Fast Breeder Reactor Research Articles

Reckless Proliferation and Guardianship Proliferation: The Fast Breeder Nuclear Reactor and the Plutonium Economy.

This article examines the history of the fast breeder reactor as a forerunner technology in three Western nuclear nations-the United States, the United Kingdom, and France-to clarify just how this charismatic technology changed the nuclear landscape. Did the adoption of fast reactors tip the nuclear balance toward a plutonium economy? And what were the interconnections between the production of plutonium for nuclear reactors and plutonium for atomic bombs? The conclusions reached differentiate between the decentralized and liberal U.S. plutonium economy from the British and the French state-controlled economies largely serving military interests. This furthermore conveys how the usage of plutonium made criticism of nuclear power both systematic and inclusive by tying together technological, political, and social risks and establishing a common bond between anti-breeder policies and environmental concerns.

Evaluation of thermal shock fracture resistance of B4C/CNT composites with a high-frequency induction-heating furnace

Abstract Boron carbide (B4C) ceramics with high thermal stability and relatively large cross-section for fast neutron absorption have been used as neutron absorbers of the control rods for the fast breeder reactors (FBR). However, cracks induced in B4C pellets due to thermal stress and swelling during neutron irradiation have been one of the practical problems, which leads to short lifetime of the control rods. It is essential to improve the mechanical properties of B4C ceramics for suppressing the cracking of B4C pellets and to enhance the thermal conductivity for relaxation of thermal stress during neutron irradiation. According to our previous study, the fracture toughness and thermal conductivity of B4C ceramics were enhanced by carbon nanotube (CNT) addition. In the present work, we report the dependence of the thermal and mechanical properties on the CNT-orientation in the sintered body prepared by hot-pressing and the enhancement of thermal shock fracture resistance of the obtained B4C/CNT composites. The thermal shock fracture resistance of hot-pressed B4C/CNT composites was evaluated by means of thermal shock test with a high-frequency induction-heating furnace. The dependence of the thermal properties on the CNT-orientation and the enhancement of thermal shock resistance of B4C ceramics were confirmed.

Simulation of Turbine-Generator

This study involved mathematical modeling of turbine generator simulator (TGS) and speed governing system for Fast Breeder Reactor (FBR). All steam turbine system utilize governor controlled valves at the inlet to the turbine to control the steam flow. The previously developed FBR and Pressurized water reactors (PWR) mathematical models (by author) [1] were connected to the same Turbine-generator model and controlled by the same method of control. Then a comparison between the dynamic performance transient response of FBR and PWR controlled plants were considered in two cases, load demand perturbation and reactivity perturbations. The steam chest and inlet piping to the turbine cylinder introduce delays between valve movement and change in the steam flow. The results show that the model is capable of simulating the dynamic transient response for FBR in two cases, load demand perturbation and reactivity perturbations.

Crystallographic phases in PbNb2O6 and Piezoelectricity

Emerging high temperature (HT) applications of piezoelectric (PE) materials in areas like nuclear Fast Breeder Reactors and rocket exhausts need PE materials with high Curie temperature. For decades, the rhombohedral (r) phase, stable form of lead metaniobate (PbNb2O6) has been assigned the space group R3m, and found non-piezoelectric. So, its study has been ignored in favour of the metastable orthorhombic PbNb2O6, showing T(Curie) > 570°C and serving HT PE applications. Our new findings on r-PbNb2O6 from DSC-TGA, impedance spectroscopy and x-ray Rietveld analysis, probe recent claims of polar R3 space group and hints of its ferroelectric phase up to ∼815°C.

