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In order to accurately calculate the photon heating distribution of fast reactor, the photon cross section processing code NGAMMA has been developed. NGAMMA can process MATXS format library and generate multi-group forms of photo-atomic cross sections, prompt photon production cross sections, delayed photon production cross sections, and neutron-photon KERMA factors for neutron-photon coupled heating calculations. An improved 21-groups photon library containing both prompt and delayed photon data is generated using the NGAMMA code based on the 1968n × 94γ MATXS format database. The photon library was verified using the fast reactor benchmark problem ZPR-6/7 and MET-1000. The results of the calculations show that the use of the 21-groups photon library generated by energy group collapse using 94-groups photon fluxes results in a significant improvement in the accuracy of the prompt photon heating calculations, especially in the non-fuel region, resulting in a relative error of less than 1% between the photon heating calculations and the MCNP calculations, compared to an error of 7% using the previous photon library. For the ZPR-6/7 benchmark problem, consideration of the delayed photons results in a 30% improvement in photon heating in the fuel region. For the MET1000 benchmark problem, the total power results of the neutron-photon coupling calculations based on the improved photon library are in good agreement with the MCNP, with a maximum relative error of no more than 3% in the total power of the fuel assembly. For reflector assemblies where photon power is a major contribution to heat release, the maximum relative error in prompt photon power is only −5.96%. Considering delayed photon significantly affects the photon heating, and for the MET1000 benchmark problem, considering delayed photon results in an overall increase of more than 30% in the photon power of the fuel assembly. • Developed a high-precision photon library processing code, NGAMMA. • The delayed photon heating was calculated distribution for fast reactors accurately. • Neutron and photon coupled heating calculations were performed for the MET1000 benchmark problem. [ABSTRACT FROM AUTHOR]

Safety assessment of fast reactors involves analyzing Hypothetical Core Disruptive Accidents (HCDA) that could result in reactor core disassembly. HCDA is generally analyzed in three distinct phases that include pre-disassembly, disassembly and post-disassembly phases. In the present work, a new coupled neutronics-hydrodynamics computer code called FA st R eactor DIS assembly (FARDIS-1) is developed in 2-D cylindrical coordinate geometry to model the super-prompt critical power excursion during the disassembly phase of HCDA in a sodium-cooled fast reactor (SFR). The modelling approach followed is essentially similar to the one used in VENUS-II code but with a few improvements. The code is benchmarked against the predictions of VENUS-II for a sample case of HCDA in the FFTF reactor model. This is followed by an analysis of the disassembly phase in a medium-sized (500 MWe) mixed oxide fuelled fast reactor. Two cases are assumed at the beginning of the disassembly phase: a sodium voided core (conservative) and a partially voided core (realistic) and the influence of several core neutronics parameters on the transients is elucidated through parametric studies. The peak power, thermal energy release, peak pressure, maximum temperature and work-potential are estimated and their trends are discussed. • A coupled neutronics-hydrodynamics computer code is developed to model the disassembly phase of HCDA in fast reactors. • Hydrodynamics module uses a Eulerian framework instead of Lagrangian approach. • Benchmarked against the predictions of VENUS-II for HCDA in FFTF reactor model. • Analysis of disassembly phase in 500 MWe MOX fuelled fast reactor for initially voided and partially voided core scenario. • Parametric studies on influence of core neutronics parameters on peak power, thermal energy and work-potential during HCDA. [ABSTRACT FROM AUTHOR]

Using dry recycled powders for uranium and plutonium mixed oxide (MOX) fuel production can reduce unnecessary storage and accountability of nuclear material in facilities. The shrinkage behavior of green compacts of dry recycled powders differs from that of conventional raw powders because the dry recycled MOX powder is obtained from the fabrication scrap of sintered pellets. The shrinkage behavior of dry recycled MOX powder has been investigated by dilatometry. Based on the shrinkage curves, sintering apparent activation energies were evaluated using the master sintering curve (MSC) and the constant rate of heating methods. The obtained values were higher than the energy evaluated for raw powder experiments. The sigmoid sintering prediction equation using the MSC function was constructed. The accumulation of data on the activation energy for various sintering conditions will lead to the wide application of this prediction formula in the future. [ABSTRACT FROM AUTHOR]

T91 steel is a candidate material for structural components in lead-bismuth-eutectic (LBE) cooled systems, for example fast reactors and solar power plants [1]. However, the corrosion mechanisms of T91 in LBE remain poorly understood. In this study, we have analysed the static corrosion of T91 in liquid LBE using a range of characterisation techniques at increasingly smaller scales. A unique pattern of liquid metal intrusion was observed that does not appear to correlate with the grain boundary network. Upon closer inspection, electron backscatter diffraction (EBSD) reveals a change in the morphology of grains at the LBE-exposed surface, suggesting a local phase transition. Energy dispersive X-ray (EDX) maps show that Cr is depleted in the T91 material near the LBE interface. Furthermore, we observed the dissolution of all Cr-enriched precipitates in this region. Although the corrosion is conducted in an oxygen deficient environment, both scanning transmission electron microscopy (STEM) and atom probe tomography (APT) reveal a thin surface oxide layer (presumably wüstite) at the LBE-steel interface. Using electron energy loss spectroscopy (EELS) in the STEM, as well as APT, the atomic scale elemental redistribution and 3D morphology of the corrosion interface is investigated. By combining results from these different techniques, several types of oxide phases and structures can be identified. Based on this detailed nano-scale information, we propose potential mechanisms of T91 corrosion in LBE. [Display omitted] [ABSTRACT FROM AUTHOR]

Sodium spray fire is a significant containment-related safety concern in sodium-cooled fast reactors (SFRs). The droplet size distribution of sodium spray is an important factor affecting the combustion intensity of sodium spray fire. Due to the unstable chemical properties of sodium, water is usually used to simulate sodium to carry out spray experiments, and the droplet size relationship between them needs to be obtained. Based on the unstable wave theory, the droplet size similarity model characterized by liquid physical property parameters and atomization pressure difference is developed and verified from the two aspects of pressure and physical property parameters, and compared with existing models. Furthermore, an experimental facility of water atomization using a nozzle to simulate a break is built and the droplet size of water atomization under the differential pressure range of 1.0 bar–6.0 bar is obtained. The droplet size similarity model is used to predict the sodium atomization droplet size at 300–600 °C under different pressures based on these experimental data. • A droplet size similarity model predicting sodium spray droplets is developed based on the theory of unstable wave. • A set of experimental equipment and experimental method for water simulation sodium leakage and spray are established. • The sodium droplet size at 300–600 °C is predicted by droplet size similarity model. [ABSTRACT FROM AUTHOR]

