Prototype Fast Breeder Reactor Research Articles

Ultrasonic imaging of projected components of PFBR

The 500MWe, sodium cooled, Prototype Fast Breeder Reactor (PFBR) is under advanced stage of construction at Kalpakkam in India. Opacity of sodium restricts visual inspection of components immersed in sodium by optical means. Ultrasonic wave passes through sodium hence ultrasonic techniques using under sodium ultrasonic scanners are developed to obtain under sodium images. The main objective of such an Under Sodium Ultrasonic Scanner (USUSS) for Prototype Fast Breeder Reactor (PFBR) is to detect and ensure that no core Sub Assembly (SA) or Absorber Rod or its Drive Mechanism is protruded in the above core plenum before starting the fuel handling operation. Hence, it is necessary to detect and locate the object, if it is protruding the above core plenum. To study the feasibility of detecting the absorber rods and their drive mechanisms using direct ultrasonic imaging technique, experiments were carried out for different orientations and profiles of the projected components in a 5m diameter water tank. The in-house developed PC based ultrasonic scanning system is used for acquisition and analysis of data. The pseudo three dimensional color images obtained are discussed and the results are applicable for PFBR. This paper gives the details of the features of the absorber rods and their drive mechanisms, their orientation in the reactor core, experimental setup, PC based ultrasonic scanning system, ultrasonic images and the discussion on the results.

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.

Deformation parameters of bedrock at a nuclear reactor site

Many engineering structures have to be founded deep into rock or on rocks, especially in a nuclear reactor site. Numerical finite-element and boundary-element analyses for studies of the stress and displacement distribution of any rock engineering projects are based on deformation parameters. Empirical relations are available to relate the rock mass deformation modulus to index properties, but the equations include some uncertainties relating to the variability of rock type and the heterogeneous nature of the rock masses. For important structures, site-specific parameters have to be evaluated to assess strains and stresses in rock due to various load combinations. Pressuremeter tests, cyclic plate load tests, uniaxial jack tests and seismic crosshole tests were conducted to evaluate in situ design parameters. The objective of this paper is to evaluate deformation parameters of a prototype fast breeder reactor from these in situ tests. For high strain deformation modulus, the pressuremeter modulus is comparable with the rock mass modulus obtained from laboratory results in association with rock quality designation. Crosshole tests results are used for evaluation of low strain modulus.

Primary containment capacity of Prototype Fast Breeder Reactor against core disruptive accident loadings

Severe accident analysis methodologies of sodium cooled fast reactors are critically re-looked in recent years and several scenarios are postulated. Development of advanced numerical and experimental techniques is carried out with the objective of determining realistic energy release, which is still evolving internationally. In view of this, a parametric study on mechanical energy release values in the range 100–1000MJ under a core disruptive accident condition, has been carried out for a 500MWe pool type Prototype Fast Breeder Reactor (PFBR) and associated mechanical consequences are assessed. The investigation focuses on issues related to structural integrity of primary containment, reactor containment building and post accident cooling capability. The assumptions made on estimation of mechanical energy release have been brought out with appropriate justifications. Fluid and structural dynamics analysis results are presented and discussed in detail depicting the loading sequences and various physical phenomena involved.The analysis carried out for PFBR indicates that primary containment has high potential to withstand the transient forces generated by energy release, corresponding to the work potential even more than 1000MJ. The sodium release to the reactor containment building through top shield penetrations under sodium slug impact phenomenon is limited with higher energy releases, mainly due to large dimensional changes of vessel associated with shorter transient duration and fall of sodium free level in the main vessel. The sodium level fall, a maximum value of 0.88m for 1000MJ is found to be acceptable for the PFBR. Hence, high uncertainties in the energy release estimation are not of great concern in PFBR safety studies. However, the deformations of decay heat exchangers immersed in the sodium pool could limit the acceptable work potential to 500MJ derived based on a separate experimental study.

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.

