattices using MOOSE-based applications: immediate accessibility (open-source) to the end user, low barrier to entry for new users, speed of mesh generation, volume preservation of meshed fuel pins, and simplification of analysis workflow when used in conjunction with MOOSE-based applications.

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.

Neutronics simulation of China Experimental Fast Reactor start-up tests using FARCOB and ERANOS 2.1 code systems

The neutronics analysis of selected start-up tests conducted from 2010 to 2011 in China Experimental Fast Reactor (CEFR) is carried out using in-house FARCOB and European ERANOS-2.1 code systems as a part of IAEA coordinated research project (CRP) on “Neutronics Benchmark of CEFR Start-Up Tests”. This exercise has the main objective of validation and verification of physical models, neutronics simulation methodology including cross-section libraries that is used for fast reactor core simulations. It is challenging to simulate such a small compact core of enriched uranium with high neutron leakage having stainless steel radial reflector by using 3-D diffusion theory codes. This benchmark analysis used a new simplified approach such as a 3-step method and experimental method for simulating temperature and sodium void reactivity coefficients. Net criticality, control rod worth, temperature and sodium void reactivity coefficients and swap reactivity could be predicted well within their measured error values. This exercise provided an additional validation of the FARCOB system against small sodium-cooled fast reactor (SFR) cores, and it can be confidently used for the design and analysis of small experimental reactor cores.

Prevalence and factors associated with sarcopenia in a cohort of patients with chronic pancreatitis

The neutronics analysis of selected start-up tests conducted from 2010 to 2011 in China Experimental Fast Reactor (CEFR) is carried out using in-house FARCOB and European ERANOS-2.1 code systems as a part of IAEA coordinated research project (CRP) on “Neutronics Benchmark of CEFR Start-Up Tests”. This exercise has the main objective of validation and verification of physical models, neutronics simulation methodology including cross-section libraries that is used for fast reactor core simulations. It is challenging to simulate such a small compact core of enriched uranium with high neutron leakage having stainless steel radial reflector by using 3-D diffusion theory codes. This benchmark analysis used a new simplified approach such as a 3-step method and experimental method for simulating temperature and sodium void reactivity coefficients. Net criticality, control rod worth, temperature and sodium void reactivity coefficients and swap reactivity could be predicted well within their measured error values. This exercise provided an additional validation of the FARCOB system against small sodium-cooled fast reactor (SFR) cores, and it can be confidently used for the design and analysis of small experimental reactor cores.

Nuclear data uncertainty propagation applied to the versatile test reactor conceptual design

The Versatile Test Reactor (VTR) currently under development is a 300 MWth sodium-cooled fast reactor (SFR) fueled with ternary metal alloy fuel, which aims to accelerate the testing of advanced nuclear fuels, materials, instrumentation, and sensors in high flux environments that are necessary to license the next generation of advanced reactor concepts. To support the VTR design process, uncertainties associated with the nuclear data has been propagated through the reactor core neutronics calculation to global parameters of interest, such as the core multiplication factor, kinetic parameters, and various reactivity feedback coefficients, following the sensitivity based uncertainty propagation approach. By folding the sensitivity coefficients, separately computed by the generalized perturbation theory code PERSENT and Monte Carlo code Serpent 2, with the variance–covariance matrices from COMMARA-2.0, we obtain the reaction-wise, isotope-wise, and overall uncertainties for each response of interest due to nuclear data uncertainty. With Serpent 2, the statistical error of the uncertainty is obtained by propagating the statistical error of the sensitivity coefficients through the same process using a newly developed uncertainty propagation method. From both codes, the overall top uncertainty contributors are found to be the cross section of Fe-56 elastic scattering, Na-23 elastic scattering, and U-238 inelastic scattering. The large contributions of the Fe-56 elastic scattering cross sections to global parameters are due to its relatively large relative uncertainty of 5–10% in nuclear data and the large volume of Fe-containing reflector assemblies in the fairly compact VTR core design. Both codes agreed well for the overall uncertainty estimates of all responses of interest, except the delayed neutron fraction, prompt neutron generation time, and the coolant density feedback coefficient, where Serpent 2 yielded a much larger value than PERSENT due to the large statistical error of sensitivity coefficients. The calculated uncertainties are also compared to those associated with other SFR cores. Another outcome of this study is a variance–covariance matrix of reactivity coefficients, which can be used in the subsequent uncertainty propagation to the system level to investigate the impact of identified uncertainties on system responses in the safety analysis.