In this paper, a thorium-based fluorinated molten salt fast neutron reactor is taken as the research object, and the utilization of thorium resources, reuse of transuranic waste (TRU), and transmutation of minor actinide (MA) in fast reactor are the research purpose. The fast reactor adopts a multi-layer core structure, in which the fuel is divided into an inner fission fuel zone and an outer breeding fuel layer, respectively loaded with fluorinated molten salts LiF+ThF 4 +XF 4 (where X is 233U, TRU, or 233U+MA) and LiF+ThF 4. In addition to serving as nuclear fuel, these two molten salts also act as coolants in the primary circuits. Through the cross sections of nuclear reactions, transmutation chain of actinide nuclides, and the particle population density formula of actinide nuclides in nuclear reaction equilibrium, the possible nuclear reactions of actinide nuclides in thorium-based fast reactors are analyzed. The calculated results show that the thorium-based fast reactor has the following characteristics: fast neutron energy spectrum, negative temperature reactivity coefficient, effective utilization of thorium resources, the ability to breed fissle fuel such as 233U, production of high-purity radioisotope 238Pu material, and the ability to incinerate Pu and MAs in the TRU spent fuel. In the 233U fuel scheme, the 233U fuel is breeded, breeding ratio of fissile fuels is 1.21, the reaction products contain very little Np, and almost no Pu, Am, and Cm long-lived radionuclides. In the TRU fuel scheme, the annual capacity to incinerate MAs and Pu is 1072.6 kg, while producing a large amount of 233U and 233Pa fuel (together 1162.5 kg/year). In the 233U+2.0 wt% TRU-MAs fuel scheme, 127.35 kg of MA waste can be incinerated annually, and their transmutation rate is 15.95%, which is equivalent to the annual production of MA nuclear waste from 4.9 conventional 1000 MWe light water reactors. Here, TRU-MAs is the mixtured MA fuels after separation of Pu from the TRU fuel. At the same time, radioactive fuel 238Pu, which can be used for space nuclear power, is produced with an annual yield of 87.89 kg and an isotopic abundance of 87.57%. If the added TRU-MAs fuel is replaced with 2.0 wt% 237Np fluorinated salt, the annual yield and isotope abundance of 238Pu are further increased to 130.24 kg and 95.4%. • Utilization of Th resources.In a molten salt fast reactor (MSFR), the fertile material is all thorium fuel. • Thorium-based MSFR can massly produce U-233 fuel, which can be used as the fissile fuel for fast reactors in the developing of thorium resources. • A large amount of Pu or MA material in nuclear waste, can be incinerated by thorium-based MSFR. • Transmutation rate of MA is relatively high, 20.0%, 21.3%, 19.0% and 15.8% for Np-237, Am-241, Am-243, and Cm-244. • In thorium-based MSFR, high-purity Pu-238 materials can be mass-produced by selecting specific MA materials. [ABSTRACT FROM AUTHOR]

The method of computer analysis of system parameters at formation of protective coating on the shell of fuel element of fast reactors is presented, which can serve as a prerequisite for development of cognitive architecture of control of nuclear power engineering processes by control of composition of gas phase in reactor core, containment, other tanks and making decisions on results of process parameters deviation from stationary function. The work was carried out to justify the prospects of using a liquid metal sublayer (LMS) instead of a gas sublayer in fuel elements, which allows, due to the high thermal conductivity of the liquid metal, to increase the width of the gap between the fuel and the shell and, therefore, to increase the fuel element life limited by the contact of the fuel with the shell. At the same time, safety in case of UTOP (uncontrolled increase of power), ULOF (termination of coolant flow through the reactor) accidents is increased due to significant decrease of temperature in the fuel center. However, the possibility of using LMS is limited by the substantial mass transfer - corrosion dissolution in the lead of the components of the fuel element shell in the hot zone and precipitation in the cold zone. In this article, the basis for computer analysis was studies on the absorption of gas (nitrogen) when forming a protective ceramic coating of zirconium nitride on the surface of the shell of a pipe-in-pipe type fuel element simulator to suppress the corrosion process from eutectic alloy lead-magnesium-zirconium. Abnormal intense decrease of nitrogen pressure (p, kPa) in expansion tank with Pb-Mg-Zr alloy at increased temperature (about 950 K) according to linear kinetic law (t, c), connected with its interaction with system components, was revealed: p, kPa =−0.0002⋅t−77.867 R2=0.942. Method of diagnostics of system parameters at formation of anticorrosive coating by nitrogen absorption is developed, empirical dependencies of process flow in wide range of temperature are built. [ABSTRACT FROM AUTHOR]

This time, several projects dedicated to fast reactors (FRs) with lead and lead-bismuth coolants, BREST-OD-300, SVBR-100, RBETS-M, BRUTs are proposed in Russia. They will have several valuable consuming properties: chemical inertness, low neutron absorption, low activation and others. Usage of lead coolant leads also to the possibility of achieving a hard-enough neutron spectrum that allows increasing the incineration probability of 241Am, 237Np and other low fissile actinides. High power FRs have large-sized cores that limits the value of neutron energy by the value of 0.5 MeV, which is insufficient for incineration of above mentioned actinides. Small and medium power reactors have smaller cores and, respectively, have harder neutron spectra. Usage of lead and low moderating innovative fuel allow further increasing neutron energy to the value inquired for low fissile actinides incineration.In the paper a possibility of obtaining a neutron spectrum with the average value of neutron energy higher than 0.5 MeV is considered. It is performed in the frame of the project of BRUTs series reactors, i.e. small power LFRs proposed in the Obninsk Institute for Nuclear Power Engineering. A scope for achieving a hard neutron spectrum in the reactor BRUTs-25 core of small sizes, D×H = 0.5×0.4 m2, is shown. Findings are that in the core fueled with Pu-Am-Np-Zr alloy and cooled with enriched lead, 208Pb, the average value of neutron energy, , is high-enough, about 0.95 MeV, as well as the share of fast neutron, E n >0.8 MeV, in the neutron spectrum is very high, about 40%. In such of conditions, 241Am and 237Np incineration probabilities in the core center are higher than 50% and values of their one-group fission cross-sections are higher than 0.7 barn. This circumstance allows burning 15-16wt% of low fissile isotopes for one campaign of BRUTs-25. The presence of 241Am in the fuel, in a quantity of 28.7 kg, allows transmuting about 8.6 kg of its mass for one reactor campaign that lasts about 3 years (1000 effective days). It means that to transmute the quantity of 241Am produced by the VVER-1000 for one year, equal to 25.8 kg, it will be needed about 3 BRUTs-25 type low power reactors operating for 1000 effective days. [ABSTRACT FROM AUTHOR]

• An optimization frame of FR design was built, and a novel strategy of encoding was suggested to support this frame. • A simplified TSP problem and enumeration on ABTR loading were employed as verification. • The loading strategy optimization of NpO2 in CEFR for Pu-238 production were performed based on this optimization frame. The neutronic design of fast reactor (FR) considers complex objectives and constraints but the operating experience of FRs is much lower than that of commercial reactors. It is complicated and time-consuming to perform optimization design manually. This study applies Genetic Algorithm (GA) and modifies the search method in coding strategy. Distinguished from previous research, a new conception of coding that maps continuous decimal integers into a loading pattern and searches using uniform sampling is developed. This coding strategy realizes the coverage of the multidimensional search space, so that it is possible to deal with the problems of continuous and discrete variables. With this coding strategy, initial designs are generated uniformly and automatically, which leads to better global convergence. Multiple design variables, objectives, and constraints are considered to cater to the requirements of practical application. With the above coding strategy and considerations, a framework that couples the SARAX neutronics analysis code and the GA module in the DAKOTA toolkit is constructed and tested on several cases for verification. It is easy to add any of the design variables, optimization objects, objectives and constraints into the framework and set the optimal range. Both the neutronic software and optimization have been fully validated and performed as a black box. These features are beneficial to the design of FRs, which is lacking of experience. Last but not the least, is a demonstration of this framework in which Pu-238 production is optimized based on CEFR. The searching result showed the excellent properties of this work. [ABSTRACT FROM AUTHOR]

