Indian Prototype Fast Breeder Reactor Research Articles

Haar wavelet for solving the inverse point kinetics equations and estimation of feedback reactivity coefficient under background noise

A new formalism is presented in this paper for solving the inverse point kinetics equations with six groups of delayed neutron precursors using Haar wavelet and estimation of feedback reactivity coefficient from the observed power transient under background noise. The Haar wavelet transforms the inverse point kinetics equations into a set of linear equations and these equations can be solved easily. Using this method the reactivity required for a desired power transient is obtained and also the feedback reactivity and the temperature coefficient of reactivity involved in the observed power transient are estimated for various background noise levels. This method is tested in two ways, i.e. (i) to estimate the reactivity required for different types of power transients in thermal reactor as well as in Indian Prototype Fast Breeder Reactor (PFBR) and (ii) to estimate the feedback reactivity and the temperature coefficient of reactivity involved in the power transients of thermal as well as Indian PFBR. In the case of Indian PFBR, the temperature coefficient of reactivity is estimated for various background noise levels in power transients. It is observed that as the noise level is reduced, the accuracy in the estimation of temperature coefficient of reactivity is increased. It is also shown that using Haar wavelet with beamforming, the temperature coefficient of reactivity can be estimated to a good accuracy even under high background noise. In this method the estimated feedback reactivity and the temperature coefficient of reactivity are found to be in good agreement with reference values. From the comparison of results it is observed that this method is efficient in estimating the reactivity required for different types of desired power transients in thermal as well as in fast reactors and this method is also efficient in estimating the temperature coefficient of reactivity under high background noise. This method is effective and simple to use.

Neutron radiation damage studies in the structural materials of a 500 MWe fast breeder reactor using DPA cross-sections from ENDF / B-VII.1

The radiation damage in the structural materials of a 500 MWe Indian prototype fast breeder reactor (PFBR) is re-assessed by computing the neutron displacement per atom (dpa) cross-sections from the recent nuclear data library evaluated by the USA, ENDF / B-VII.1, wherein revisions were taken place in the new evaluations of basic nuclear data because of using the state-of-the-art neutron cross-section experiments, nuclear model-based predictions and modern data evaluation techniques. An indigenous computer code, computation of radiation damage (CRaD), is developed at our centre to compute primary-knock-on atom (PKA) spectra and displacement cross-sections of materials both in point-wise and any chosen group structure from the evaluated nuclear data libraries. The new radiation damage model, athermal recombination-corrected displacement per atom (arc-dpa), developed based on molecular dynamics simulations is also incorporated in our study. This work is the result of our earlier initiatives to overcome some of the limitations experienced while using codes like RECOIL, SPECTER and NJOY 2016, to estimate radiation damage. Agreement of CRaD results with other codes and ASTM standard for Fe dpa cross-section is found good. The present estimate of total dpa in D-9 steel of PFBR necessitates renormalisation of experimental correlations of dpa and radiation damage to ensure consistency of damage prediction with ENDF / B-VII.1 library.

Impurity analysis and dissolution behaviour of plutonium bearing impure MOX scrap: an assessment of recycling feasibility

A significant quantity of plutonium bearing scrap was getting generated during the fabrication of (U,Pu)O2 mixed oxide fuel for Indian Prototype Fast Breeder Reactor. This document assesses the feasibility of carrying out recycling of the impure scrap to permit the efficient plutonium recovery and re-use. The impure scrap which constitutes nearly 5 % of the total throughput was examined for impurities present and dissolution behaviour in HNO3. On the basis of experimental results obtained recycling methodologies are recommended. This study reveals that not all types of the impure scrap need aqueous recycling methods; some can be dry recycled.