Core Materials for Sodium-Cooled Fast Reactors: Past to Present and Future Prospects

Abstract The Sodium-Cooled Fast Reactor (SFR) is the most promising and technologically evolved among the six nuclear reactor concepts selected by the Generation IV International Forum. Although it is compatible with a closed fuel cycle for sustainable energy production, its competitiveness depends on being able to achieve high fuel burn-up up to 200 GWd/t. This would enable efficient fuel utilization to minimize fuel cycle costs. Hence, development of optimized core materials, especially for fuel cladding tubes that are subjected to extreme conditions of intense fast-spectrum neutron irradiation, high temperatures, and mechanical and chemical fuel–cladding interactions, continues to be a priority worldwide. Resistance to void swelling, irradiation creep, and embrittlement are required to be enhanced to minimize dimensional changes and loss of ductility in core components, which limit achievable burn-up. Current SFR core materials comprise austenitic stainless steels (SS) and ferritic-martensitic (FM) steels. Cold-worked austenitic SS grades 304, 316, 316Ti, niobium-stabilized grades FV548, EI-847, EP-172, and Ti-modified 15Cr-15Ni SS, such as D9, 1.4970, 15/15Ti SS, and ChS-68, have been used for fuel cladding and ducts in SFRs built to date, and they withstand neutron irradiation damage up to 80–100 dpa. Subsequent improvements have been made by optimizing minor elements (titanium, silicon, and phosphorous) and multi-stabilization to develop advanced grades such as Si-modified 15-15Ti, Indian Fast Reactor Advanced Clad (IFAC-1), PNC-316, EK-164, and HT-UPS, which could support burn-up of 150 GWd/t. In the case of FM steels, several commercial grades are found suitable in view of their inherent void swelling resistance to 200 dpa, and are being developed for improved creep strength by oxide dispersion strengthening to realize burn-up of 200–250 GWd/t. This article presents the development of structural materials for SFR fuel pin cladding and ducts and associated enhancement of permissible fuel burn-up. Indian and international experience resulting from extensive in-pile and out-of-pile mechanical testing and fuel–cladding interactions has been covered.

A design study on a metal fuel fast reactor core for high efficiency minor actinide transmutation by loading silicon carbide composite material

ABSTRACT A metal fuel fast reactor core for high efficiency minor actinide (MA) transmutation was designed by loading silicon carbide composite material (SiC/SiC) which can improve sodium-cooled fast reactor (SFR) core safety characteristics such as sodium void reactivity worth and Doppler coefficient due to neutron moderation. Based on a 750 MWe metal fueled SFR core concept designed in a prior work, the reactor core loading fuel subassemblies with SiC/SiC wrapper tubes and moderator subassemblies was designed. To improve the reactor core safety characteristics efficiently, three layers of SiC/SiC moderator subassemblies were loaded in the core by replacing 108 out of 393 fuel subassemblies with the moderators. The reactor core with approximately 20 wt% MA-containing metal fuel satisfied all safety design criteria and achieved the MA transmutation amount as high as 420 kg/GWe-y which is twice as high as that of the axially heterogeneous core with inner blanket and upper sodium plenum, and two-thirds of that of the accelerator-driven system.

Experimental Investigation of the Friction Factor of Wire-Wrapped Bundles at Low Velocities Under Natural Circulation Flow Conditions

The sodium flow resistance in sodium-cooled fast reactor cores experiencing natural circulation conditions was measured for wire-wrapped 19- and 37-pin bundles using low-velocity water flows with Re <1000 and Re <750, respectively. The measurements were compared with predictions of existing wire-wrapped bundle friction factor correlations. The results show that the existing correlations usually underestimate the friction factors in the transition flow regime particularly for those with small transition Reynolds numbers from laminar to turbulent flow. The reason for the underestimation is that the transition Reynolds numbers observed in this study were much smaller than the predictions of all the existing correlations, and as a result, the transition flow at the small Reynolds number was treated as laminar or quasi-laminar flow by the correlations. In addition, the quasi turbulence in the early stage of transition flow should have a significant influence on flow resistance.

Neutron Balance Features in Breed-and-Burn Fast Reactors

Since nuclear power plants were first developed over 80 years ago, most have had thermal reactors with enriched uranium as the main fuel. As a result, huge amounts of depleted uranium must be stockpiled and stored. Fast reactors operating in breed-and-burn (B&B) conditions are one potential approach to reducing the need for storage facilities to hold spent nuclear fuel and maximize the effective use of uranium resources. Sustaining B&B operating conditions depends on the choice of core materials. Here, we perform neutron balance analyses to determine the best combination of fuel and coolant material. We identify the features of the reactor core needed to maintain B&B operating mode by the amount of neutron loss from the core and coolant heat removal capability. Finally, we demonstrate that a large or small sodium-cooled fast reactor core with metallic or enriched nitride fuel, respectively, can sustain the chain reaction and B&B operation mode.

Assessment of the neutron flux stability in a high power BN-type reactor in terms of modal spatial kinetics

The article discusses the neutron flux stability in the core of a high-power sodium-cooled fast reactor (of the BN-type) without feedbacks. The importance of this problem for high-power BN-type reactors is associated with the specific features of the layout of their cores, including a large diameter and height/diameter ratio about 5. The technique used to substantiate the stability of neutron fields is based on the analysis of the spectrum of the matrix of the system of spatial kinetics equations describing the core of a high-power BN-type reactor without feedbacks. A computational model of the spatial kinetics of a high-power BN-type reactor has been developed in the modal approximation based on the representation of an unsteady flux as a sum of orthogonal functions multiplied by time-dependent amplitudes. The eigenfunctions of the conditionally critical problem are used in the diffusion approximation, which in the discrete case form a complete system. The spectrum of the matrix of the system of ordinary differential equations describing the spatial kinetics of the reactor has been calculated. It is shown that the neutron flux in the core of a high-power BN-type reactor without feedbacks is stable. Test calculations have illustrated the damping of perturbations of the power distribution for a reactor in a critical state.