In this paper, design studies for a 50 MW th Molten Salt Reactor (MSR) have been carried out for the eutectic point of the molten salt mixture. Neutronic calculations are performed with the 19.75% enriched uranium and 100% 7Li isotope contained. The MCNP6 nuclear code was used with the ENDF/B-VIII nuclear data library to determine geometry dimensions and criticality. The time-evolution of Pu and other heavy isotopes in the reactor are calculated with the interface code MCNPAS. Four models are investigated with different reactor vessel materials: Model 1: Ni alloy (NiCrW-Hastelloy steel), Model 2: Beryllium, Model 3: Graphite and Model 4: Silicon Carbide (SiC). For the respective models, time-dependent criticality calculations are performed with a startup criticality value of k eff = 1.0262, 1.0298, 1.0404, and 1.0223. The 235U consumed for the corresponding models over the 10 years of reactor operation are 257.9 kg, 258.8 kg, 259.4 kg and 257.9 kg, respectively. At the same time, the 238U consumptions are 400.6 kg, 380.0 kg, 379.8 kg, and 400.9 kg, respectively. The amount of the higher quality new fuel (239Pu) produced for 10 years is calculated as 129.10 kg, 127.41 kg, 122.95 kg and 128.66 kg, for the respected models. • Study of a 50 MWth Molten Salt Fast Reactor (MSFR) using the MCNP neutronic code. • Performed extended reactor operation (over 10 years). • Investigated 4 different reactor vessel materials (Ni-alloy, Be, graphite and SiC). • Conducted 19.75% enriched U and 100% 7Li eutectic molten salt mixture. • Produced high-quality plutonium and heavy isotopes. [ABSTRACT FROM AUTHOR]

• Potential absorbers for Gen-IV fast reactors are evaluated in detailed designs. • Moderators are used to optimizing control rods. • Neutronic, safety, and waste management of innovative designs are assessed. The characteristics of Generation-IV fast reactors strongly impact the design of its control rods. The traditional control rod is a cluster of vented pins with boron carbide (B 4 C) as the absorber. Due to the gas release, the neutron-induced swelling, the melting risk, and the high loss of the reactivity worth, B 4 C impact the safety performance of control rods and thus limit their operating lifetime. Various alternative absorbers to B 4 C are assessed in detailed designs according to their neutronic performance, safety performance, and waste management. The hafnium diboride (HfB 2) process a high thermal conductivity, a high melting point, and a small neutron-induced swelling, which contributes to an enhanced safety performance. The europium oxide (Eu 2 O 3) process a very small loss of the absorption ability for a long-time utilization. Hafnium hydride (HfH 1.62) has a good neutronic performance, but its hydrogen desorption issue should be further investigated. The absorption cross-section decreases generally with the neutron incident energy. Moreover, an important spatial self-shielding effect is found in the control rod, which leads to a peak temperature and peak burnup in the outer zone while a non-optimized utilization of absorber in the inner zone. In this work, the hydrogen-containing moderator is used in independent pins to replace some absorber pins in control rods. The moderator is able to increase significantly the absorption ability of Eu 2 O 3 and of gadolinium oxide (Gd 2 O 3). The moderator optimizes the utilization of absorbers, such as HfB 2 and B 4 C, and save their investment. By homogenizing the absorption distribution, the application of moderators has only a limited influence on safety performance or waste management. [ABSTRACT FROM AUTHOR]

• New SCALE/AMPX capability has been developed for advanced fast reactor analysis. • New analytic probability table method based on narrow resonance approximation was developed. • The AMPX 1597-group library have been developed for fast reactor analysis. Reaction rate analysis was performed to verify the AMPX 1597-group library and cross section processing. • The SCALE results are consistent with the continuous energy Monte Carlo results. The SCALE/AMPX multigroup (MG) cross section processing procedure has been updated to minimize reactivity differences for fast reactor designs and boiling water reactors (BWRs) with very high void fractions to provide excellent agreement with continuous-energy reference calculations. SCALE MG calculations are widely applied to thermal spectrum light-water reactor (LWR) systems as well as fast spectrum metallic systems of interest to National Nuclear Security Administration. With growing interest from industry and regulators in applying SCALE for the design of fast spectrum reactors, both sodium and molten salt, and in the licensing of power up rates for BWRs, it is desirable to review the SCALE/AMPX procedure for unresolved resonance self-shielded data and high energy neutron spectra. The data were improved by generating MG unresolved resonance data based on the analytic probability table method with the narrow resonance approximation as well as the use of a very fine group structure typically used in fast system analysis. This study focused on verifying the probability table and the SCALE/AMPX MG cross section processing procedure by performing reaction rate analysis and benchmark calculations for various fast reactor and high void BWR systems. Results indicate that the improved SCALE/AMPX MG cross section processing provides excellent results for the fast reactor analysis. [ABSTRACT FROM AUTHOR]

• The probability table method was adopted and the self-shielding factors for different reaction types were distinguished. • The Energy-collapsing SuPer-Homogenization technique for the group condensation was used. • The accuracy of Doppler constant and power/reaction-rate distribution was significantly improved. For better accuracy, the few-group cross-section generation module, named TULIP, in the SARAX code system was further improved. In the new TULIP, the pure deterministic method was used for the spectrum calculation and self-shielding treatment. The improvements contained two parts. One was the method for resonance treatments in the unresolved resonance region. The probability table method was adopted and the self-shielding factors for different reaction types were distinguished. The background cross-section interpolation points of major actinides were refined for the Bondarenko interpolation. These techniques significantly improved the accuracy of self-shielded cross-sections and contributed to better Doppler constant calculation. The other improvement was the method of Energy-collapsing SuPer-Homogenization technique for the group condensation. It preserved the reaction rates conservation in the multigroup-fewgroup equivalence and improved the accuracy in the strong heterogeneous problems. Several numerical and experimental benchmarks have been done to test the improvements. The results showed that the accuracy of homogenized cross-sections, k -infinity and Doppler constant in subassembly calculation was significantly improved, also for the k -effective, power distribution and reaction rate distribution in the core calculation. [ABSTRACT FROM AUTHOR]

• The design of the Core Assembly Deformation Test Facility was described in detail. • The conditions of assembly bending and thermal bowing tests were introduced briefly. • The heating and measurement technologies and the layout of sensors were introduced. China Demonstration Fast Reactor (CFR600) is under construction, and the Sub-assembly Deformation Analysis Program (SDAP) developed by China Institute of Atomic Energy is used to optimize the core design and operation management of CFR600. The Core Assembly Deformation Test Facility (CADTF), designed and built by Xi'an Jiaotong University, was selected to carry out a series of assembly bending and thermal bowing tests to provide validation problems for the SDAP. The specimen assembly was specially designed for better non-intrusive optical temperature measurement of the inner wrapper tube. The automatic heating control method and a lot of new measurement technology were applied into the tests to gain abundant and accurate test data. [ABSTRACT FROM AUTHOR]