Primary system thermal hydraulics of future Indian fast reactors

As a follow-up to PFBR (Indian prototype fast breeder reactor), many FBRs of 500MWe capacity are planned. The focus of these future FBRs is improved economy and enhanced safety. They are envisaged to have a twin-unit concept. Design and construction experiences gained from PFBR project have provided motivation to achieve an optimized design for future FBRs with significant design changes for many critical components. Some of the design changes include, (i) provision of four primary pipes per primary sodium pump, (ii) inner vessel with single torus lower part, (iii) dome shape roof slab supported on reactor vault, (iv) machined thick plate rotating plugs, (v) reduced main vessel diameter with narrow-gap cooling baffles and (vi) safety vessel integrated with reactor vault. This paper covers thermal hydraulic design validation of the chosen options with respect to hot and cold pool thermal hydraulics, flow requirement for main vessel cooling, inner vessel temperature distribution, safety analysis of primary pipe rupture event, adequacy of decay heat removal capacity by natural convection cooling, cold pool transient thermal loads and thermal management of top shield and reactor vault.

CFD investigation of effect of helical wire-wrap parameters on the thermal hydraulic performance of 217 fuel pin bundle

In fast reactors, the fuel pins are wound with helical wire-wrap spacer to provide support for the fuel pins and to provide space for sodium coolant to flow through the bundle. Due to the helical wire-wrap spacer, the coolant not only flows in axial direction but also in a transverse direction. This transverse flow provides better mixing of coolant among the sub channels and due to this, the heat transfer coefficient of the coolant increases. But, the frictional resistance to flow also increases. The effect of helical wire wrap parameters on the flow and temperature distributions of sodium in heat generating 217 pin fuel bundle and the variation of friction factor and Nusselt number as a function of helical wire parameters have been predicted. Toward this, the statistically averaged 3-dimensional conservation equations of mass, momentum and energy are solved along with k–ε turbulence model by varying the helical pitch and diameter of the spacer wire using customized Computational Fluid Dynamics (CFD) code CFDEXPERT. The geometric details of the fuel pin bundle and heat flux are similar to that of the Indian Prototype Fast Breeder Reactor (PFBR) that is currently in an advanced stage of construction.Based on detailed parametric study, it is found that the transverse flow induced by wire-wrap increases if the helical pitch is reduced. As a consequence of this, the friction factor and the Nusselt number increase with reduction in helical pitch. When the wire diameter is reduced, the pin bundle becomes tighter with high resistance for transverse flow. As a result of the reduced cross flow, the friction factor as well as Nusselt number decrease when wire diameter is reduced. Based on the computational data, suitable correlations have been derived for estimation of Nusselt number in 217 pin fuel bundle for a range of practical interest in fast reactor core design. It is also seen that the clad temperature decreases with decrease in helical pitch and increase in wire diameter due to increase in the transverse flow and associated enhancement in heat transfer coefficient. It is seen that the sodium temperature difference between the central sub-channels in the various rows of pin bundle and the peripheral sub-channels at the same hexagonal face is lower for shorter helical pitch and larger helical wire diameter which is attributed to the enhanced heat exchange due to higher cross flow in these cases.

Markov analysis for time dependent success criteria of passive decay heat removal system

Safety systems deployed in nuclear industry are generally required to operate for a particular mission time. Most of such systems employ redundancy to ensure their high availability over the stipulated mission time. However, availability of a system with redundant configuration depends upon success criteria and is application-specific. The Safety Grade Decay Heat Removal System of Indian Prototype Fast Breeder Reactor is required to operate with different success criteria during the specified mission time on account of steady decline in decay heat produced by the reactor core.In this paper, Markov analysis is carried out to evaluate the availability of the system under both continuous and periodic monitoring schemes. The study estimates the upper bound and lower bound for mean unavailability of SGDHR system for the specified mission time. Sensitivity analysis of the system attributable to important parameters is also carried out. The study provides a comprehensive approach to model scenarios with time dependent success criteria and provides an insight on the factors affecting availability of such systems.

Determination of total gas content and its composition in Indian PFBR blanket pellets

Total gas content and its composition are important specifications for sintered nuclear fuel pellets particularly in the case of fast breeder reactor fuels. Most commonly, total gas content and its composition is determined by hot vacuum extraction-quadrupole mass spectrometry (HVE-QMS). A number of parameters in this methodology such as temperature, duration of heating for quantitative extraction of evolved gases, total volume of the system, gas analysis conditions etc. need to be optimized for reliable measurements. In addition, sensitivity factors for various gases like H2, CH4, N2, CO, O2 and CO2 in quadrupole mass spectrometry required for quantification of results have been determined and validated employing reference gas mixtures of known composition. Employing these optimized conditions total gas content and its composition in blanket pellets (uranium oxide pellets) of Indian prototype fast breeder reactor was determined employing HVE-QMS. The relative expanded uncertainty (at a coverage factor k = 2) in the measurement of total gas content excluding hydrogen was estimated as per ISO guidelines and it was found to be 9.2 %.

