Development of fluorocarbon rubber for backup seals of sodium cooled fast breeder reactor

The development of a fluorohydrocarbon rubber compound for static backup seals of 500 MWe, Prototype Fast Breeder Reactor (PFBR) is depicted. Variations of a previously developed Viton A-401C based formulation were subjected to processability tests, accelerated heat ageing in air, mechanical characterization and production trials. Finite element analysis and literature data extrapolation were combined with long term ageing to ascertain the life (minimum 10 years) of chosen formulation in reactor under synergistic influences of 110 °C, 23 mGy/h (γ dose rate) and air considering postulated accidental conditions. Validation of test seals and quality assessment indicate that composition and properties of the validated laboratory compound has been translated effectively to the reactor seals, installed recently in PFBR. The tensile and hardness specimens indicated negligible degradation and exceptional thermo-oxidative stability of the seal compound during ageing (32 weeks at 140/170/200 °C) even though interesting manifestations of cross-link exchange and ionic interactions were observed. Compression set results, showing definite trends of change under ageing and stain, were used in Arrhenius and Williams Landel Ferry equations for realistic life prediction. The development provides a foundation to simplify and standardize the design, development and operation of major elastomeric sealing applications of Indian nuclear reactors based on a few qualified compounds.     

Coatings for fast breeder reactor components

Several types of metallurgical coatings are used in the unique environments of the fast breeder reactor. Most of the coatings have been developed for tribological applications, but some also serve as corrosion barriers, diffusion barriers or radionuclide traps.The tribological coatings must meet a variety of performance criteria for friction coefficients, wear rates, galling resistance and self-welding resistance in liquid sodium. In addition, most applications require the coating to exhibit long-term resistance to sodium corrosion, resistance to damage by thermal cycling at temperatures up to 625°C and, for core applications, resistance to irradiation damage to neutron fluences of 6 × 10²² neutrons cm^-2 or more.The materials that have consistently given the best performance as tribological coatings in the breeder reactor environments have been coatings based on chromium carbide, nickel aluminide or Tribaloy 700 (trademark of Cabot Corporation) (a nickel-base hardfacing alloy). Other coatings that have been qualified for limited applications include chromium plating for low temperature galling protection and nickel plating for radionuclide trapping.Processes now employed include detonation gun coating, diffusion coating, electrospark deposition coating and electroplating. Several plasma spray processes, sputtering and chemical vapor depositions have been candidates for use on some reactor components, but did not pass the required qualification tests or were not economically competitive with alternative coating methods.     

Materials science research for sodium cooled fast reactors

The paper gives an insight into basic as well as applied research being carried out at the Indira Gandhi Centre for Atomic Research for the development of advanced materials for sodium cooled fast reactors towards extending the life of reactors to nearly 100 years and the burnup of fuel to 2,00,000 MWd/t with an objective of providing fast reactor electricity at an affordable and competitive price.     

Mixed plutonium-uranium carbide fuel in fast breeder test reactor

This paper describes the development of the indigenous plutonium-uranium mixed carbide nuclear fuel for the fast breeder test reactor. The fuel has performed satisfactorily and produced, for the first time in our country, nuclear electricity from a fast reactor. The experience and knowledge gained in the fuel development has provided great confidence for undertaking a programme on utilization of fast reactor technology for power production.     

On a leak-before-break assessment methodology for piping systems of fast breeder reactor

The Leak-Before-Break (LBB) concept has an effect on the safety design of Fast Breeder Reactors (FBRs), and thus its assessment has been one of the most significant issues. In the case of a commercial-scale FBR, since the main loads are the thermal expansion and thermal transient stresses, ferrite steel with a low thermal expansion rate has been a candidate material. Moreover, thin-walled and large-diametric pipes have been used to reduce the number of loops, which might also result in an economical advantage. A conventional LBB assessment method is insufficient to consider these characteristics, thus an advanced method is required. In this context, in the present paper, the following points were proposed to apply the LBB assessment method to ferrite steel pipes with thin walls and large diameters: (1) The surface resistance correction factor against a flow through penetrated cracks was improved for a reasonable leakage assessment under low-pressure. (2) The R6 method was applied to an unstable fracture assessment for postulated cracks. (3) A buckling assessment was introduced in determining the critical crack length for elbows. The applicability of this proposed method has been verified through an LBB assessment on typical ferrite steel pipes.     

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.     