Challenges and Safety in Erection and Commissioning of 280/85 Tons Single Failure Proof EOT Crane at PFBR

The Electric overhead traveling (EOT) crane installed in Prototype Fast Breeder Reactor (PFBR), Reactor Containment Building (RCB) is the largest capacity EOT crane installed in any nuclear power plants in India with some unique features of its own. It has been built with a single failure proof concept which is one of the prime requirements of a nuclear facility as per the standards of NUREG/IS. It is a double bridge girder type; the girder used in this crane is a single structure having a length of 34.8m without any joints as designed by IGCAR/BHAVINI (first of its kind). The span of the EOT crane is 33.4m and installed at a height of 42m from the ground level with the lifting height of 42.5m. This crane was manufactured by an Indian company M/s. UNIQUE CRANE, NASIK. Two hooks (280 tons / 85 tons) are connected in the same trolley sharing the Long Travel (LT). Load testing of the complete crane was done at factory before transporting it to PFBR site. The components of the crane were assembled in PFBR site. The erection methodology adopted was distinct w. r. t the site condition, as the components which were handled were very heavy, available space for installation was restricted and the head room was also very less. This erection has been completed successfully and very safely without any notable incidents. This activity is a biggest achievement / milestone in the nuclear industry in India. Later the EOT crane was commissioned, load tested with 350 tons load (negligible deflection) and being operated successfully without any incidents so far. This was achieved by accurate erection procedure, stringent Quality Assurance Plan (QAP), Job Hazard Analysis (JHA) and Safe Operation Procedure (SOP). Many challenges were faced during manufacturing, testing, erection and load testing at site. This paper deliberates those challenges.

Sensitivity to sterile neutrino mixing using reactor antineutrinos

The reactor antineutrinos are used for the precise measurement of oscillation parameters in the 3-neutrino model, and also used to investigate active-sterile neutrino mixing sensitivity in the 3+1 neutrino framework. In the present work, we study the feasibility of sterile neutrino search with the indian scintillator matrix for reactor anti-neutrino (ISMRAN) experimental set-up using electron antineutrinos ({overline{nu }}_e) produced from reactor as a source. The so-called 3+1 scenario is considered for active-sterile neutrino mixing, which leads to projected exclusion curves in the sterile neutrino mass and mixing angle plane. The analysis is performed considering both the reactor and detector related parameters. It is found that, the ISMRAN set-up can observe the active-sterile neutrino mixing sensitivity for sin ^{2}2theta _{14} ge 0.064 and varDelta m^{2}_{41} = 1.0hbox {eV}^2 at 90% confidence level for an exposure of 1 ton-year by using neutrinos produced from the DHRUVA reactor with thermal power of 100 hbox {MW}_{{th}}. It is also observed that, there is a significant improvement of the active-sterile neutrino mixing parameter sin ^{2}2theta _{14} to sim 0.03 at the same varDelta m^{2}_{41} by putting the ISMRAN detector set-up at a distance of 20 m from the compact proto-type fast breeder reactor facility with thermal power of 1250 hbox {MW}_{{th}}.

Ultrasonic Testing as Alternative for Radiography for Evaluation of PFBR Blanket Pin End Plug Weld Integrity

Radial blanket pins of Prototype Fast Breeder Reactor (PFBR) are made by encapsulating Deeply Depleted UO2 (DDUO2) pellets in a D9 austenitic stainless steel clad tube. The ends of the tubes are sealed by TIG welding of end plugs at both the ends. This is a critical weld as any failure will lead to spread of radioactivity in sodium coolant circuit of PFBR. At present, integrity of these welds is ensured by 100% radiography by using double wall single image full compensating block tangential radiography method. The present work aims at understanding the nature of defects encountered in the end plug welds and developing an Ultrasonic Testing (UT) methodology in lieu of radiography for qualification of the end plug welds. Typical defects have been examined by metallography and statistical analysis of weld zone has been carried out. A detailed analysis on various parameters affecting detection of tight lack of penetration (LOP) defects in radiography has also been carried out. A UT technique based on B-scan imaging in pulse-echo mode has been explored for inspection of end plug weld. Testing has been carried out with a highly damped spherically point focused 20 MHz transducer to produce 450 shear wave in weld joint area. End plug welds with different types of natural defects have been generated by varying different TIG welding parameters. UT and radiography of the same have been carried out. LOP has been detected with good sensitivity and UT results show good correlation with radiography.

