Commercial Fast Breeder Reactor Research Articles

Design and Suitability Analysis of FPGA-Based TCP/IP Offload Module for Control Systems of FBRs

Transmission Control Protocol/Internet Protocol (TCP/IP) offload modules play a crucial role in India’s first commercial fast breeder reactor (PFBR) distributed digital control system architecture by providing networking capabilities to the real-time control systems used in safety-critical and safety-related applications of the PFBR. These modules enable operators to remotely control various mechanisms, facilities, and process loops via real-time control systems from the main control room display stations. The malfunctioning of these modules may hamper the overall plant’s availability; hence to improve further upon safety, security, reliability, and performance, these modules have been indigenously developed for upcoming fast breeder reactors in India. This paper proposes methods to justify the suitability of these communication modules as per the D-25 design guidelines issued by the Atomic Energy Regulatory Board of India. The design and testing aspects are covered, including an approach toward hardware descriptive language-based design and testing with respect to the IEEE standard and the TCP/IP protocol specifications. Module design validation based on ethernet frame transactions is detailed in this paper. The test platform design, which includes the diagnostics software design for CPUs designed with these TCP offload modules, and the design of a graphical user interface to detect and record TCP connection anomalies are detailed. The network interface response to postulated events of TCP packet corruption, packet drop, packet delay, out-of-order packets, duplicate packets, and heavy traffic is characterized.

Study on startup characteristics of prototype once-through steam generator for China fast reactor

Once-through steam generator (OTSG) is a vital heat transfer equipment of China Fast Reactor-600 (CFR-600), and its startup characteristics are of great importance to the safe operation of nuclear power plant. In this paper, first of all, the Thermal-hydraulic analysis Code Of Sodium heated once-through Steam generator (TCOSS) has been further optimized, and the code was introduced in the text. Secondly, the Steam Generator Test Facility (SGTF) and India Commercial Fast Breeder Reactor (CFBR) were selected for steady-state verification, and the feedwater flow rate increase condition of the prototype OTSG experiment was selected for transient verification. All the verifications were in good compliance, which proved the reliability of the TCOSS code for subsequent calculation of startup conditions. Finally, taking the prototype OTSG of CFR-600 as the research object, and the transient characteristics of cold startup and hot startup have been predicted and analyzed. The results show that both processes of cold startup and hot startup for prototype OTSG can be realized stably. The simulation results of startup processes showed a good agreement with expectations, which can provide support for the development of CFR-600 steam generator.

On Some Differences of the Thermoelectrical Effects in BN-600 and Superphenix

Abstract The paper discusses specific electromagnetic effects observed in the large circulating masses of liquid metal inherent in the heat transport systems of the commercial fast breeder reactors. It was explained why the criterion Lu of the magnetic field self-excitation in the liquid metal coolant has limiting conditions, determined by the balance nature of the thermoelectric currents circulating in reactor.

Verification of the FBR fuel bundle–duct interaction analysis code BAMBOO by the out-of-pile bundle compression test with large diameter pins

The BAMBOO computer code was verified by results for the out-of-pile bundle compression test with large diameter pin bundle deformation under the bundle–duct interaction (BDI) condition. The pin diameters of the examined test bundles were 8.5mm and 10.4mm, which are targeted as preliminary fuel pin diameters for the upgraded core of the prototype fast breeder reactor (FBR) and for demonstration and commercial FBRs studied in the FaCT project. In the bundle compression test, bundle cross-sectional views were obtained from X-ray computer tomography (CT) images and local parameters of bundle deformation such as pin-to-duct and pin-to-pin clearances were measured by CT image analyses. In the verification, calculation results of bundle deformation obtained by the BAMBOO code analyses were compared with the experimental results from the CT image analyses. The comparison showed that the BAMBOO code reasonably predicts deformation of large diameter pin bundles under the BDI condition by assuming that pin bowing and cladding oval distortion are the major deformation mechanisms, the same as in the case of small diameter pin bundles. In addition, the BAMBOO analysis results confirmed that cladding oval distortion effectively suppresses BDI in large diameter pin bundles as well as in small diameter pin bundles.