Highlights • A first order perturbation theory is developed for material displacement worth evaluation. • A FOP theory code RAINBOW is developed to evaluate assembly bowing reactivity in SFR. • The RAINBOW code has been verified against MCNP6 and PROTEUS-SN. • RAINBOW calculation is several orders of magnitude more efficient than MCNP6. Abstract A perturbation theory code RAINBOW was developed to evaluate reactivity feedback effects due to assembly bowing in sodium-cooled fast reactors. During startup and nominal operations, bowing of the core assemblies is induced by thermal gradients and the core geometry is distorted. Depending on the design of the core restraint system, the core can be characterized as having a "free flowering" or a "limited free bow" design. In either case, accurately quantifying the reactivity feedback is required to support the licensing basis of the reactor system through limiting transient and severe accident analyses. Within the RAINBOW code, assembly bowing is modeled by shifting axially discretized assembly segments, and the heterogeneous assembly configurations are represented using an unstructured finite element mesh. This unique approach provides a convenient and general form for tracking the material movements within a heterogeneous hexagonal assembly design. Additionally, this methodology eliminates the need to consider numerous intersections of geometrical meshes and the shifting of material boundaries. The forward and adjoint neutron flux distributions that are required for perturbation theory are reconstructed using a combination of core-wide homogeneous-assembly solutions from the VARIANT transport code and the unstructured, heterogeneous single-assembly solutions from the PROTEUS-SN transport code. Code-to-code verification tests of this methodology were performed using two- and three-dimensional core models. The perturbation theory results agreed well with the reference MCNP6 and PROTEUS-SN solutions, which were obtained through direct eigenvalue calculations of the unperturbed and perturbed core states. Finally, by comparing solution metrics, RAINBOW offers a similar degree of accuracy to MCNP6 with a several order of magnitude reduction in computational time and resources. [ABSTRACT FROM AUTHOR]

Highlights • A deterministic methodology is proposed for control rod modeling in fast reactors. • The lattice level can accurately calculate self-shielding effects in control rods. • The core level is numerically and experimentally validated in different reactors. • The simulation accuracy in the control rods is improved with this methodology. Abstract High accuracy neutronic simulations are required for the development of Generation-IV sodium cooled fast reactors (SFR). This paper therefore focuses on the development and validation of an improved calculation method in APOLLO3® deterministic code to predict the reactivity control system in SFR. This method relies on the "lattice-core" paradigm and the generation of few group effective cross-sections. Those are computed over small lattice geometries with detailed descriptions of structures but no heterogeneity correction method. Two models are considered at the core level: the first one relies on homogeneous descriptions for all assemblies (fuel, control rods, reflector, diluent, etc.), whereas the second one takes into account the heterogeneity of absorber pins (semi-heterogeneous description). In order to assess the robustness of calculation methods, a validation work is achieved for three distinct SFR cores. A numerical validation is first performed for the SFR-3600-MOX core taken from the international WPRS benchmark. Then an experimental validation work is conducted for the control rods measurements from SUPERPHENIX start-up experiments. The last case of application is the ASTRID CFV core. At the cross-sections generation stage, the results show a good coherence with reference Monte Carlo calculations for both spatial and energy distributions. At the core level, it is shown that the homogeneous description of the control rods induces an overestimation of the reactivity worth while the semi-heterogeneous model significantly improves the results. [ABSTRACT FROM AUTHOR]

Highlights • Calculation schemes for absorber depletion in sodium fast reactors are proposed. • The variation of spatial self-shielding effect is important for absorber depletion. • Significant improvements are found for the approach with updating of cross-sections. • Different designs are validated from cluster to core configurations. Abstract Neutronic simulations with high accuracy are required for the development of future sodium fast reactors. The "lattice-core" paradigm used for deterministic codes is able to save computation resources and improve calculation speed, but it defies the accurate depletion calculation. This paper focuses on the development and validation of an improved method, in the deterministic code APOLLO3®, for the depletion calculation of control rods. The absorber depletion is validated for different control rods designs from cluster type configuration to core configuration. The lattice calculation in APOLLO3® shows a good adaptability to complex geometries with alternative absorbers and moderators. The self-shielding effect and its variation under neutron irradiation is the key issue for the accurate simulation of control rods. Therefore, updating cross-sections according to the burn-up improves significantly the accuracy on the depletion calculation. [ABSTRACT FROM AUTHOR]

• Lumped mass stick model for a pool-type fast reactor developed. • Simplified model verified through comparison of dynamic properties and response-history analyses. • Influence of fluid–structure interaction found small. • Substantial response reduction due to seismic isolation may enable standardized reactor designs. Nuclear energy is being considered as a frontrunner towards achieving the zero carbon goal. A sustainable use of nuclear energy is possible when a reactor design is deployed across regions with varying seismic hazard. Seismic isolation technology has been proven to reduce earthquake loads in the structures, and is being seen as a key towards the standardized reactor designs. This paper investigates the influence of seismic isolation on the response of a generic fast reactor (GFR) similar to the prototype fast breeder reactor (PFBR) currently under commissioning at Kalpakkam, India. This reactor comprises a head-supported main vessel that houses other reactor components (e.g., core, inner vessel). The major vessels of the reactor are partially filled with liquid sodium. In order to understand the influence of seismic isolation, simplified lumped mass stick models (LMSMs) for the GFR components were developed. These models were based on static analysis results obtained using corresponding finite element (FE) models. The component models were assembled to generate the LMSM and FE model for the GFR. The LMSM was verified through the comparison of dynamic properties obtained using the FE model, and through the comparison of response-history analysis results. These models considered fluid masses lumped at respective shells, thereby ignoring the fluid–structure interaction (FSI) effects. A representative system was analyzed with and without considering FSI effects. These effects did not influence acceleration response significantly. Fixed-base (FB) and isolated base (IB) configurations of the LMSM for the GFR were subjected to a series of response-history analyses. Seismic isolation led to a substantial reduction (e.g., up to 10 times) in the median floor accelerations, particularly for lower sliding friction and longer sliding period. The floor spectral ordinates for the IB configuration of GFR subjected to 1.00 g shaking were less than that for the FB configuration of GFR subjected to 0.25 g shaking at most periods, suggesting that the PFBR can be deployed at locations with seismic hazard much greater than that at Kalpakkam if suitable isolation systems are used. [ABSTRACT FROM AUTHOR]

The effect of void propagation to sodium void reactivity in fast reactors with sodium-plenum has been investigated. The sodium boiling occurs around the top of reactor core, and propagates in the sodium plenum. In the transient analysis of severe accidents the sodium void reactivity in sub-regions of the reactor is calculated considering the void propagation and it is seen the reactivity is different from the void reactivity postulated that only sodium in a sub-region is boiling. We found that the two results of void reactivity are different by up to twice depending on the location of sub-region in the sodium plenum. [ABSTRACT FROM AUTHOR]

As the first sodium cooled fast reactor in China, China Experimental Fast Reactor (CEFR) has many characteristics, such as high power rating, high neutron flux and high neutron leakage as its hard spectrum. These leakage neutrons can be used to produce 89 Sr by fast neutrons (n, p) reaction. By means of some special designs, including designing optimized target assembly, choosing reasonable irradiation location and appropriate irradiation cycle length, in order to improve specific activity of 89 Sr. The MNCP code was used to the calculations coupled ORIGRN2 procedure. The higher activity of 89 Sr can be obtained by these optimization designs, so it is feasible to produce 89 Sr in the fuel region of CEFR. [ABSTRACT FROM AUTHOR]

This paper introduces the verification and validation (V&V) work on the newly developed fast reactor code SARAX. The verifications were done based on numerical benchmark calculations and code-to-code comparisons with the Monte-Carlo code. The validation works were performed and compared with evaluated reactor physics experiments. In this paper, results from different calculation schemes are discussed, and the uncertainty from nuclear data is quantified. The results show that SARAX has good accuracy in modeling a sodium fast reactor regardless of the integration parameters, such as k eff , or the distributed parameters like the sub-assembly power/reaction rate. The uncertainty quantification shows that, for a MOX-fueled SFR, the greatest uncertainty in contributed nuclear data is from the inelastic scattering cross section of U-238, which requires further improvement to reduce the uncertainty of k eff . [ABSTRACT FROM AUTHOR]