Evaluation of cobalt free coatings as hardfacing material candidates in sodium-cooled fast reactor and effect of oxygen in sodium on the tribological behaviour

The feedback produced by operating Sodium-cooled Fast Reactors (SFRs) has shown the importance of material tribological properties. Where galling or adhesive wear cannot be allowed, hardfacing alloys, known to be galling-resistant coatings, are usually applied on rubbing surfaces. The most used coating is the cobalt-base alloy named Stellite 6® because of its outstanding friction and wear behaviour. Nevertheless, cobalt is an element which activates in the reactor leading to complex management of safety during reactor maintenance and mainly decommissioning. As a consequence, a collaborative work between CEA, EDF and FRAMATOME has been launched for selecting promising cobalt-free hardfacing alloys for the 600 MWe Sodium-cooled Fast breeder reactor project named ASTRID. Several nickel-base alloys have been selected from literature review then deposited by Plasma Transferred Arc or Laser Cladding on 17Cr austenitic stainless steel 316L(N) according to RCC-MRx Code (AFCEN Code). Among the numerous properties required for qualifying their use as hardfacing alloys in SFR, good corrosion behaviour and good friction and wear behaviour in sodium are essential. First results on these properties are shown in this article. Firstly, the corrosion behaviour of all coatings was evaluated through exposure tests in purified sodium for 5000 h at 400 °C. All coatings showed an acceptable corrosion behaviour in sodium. Finally, the friction and wear properties of one alloy candidate, NiCrBSi alloy, were studied in sodium in a dedicated designed facility. The influence of the oxygen concentration in sodium on the friction and wear properties was evaluated.

Austenitization and Tempering Temperatures Effects on EUROFER 97 Steel

EUROFER 97 is considered as standard steel for the nuclear applications in the case of high radiation density for first wall of a fast breeder reactors. Based on such consideration the microstructural behaviour of EUROFER 97 after thermo-mechanical processing is fundamental, since such materials properties are interesting also for innovative solar plants. In this paper the effect of thermo-mechanical behavior on the mechanical properties of EUROFER 97 has been analyzed by hot rolling followed by heat treatment on laboratory scale.A strong effect of reheating conditions before rolling on the material strength, due to an increase of hardenability following the austenite grain growth is found. A limited effect of the hot reduction and of the following tempering behavior is found in the considered deformation-range investigated. A loss of impact toughness is detected together with the hardness improvement.Mechanical properties are depending on the tempering temperature and an improvement of tensile yield stress (YS) and ultimate stress (UTS) was determined in tensile test carried on at T=550°C and T=650°C, e.g.: YS increase from about 400 MPa for standard EUROFER 97 [1] to about 550 MPa in samples treated by a tempering temperature of 720°C instead of the standard 760°C for EUROFER 97. Same trend has been found for UTS results.

Application of diffusion approximation in the calculations of reactor with cavities

The importance of calculation of radiation fields inside in-reactor cavities is associated with the necessity to simulate the emergency modes in fast breeder reactors (FBR), as well as reactor states with different coolant levels in special dedicated channels of passive feedback devices in lead-cooled fast reactors (LFR) of BREST type or in sodium cavities in sodium-cooled fast reactors (SFR). The Last Flight (LF) method (Bell and Glesston 1974, Davison 1960, DOORS3.2 1988, Mynatt et. al. 1969, Rhoades and Childs 1988, Rhoades and Sipmson 1997, SCALE 2009, Voloschenko et. al. 2012), or the method of the unscattered component is widely known and is commonly used in computer codes based on the method of spherical harmonics for obtaining solution in a gas medium at a certain distance from the calculated volume domain (DORT (Rhoades and Childs 1988), TORT (Rhoades and Sipmson 1997) and others (SCALE 2009)). The practice of its application (DOORS3.2 1988) demonstrated that acceptable results are achievable at considerable distances from the surface separating dense and gas media (more than two meters). Obtaining high-quality solution is not guaranteed for cavities within the calculation area. In addition, it is desirable to implement the cavities calculation methodology within the framework of the approximations used in reactor calculations introducing certain specific features. In particular isotropy of the neutron flux density and the necessity of forced introduction of a “conditional” calculation cell on the boundary surface of the void cavity are assumed in the diffusion approximation. If the LF method is oriented on the connection of the source point with the detection point, then it is necessary to determine in the calculation of neutron field in the cavities the neutrons escaping the surface area of the source and neutrons reaching a certain surface area of the cavity. In order to solve the problem, the authors suggested using the approximate solution presented in the paper. Thus, an algorithm for calculation of in-reactor cavities using the diffusion approximation was developed and implemented by the authors.