Estimation of hydrogen mass flux from water side to sodium side in a sodium heated Mod 9Cr1Mo steam generator

Heat, produced in a SFR (Sodium cooled Fast Reactor) is exchanged to a sodium circuit, where it is further transferred from sodium to water in a steam generator (SG). SG, typically made of ferritic steel, is a shell and tube heat exchanger with sodium on shell side and water/steam on tube side. Water/steam, flowing through the SG tube, reacts with the tube material, producing magnetite and hydrogen. Most of the hydrogen, formed in the water side, permeates into sodium through the SG tube wall and becomes the major impurity in sodium. The mass flux of hydrogen, permeating into sodium, is particularly important in the design of cold trap (CT) which is used for purifying sodium. Experiments were performed with a model SG made of Mod 9Cr1Mo having 19 tubes, in Steam Generator Test Facility (SGTF) at Indira Gandhi Centre for Atomic Research (IGCAR), to quantify hydrogen flux. Material, structural layout and operating conditions of SG of Prototype Fast Breeder Reactor (PFBR) were simulated. This paper brings out the data and experiences gained through the experiments.

Design of instrumentation for control and safety rod drive mechanisms of prototype fast breeder reactor

Prototype Fast Breeder Reactor (PFBR) is a 500 MWe, sodium cooled, plutonium oxide-uranium oxide fuelled, pool type fast breeder nuclear reactor under construction in India. Control and Safety Rod Drive Mechanism (CSRDM) is part of one of the shutdown systems and is used to drive neutron absorber rods up and down. Instrumentation and control for CSRDM realises significant functions like power control, power set back and controlled shutdown. A study was carried out into incidents during the operation of 40 MWt experimental reactor called Fast Breeder Test Reactor and testing of a prototype CSRDM. Significant incidents in FBTR like uncontrolled withdrawal of absorber rod, movement in unintended direction, partial absorber rod drops are found to have occurred. Corrective measures are taken in the I&C of PFBR. PFBR I&C for CSRDM is presented. I&C designers for nuclear systems may gainfully make use of the experiences described in the paper.

Evaluation of Speed Lowering of Primary Sodium Pump into Sodium Pool for 500 MWe PFBR Based on Creep Damage

Abstract In 500 MWe Prototype Fast Breeder Reactor (PFBR), the Primary Sodium Pump (PSP) contained in the primary circuit circulates sodium through the core. The PSP is supported on the roof slab. The space between the roof slab and sodium free surface is filled with argon. PSP will be lowered slowly, using a flask, into the main vessel (MV) after the completion of preheating and sodium filling of the main vessel. Pump inside the flask will be at ambient condition before lower ng into the sodium pool. While lowering PSP into MV sodium, one of the critical welds near the shell-flange junction at the discharge nozzle will experience thermal stress due to the hot shock offered by the sodium. The governing failure mode to ensure the structural integrity of the PSP is the creep damage developed in the weld near the junction of flange due to the discontinuity stresses that would be locked up the weld while lowering. This paper addresses the thermo-mechanical analysis carried out towards establishing the preheating temperature and speed of lowering the pump in the sodium pool. The PSP may be take in out for inspection/maintenance and lowered it back into the sodium pool around 5-10 times during the plant life. Based on this investigation, the PSP preheat temperature and lowering rate has been recommended.

Creep-fatigue Damage Analysis of Lower Stay Plate in Control Plug

Abstract In Prototype Fast Breeder Reactor, the lower stay plate made of SS 316 LN in control plug has cut opening in three locations for the purpose of easiness of assembling the sampling tubes through the lower stay plate. These openings are then covered with appropriate cover plates bolted with the stay plate. The stay plate along with cover plate will be seeing thermal transients during reactor scrams. The thermal transient will cause the membrane stress due to global temperature increase in cover plate and bending stress due to temperature gradient across the thickness. This paper aims the creep and fatigue damages estimation of the stay plate and covers plate. The material of stay plate and cover plate is SS 316 LN. The stay plate and cover plate is analysed using CAST3 M FE software developed by CAE France. A small square segment including a bolt and portion of junction between stay plate and cover plate is modeled with 8 noded, CUB8 elements. The fatigue damage for 851 number of thermal transient has been calculated. The creep damage due to the thermal stress has been evaluated. The interaction of creep-fatigue damage has also been evaluated as per RCC-MR code rule and is found within the limit.