Development of transition metal joint for steam generator circuit of prototype fast breeder reactor

Abstract

A transition metal joint between type 304 stainless steel and 2·25Cr–1Mo steel, with Alloy 800 as the transition piece, is being developed for application in the steam generator circuit of the 500 MW prototype fast breeder reactor. As part of this programme, the hot cracking susceptibility of Inconel 82/182 and of 16–8–2 welding consumables were compared and the microstructure and mechanical properties of butt welds between type 304 stainless steel and Alloy 800, welded by the two consumables, were studied to select the appropriate welding consumables for this joint. It is recommended that the 16–8–2 consumable should be used for welding this joint because of its lower microfissuring tendency and reduced mismatch in the coefficient of thermal expansion across the joint, although this would mean a slight adverse effect on the elevated temperature mechanical properties. Further, to select the optimum post-weld heat treatment (PWHT) of the joint between Alloy 800 and 2·25Cr–1Mo steel, welded with Inconel 82/182 welding consumables, the effect of PWHT on the microstructure and mechanical properties was studied. Decreasing the PWHT temperature was found to improve the mechanical properties and the microstructural condition of this joint.

MST/842     

Safety system consideration of a supercritical-water cooled fast reactor with simplified PSA

The probabilistic safety of the supercritical-water cooled fast reactor (SCFR) is evaluated with the simplified probabilistic safety assessment (PSA) methodology. SCFR has a once-through direct cycle where all feedwater flows through the core to the turbine at supercritical pressure. There are no recirculation loops in the once-through direct cycle system, which is the most important difference from the current light water reactor (LWR). The main objective of the present study is to assess the effect of this difference on the safety in the stage of conceptual design study. A safety system configuration similar to the advanced boiling water reactor (ABWR) is employed. At loss of flow events, no natural recirculation occurs. Thus, emergency core flow should be quickly supplied before the completion of the feedwater pump coastdown at a loss of flow accident. The motor-driven high pressure coolant injection (MD-HPCI) system cannot be used for the quick core cooling due to the delay of the emergency diesel generator (D/G) start-up. Accordingly, an MD-HPCI system in an ABWR is substituted by a turbine-driven (TD-) HPCI system for the SCFR. The calculated core damage frequency (CDF) is a little higher than that of the Japanese ABWR and a little lower than that of the Japanese BWR when Japanese data are employed for initiating event frequencies. Four alternatives to the safety system configurations are also examined as a sensitivity analysis. This shows that the balance of the safety systems designed here is adequate. Consequently, though the SCFR has a once-through coolant system, the CDF is not high due to the diversity of feedwater systems as the direct cycle characteristics.     

Liquid sodium—The heat transport medium in fast breeder reactors

Liquid sodium is used as a coolant in fast breeder reactors on account of its excellent heat transfer properties. It must, however, be in the pure form to be compatible with structural materials. Techniques for its purification to nuclear grade and its characterization had to be developed in our laboratory before we could embark on an R&D programme. It is essential to monitor hydrogen at ppb levels in the sodium circuits of the fast reactor in order to detect water leaks in the steam generator in a timely manner. Similarly it is useful to make on-line measurements of oxygen and carbon at trace levels. Electrochemical sensors have been developed in our laboratory for this purpose. These compact sensors work on the principle of concentration cells and are far superior to devices used elsewhere for this purpose. Corrosion of structural materials in the sodium environment depends on the oxygen content of sodium. In order to understand the mechanism of this corrosion, one must have a good grasp of the thermochemistry of the ternary systems, Na-M-O, where M stands for the alloying constitutents of stainless steels. The phase diagrams of most of these systems were established in our laboratory. A specially designed sodium loop is used in the study of corrosion, activity transport and kinetics of sodium-water reaction.     

Development of fuels and structural materials for fast breeder reactors

Fast breeder reactors (FBRs) are destined to play a crucial role in the Indian nuclear power programme in the foreseeable future. FBR technology involves a multi-disciplinary approach to solve the various challenges in the areas of fuel and materials development. Fuels for FBRs have significantly higher concentration of fissile material than in thermal reactors, with a matching increase in burn-up. The design of the fuel is an important aspect which has to be optimised for efficient, economic and safe production of power. FBR components operate under hostile and demanding environment of high neutron flux, liquid sodium coolant and elevated temperatures. Resistance to void swelling, irradiation creep, and irradiation embrittlement are therefore major considerations in the choice of materials for the core components. Structural and steam generator materials should have good resistance to creep, low cycle fatigue, creep-fatigue interaction and sodium corrosion. The development of carbide fuel and structural materials for the Fast Breeder Test Reactor at Kalpakkam was a great technological challenge. At the Indira Gandhi Centre for Atomic Research (IGCAR), advanced research facilities have been established, and extensive studies have been carried out in the areas of fuel and materials development. This has laid the foundation for the design and development of a 500 MWe Prototype Fast Breeder Reactor. Highlights of some of these studies are discussed in this paper in the context of our mission to develop and deploy FBR technology for the energy security of India in the 21^st century.     