A computer code for the analysis of the overall irradiation performance of a fast reactor mixed-oxide (MOX) fuel element was coupled with a specialized code for the analysis of fission product cesium behaviors in a MOX fuel element. The coupled code system allowed for the analysis of the radial and axial Cs migrations, the generation of Cs chemical compounds and fuel swelling due to Cs-fuel-reactions in association with the thermal and mechanical behaviors of the fuel element. The coupled code analysis was applied to the irradiation performance of a fast reactor MOX fuel element attaining high burnup for discussion on the axial distribution of Cs, fuel-to-cladding mechanical interaction owing to the Cs-fuel-reactions and the Cs-Mo-O compound formed in the fuel-to-cladding gap by comparing the calculated results with experimental data obtained from post irradiation examinations. [ABSTRACT FROM AUTHOR]

In a pool-type fast breeder reactor, the shutdown by SCRAM involves crash cooling of primary sodium at the rate of 480 K/h. This high cooling rate can develop high thermal stresses at the thicker junctions that are immersed in sodium, due to a lower Fourier number and a higher Biot number, which results in higher Fitz number. Thermo-mechanical analysis of various critical locations within sodium with the transient loading, and parametric studies with varying cooling rates and thicknesses has been carried out. Due to the drop in sodium temperature during crash cooling, there is a fall in the sodium level, resulting from the volumetric contraction of sodium. Thus the locations of the hot pool components that are at the free level interface of sodium and argon experience an abrupt shift in the ambience, i.e., from sodium to argon cover gas. A numerical model has been developed to simulate the complex loading condition using CAST3M code. The model has been validated with theoretical equations and the overall heat balance equations. Transient analysis of all the hot pool components has been performed to estimate the structural damage due to thermo-mechanical behavior during crash cooling. Based on this study, it is concluded that for the future design of fast reactors with 60 years of design life, controlled cooling is not essential from the structural damage point of view. The recommendation simplifies the design of decay heat removal system. [ABSTRACT FROM AUTHOR]

In this paper thorium fuel feasibility in large scale Gas Cooled Fast Reactor (GCFR) is investigated. The neutronics benchmark used in this study, GFR2400, corresponds to a 2400 MWth GFR concept proposed by the French CEA. MCNPX computational code is used to design a 3D heterogeneous model of the GFR2400 core. A detailed feasibility analysis of the performance of thorium fuel cycle is performed by using thorium as an alternative fertile fuel for the natural uranium vector of the reference core design. The most essential neutronic parameters characterizing the core are determined both for beginning of life (BOL) conditions as well as during burnup. Also, a three-dimensional core cycle-by-cycle simulations is performed to allow explicit characterization of the core behavior and safety-related parameters during both open and equilibrium cycles. The thorium-based core shows favorable neutronic characteristics with an acceptable control and safety parameters for both BOL and open cycle states. The depressurization reactivity effect and core expansion coefficients (axial and radial) show improvement compared to the uranium-based core. However, this improvement is compensated by the deterioration in the effective delayed neutron fraction (β-eff) and the Doppler reactivity Effect. The results of isotopic transmutation and fuel burnup confirm the capability of the core to work in both open and closed cycles and to self-recycle its own MA vector and plutonium of LWRs. However-as is the case in other fast reactors including the uranium-based GFR2400 core, the fuel cycle closure causes safety related parameters to degrade. [ABSTRACT FROM AUTHOR]

The Experimental Breeder Reactor-II was a sodium-cooled, metal-uranium-fueled fast-neutron reactor designed and built by Argonne National Laboratory. The reactor achieved initial criticality September 30, 1961, and continued operation until 1994. The reactor thermal power limit was 62.5 MW with a corresponding electrical output of 19 MW. To preserve important EBR-II reactor physics information, a benchmark evaluation is underway for proposed inclusion in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. Coupled with development of the reactor physics benchmark evaluation, sensitivity analysis has been performed. Individual nuclide cross-section sensitivities for heterogeneous and homogeneous fuel assembly models were calculated using the multi-group adjoint method, the iterated fission probability (IFP) method, and the Contribution-Linked eigenvalue sensitivity/Uncertainty estimation via Track length importance Characterization (CLUTCH) method. Good agreement between the three methods was observed except for the sodium total and elastic cross-section and the uranium-235 total cross-section sensitivity coefficients where the multi-group adjoint method produced results approximately 14% and 7% lower than the IFP and CLUTCH methods, respectively. [ABSTRACT FROM AUTHOR]

This paper describes a simple analytical technique for determining the specific activity of plutonium (Pu) present in the irradiated fuel solution without having the knowledge of isotopic composition. The Pu of unknown isotopic composition present in the irradiated fuel solution of spent fuel discharged from nuclear reactor was quantitatively oxidized to Pu (VI) with Ce (IV) in the nitric acid medium and the optical density was measured at a wavelength of 830 nm subsequently. The gross alpha activity of the Pu in the sample was determined after purification from fission products and minor actinides by solvent-extraction method. A spectrophotometric measurement of different samples with known concentration of Pu (VI) free from fission products and minor actinides was performed for obtaining the molar extinction coefficient. The experimental data such as optical density, molar extinction coefficient and gross alpha activity were employed to calculate the alpha specific activity of Pu of un-known isotopic composition. The alpha specific activity of Pu obtained by the method studied was in excellent agreement with the value calculated through the isotopic composition. The error involved in determination of specific activity of Pu by this method is within ±1.7%. The alpha specific activity determined by this procedure was successfully employed in determining the Pu concentration of plant samples by radiometric counting technique in spent nuclear fuel reprocessing plants. [ABSTRACT FROM AUTHOR]

Fast reactor concept like the 2400 MW GCFR is considered a promising candidate for future sustainable and economic nuclear power systems. It is a highly innovative system with advanced geometry and fuel materials (Ceramic fuel pellets of mixed uranium-plutonium carbide within fuel pins). The active fuel region of this reactor is divided radial to two core zones namely, inner core fuel assemblies (IC), and outer core fuel assemblies (OC). In the present study Monte Carlo transport code, MCNPX is used to design heterogeneous and homogeneous models for inner and outer core fuel assembly. The deigned models have been used to simulate the neutronic behavior and fuel transmutation during open cycle operation and to study the effect of Homogeneity inside the fuel assembly. The results of heterogeneous and homogeneous models were compared with each other and good agreement was found. [ABSTRACT FROM AUTHOR]

The present paper describes an evaluation of power reactivity loss data obtained from the Japanese prototype fast breeder reactor Monju. The most recent analysis on the power reactivity loss measurement (Takano et al, 2008) is updated in consideration of the following findings: (a) in-core temperature distribution effect, (b) crystalline binding effect, (c) logarithmic averaging of the fuel temperature, (d) localized fuel thermal elongation effect, (e) updated Japanese Evaluated Nuclear Data Library, JENDL-4.0, and (f) refined corrections on the measured values. The influences of these updates are quantitatively identified and the most precise and probable calculation-to-experiment (C/E) C/E value is derived and a thorough uncertainty evaluation is conducted. These reveal that analysis overestimates measurement by 4.6% with a measurement uncertainty of 2.0%. This discrepancy is reduced to as small as 1.2% when the core bowing effect is taken into account, which implies the importance of this effect in the calculation of power reactivity loss. [ABSTRACT FROM AUTHOR]