Effect of Prior Cold Work on Creep Rupture and Tensile Properties of 14Cr-15Ni-Ti Stainless Steel

Abstract Titanium bearing 14Cr-15Ni stainless steel is the material chosen for the fabrication of the fuel clad tubes used in the Prototype Fast Breeder Reactor which is under construction at Kalpakkam. The superior swelling resistance and creep strength of this alloy is due to the intragranular precipitation of fine titanium carbides especially on dislocations. Cold work on these austenitic steels can improve both the creep strength and swelling resistance. A prior cold work (PCW) level of 20±4% is hence prescribed for materials used in the core of Fast breeder reactors (FBRs). Variation of tensile and creep properties within this cold work range is required to be evaluated, to check if the range of cold work can be further narrowed for better performance of the material. In this work, the cold work levels were varied in the range of 16% to 24%. Creep tests were carried out at 923 K and 973 K. There was no significant change in rupture lives with increase in PCW at 973 K. Increasing the level of prior cold work increased the rupture life significantly at 923 K. TEM studies revealed extensive recovery of the samples tested at 973 K. Samples tested at 923 K showed that extensive precipitation retained dislocation substructure.

Analysis of Interfacial Microsstructure of Post Weld Heat Treated Dissimilar Metal Weld of Type 316LN/C-Steel Joint

In Prototype Fast Breeder Reactor (PFBR), the main vessel which contains the primary sodium and supports the core is suspended from the roof slab. The materials for construction for main vessel and roof slab are type 316LN austenitic stainless steel and Carbon steel of grade A48P2, respectively. As the materials of construction are different, a transition joint between austenitic stainless steel and C-steel is necessary. In this investigation the effect of post-weld heat treatment (PWHT) on the interfacial microstructure of as-welded and PWHTed type 316LN/C-steel joint welded with Inconel 182 was investigated. These joints were PWHTed to various temperatures between 898 to 973K for 1h and results were evaluated. From the above results, different methods to temper the martensitic structure or to change to an equilibrium structure without PWHT are also presented.

High Temperature Materials for Nuclear Fast Fission and Fusion Reactors and Advanced Fossil Power Plants

Development of materials plays a crucial role in the economic feasibility of fast nuclear fission and fusion power plant. In order to meet this objective, one of the methods is to extend the fuel burnup and decreasing doubling time. The burnup is largely limited by the void swelling and creep resistances of the fuel cladding and wrapping materials. India's 500 MWe Prototype Fast Breeder Reactor (PFBR) is in advanced stage of construction. The major structural materials chosen for PFBR with MOX fuel are alloy D9 as fuel clad and wrapper material, 316LN austenitic stainless steel for reactor components and piping and modified 9Cr-1Mo steel for steam generator. In order to improve the burnup further, titanium, phosphorous and silicon contents in alloy D9 have been optimized for better swelling and creep resistances to develop modified version of alloy D9 as IFAC-1. Creep resistance of inherently void swelling resistance 9Cr-ferritic steel has been improved with the dispersion of nano-size yttria to develop oxide dispersion strengthened (ODS) steel clad tube with long- term creep strength, similar to D9, for increasing the fuel burnup. Development of modified 9Cr-1Mo steel clad tube and 9Cr-1Mo steel wrapper for future metallic fuel reactors being developed for reducing the doubling time are in progress. Extensive studies on resistance of this new generation core materials to void swelling are also under progress along with material development. Improved versions of 316LN stainless steel with nitrogen content of about 0.14 wt.% having higher creep strength to increase the life of fast reactor and modified 9Cr-1Mo steel with reduced nitrogen content and controlled addition of boron to improve type IV cracking resistance for steam generator are other developments. India's participation in ITER programme necessitates the development of India-specific RAFM steel for Test Blanket Module (TBM). A comprehensive research programme is being carried out to develop India-specific 9Cr-W-Ta RAFM steel with the optimization of tungsten and tantalum contents for better combination of strength and toughness. Based of the extensive mechanical tests including impact, tensile, creep and fatigue on four heats of RAFM steels having tungsten in the range 1 – 2 wt. % and tantalum in the range 0.06 -.014 wt., the RAFM steel having 1.4 wt. % tungsten with 0.06 wt. % tantalum is found to possess better combination of strength and toughness. This steel is considered as India-specific RAFM steel and TBM is being manufactured by this RAFM steel. To limit the emission of green house gases, a research and development programme has been initiated to develop advanced ultra super critical fossil fuel fired thermal power plants working at temperature of around 973 K and pressure of 300 bar. High temperature creep strength super 304H austenitic steel and Inconel 617 superalloy tubes are indigenously developed for this purpose.