Numerical modeling of thermodynamic processes in a cooled magnetic-fluid seal

The results are given from numerical computations of the velocity and temperature fields in a magnetic-fluid sealing layer under one tooth of a cooled multistage magnetic-fluid seal.Translated from Inzhenerno-Fizicheskii Zhurnal, Vol. 51, No. 3, pp. 368–375, September, 1986.     

单金属密封的结构优化及其密封性能研究

针对油气田勘探中,复杂的钻井工况导致动密封工作性能极不稳定的问题,结合单金属密封结构和井底高压环境,利用有限元方法对单金属密封受压情况下的接触压力进行分析。用雷诺方程计算单金属动密封的泄漏率,以减小最大接触压力和泄漏率为优化目标,利用正交试验和F评价方法对单金属密封结构参数进行优化,得到密封结构参数对密封面接触压力和泄漏率的影响情况,并将每个水平数对应的优化目标计算结果分别取平均值,得到不同水平影响下接触压力和泄露率平均值的变化趋势,从而确定密封结构的最优水平值,并借助有限元仿真对优化前后的密封性能进行对比。最后根据优化前后的密封结构参数加工2套密封试件,进行密封实验。仿真分析和实验结果表明：高压工况下优化前的密封面内侧磨损严重,钻井液颗粒容易侵入密封面;而优化后密封面的最大接触压力有所降低,动密封面的最高温度和泄漏率明显降低。研究结果对改进单金属密封的工作性能、提高井下动密封的可靠性有重要的现实意义。     

Numerical optimization of conduction cooled current leads for low current applications

The basic design principles of current leads for superconducting magnets are well established but HTS materials and conduction cooled systems call for new numerical methods. In this paper the design of current leads was formulated as an optimization problem. Both time integration and finite differencing were examined as possible ways to compute the temperature distribution inside the leads. Three examples about optimization of conduction cooled as well as gas cooled systems are presented. First, the design of tubular normal conducting gas cooled current leads was studied. Second, normal conducting leads cooled with a two-stage cryocooler were examined. Third, the optimization was applied to current leads consisting of HTS tapes at the low temperature end of a normal conducting bar. The study took into account the magnetic field and temperature dependent voltage-current characteristics of the anisotropic Bi-2223 material. The results are compared with traditional analytical ones and the numerical optimization is shown to be an efficient design tool for both normal conducting and HTS current leads.     

高温气冷堆氦气透平循环流动阻力特性分析

采用更加真实的阻力模型分析了流动阻力对10MW高温气冷堆(HTR-10)氦气透平循环特性的影响规律。分析结果表明,高温气冷堆氦气透平循环的压力损失主要由局部阻力和摩擦阻力组成。10MW高温气冷堆闭式氦气透平循环(HTR-10GT)发电系统在实际充装量调节及额定工况下,氦气在部件连接管道的局部压降占82.4%,沿程阻力压降占17.6%。氦气充装量减小时,局部压损系数不变而沿程阻力系数增大,导致循环效率降低;当充装量由100%降低到30%,连接管道的局部压降份额下降约20%,系统效率下降15%左右。随着充装量的减小,做功部件的进出口压力随充装量的变化线性变化,压气机的压比略有增大,透平的膨胀比有较大幅度的非线性增大。     

A study of reactor neutrino monitoring at the experimental fast reactor JOYO

We carried out a study of neutrino detection at the experimental fast reactor JOYO using a 0.76 tons gadolinium loaded liquid scintillator detector. The detector was set up on the ground level at 24.3 m from the JOYO reactor core of 140 MW thermal power. The measured neutrino event rate from reactor on-off comparison was 1.11±1.24(stat.)±0.46(syst.) events/day. Although the statistical significance of the measurement was not enough, backgrounds in such a compact detector at the ground level were studied in detail and MC simulations were found to describe the data well. A study for improvement of the detector for future such experiments is also shown.     

反应堆用磁致伸缩材料设计

简要介绍了磁致伸缩现象及几类常见的磁致伸缩材料,并对反应堆用磁致伸缩材料做了详尽的设计;特别是对磁致伸缩材料的抗辐照性能、磁致伸缩性能及高温性能做了详尽地分析.     

Data acquisition and postprocessing strategies for fast quantitative sodium imaging

An efficient scheme for quantitatively mapping the three-dimensional distribution of the sodium ion in vivo using magnetic resonance imaging is described. To make the methodology totally quantitative, the data acquisition scheme is performed with very short echo times and negligible T₁ saturation. Removal of signal variation due to imperfect radiofrequency (RF) response is accomplished using RF inhomogeneity maps acquired during each study. The high efficiency of the k-space trajectories allows the entire data collection process to be performed in under 10 min. The theory underlying the data collection and processing scheme is described along with representative examples acquired at 1.5 and 3.0 T. © 1997 John Wiley & Sons, Inc. Int J Imaging Syst Technol, 8, 544–550, 1997