Breeder reactors are considered as a unique tool for fully exploiting natural resources. Fast breeder reactors based on thorium fuel can enhance inherent safety. Fluoride salt has good performance as a coolant in high-temperature nuclear systems. However, there is some doubt about the fuel breeding ability using fluoride salt coolant for fast spectrum due to its moderating ability. The aim of this study was to choose a proper fluoride salt mixture for Liquid-salt-cooled Solid-fuel Fast Reactor (LSFR) based on thorium-uranium fuel and give parametric studies to provide a design window for flexible self-sustaining core design. Infinite assembly model was used to analyze the salt selection from five candidate fluorides for fast spectrum as coolant. Combining neutron balance analysis with linear least squares fitting method based on 0D model, parametric studies at the neutron balance equation unique solution with burn-up for several parameters such as fuel volume fraction, removing fission gases process, total neutron losses and power density were presented in this paper. It was found that BeF 2 -NaF was a promising coolant in the five candidate fluoride salt mixture. This study proved that the design of a self-sustaining core for fluoride-salt-cooled fast breeder based on thorium fuel is achievable. A design window was found in the definition of a self-sustaining core for various fuel volume fractions and neutron loss fractions. Core design and fuel management strategy will be given in future. [ABSTRACT FROM AUTHOR]

A new cross section adjustment method has been derived in which systematic errors in measured data and calculated results of neutronics characteristics are estimated and removed in the adjustment. Bias factors which are the ratio between measured data and calculated results are used to estimate systematic errors. The difference of the bias factors from unity is caused generally by systematic errors and stochastic errors. Therefore by determining whether the difference is within the total stochastic errors of measurements and calculations, systematic errors are estimated. Since stochastic errors are determined for individual confidence levels, systematic errors are also dependent to the confidence levels. The method has been applied to cross section adjustments using 589 measured data obtained from fast critical assemblies and fast reactors. The adjustments results are compared with those of the conventional adjustment method. Also the effect of the confidence level to the adjusted cross sections is discussed. [ABSTRACT FROM AUTHOR]

The Heat eXchanger Tube Rupture (HXTR) accident, which is also called Steam Generator Tube Rupture (SGTR) accident, is important and should be considered as a safety issue in the design and safety assessment of Heavy Liquid Metal (HLM) cooled reactors. The HLM cooled reactor is a kind of advanced reactor which is meaningful for sustainable energy development. In this paper, the research progress on HXTR (or SGTR) accident of HLM cooled reactors from 1992 to 2016 is reviewed. Currently, because of the restrictions of experimental facilities and simulation tools for HXTR accident, the main phenomena, which should be concerned, were not all well studied. For experimental research, the most studied problems were the investigations of the pressure evolution, vapor transmission, fragmentation behavior, physics of the thermal and hydraulic interactions between HLM and water; And for simulation research, the pressure evolution and steam transmission were the mostly studied. For further research of HXTR accident of HLM cooled reactors, the propagation of the pressure waves and sloshing of the primary coolant pool with mechanical impact of the heavy liquid metal on structures should be more studied. And as the basis, the experimental facilities and numerical tools for HXTR accident should be improved firstly. [ABSTRACT FROM AUTHOR]

In this paper the dynamic thermal hydraulic fast reactor simulation code CHD is presented. The code is built around a scriptable object-oriented framework in the programming language Python to be able to flexibly describe different reactor geometries including thermal-hydraulics models of an arbitrary number of coolant channels as well as pumps, heat-exchangers and pools etc. In addition, custom objects such as the Autonomous Reactivity Control (ARC) system for enhanced passive safety are modeled in detail. In this paper we compare the performance of the CHD code with other similar fast reactor dynamics codes using a benchmark study of the European Sodium cooled Fast Reactor (ESFR). The results agree well, both qualitatively and quantitatively with the code benchmark. In addition, we demonstrate the code’s ability to simulate the long-term asymptotic behavior of a neutronically shut down reactor in an unprotected loss of flow scenario using a model of the Advanced Burner Reactor (ABR). [ABSTRACT FROM AUTHOR]

India is pursuing a three stage nuclear power program employing its natural uranium and Thorium reserves. FBRs form the second stage of the program linking the first phase with natural U and third phase with Th fuels. The saga of FBR technology development in India is presented in this paper. The valuable experience in design and operation of Fast Breeder Test Reactor (FBTR) is recounted briefly. The R&D for the techno-economic demonstration of Prototype Fast Breeder Reactor (PFBR) is explained including technology development for addressing manufacturing challenges and engineering full-scale qualification. Finally a brief projection is made on the developments for FBR600 with improved safety and economics as well as metal fuel technology for future experimental and power reactors to follow. [ABSTRACT FROM AUTHOR]

In the present study, the feasibility of employing the indigenously developed ferroboron alloy (Fe-15 wt.%B) as an alternate neutron shield material in combination with 9Cr-based ferritic steel (P91) clad in future Indian fast breeder reactors (FBR), has been investigated from a metallurgical perspective. Towards this goal, a series of diffusion couple experiments have been conducted at three different temperatures namely, 600, 700 and 800 °C for time durations up to 5000 h. The thickness of interaction layer has been monitored using standard metallographic procedures. The experiments revealed that ferroboron/P91 combination exhibited a tendency to form complex intermetallic borides at the interface. The structural and microstructural characterization of the interface confirmed that the reaction layer consists predominantly of borides of Fe and Cr of type FeB, Fe 2 B, (Fe,Cr) 2 B and (Fe,Cr)B. The measured variation of interaction layer thickness as a function of time and temperature have been modelled in terms of diffusion mediated interaction. The growth kinetics of borided layer has followed the parabolic law at each temperature, and the apparent activation energy for boride layer formation is found to be of the order of 115 kJ mol −1 . This indicates that the kinetics of boriding could be governed by diffusion of B into the P91 matrix. Based on the findings of present study, an extrapolative estimate of the clad attack thickness at 550 °C for 60 years of operating time has been made and it turns out to be 210 ± 15 μm, which is less than the clad thickness of FBR shielding subassembly (4 mm) [1]. Thus, this study confirms that at testing temperatures from 550 to 600 °C, the ferroboron/P91 steel combination can be safely employed for shielding subassembly applications in fast reactors. [ABSTRACT FROM AUTHOR]

Research with fast reactors has drawn interest in the investigation of their characteristics, their applications involving fast neutrons, and such. In this study, to examine the feasibility of the fast core design with the use of three low-enriched uranium (LEU) fuels (Al-alloy LEU, Metal-LEU and U10Mo) at the Kyoto University Critical Assembly (KUCA), numerical simulations with the MCNP6.1 code are conducted with 3-D assembly model of one fuel assembly in the infinite lattice of fuel assemblies to determine 1) performance of the candidate LEU fuels; 2) effect of reflectors and 3) combination of fuel and other plates in the fuel region. Based on the results, the reactor characteristics are analyzed through the full 3-D core model. Consequently, the fast reactor meeting the safety regulations at KUCA is found feasible with the use of the Metal-LEU fuel together with graphite and lead plates, while the other LEU fuels present difficulty in achieving criticality in the designed fast reactor. Finally, the design analyses could offer basic data for onward plans and operation of the KUCA core for fast reactor experiments. [ABSTRACT FROM AUTHOR]