A Comparison of Creep Deformation and Rupture Behaviour of 316L(N) Austenitic Stainless Steel in Flowing Sodium and in Air

Abstract Type 316L(N) austenitic stainless steel is used for the fabrication of Proto-type Fast Breeder Reactor (PFBR) under advanced stage of construction at Kalpakkam, India. The influence of flowing sodium on creep rupture behaviour of the steel has been investigated and results are compared with those obtained on carrying out creep test in air environment. Creep test on the material both in flowing sodium and in air were carried out at 873 K over a stress range of 225- 305 MPa. Sodium velocity across the creep specimen was maintained around 2.5 m/s. The testing environment was found not to change the rate of steady state creep deformation significantly. The tertiary stage of creep deformation of the steel was found to change significantly by the testing environment. The tertiary stage of creep deformation in the steel started much later for testing in sodium environment than that in air environment. The steel possessed higher creep rupture life for testing in liquid sodium environment that in air environment. Higher creep rupture strength of the material in flowing sodium was companied with higher creep rupture elongation. Optical micrographic investigation of the creep ruptured specimen in air showed extensive intergranular creep cavitation both in interior as well as on the specimen surface, whereas specimen tested in sodium showed relatively less creep cavitation. Almost no oxidation was observed on the specimen surface creep tested in flowing sodium and also no evidence of surface damage due to possible carburization and decarburization was noticed. SEM fractrographs of the creep ruptured specimens tested in flowing sodium showed predominantly ductile dimple failure whereas predominantly intergranular creep failure was observed in the creep ruptured specimen tested in air.

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.

Quantification of plutonium heterogeneity in (U, Pu) mixed oxide fuel using Passive Gamma Scanning

During the fabrication of (U, Pu)O2 mixed oxide (MOX) fuel for the Prototype Fast Breeder Reactor (PFBR), which is being built at Kalppakkam, India, it was necessary to achieve homogeneous distribution of plutonium in UO2 matrix for good thermo mechanical behavior of the fuel in reactor as well as for better solubility of the spent fuel during reprocessing. Since the conventional techniques used for estimating plutonium homogeneity in MOX are destructive in nature, they can be performed only on representative samples. The reliability of these techniques is less in large scale MOX fabrication facilities where MOX blends are prepared in large quantity. Passive Gamma Scanning of the fuel pin is carried out as a quality control check for PuO2 content in MOX fuel pin for PFBR. A new methodology has been developed in analyzing the Passive gamma scans to estimate the heterogeneity of plutonium distribution in MOX fuel pins for PFBR.

Helium Leak Detection of Vessels in Fuel Transfer Cell (FTC) of Prototype Fast Breeder Reactor (PFBR)

Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500MW Prototype Fast Breeder Reactor (PFBR) at Kalpak am, Chennai. In this very important and prestigious national programme Special Product Division (SPD) of M/s Kay Bouvet Engg.pvt. ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies like- Under Water trolley (UWT), Airlocks (PAL, EAL) Container and Storage Rack (CSR) Vessels in Fuel Transfer Cell (FTC) etc for PFBR. SPD of KBEPL caters to the requirements of Government departments like – Department of Atomic Energy (DAE), BARC, Defense, and Government undertakings like NPCIL, BHAVINI, BHEL etc. and other precision Heavy Engg. Industries. SPD is equipped with large size Horizontal Boring Machines, Vertical Boring Machines, Planno milling, Vertical Turret Lathe (VTL) & Radial drilling Machine, different types of welding machines etc. PFBR is 500 MWE sodium cooled pool type reactor in which energy is produced by fissions of mixed oxides of Uranium and Plutonium pellets by fast neutrons and it also breeds uranium by conversion of thorium, put along with fuel rod in the reactor. In the long run, the breeder reactor produces more fuel then it consumes. India has taken the lead to go ahead with Fast Breeder Reactor Programme to produce electricity primarily because India has large reserve of Thorium. To use Thorium as further fuel in future, thorium has to be converted in Uranium by PFBR Technology.