A robust thermal model to investigate radial propagation of core damage due to total instantaneous blockage in SFR fuel subassembly

A robust enthalpy based thermal model has been developed to investigate the sequence of various phenomena that take place during total instantaneous blockage (TIB) in a fuel subassembly (SA) of a sodium cooled fast reactor. The (liquid–solid) interface movement in the multiphase and multi-material heat transfer is predicted by the Voller’s algorithm employing a dynamic node deletion procedure for realistic simulation of density difference between immiscible fuel and steel. Employing this model, the scenario of TIB in the fuel SA of Indian Prototype Fast Breeder Reactor (PFBR) has been analyzed. The focus of the investigation has been, (i) detection of the event by online monitoring of sodium outlet temperature from the neighboring subassembly and (ii) determination of number of subassemblies that are likely to get damaged severely before reactor shut down, which is an important parameter to define thermal load on core-catcher. Detailed parametric studies have been performed to bring out the effects of various parameters that influence inter-subassembly heat transfer and damage propagation. The parameters include, subassembly power (and hence reactor power), hexcan wall thickness, thermal conductivity of hexcan material and thermocouple time constant. It is found that during nominal power condition of the reactor, thermal damage propagates to only one row of SA suggesting that the thermal load on core-catcher during the event of TIB in one fuel SA can be that corresponding to decay power of seven SA. Damage of the blocked SA is completed within ∼20s. Reactor SCRAM from neighboring SA thermocouple takes place at 55s after the TIB and the residual thickness of hexcan wall at the time of reactor SCRAM is 55%. At low power conditions of the reactor (and hence at low SA power ratings) detection capability improves and the residual thickness of neighboring hexcan during reactor SCRAM increases. Thicker hexcan does not enhance detection. Increase in thermal conductivity of hexcan material enhances early detection of the event and improves the residual thickness. Reduction in time constant of thermocouple does not significantly help in early detection of TIB event. The results predicted for PFBR fuel SA are found to agree qualitatively with that reported for other reactors.

Uncertainty evaluation of reliability of safety grade decay heat removal system of Indian prototype fast breeder reactor

Deterministic and probabilistic safety assessment of nuclear power reactor technology is very important in assuring that the design is robust and safety systems perform as per requirement. The parameters required as input data for such analysis have uncertainties associated with them. Their impact is to be assessed on the results obtained for such analyses and it affects the overall decision making process.Safety Grade Decay Heat Removal System (SGDHRS) is one of the safety systems in fast breeder reactors and itremoves decay heat after reactor shutdown. It is a critical safety system; hence failure frequency for SGDHR is targeted to be less than 1.0×10−7 per reactor year. By bringing diversity in some of the components of SGDHRS, such as sodium-to-sodium decay heat exchanger (DHX), sodium to air heat exchanger (AHX) and valves, one can achieve the targeted low failure frequency of SGDHRS. We perform uncertainty analysis of the reliability of such SGDHRS here. Uncertainty in failure rate (of components of SGDHRS) is assumed to follow the log-normal distribution with error factor of three. Monte Carlo method of sampling is used in MATLAB environment. Results are obtained in terms of mean, median and standard deviation values of failure frequency. Percentile and confidence interval analysis of mean values are also obtained. These provide 95 and 98 percentile and confidence interval values of 98%, 99% and 99.8%. It is found that error factor of failure frequency of SGDHRS is found to be less than 3 in all the cases except the one in which DHX, AHX and Valves are designed with diversity in design. It is to be noted here that error factor of all input parameters distribution is taken as 3.