Effect of Applied Energy on the Microstructure, Texture, and Mechanical Properties of Short-Circuit Metal Inert Gas-Welded Modified Cr-Mo Steel Joints

Modified 9Cr-1Mo is a ferritic-martensitic steel widely used in steam generators of fast breeder reactors in the nuclear industry due to their enhanced creep resistance. In the present investigation, P91 steel was welded without preheat using three different heat inputs by pulsed short-circuit metal inert gas welding process using a ER90S-B9 wire electrode for the first time. The microstructural developments were characterized by using optical and electron microscopy, while the residual stresses were measured by X-ray diffraction using the sin2ψ method. The grain size variation in the fusion zone/heat-affected zone was effectively studied using electron backscattered diffraction to bring out the changes quantitatively. The weld zone had a consistent texture, while the heat-affected zone was random. The grain size tends to increase with the increase in heat input which leads to reduced joint strength. It is evident that at highest heat input, the weld microstructure shows substantial precipitation of M23C6-type carbides. The residual stresses were near compressive in the weld-HAZ, while maximum tensile stresses were found for the highest heat input.

Results of tensile, hardness and bend tests of modified 9Cr 1Mo steel welds: Comparison between cold wire and hot wire gas tungsten arc welding (GTAW) processes

Hot wire Gas Tungsten Arc Welding (GTAW) process is considered in addition to conventional cold wire GTAW process for the manufacture of Steam Generators (SG) for India's Prototype Fast Breeder Reactor (PFBR). During manufacture, it is an obligatory requirement to weld one test coupon for every 20 m of weld length on the actual SG for ascertaining the properties during fabrication. Weld pads thus prepared were used for characterization of hardness, bend and tensile properties of modified 9Cr 1Mo steel welds produced by hot wire and cold wire GTAW processes and the test results were compared. The results revealed that hardness, bend and tensile properties of modified 9Cr 1Mo steel welds of hot wire and cold wire GTAW processes are similar and no genuine difference exist between these two welding processes. Further, the advantage of high quality with excellent weld profile and high productivity has been realized using hot wire GTAW process.

Development of ultrasonic guided wave inspection methodology for steam generator tubes of prototype fast breeder reactor

An ultrasonic guided wave based methodology is developed for inspection of steam generator tubes of the prototype fast breeder reactor. To this aim, axisymmetric longitudinal mode (L(0,2)) at the frequency of 250 kHz is optimized using 3D-finite element simulation and experiments. The group velocity of mode L(0,2) at 250 kHz is found to be 5387 m/s. First, the long range propagation of the L(0,2) mode at 250 kHz is examined and the mode is found to propagate over a distance of 45.6 m with a sufficiently good SNR. Secondly, the detection of multiple defects such as circumferential, axial, partial-pinholes and tapered defects lying in the same line of sight is investigated using 3D-finite element simulation and the results obtained are validated experimentally for the first three cases. The sensitivities achieved are 0.23 mm depth (10%WT) for circumferential, axial and tapered defects and for partial-pinholes: 1 mm diameter and 1.38 mm depth (60%WT). Thirdly, 3D-FE simulations with ID and OD pinhole defects are performed which show that the ID and OD defects are detected by L(0,2) with a fairly similar sensitivity. Finally, study on the thermal expansion bend (with three successive bends) shows that the bend does not have much influence on the mode and the multiple circumferential defects considered in the bend are detected with good sensitivity.