Axial and radial shield optimization for CFBRs with ferro-boron

A series of Commercial Fast Breeder Reactors (CFBRs) are planned to be constructed in India. Ferro-boron is proposed as neutron shield in CFBR cores. This work presents the optimization studies carried out towards reducing volume and weight of shields with ferro-boron, keeping in view the requirements of maintaining secondary sodium activity below permissible limits and improving neutron monitoring efficiency at the above core location. Three shield configurations have been considered, all with reduced radial ferro-boron shields as compared to Prototype Fast Breeder Reactor (PFBR). They differ in their axial shield configuration. The studies have shown that the best configuration is with axial stainless shields above core-1 and boron carbide (B4C) shielding over blankets. This configuration preserves the secondary sodium activity and also gives rise to 50% increase in neutron detection efficiency at the above core location. The material cost reduces nearly 5 times and the advantage of weight reduction is considerable.

A CFD based approach for thermal hydraulic design of main vessel cooling system of pool type fast reactors

A computational fluid dynamics (CFDs) based approach is proposed for the thermal hydraulic design of the main vessel cooling system for pool type sodium cooled reactors. Usage of the proposed method is demonstrated by applying it to a future Indian commercial fast breeder reactor. Towards quantifying the amount of sodium flow rate for the main vessel cooling system, two-dimensional CFD investigations have been performed. The conjugate conduction–convection models adopted for this purpose are validated against sodium experiments available in literature. The required flow fraction has been determined to be 2.6% of core flow, which is 175.6kg/s at full power conditions. The heat loss from the hot pool to the cold pool through the main vessel cooling system is estimated to be 10.6MW at full power and 3.7MW at 20% power conditions. By detailed three-dimensional CFD studies, the effect of ovality in the main vessel cooling annuli due to manufacturing tolerances has been assessed and the associated circumferential temperature difference in the main vessel is determined to be 14°C, which is less than the permissible upper limit of 30°C. The uniformity of sodium flow in the cooling annulus has been investigated by a three-dimensional hydraulic analysis with a view to identify a suitable passive device that can render a uniform velocity distribution. To compare the effectiveness of various devices, a flow mal-distribution index is defined. Detailed parametric studies have been carried out to identify an appropriate porous jet breaker plate configuration.

Design of Advanced Integrated Reactor Protection System for Commercial Fast Breeder Reactor

Shutdown systems are one of the most important one for the safety of the reactor core. In case of any abnormal event or any parameter going abnormal or otherwise, it trips the reactor. Hence this system should be highly reliable. The reliability of the Present shutdown system could not be improved beyond a certain limit, due to many interconnections between the sub systems in the safety chain. i.e. Different sensors, processing electronics, safety logic are located physically at different and far off locations connected with cables, passing through cable trays, penetrations and buildings. New design configuration has been proposed, where there is only single system (PCB) comprising of processing electronics and comparison with thresholds, with voting logic built into the control rods itself. This eliminates the need for one exclusive safety logic system. With this design, the reliability promised to be more than the existing system, without the need for more than one shutdown system. With this design, the reliability promised to be more than the existing system, without the need for more than one shutdown system. The thesis discusses the details of the design made and improvements in the reliability factor.

Thermal simulation of sodium heated once through steam generator for a fast reactor

In typical sodium cooled fast reactors, steam generators (SG) transports heat from hot liquid sodium to high pressure feed water. Feed water gets heated and converted to superheated steam and is used to roll the turbine. In SG, tubes are in triangular pitch pattern and shell in cylindrical shape, making some non-uniform gap at periphery allowing sodium to by-pass resulting radial thermal difference. Estimation of this thermal difference is very important in order to design a provision to accommodate the maximum differential thermal expansion among tubes and shell and to resist maximum thermal cycles faced by SG in its design life. A detailed three-dimensional thermal analysis of a sector model of SG with 30 m long tube is performed by combining a commercial Computational Fluid Dynamic code with in-house one-dimensional code DESOPT. The analysis is also validated with in-house experimental test facility results. From the analysis, the radial and axial temperature distributions and the differential temperature between tubes and shell are estimated for commercial fast breeder reactors SG for normal operation and typical plugged tube conditions and the results are highlighted in this paper. The results would form input for the estimation of differential thermal expansion of shell and tubes of SG and design of tube-sheets.