The safe operation of nuclear reactors cooled by liquid metal is an important issue to be addressed for the development of nuclear technology and presents specific aspects related to the properties of the coolant. While the main parameters influencing safety for water-cooled reactors are reasonably well known, DHR systems for reactors cooled by liquid metal present additional challenges related to coolant freezing, since solidification temperature is higher than the temperature of the final heat sink. If the primary coolant solidifies obstructions of the primary flow path can occur, inhibiting natural circulation and core cooling. This paper presents an innovative passive safety system for decay heat removal which passively delays the coolant freezing. The system adopts noncondensable gases to passively control the power removed from the primary system and to delay freezing in the long term while keeping primary system temperatures below an acceptable limit. The system is simulated by means of the Relap5-3D computer code for a loss of offsite power of the innovative lead-cooled reactor ALFRED. A sensitivity analysis has been carried out in order to study the effect of noncondensable gas pressure on the performance of the passive decay heat removal system, and in particular on the primary coolant temperatures. [ABSTRACT FROM AUTHOR]

The paper presents results of material testing of the full-scale corium (melt of fuel and structural materials of the reactor core), obtained in the in-pile experimental device. Tests were carried out in the research reactor IGR (Impulse Graphite Reactor) at simulation of the final stage of severe accident of the fast power reactor with sodium coolant. The investigations include metallographic analysis of structure condition of corium ingot fragments. They are partly supplemented by the results of X-ray phase analysis of the powders made from those fragments. [ABSTRACT FROM AUTHOR]

The coupled numerical analysis of mechanical and thermal-hydraulic behaviors was performed for a wire-wrapped fuel pin bundle subassembly irradiated in a fast reactor. For the analysis, the fuel pin bundle deformation analysis code BAMBOO and the thermal-hydraulic analysis code ASFRE exchanged the deformation and temperature analysis results through the iterative calculations to attain convergence corresponding to the static balance between deformation and temperature. The analysis by the coupled code system showed that the radial distribution of coolant temperature in the subassembly tended to flatten as a result of the fuel pin bundle deformation governed by cladding void swelling and irradiation creep. Such flattening of temperature distribution was slightly observed as a result of fuel pin bowings due to the cladding-wire interaction even when no bundle-duct interaction occurred. The effect of the spacer wire-pitch on deformation and thermal-hydraulics was also investigated in this study. [ABSTRACT FROM AUTHOR]

To render future SFR more robust and safe, certain BDBE have been considered in the recent years. A Core Disruptive Accident leading to a whole core meltdown scenario has gained the interest of researchers. Various design concepts and safety measures have been suggested and incorporated in design to address such a low probability scenario. A core catcher concept, in particular, has proved to be inevitable as an in-vessel core retention device in SFR for safe retention of core debris arising out after the severe accident. This study aims to analyse the cooling capability of the innovative design concept of core catcher to remove decay heat of degraded core after the accident. First, the capability of single collection tray is established and then the study is extended to two and three collection trays with different design concepts. Transient forms of governing equations of mass, momentum and energy conservations along with k-ε turbulence model are solved by finite volume based CFD solver. Boussinesq approximation is invoked to model buoyancy in sodium. The study shows that a single collection tray is capable of removing up to 20 MW decay heat load in a typical 500 MWe pool type SFR. Further, studies are carried out to improve the natural circulation of sodium around the source, in the lower plenum and to distribute core debris of the whole core to multiple collection trays. It is found that the double and triple collection trays can accommodate decay loads up to 29 MW. Provision of openings in the collection trays has proved to be effective in improving the heat transfer and sodium flow as well as in distributing the core debris to the successive plates. Detailed analysis of isotherms and velocity field in the multiple collection trays with openings reveals improved natural circulation in and around the collection trays in the lower plenum. A three tray core catcher device with multiple openings can accommodate the decay heat of whole core meltdown, without exceeding the safe temperature limits of structural material. Special attention has been paid to the space constraint in the lower plenum while incorporating the three tray core catcher device. [ABSTRACT FROM AUTHOR]

The biggest challenge of implementing GEN IV technologies in the nuclear fuel cycle is their safety. The challenges are very broad, but one of the most significant ones is the effectiveness of the reactivity control system. The Gas-cooled Fast reactor requires a reactivity control system which provides sufficient reactivity worth to compensate the local neutron zones in the core. Due to the early development stage of the pin type GFR 2400 core, only the homogenous material composition of the system of control rods was published so far. In this paper the performance of the reactivity control system of the GEN IV Gas-cooled Fast Reactor (GFR), was investigated. The presented analyses were performed for the pin type GFR 2400 core by the MCNP5 and KENO6 stochastic calculation codes. To analyse the effects of geometry self-shielding, a simple heterogeneous control rod design was created and evaluated through a special procedure. Since the worth of the created control rod design is lower compared to the homogeneous configuration, moderator materials were accommodated in the control rod structure. The main goal of this study was to create a design which is, in terms of reactivity, comparable with the reference homogenous design. The presented paper deals with 6 proposed geometry configurations in combination with 8 moderator materials. The best properties were found for the hydride moderators, ZrH 2.0 , ZrH 1.5 and LiH, unfortunately these materials have undesirable properties at high temperatures. The BeO and Li 2 O moderators showed worse properties but their impact on the control rod could be enhanced by increasing their mass proportion in the assembly. [ABSTRACT FROM AUTHOR]

Ensuring post accident decay heat removal with high degree of reliability following a Core Disruptive Accident (CDA) is very important in the design of sodium cooled fast reactors (SFR). In the recent past, a lot of research has been done towards the design of an in-vessel core catcher below the grid plate to prevent the core debris reaching the main vessel in a pool type SFR. However, during an energetic CDA, the entire core debris is unlikely to reach the core catcher. A significant part of the debris is likely to settle in core periphery between radial shielding subassemblies and the inner vessel. Failure of inner vessel due to the decay heat can lead to core debris reaching the main vessel and threatening its integrity. On the other hand, retention of a part of debris in core periphery can reduce the load on main core catcher. Towards achieving an optimum design of SFR and safety evaluation, it is essential to quantify the amount of heat generating core debris that can be retained safely within the primary vessel. This has been performed by a mathematical simulation comprising solution of 2-D transient form of the governing equations of turbulent sodium flow and heat transfer with Boussinesq approximations. The conjugate conduction-convection model adopted for this purpose is validated against in-house experimental data. Transient evolutions of natural convection in the pools and structural temperatures in critical components have been predicted. It is found that 50% of the core debris can be safely accommodated in the gap between radial shielding subassemblies and inner vessel without exceeding structural temperature limit. It is also established that a single plate core catcher can safely accommodate decay heat arising due to ∼70% of the core debris by establishing natural circulation in the lower sodium pool. The influence of heat removal rate by natural circulation on availability of the number of decay heat exchangers (DHX) dipped in the upper pool is also analyzed. It is seen that the temperatures in the inner vessel, source plate and the maximum debris temperature do not increase significantly even when the DHXs are deployed 5 h after the accident, demonstrating the benefit of large thermal inertia of the pool. [ABSTRACT FROM AUTHOR]

The uncertainty of minor actinides transmutation in fast reactors has been estimated for individual nuclides of minor actinides (MA). The burn-up dependent sensitivities of MA number densities are calculated using the generalized perturbation theory, and a method is derived to transform the sensitivities to the sensitivities of minor actinides transmutation. The sensitivities are used to obtain the uncertainties of minor actinides transmutation together with the cross-section covariance data of JENDL-4. The method is applied to the two MA transmutation fast reactor cores with MA content of 6% and 11%. It is seen that the uncertainties of the total MA transmutation amounts are about 3% and 4% for the two cores because the uncertainties of Np-237 and Am-241 transmutations which mainly contribute the total transmutation are small (about 3%). The contribution to the uncertainties is mainly caused by the Np-237 and Am-241 capture cross-sections. The transmutation uncertainties for Np-239, Am-233, Cm-243, and Cm-245 are rather large. However the transmutation amounts of these nuclides are very small, and the large uncertainties are not the problem in practice. However, it is found that the transmutation amount of Cm-244 has large uncertainties of 10% and 13% for the cores with MA content of 6% and 11%, and this large uncertainties have to be considered when MA fuels are fabricated and reprocessed because Cm-244 produces large decay heat and neutron emission. [ABSTRACT FROM AUTHOR]