Development of gas entrainment mitigation devices for PFBR hot pool

The phenomenon of argon gas entrainment into primary sodium in the hot pool of a liquid Sodium cooled Fast Reactor (SFR) is investigated. A combined experimental and computational approach is proposed. Physical entrainment of argon into sodium is highly geometry dependent and the present study is focused on the Indian Prototype Fast Breeder Reactor (PFBR). The computational model which is used for parametric studies is validated against a 1/4 scale water model tests. Based on systematic experiments, the free surface velocity for onset of gas entrainment is determined to be 0.32m/s, and the corresponding threshold value for the surface vorticity has been estimated as 8.3s−1.A horizontal baffle plate attached to inner vessel is found to reduce the free surface velocity/vorticity effectively. The radial width of the baffle plate and its elevation has been optimized to be 0.125m and 0.315m from free surface by computational studies. The optimized device has been tested in the water model to confirm its satisfactory performance before its recommendation for PFBR.

3D core burnup studies in 500 MWe Indian prototype fast breeder reactor to attain enhanced core burnup

Fast breeder reactors are capable of producing high fuel burnup because of higher internal breeding of fissile material and lesser parasitic capture of neutrons in the core. As these reactors need high fissile enrichment, high fuel burnup is desirable to be cost effective and to reduce the load on fuel reprocessing and fabrication plants. A pool type, liquid sodium cooled, mixed (Pu–U) oxide fueled 500MWe prototype fast breeder reactor (PFBR), under construction at Kalpakkam is designed for a peak burnup of 100GWd/t. This limitation on burnup is purely due to metallurgical properties of structural materials like clad and hexcan to withstand high neutron fluence, and not by the limitation on the excess reactivity available in the core. The 3D core burnup studies performed earlier for approach to equilibrium core of PFBR is continued to demonstrate the burnup potential of existing PFBR core. To increase the fuel burnup of PFBR, plutonium oxide enrichment is increased from 20.7%/27.7% to 22.1%/29.4% of core-1/core-2 which resulted in cycle length increase from 180 to 250 effective full power days (efpd), so that the peak fuel burnup increases from 100 to 134GWd/t, keeping all the core parameters under allowed safety limits. Number of diluents subassemblies is increased from eight to twelve at beginning of life core to bring down the initial core excess reactivity. PFBR refueling is revised to accommodate twelve diluents. Increase of 10B enrichment in control safety rods (CSRs) and diverse safety rods (DSRs) is made along with a design change in CSRs to increase the reactivity worth of absorber rods for ensuring safe reactor operation. This new enhanced burnup core can be used in the future with advanced structural materials for clad and hexcan, which can withstand higher neutron fluence and gamma radiation.

Perturbation analysis of prototype fast breeder reactor equilibrium core using IGCAR and ERANOS code systems

First order approximation in standard perturbation theory has been used for the safety analysis of Indian prototype fast breeder reactor (PFBR). Material void and Doppler worth distribution obtained with first order approximation has been used to find the response of PFBR core during reactivity transients like unprotected loss of flow accident (ULOFA) and unprotected transient over power accident (UTOPA) up to sodium voiding. The validity of first order approximation applied to larger perturbations during such accidents has been tested by developing an exact perturbation code PERTX. 2D diffusion code ALCIALMI and ABBN-93 evaluated nuclear data has been used for the analysis.European Reactor Analysis Optimized Calculation System (ERANOS) is a code system used for reliable neutronic calculations of current and advanced fast reactor cores. The results of perturbation analysis of PFBR done with IGCAR code systems have been compared with those obtained with ERANOS 2.1 and JEFF 3.1 nuclear data. The present study reveals the adequacy of the IGCAR perturbation methods in 2D for the safety analysis of PFBR core involving transients with fuel, structure and coolant temperatures change.

A fundamental approach to specify thermal and pressure loadings on containment buildings of sodium cooled fast reactors during a core disruptive accident

Reactor Containment Building (RCB) is the ultimate barrier to the environment against activity release in any nuclear power plant. It has to be designed to withstand both positive and negative pressures that are credible. Core Disruptive Accident (CDA) is an important event that specifies the design basis for RCB in sodium cooled fast reactors. In this paper, a fundamental approach towards quantification of thermal and pressure loadings on RCB during a CDA, has been described. Mathematical models have been derived from fundamental conservation principles towards determination of sodium release during a CDA, subsequent sodium fire inside RCB, building up of positive and negative pressures inside RCB, potential of in-vessel sodium fire due to failed seals and temperature evolution in RCB walls during extended period of containment isolation. Various heating sources for RCB air and RCB wall and their potential have been identified. Scaling laws for conducting CDA experiments in small-scale water models by chemical explosives and the rule for extrapolation of water leak to quantify sodium leak in reactor are proposed. Validation of the proposed models and experimental simulation rules has been demonstrated by applying them to Indian prototype fast breeder reactor. Finally, it is demonstrated that in-vessel sodium fire potential is very weak and no special containment cooling system is essential.