Reactor dynamics of in-pin fuel motion in fast breeder reactors

Abstract In-pin fuel motion is a multi-phase hydrodynamic movement of molten fuel inside the fuel pins of a fast reactor during a core disruptive accident. This study focuses on the potential consequences of this phenomenon on fast breeder reactor dynamics during an unprotected transient overpower (UTOP) accident. An in-house developed solver is benchmarked with the CABRI-E9bis test to addresses the complex interplay of hydrodynamic forces. It is deduced that in slow overpower, molten fuel hydrodynamics is dominated by gravity and solidification blockage. Next, the solver is integrated with an in-house reactor dynamics code ‘PREDIS’ to simulate UTOP under the effect of in-pin fuel motion. A 500 MWe mixed oxide fuel based fast reactor is chosen for analyses. Results show that in-pin fuel motion generates a negative reactivity feedback of the order of −0.45 $. This results in a reduced peak power (182%) and stabilized state power (175%). Damage to the reactor core is contained more effectively. It is found that in a beginning of equilibrium core (BOEC), high internal pressure of the fuel pin does not permit molten fuel vaporization at high temperatures. Higher reactivity insertion rates of up to 20 pcm/s are also countered with in-pin fuel motion. From this study, it is inferred that in-pin fuel motion feedback enhances the inherent safety features of fast reactors during UTOP. It scales in magnitude with the fuel Doppler, and is therefore an important parameter for future safety analyses of oxide fuelled fast reactors.

Chemical interaction of D9 alloy clad with B4C in liquid sodium: Studies employing XPS, XRD, and SEM

AbstractA cell representing typical control rod subassembly of the prototype fast breeder reactor is designed to probe the liquid sodium‐mediated chemical interaction of boron carbide (B4C) control rod with the D9 alloy clad. The cell was equilibrated at 973 K for 5000 hours with liquid sodium in the annular gap. XRD shows the formation of Cr3C2 and Fe2B along with oxides of Ni, Cr, and Fe. XPS studies reveal the diffusion of boron and carbon up to a depth of about 160 and 120 μm, respectively. A boron‐rich region is observed up to a depth of about 40 μm which consists of B3+ arising from oxides and a nearly constant elemental boron region, extending up to a depth of 160 μm. The highly reacted zone extends up to a depth of about 40 μm consisting of oxides of most of the elements of D9 clad exhibiting a higher valence states. As the cell is helium leak tested, oxygen out gassed from B4C at 973 K played a major role in chemical oxidation of the constituents of the D9 alloy which can be minimized by using high‐density B4C.

Stability characteristics of metal fueled fast reactor cores

Abstract Linear stability analysis using transfer function method is carried out for both metal and MOX (Mixed Oxide) fueled Fast Breeder Reactors (FBR). The better heat transfer properties of metal fueled reactor gives a prompt feedback and hence a lower feedback time constant. The stability of each system is demonstrated by plotting the Nyquist diagram. The stability margin is estimated in terms of phase margin calculated from the Nyquist diagram.

Examining Stress Relaxation in a Dissimilar Metal Weld Subjected to Postweld Heat Treatment

Abstract Dissimilar metal welds are often required in nuclear power plants to join components made from austenitic steels to those from ferritic steels, particularly in fast breeder reactor plants, in order to join the intermediate heat exchanger to the steam generator. The process of welding alters the microstructure of the base materials and causes residual stresses to form, both because of the change in the microstructure and the differing thermal histories in various regions. Postweld heat treatment (PWHT) is required to relieve the residual stresses and achieve preferable microstructural gradients across the weld joint. Therefore, in order to arrive at the optimal PWHT process, it is necessary to investigate the effects of heat treatment on the joint integrity, microstructure, and residual stress relaxation in the welds. To investigate the effect of PWHT on the residual stress relaxation and corresponding alteration of microstructure across a welded joint, a dissimilar weld between modified 9Cr-1Mo steel and austenitic stainless steel AISI 316LN was made using autogenous electron beam welding. To achieve this, the welding process was first modeled numerically using finite element analysis, and the residual stress predictions were validated by experimental investigation using neutron diffraction. The validated model was then used to study the residual stress relaxation through the simulation of PWHT. The predicted stress relaxation was compared with contour method measurement of residual stresses in the actual welded plate subjected to PWHT. The results indicate that, although some relaxation of residual stresses occurred during PWHT, there is still a significant portion of highly localized residual stresses left in the specimen.

