An investigation of inlet blockage in the central subassembly (SA) of a medium-sized sodium-cooled fast reactor (SFR) is carried out. To analyze the accident scenario, a sodium boiling model is developed based on two-fluid approach. The model is added to the thermal model based on the single-pin assumption developed in the previous study. The developed model is validated with CABRI BI1 loss of flow experimental data. Having established the validity of the model, a detailed analysis of inlet blockage in the central SA of a medium-sized SFR is carried out. The model is able to predict the accident detection time by temperature monitoring thermocouple located at the SA outlet. Boiling front evolution, temperature evolution of the core components, dryout time, and clad melting time are predicted. A parametric study on the effect of growth of blockage at the SA inlet on time of detection and boiling is carried out.

The design of the next generation of fast breeder reactors has commenced, with the main targets being enhanced safety and improved economy. Nuclear heat generated in the fuel subassembly of fast reactors is removed by circulating sodium through the core using centrifugal pumps. The primary sodium pumps (PSPs) used are large-capacity pumps, and the design of these pumps is different from that of traditional pumps. Though many works have been reported for the performance prediction of centrifugal pumps, most of these works have been carried out in a decoupled way, and only a few works have been reported where the pump is modeled with all the associated geometric structures. Centrifugal pumps are prone to a phenomenon called suction recirculation, which occurs when pumps are operated significantly below the best efficiency point. This suction recirculation has a strong potential to damage the impeller. Correlations given in the literature for the prediction of the onset of recirculation cannot be used for complicated inlet geometries, and three-dimensional computation fluid dynamics (CFD) investigations are most suited for such applications. Many devices have been reported in the literature to reduce the intensity of (or to suppress) suction recirculation. Webs provided in the suction plenum will modify the velocity distribution at the impeller inlet and also can influence suction recirculation. In this work, the centrifugal pump used for primary sodium pumping for fast reactor applications is simulated using CFD techniques in an integrated way. The frozen rotor approach is used to simulate the impeller-diffuser hydraulics. The effect of flow hydraulics in the suction plenum, flow distribution in the standpipe-pump gap, and flow conditions in the pool on the performance characteristics of PSPs are simulated. The flow rate for the onset of suction recirculation is predicted and compared with correlations available in the literature. Simulations are carried out to study the effect of webs on suction recirculation. The effects of the number of webs and the web geometry are also studied.

Due to the presence of sodium, it is a challenging task to achieve the reliable and safe operation of steam generators in a sodium-cooled fast reactor (SFR). Water flow oscillations in a two-phase flow system worsen the tube integrity. An accurate prediction of two-phase pressure drop is essential in designing steam generators to operate in a stable regime. Toward this, experiments have been carried out on an industrial-size 19-tube model sodium-heated steam generator of 5.5-MW capacity to understand two-phase pressure drop characteristics at various operating conditions. The measured data are used to estimate the two-phase frictional pressure drop. The concept of a two-phase friction multiplier has been used in the present study. A significant variation in the two-phase frictional multiplier is seen with steam quality, whereas the variation of the two-phase friction multiplier is insignificant at saturated steam condition. Based on the experiments, complemented by computational model, a correlation has been developed for the two-phase frictional multiplier as a function of steam quality for sodium-heated once-through straight-tube steam generators.

An ingenious prototype Fuel Assembly Inner DUct Structure (FAIDUS) design is conceptualized under the Generation IV safety philosophy to prevent large-scale corium pool formation following total coolant flow blockage accident. The early failure of the duct wall provides a built-in guideway for molten material relocation out of the core region, ensuring inherent safety robustness. Therefore, understanding the duct wall failure mechanism is critical for evaluating accident mitigation potential of the FAIDUS design. In this regard, a transient enthalpy-based multiphase computational model is developed to investigate the heat transfer and fluid flow behavior involved during the duct wall failure in 3-D domain. The numerical model is employed initially to simulate wall failure in the EAGLE ID1 in-pile test. The model results are validated against the in-pile test data. Subsequently, FAIDUS duct wall failure in a prototype fast reactor fuel subassembly (SA) is investigated employing the present model. The study reveals that the large heat flux from the molten pool to the duct is the main reason for the duct wall failure. The duct failure starts with an initial rupture across the corners of the hexagonal-shaped duct and progresses to total failure of the duct wall by the expansion of the rupture over the whole duct wall area with time. The outer hexcan wall sustains only minor thermal damage during the duct wall failure process. The present study explores the associated thermal-hydraulics, molten pool dynamics, crust formation behavior, and event sequences involved in duct wall failure in great detail. The simulation findings suggest that the FAIDUS SA design facilitates premature failure of the slim duct wall prior to the hexcan wall failure. This could potentially enable early discharge of the molten corium away from the core region, limiting the accident progress.