Study of Neutronic Performance of Sodium-Cooled Fast Reactor Design for Various Output Power Using Radial Fuel Shuffling Strategy

Study of neutronic performance analysis of Sodium-Cooled Fast Reactor (SFR) design based on the variations of output power using radial fuel suffling strategy has been performed. SFR is a type of the generation IV reactor that is widely researched for application on a commercial scale. The reactor design uses natural uranium as a fuel and Sodium as a coolant. The research has been carried out by using the SRAC code and JENDL-3.2 as a library with two dimensional R- Z suffling strategy of cylinder core for variations of power output 300, 350, 400, 450 and 500 MWTh. The neutronic parameters such as a multiplication factor (k-eff and k-inf) and burn-up analysis are observed. The SFR core is separated into 10 regions having the same volume fraction in the radial direction. At beginning of burn-up process, the reactor core is only filled with natural uranium fuel called fresh fuel and prepared for the one cycle of 10 years of burn-up time. The burn-up result in the first region is shifted into the second region, the burn-up result in the second region is shifted into the third region, and so on until burn-up result into the tenth region. The burn-up result in the tenth region is removed from the reactor core, then the first region can be filled with fresh natural uranium fuel and so on up to 10 times fuel cycles of reactor operation. The neutron calculation results indicate that the multiplication factors (k-eff and k-inf) in a critical condition are occurred on the 300 MWTh of output power. Overall, the output power of 300 MWTh has requirements and a greater chance of being operated for SFR designed.

Calculation studies of coated particles performance in sodium-cooled fast reactor

Abstract This paper presents results of calculation studies of the viability of coated particles in the conditions of the reactor core on fast neutrons with sodium cooling, justifying the development of the concept of the reactor BN with microspherical fuel. Traditional rod fuel assemblies with pellet MOX fuel in the core of a fast sodium reactor are directly replaced by fuel assemblies with micro-spherical mixed (U,Pu)C-fuel. Due to the fact that the micro-spherical (U, Pu)C fuel has a developed heat removal surface and that the design solution for the fuel assembly with coated particles is horizontal cooling of the microspherical fuel, the core has additional possibilities of increasing inherent (passive) safety and improve the competitiveness of BN type of reactors. It is obvious from obtained results that the microspherical (U, Pu)C fuel is limited with the maximal burn-up depth of ∼11% of heavy atoms in conditions of the sodium-cooled fast reactor core at the conservative approach; it gives the possibility of reaching stated thermal-hydraulic and neutron-physical characteristics. Such a tolerant fuel makes it less likely that fission products will enter the primary circuit in case of accidents with loss of coolant and the introduction of positive reactivity, since the coating of microspherical fuel withstands higher temperatures than the steel shell of traditional rod-type fuel elements.

Optimization of multi-group energy structures for diffusion analyses of sodium-cooled fast reactors assisted by simulated annealing – Part I: Methodology demonstration

This study presents an approach to the selection of optimal energy group structures for multi-group nodal diffusion analyses of Sodium-cooled Fast Reactor cores. The goal is to speed up calculations, particularly in transient calculations, while maintaining an acceptable accuracy of the results.In Part I of the paper, possible time-savings due to collapsing of energy groups are evaluated using 24-group energy structure as a reference. Afterwards, focusing on energy structures with a number of groups leading to significant calculation speedups, optimal grid configurations are identified. Depending on a number of possible energy grid configurations to explore, the optimization is conducted by either a direct search or applying the simulated annealing method. Speedup and optimization studies are performed on a selected case of the Superphénix static neutronic benchmark by using the nodal diffusion DYN3D code. The results demonstrate noticeable improvements in DYN3D performance with a marginal deterioration of the accuracy.

A study on the radial expansion reactivity of a metal-fueled sodium-cooled fast reactor via a physics experiment

For a metal-fueled SFR (Sodium-cooled Fast Reactor), the negative reactivity feedback due to the thermal expansion of the support grid which is known as a radial expansion reactivity, provides a principal safety mechanism in most severe accidents. In this paper, an experimental model of the radial expansion reactivity in a metal-fueled SFR is proposed, and experimental models are established in a collaboration between Korea Atomic Energy Research Institute (KAERI) and Institute for Physics and Power Engineering (IPPE). Modeling of the radial expansion reactivity in a realistic experiment condition is very difficult due to physical limitation of experimental facility. Hence, a sparse perturbation method suggested by authors was employed for reasonable modeling of the radial expansion phenomenon. Measured values were compared with calculated ones using the MCNP transport code with the ENDF/B-VII.0 library. The energy-dependent radial expansion reactivity components of the planned experimental model showed good agreement with those of the target Prototype Gen IV SFR core. Errors of C/M (calculation/measurement ratios) varied from −4.3% to 9.2%. Considering the 1 σ uncertainty in the C/M, 5% in maximum, errors of calculation/measurement ratios were within 2 σ uncertainty range.