Passive system reliability analysis using Response Conditioning Method with an application to failure frequency estimation of Decay Heat Removal of PFBR

Innovative nuclear reactor designs include passive means to achieve high reliability in accomplishing safety functions. Functional reliability analyses of passive systems include Monte Carlo sampling of system uncertainties, followed by propagation through mechanistic system models. For complex passive safety systems of high reliability, Monte Carlo simulations using mechanistic codes are computationally expensive and often become prohibitive. Passive system reliability analysis using recently proposed Response Conditioning Method, which incorporates the insights obtained from approximate solutions like response surfaces in simulations to obtain computationally efficient and consistent probability estimates, is presented in this paper. The method is applied to evaluate the reliability of passive Decay Heat Removal (DHR) system of Indian Prototype Fast Breeder Reactor (PFBR). The accuracy as well as efficiency of the method is compared with direct Monte Carlo simulation. The variability of the reliability values is estimated using bootstrap technique. The system abilities, to prevent critical structural damage as well as to ensure operational safety, are quantitatively ascertained. The system functional failure probabilities are integrated with hardware failure probabilities and the inclusion of passive system unreliability in Probabilistic Safety Assessment is demonstrated.

Development of process for aluminising nickel alloy 718 strips for tube bundle support structures in liquid sodium–water steam generators

The tubes in liquid sodium–water steam generators of the Indian prototype fast breeder reactor (PFBR) will be supported by corrugated nickel alloy 718 strips. Aluminisation of nickel alloy 718 strips has been chosen for this application because of the excellent performance of aluminide coatings in reducing impact fretting wear of the tubes due to flow induced vibrations and compatibility of the coating with liquid sodium at the operating temperature of the steam generators. Aluminisation of nickel alloy 718 strips for steam generator tube bundle support structures has been developed using a procedure involving thermal spraying of aluminium followed by diffusion heat treatment in vacuum atmosphere. One of the advantages of the technique is that it will coat only the desired surfaces of the strips, whereas in conventional pack cementation process, significant precautions have to be ensured. Furthermore, this process has enabled aluminisation to be carried out at a much lower cost than the conventional process of pack aluminising. The problems encountered during the initial trials and technology development, such as coating thickness and distortion, are discussed. A process flow chart for this procedure to take the job on an industrial scale is also reported. This process (under patenting) has been adopted for the aluminisation of corrugated strips for the support structures of one steam generator module and the steam generator for a test facility during the technology development phase, as also of all the steam generators being fabricated for the PFBR.

Hardfacing of austenitic stainless steel with nickel-base NiCr alloy

Fast breeder reactors components encounter adhesive/abrasive wear nature and erosion. Hardfacing by weld deposition improves resistance to high temperature wear, especially galling, of mating surfaces in sodium. Based on induced radioactivity rate and shielding considerations, nickel-base E NiCr-B hardfacing alloy was chosen to replace cobalt-base Stellite alloys. Studies on this hardface deposit on austenitic stainless steel substrate demonstrated that deposits after exposure at service temperatures up to 823 K would retain adequate hardness (> RC 40) at end design service-life of 40 years. Also, the plasma transferred arc welding process was chosen for hardfacing to ensure that width of dilution zone could be controlled by optimising deposition parameters. The selection of nickel-base hardfacing alloy and deposition process used for development of hardfacing technology for various components of the Indian prototype fast breeder reactor is discussed.