The Indian 40 MWt experimental Fast Breeder Test Reactor (FBTR) operating at Kalpakkam has different systems for fuel pin failure detection. It uses a Gaseous Fission Product Detection (GFPD) system to detect the dry rupture phase of fuel pin clad failure, and it also has a delayed neutron detection (DND) system in each primary loop (east and west) for wet rupture phase detection. In 2011, a series of delayed neutron (DN) signal measurements were performed in FBTR to assess the sensitivity and localisation capabilities of the DND system. A special assembly with 19 perforated fuel pins of natural uranium-nickel metal alloy was used as a fission product source (FPS) for this test. In this paper, an integrated analysis has been carried out to simulate the experimental observations by using both neutronics and thermal hydraulics calculations. The Prompt Recoil Model (PRM) and modified Non-Recoil Model (NRM) with isotopic hold-up time are used to estimate the DN precursor release rate from the perforated fuel pin to the coolant sodium. The time-dependent activity is evaluated considering hydraulic dilution and decay of the DN precursors. To get the hydraulic dilution of DN precursors during their transport to the detector, a 3D CFD analysis of FBTR core with entire pool sodium has been performed using the commercial code ANSYS FLUENT. Monte Carlo modelling of the DND system is done for DN signal estimation by considering the spatial distribution of the DN source around the detectors. Results showed that a modified nonrecoil DN precursor release model coupled with the neutronics-hydraulics simulation gives better prediction of DN signal in FBTR, and hence, this methodology can be extended for generating the contrast ratio for core locations where measurements are not performed.

The early localization of a fuel subassembly with a failed (wet rupture) fuel pin is very important in reactors to limit the associated radiological and operational consequences. This requires a fast and reliable system for failure detection and their localization in the core. In the Prototype Fast Breeder Reactor, the system specially designed for this purpose is Failed Fuel Location Modules (FFLM) housed in the control plug region. It identifies a failed sub-assembly by detecting the presence of delayed neutrons in the sodium from a failed sub-assembly. During the commissioning phase of PFBR, it is mandatory to demonstrate the FFLM effectiveness. The paper highlights the engineering and physics design aspects of FFLM and the integrated simulation towards its function demonstration with a source assembly con-taining a perforated metallic fuel pin. This test pin mimics a MOX pin of 1 cm2 of geometrical defect area. At 10% power and 20% sodium flow rate, the counts rate in the BCCs of FFLM system range from 75 cps to 145 cps depending upon the position of DN source assembly. The model developed for the counts simulation is applicable to both metal and MOX pins with proper values of k-factor and escape coefficient.(c) 2023 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the CC BY license (http://creativecommons.org/licenses/by/4.0/).

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.

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

The Unprotected loss of flow accident (ULOFA) is a severe accident that may lead to core disruption in a sodium cooled fast reactor (SFR). It is investigated as a part of the defense in depth concept in nuclear safety. A coupled thermal hydraulics neutronics model is required to model the severe accident scenario where reactivity is affected by the changes in the temperature of the core components, coolant boiling, and material distribution during the transient. The coupled code previously developed for the analysis of total instantaneous blockage (ASTRA) is improved for the study of ULOFA by adding various models, viz., (i) a one-dimensional two-fluid sodium boiling model, (ii) a simple primary hydraulics model, (iii) a simple clad motion model, and (iv) an improved point kinetics solver. The boiling model is validated with different loss of flow experiments. The improved code is used to analyze ULOFA in a 500 MWe medium size sodium cooled fast reactor up to the onset of fuel melting. At 21.5s, coolant boiling is initiated in the central channel. The temperature evolution of the core components, reactivity feedbacks, power evolutions, and coolant voiding propagation are calculated. The code predicts early fuel melting at 24.9s due to the power excursion by the early voiding in the fuel channels compared to the study without including the boiling model. Boiling is started in five out of ten representative fuel subassembly (SA) channels at the time of fuel melting.