This paper reports development of a process named MicroWave Direct Oxidation (MWDO) for recycling the off-specification (U,Pu)O 2 mixed oxide (MOX) fuel pellets generated during fabrication of typical fast reactor fuels. MWDO is a two-stage, single-cycle process based on oxidative pulverisation of pellets using 2450 MHz microwave. The powder sinterability was evaluated by bulk density and BET specific surface area. The oxidised powders were analyzed for phases using XRD and stoichiometry by thermogravimetry. The sinterability was significantly enhanced by carrying out oxidation in higher oxygen partial pressure and by subjecting MOX to multiple micronisation-oxidation cycles. After three cycles, the recycled powder from (U,28%Pu)O 2 resulted surface area >3 m 2 /g and 100% re-used for MOX fabrication. The flow sheet was developed for maximum utilization of recycled powder describable by a parameter called Scrap Recycling Ratio (SRR). The process demonstrates smaller processing cycle, better powder properties and higher oxidative pulverisation over conventional method. [ABSTRACT FROM AUTHOR]

Metallic fuel has an important historical significance in the development of nuclear reactors and continues to be relevant to the progression of advanced test and power reactors. A number of models, ranging from empirical to mechanistic, have been developed and implemented in various fuel performance codes to describe U-Zr and U-Pu-Zr fuel and typical fast reactor cladding materials. One challenge of using these models to simulate fuel performance is the inevitable tangling of coupled phenomena that can cloud proper implementation, calibration, and eventual utilization of new models. In an effort to provide a baseline capability that will facilitate the use of advanced models, new capabilities have been implemented into the fuel performance code BISON specific to metallic fuel simulations, ranging from materials properties, fission gas release and swelling calculations, coolant channel models, and cladding correlations. These models have been applied to the X441/X441A EBR-II experimental assembly data, a set of irradiated metallic UPuZr fuel rods of varying pin designs. The models implemented in BISON are able to capture the general trend of the expected response of the fuel and cladding to irradiation in EBR-II, especially when considering the spread in experimental measurements and the uncertainties inherited from the historical material models. Ultimately, the models outlined here provide the baseline capabilities on which new models can build upon in order to improve the prediction of metallic fuel performance simulations in off-normal designs or operations. [ABSTRACT FROM AUTHOR]

For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named the stationary liquid fuel fast reactor (SLFFR) has been proposed based on a stationary molten metallic fuel. A compact core design of a 1000 MWt SLFFR has been developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches have been adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses have been performed to evaluate the steady-state performance characteristics. The analysis results indicate that the SLFFR of a zero TRU conversion ratio is feasible while satisfying the conservatively imposed thermal design constraints. A theoretical maximum TRU consumption rate of 1.01 kg/day is achieved with uranium-free fuel. Compared to the solid fuel reactors with the same TRU conversion ratio, the core size and the reactivity control requirement are reduced significantly. The primary and secondary control systems provide sufficient shutdown margins, and the calculated reactivity feedback coefficients show that the prompt fuel expansion coefficient is sufficiently negative. [ABSTRACT FROM AUTHOR]

Safety characteristics have been evaluated for the stationary liquid fuel fast reactor (SLFFR) proposed for effective burning of hazardous TRU elements of used nuclear fuel. In order to model the geometrical configuration and reactivity feedback mechanisms unique to SLFFR, a multi-channel safety analysis code named MUSA was developed. MUSA solves the time-dependent coupled neutronics and thermal-fluidic problems. The thermal-fluidic behavior of the core is described by representing the core with one-dimensional parallel channels. The primary heat transport system is modeled by connecting compressible volumes by liquid segments. A point kinetics model with six delayed neutron groups is used to represent the fission power transients. The reactivity feedback is estimated by combining the temperature and density variations of liquid fuel, structural material and sodium coolant with the corresponding axial distributions of reactivity worth in each individual thermal-fluidic channel. Preliminary verification tests with a conventional solid fuel reactor agreed well with the reference solutions obtained with the SAS4A/SASSYS-1 code. Transient analyses of SLFFR were performed for unprotected transient over-power (UTOP), unprotected loss of heat sink (ULOHS) and unprotected loss of flow (ULOF) accidents. The results showed that the thermal expansion of liquid fuel provides sufficiently large negative feedback reactivity for passive shutdown of UTOP and ULOHS. The ULOF transient is also terminated passively with the negative reactivity introduced by the gas expansion modules installed at the core periphery. The maximum coolant temperatures have sufficient margins to the sodium boiling point for all three unprotected transient scenarios. [ABSTRACT FROM AUTHOR]

A base-supported partially-filled fluid–shell system representing the pressure vessel of a fast reactor is studied. The shell is made of two cylindrical parts joined through a conical part. The shell is characterized using Donnell's theory, while the fluid is modeled using a velocity potential approach. Compatibility and interface conditions led to the energy functional, minimization of which yielded natural frequencies. The approach was validated against previously studied joined systems. Parametric studies revealed that vibration frequencies of the fluid-shell system (i.e., bulging frequencies) are comparable to shell-only vibration frequencies if only the lowermost part is filled. Vibration frequencies of the free fluid surface (i.e., sloshing frequencies) do not change materially with fluid height if most of the vessel is filled. Present study did not consider the coupling between bulging and sloshing modes, which may need to be considered for a completely-filled vessel with a large base radius (e.g., > 3 m) as the two sets of frequencies for such systems can be comparable. • Closed-form solution for a cylindrical–conical–cylindrical fluid–shell system. • Minor modification in solution parameters yield results for other joined systems. • Frequencies depend significantly on level of fluid and base radius. • Sloshing frequencies unaffected by fluid height for a completely filled system. [ABSTRACT FROM AUTHOR]

The study of the nuclear reaction rate distribution in the MOX core of China Experimental Fast Reactor is a necessary condition for obtaining the operating license. From the characteristics of core structure, both the principle of experimental and experimental system are studies, emphasized on the scheme design such as the size of foils, the irradiation power and counting rate. Firstly, based on the Monte Carlo Code（MCNP）and NJOY codes, the distribution of reaction rates is obtained. MCNP is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron, and so on. The NJOY nuclear data processing system is a comprehensive computer code package for producing to cross sections and related nuclear parameters from ENDF/B VIII.0 evaluated nuclear data. Secondly, the activity measurement of foils was obtained by HPGE, then the reaction rate is determined by data processing. It is worth mentioning that the experimental reactivity introduction has a negligible effect on the reactor core by MCNP. Finally, a set of activation experiment scheme that is suitable for CEFR MOX core is formed, which serves as a guide to carry out activation method experiments. • From the characteristics of core structure, both the principle of experimental and experimental system are studies, emphasized on the scheme design such as the size of foils, the irradiation power and counting rate. • The NJOY nuclear data processing system is a comprehensive computer code package for producing to cross sections and related nuclear parameters from ENDF/B VIII.0 evaluated nuclear data. • A set of activation experiment scheme that is suitable for CEFR MOX core is formed, which provides technical guidance to carry out activation method experiments. [ABSTRACT FROM AUTHOR]