Overview of pool hydraulic design of Indian prototype fast breeder reactor

Thermal hydraulics plays an important role in the design of liquid metal cooled fast breeder reactor components, where thermal loads are dominant. Detailed thermal hydraulic investigations of reactor components considering multi-physics heat transfer are essential for choosing optimum designs among the various possibilities. Pool hydraulics is multi-dimensional in nature and simple one-dimensional treatment for the same is often inadequate. Computational Fluid Dynamics (CFD) plays a critical role in the design of pool type reactors and becomes an increasingly popular tool, thanks to the advancements in computing technology. In this paper, thermal hydraulic characteristics of a fast breeder reactor, design limits and challenging thermal hydraulic investigations carried out towards successful design of Indian Prototype Fast Breeder Reactor (PFBR) that is under construction, are highlighted. Special attention is paid to phenomena like thermal stratification, thermal stripping, gas entrainment, inter-wrapper flow in decay heat removal and multi-physics cellular convection. The issues in these phenomena and the design solutions to address them satisfactorily are elaborated. Experiments performed for special phenomena, which are not amenable for CFD treatment and experiments carried out for validation of the computer codes have also been described.

A comparative CFD investigation of helical wire-wrapped 7, 19 and 37 fuel pin bundles and its extendibility to 217 pin bundle

Preliminary investigations of sodium flow and temperature distributions in heat generating fuel pin bundles with helical spacer wires have been carried out. Towards this, the 3D conservation equations of mass, momentum and energy have been solved using a commercial computational fluid dynamics (CFD) code. Turbulence has been accounted through the use of high Reynolds number version of standard k– ɛ model, with uniform mesh density respecting wall function requirements. The geometric details of the bundle and the heat flux in are similar to that of the Indian Prototype Fast Breeder Reactor (PFBR) that is currently under construction. The mixing characteristics of the flow among the peripheral and central zones are compared for 7, 19 and 37 fuel pin bundles and the characteristics are extended to a 217 pin bundle. The friction factors of the pin bundles obtained from the present study is seen to agree well with the values derived from experimental correlations. It is found that the normalized outlet velocities in the peripheral and central zones are nearly equal to 1.1–0.9, respectively which is in good agreement with the published hydraulic experimental measurements of 1.1–0.85 for a 91 pin bundle. The axial velocity is the maximum in the peripheral zone where spacer wires are located and minimum in the zones which are diametrically opposite to the respective zone of maximum velocity. The sodium temperature is higher in the zones where the flow area and mass flow rates are less due to the presence of the spacer wires though the axial velocity is higher there. It is the minimum in the peripheral zones where the circumferential flow is larger. Based on the flow and temperature distributions obtained for 19 and 37 pin bundles, a preliminary extrapolation procedure has been established for estimating the temperatures of peripheral and central zones of 217 pin bundle.

Functional reliability analysis of Safety Grade Decay Heat Removal System of Indian 500 MWe PFBR

Passive systems are increasingly deployed in nuclear industry with an objective of increasing reliability and safety of operations with reduced cost. Methods for assessing the reliability of thermal–hydraulic passive systems, that is systems with moving working fluid, address the issues in natural buoyancy-driven flow that could result in a failure to meet the design safety limits under accident scenarios. This is referred as design functional reliability. This paper presents the results of functional reliability analysis carried out for the passive Safety Grade Decay Heat Removal System (SGDHRS) of Indian Prototype Fast Breeder Reactor (PFBR). The analysis is carried out based on the overall approach reported in the Reliability Methods for Passive System (RMPS, European Commission) project. Functional failure probability is calculated using Monte-Carlo method and also with method of moments.

Impact toughness of 316 stainless steel weld metal on elevated temperatures aging

Shielded metal arc welding electrodes of a modified E316-15 austenitic stainless steel, for service at 673–823 K with delta ferrite in the range of 3–7 ferrite number, have been developed indigenously for welding of 316L(N) stainless steel structural materials for the Indian Prototype Fast Breeder Reactor. Delta ferrite content in weld metals for high temperature service is restricted for limiting the formation of embrittling secondary phases during service. To study the effect of high temperature exposure on microstructure and mechanical properties, the 316 weld metal was aged at three different temperatures of 923, 973 and 1023 K, for various durations up to 500 h. The activation energy for the transformation of delta ferrite has been estimated to analyse the mechanism associated with the micro structural changes that led to the deterioration in toughness on elevated temperature aging of this weld metal.