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.

The innovative Fuel Assembly Inner DUct Structure (FAIDUS) design inherently offers a quick relocation path for the molten core material to descend downwards via the inner duct and arrive at the foot of the Subassembly (SA) following the local core meltdown during Total coolant Flow Blockage (TFB) accident. With its high thermal energy and continual heat generation, the amassed molten mass targets the surrounding foot structure, causing the foot wall to fail. The thermal damage progression and discrete failure of the foot wall govern the later phase of the melt relocation that signals the accident state. To understand the thermal damage progression of SA foot, transient numerical simulations are performed using a commercial computational fluid dynamics solver, ANSYS Fluent. An enthalpy-porosity approach based solidification/melting model is employed to predict the wall failure mechanism and associated event sequences. The computational model is validated against the wall failure test results available in the literature. Subsequently, the foot wall failure analysis is undertaken using a 2-D axisymmetric physical model for various plausible blockage occasions that initiate the TFB accident near the SA foot. The numerical solution endeavors to explicate the transient behavior of molten pool, convection-diffusion controlled phase interface development, thermal transport mechanism, liquid front propagation across the foot cross-section, and associated thermal hydraulics. The present findings indicate that the standard SA foot design perpetuates the accident scenario by releasing the molten mass through the ruptured wall segment after failure. To address this, an improved slotted discriminator-based foot design is proposed in the present study, and the effectiveness of the modified foot design in realizing a positive melt relocation to the core catcher is examined. The simulation results suggest that the modified foot design permits the intended melt release to the core catcher, confirming the Controlled Material Relocation (CMR) strategy and restraining the accident propagation to neighboring SA.

Studies on local flow blockages are of great importance in Sodium cooled Fast Reactors (SFRs). Total instantaneous blockage (TIB) at the inlet of a subassembly (SA) of an SFR is a beyond design basis event that is considered as a theoretical envelope of all small blockages. The extent of core damage propagation is an essential input for the thermal design of the core catcher, which is provided for the collection and cooling of molten core. This study aims to conduct a coupled thermal hydraulics and neutronics analysis of TIB in a medium sized SFR up to the end of fuel melting in the blocked SA. Towards this, a coupled thermal hydraulics and neutronics model is developed by solving heat transfer equations and point kinetics equations numerically. A one-dimensional (1-D) molten clad film motion model is incorporated for simulating the clad relocation dynamics. Using the developed models, a coupled analysis of TIB in a medium sized SFR is carried out. The time of occurrence of key events such as sodium boiling, clad melting, and fuel melting are predicted. The times of detection of the accident by the core temperature monitoring system of neighbouring SA and by the overpower detection signals are also predicted. The time of the TIB detection is earlier than that predicted by a previous study which does not consider reactor power change during the transient. The rise in the sodium outlet temperature beyond the threshold level for the TIB detection is observed for all the fuel SAs before the end of fuel melting in the blocked SA. Thus, assurance towards early detection of TIB by core temperature monitoring system is strengthened.(c) 2022 Elsevier Ltd. All rights reserved.

This study reports the pressure effect on structural stability of neutron irradiated ferroboron systems. Ferroboron, a mixture of boron and iron, has been found to have three phases, i.e., FeB, Fe2B and Fe3B. Studies have been conducted on single-phase Fe2B and ferroboron. Fe2B adopts tetragonal structure at ambient and undergoes structural transition to orthorombhic phase at 6 GPa. Further, Fe2B is irradiated with neutrons with a fluence of ~1017 n cm–2 and yields bulk modulus of 254 GPa, which is 16% enhancement as compared to unirradiated sample. The defects are estimated by the use of SRIM code. Total displacement per atom (dpa) in Fe2B for the irradiation fluence is found to be 5.53×10−5. The study also shows that phase transition seen in pristine Fe2B is inhibited upon neutron irradiation under pressure up to 24 GPa. Similar result was obtained on ferroboron mixture, irradiated with a neutron fluence of 8.18×1021 n cm−2 with dpa of 2.8. The irradiated sample is found to be stable up to 16 GPa.