Sodium removal from sodium wetted under sodium ultra sonic scanner

Sodium is used as a coolant in primary and secondary heat transport systems of Liquid Metal cooled Fast Breeder Reactors (LMFBR). The Prototype Fast Breeder Reactor (PFBR) is in advanced stage of construction at Kalpakkam, India. Under Sodium Ultra Sonic Scanner (USUSS) is used to image the components dipped in sodium and more specifically the gap between Sub Assembly (SA) heads and the control plug. This is essential to confirm that no SA is protruding from its position. USUSS has to be qualified in sodium before its use in the reactor. After sodium testing, sodium adhering to the USUSS needs to be removed to prevent reaction with oxygen and moisture present in air when it is taken out. The sodium removal process is based on the use of moist carbon-di-oxide (CO2) at a temperature of about 353K. The reaction rate is controlled by varying the ratio of CO2 to moisture. Hydrogen released during the reaction of sodium with moisture is continuously monitored to ensure safe limits of hydrogen concentration. This paper describes the details of the experimental setup used in a cleaning pit for the removal of sodium adhering on USUSS and the technical details of the cleaning process.

Behaviour of steam generator bolted flange joints at elevated temperatures

Manholes with flanged joints are provided in the sodium heated once through straight tube steam generators of Prototype Fast Breeder Reactor (PFBR). The material of construction of steam generator is modified 9Cr 1Mo grade steel with studs made of 1 Cr 0.5 Mo steel. During the commissioning of model steam generator, the flanged joint lost its leak tightness at elevated temperature due to differential thermal expansion between flange and stud. A design modification incorporating sleeves of higher thermal expansion coefficient is discussed in this paper.

Mechanical behavior of fluoroelastomer considering long term ageing

Stress-elongation behavior of a Viton A-401C based compound established for backup seals of 500MW(e), Prototype Fast Breeder Reactor is depicted. Stress-softening effects are negligible during room temperature (RT) or 110°C measurements on unaged samples at strain rates of 50mm/min and 500mm/min because of low filler content. Relaxation is observed during RT property determination at lower strain rate. Stiffening of behavior at 50mm/min and 110°C is attributed to Joule-Gough effect and absence of relaxation because of increased molecular chain mobility. The two zone stress-elongation behavior (determined by chemical cross-links up to ∼85% strain and by ionic interactions beyond) propounded in the article by combining air-aged specimen data (32 weeks at 140/170/200°C) provides a behavior map of bisphenol cured, low filler, fluoroelastomers made of vinylidene fluoride and hexafluoropropylene. This could be extended to peroxide cured fluorocarbon rubbers for verifications and providing qualified compounds of better grades for critical nuclear elastomeric sealing applications.

Uncertainty analysis of the prototype FBR Monju with the JENDL-4.0 nuclear data set

This paper discusses the uncertainty analysis of the criticality of the prototype Fast Breeder Reactor Monju with JENDL-4.0 nuclear data. The JENDL-4.0 nuclear data are more physically correct and complete than previous JENDL releases. The well-established ERANOS code was used. Since no cross section libraries based on JENDL-4.0 exist for ERANOS, these libraries were generated as part of the work. To validate the newly made libraries, benchmark calculations were performed. A model of the Monju reactor was prepared, based exclusively on publicly available information. The criticality of May 2010 of Monju was adequately calculated. A perturbation analysis identified the most important differences between JENDL-3.3 and JENDL-4.0. A strong contribution of 241Am was found. In the subsequent uncertainty analysis, it was found that the overall uncertainty is roughly the same between JENDL-3.3 and JENDL-4.0. This result was not as expected. Two main causes have been identified: firstly, in JENDL-4.0 covariance data has been added for isotopes and energy ranges for which there previously was no covariance data. Secondly, for some isotopes the covariances in JENDL-4.0 are larger than in JENDL-3.3. For several main reactions and isotopes, the uncertainty is largely reduced between JENDL-3.3 and JENDL-4.0. But for many isotopes with a non-negligible contribution the uncertainty is increased in JENDL-4.0. The balance is a slight increase in the overall uncertainty when JENDL-4.0 is used.