Numerical approach to study the thermal-hydraulic characteristics of Reactor Vessel Cooling system in sodium-cooled fast reactors

The main vessel plays an important role in containing the entire primary sodium for pool-type sodium-cooled fast reactors (SFRs). The Reactor Vessel Cooling System (RVCS) has great effect on cooling the main vessel. However, little attention has been given to the study on transient characteristics of RVCS in the previous SFR research. Thus, a home-made one-dimensional (1-D) code named Reactor Vessel Cooling system Analysis Code for Sodium-cooled fast reactor (VECAS) is proposed to evaluate the thermal-hydraulic characteristics for SFR. The detailed models of the developed VECAS are presented in this paper. Moreover, the developed models have been validated against an experimental study. Numerical data of the main vessel cooling circuit are compared with the measurements of the Demonstration Fast Breeder Reactor (DFBR). The simulation results are in good agreement with the experimental data. Furthermore, the validated VECAS is coupled with the Transient Thermal-Hydraulic Analysis Code for Sodium-cooled fast reactors (THACS). The transient characteristics of RVCS in China Experimental sodium-cooled Fast Reactor (CEFR) are simulated by the coupled code. Steady analysis shows that the main vessel is cooled effectively. The peak temperature appears at the top of the main vessel lower than the permissible upper temperature limit. During the transient analysis, VECAS has predicted a reverse flow in RVCS, which contributes to the core cooling. Furthermore, sensitivity analysis of the main parameter has also been performed. Therefore, it can be concluded that coupled VECAS has the ability to evaluate the thermal-hydraulic characteristics as well as the decay heat removal capacity of RVCS. The coupled code could provide references and technical supports for the design and optimization of the pool-type sodium-cooled fast reactor.

Plasma-sprayed alumina coating on Inconel 600 for fast reactor neutron detector application

ABSTRACTAlumina coating, with high electrical resistivity, has been proposed over Inconel 600 rings for application in fission chamber neutron detectors, operating at 923 K in fast breeder reactor. For improving the thermal cycling life and adhesion strength of alumina coating, selection of the suitable coating technique and bond coat, with appropriate spray parameters was pursued. In the present study, Ni–50%Cr bond coat was deposited over Inconel 600 by the high velocity oxygen fuel technique, while the top alumina coat was deposited using air plasma spray process. A dense coating with minimum porosity and a stable insulation resistance (600 V DC) of ∼1011 ohm at room temperature and ∼108 ohm at 923 K was achieved. Thermal cycling studies on coated samples and rings showed good thermal cycle life tested up to 100 cycles. The coating exhibited good bond strength and only marginal decrease in bond strength was observed with increase in thermal cycling.

Development of a high sensitive magnetic flux leakage instrument for imaging of localised flaws in small diameter ferromagnetic steel tubes

The authors report the development of a high sensitive, 16 channel magnetic flux leakage (MFL) instrument for imaging of localised flaws in small diameter ferromagnetic steel tubes. The instrument consists of a magnetising unit of two bobbin coils wound on a ferrite core, a sensing unit of flexible giant magneto-resistive (GMR) sensor array and an image visualisation unit. 3D-non-linear finite-element modelling has been carried out to optimise the inter-coil spacing of the bobbin coils and to identify the number of GMR elements and their locations between the bobbin coils. A series of experiments has been carried out to evaluate the flaw detection performance of the MFL instrument. Studies reveal that the instrument can image as small as 1.1 mm diameter hole and 0.46 mm deep shallow outer diameter circumferential notch in 17.2 mm diameter Mod. 9Cr-1Mo steel tubes used as steam generator material in prototype fast breeder reactor, with a signal-to-noise ratio better than 8 dB. Studies also indicate that intensity of MFL signal measured by the instrument increases with increase in hole diameter and notch depth. The instrument is portable and easy to use for field inspection of ferromagnetic steel components.