This chapter discusses the various methods and codes used in analyzing thermal (light-water reactors) and fast reactor (sodium-cooled fast reactors) cores. Multigroup nuclear data libraries and details of lattice physics calculations to account spatial and energy self-shielding effects in the resonance analysis are discussed. The whole-core calculations for estimating steady-state neutronics parameters for plant operation are also given. In addition, a brief outline of safety analysis and severe accident analysis in fast reactors is given at the end. The aim of reactor analysis is to obtain the neutron density distribution as a function of space, energy, scattering angle, and time. The neutron balance equation is formulated in the region of interest and solved for the fluxes. The problem is divided into two main sequences. First, the reactor core is considered to be an ensemble of small units. The neutron transport is treated in a hyperfine energy structure over this unit representative cell. The spatial and energy-dependent flux is then used to homogenize the unit cell properties to derive homogenized cell cross sections called lattice parameters. This representative cell is treated either in a one-dimensional (1-D) or two-dimensional (2-D) geometry. The inputs required for the lattice-level calculations are the energy-dependent cross-section set called as nuclear physics data and the geometry of the lattice. In the next step, the whole three-dimensional (3-D) reactor core is treated as a periodic arrangement of these unit representative cells with homogenized lattice properties. The global parameters are derived from the solution of flux over the 3-D core. The criticality problem is treated as a steady-state solution by formulating a time-independent neutron balance equation. The transient solutions are obtained by introducing a time-dependence in the balance equations.

Neutron diffusion equation is derived for a homogeneous medium by using Fick's law of neutron diffusion. The one-speed neutron diffusion equation is solved analytically for various homogeneous nonmultiplying and multiplying media in different geometries and the shape of neutron flux is discussed. The concept of criticality and its dependence on material and geometrical properties is briefly outlined. Neutron life cycle in a thermal reactor is given and explained qualitatively through the four-factor formula for neutron multiplication in an infinite medium. Thermalization effects of neutron energy spectrum in thermal reactors are given in brief. This chapter also discusses neutron multigroup diffusion equation and its solution methods. The weighting flux required for computing multigroup constants is obtained by solving neutron slowing down equation for different types of media containing heavy (fertile and fissile) and light (moderator) nuclides.

Detection of dry and wet fuel pin clad rupture is very important to ensure safe operation of fast reactors. An in vessel concept of delayed neutron detection system has been adopted for detecting wet rupture in a pool type MOX fueled 500 MWe Prototype Fast Breeder Reactor (PFBR) which is currently in the advanced stage of commissioning in India. It has 8 DND modules installed in the hot pool just above the intermediate heat exchangers inlet windows; each module has three 0.2 cps/nv sensitivity high temperature fission chambers which is surrounded by graphite moderator and B4C shield. In this paper, we propose a revised simulation methodology of DN signal estimation involving various steps of neutronics and thermal hydraulics calculations, which is considered to be a pre-requisite for the program of experimental demonstration of bulk DND system. A special DN source sub-assembly simulating a failed MOX pin of 1 cm2 defect area has been manufactured using natural uranium metal fuel slugs with a perforated clad. Prompt recoil model is employed to compute the release of fission products through the perforated clad. The DN source strength is estimated by solving 52 coupled radioactive decay equations for 52 parent and daughter (DN precursor) nuclides. The CFD analysis using PHOENICS code is performed to simulate mixing and decay of DN precursors in the sodium pool during their transport. Neutron transport calculation with distributed DN sources around the DND block is done to estimate the DN signal in the HTFC detectors using the fission equivalent neutron flux concept. The sensitivity of DN signal for different core locations is studied in detail at 20% thermal power level with 50% flow. The DN signal variation with power is also studied. This research work is performed to have a theoretical database prior to the actual calibration experiment of bulk DND system in PFBR after attaining first criticality.Y