Steam generator test facility—A test bed for steam generators of Indian sodium cooled fast breeder reactors

Abstract Second stage of India's three stage nuclear power programme consisting of the development and operation of fast breeder reactors (FBRs) commenced with the construction, commissioning and operation of fast breeder test reactor (FBTR) at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam. FBTR operating experience and better understanding of the technologies involved has enabled India to design a 500 MWe prototype fast breeder reactor (PFBR). Construction of PFBR is in advanced stage at Kalpakkam. Steam generator (SG) is one of the critical components of a fast reactor power plant. It is a once-through SG housing both water at high pressure and sodium at low pressure separated by a tube wall. Any defect to this barrier resulting in an internal leak allowing sodium and water to react could severely affect the plant availability. A 5.7 MWt capacity steam generator test facility (SGTF) has been set up in Indira Gandhi Centre for Atomic Research (IGCAR) to test a 5.5 MWt sodium heated once through SG of 19 tubes similar to the indigenously designed 157 MWt, 547 tube SG of PFBR. The test programme includes the evaluation of thermal hydraulic performance of SG, instability mapping, measurement of tube bundle vibrations etc. Most of the planned experiments have been completed successfully. The heat transfer experiment has proved that the SG has a heat transfer area margin of 12.25%. This experiment has also helped in validating the design codes and reducing the area margin for SGs of future FBRs. Experiment on flow induced vibration of tube bundles has shown that the bending stress due to vibration is much less than the endurance limit of the material. The results of the experiments and the valuable lessons learnt from the operating experience have been successfully applied to PFBR and will be applied to future commercial fast breeder reactors. The facility is also planned as a test bed for SGs of Indian FBRs and to assess improved designs of auxiliary equipment used in FBRs. This paper brings out various experiments conducted and the experience gained during the operation of the facility.

Design and development of thick plate concept for rotatable plugs and technology development for future Indian FBR

► Thick plate concept and design for rotatable plugs of future Indian FBR described. ► Salient aspects of 800 mm thick plate welding technology development presented. ► Non-destructive examination (NDE) and destructive testing aspects discussed. ► Ultrasonic wave is attenuated less in narrow gap carbon steel weld than base metal. ► Recommendation made regarding RCC-MR code rules for NDE of carbon steel weld. Prototype fast breeder reactor (PFBR) is at an advanced stage of construction in Kalpakkam, India. Top shield (consisting of roof slab and rotatable plugs) of PFBR is of box type construction. It is a Class-1 component being a part of the primary leak-tight boundary of the reactor assembly. In the future commercial fast breeder reactor (CFBR), it is planned to adopt thick plate concept for rotatable plugs. The thickness of the plates required for the rotatable plugs of CFBR was arrived at based on finite element analysis considering mechanical and thermal loads. The main advantage of using thick plate for the rotatable plugs is that it eliminates the possibility of lamellar tearing which exists in the box type construction. The successful realisation of this concept necessitates the indigenous development of thick plate narrow gap welding technology. Hence a technology development programme for realising 800 mm thick narrow gap welds using submerged arc welding process was undertaken. Three numbers of welded joints were made to demonstrate the successful development of thick plate narrow gap welding technology. Repair procedure for the weld was also established. Non-destructive examination and destructive testing were carried out and the results were analysed critically. The attenuation of ultrasonic wave in narrow gap weld as compared to that in the base metal was examined. This paper discusses the need and conceptual design of thick plate concept for rotatable plugs in CFBR, the approach adopted for technology development of thick plate narrow gap welding and the salient experiences gained during the developmental project. Recommendations are made regarding the RCC-MR code rules for the non-destructive volumetric examination of welds of thick carbon steel plates.

Enhancement of proliferation resistance properties of commercial FBRs by material barriers

The possibility to enhance proliferation resistance of discharged plutonium in the blanket of fast breeder reactor (FBR) has been investigated in terms of reactor core design aspects. It can be achieved by increasing the fraction of even-mass-number plutonium isotopes in the discharged plutonium. Although it can be achieved by a radial blanket-free core, it also decreases the breeding ratio. By loading the blanket with plutonium (Pu) or minor actinides (MAs) or Pu-MA combination, it is possible to reduce the attractiveness of discharged plutonium. However, the loading material in the blanket should be kept in balance with achievable core performances, fuel fabrication, and fuel handling, accordingly and hence the criteria needed to comply with. It is found that applying the attractiveness or FOM criteria that using the combination of even-mass-number plutonium isotopes need more material to be loaded than the one needed when applying Pellaud’s or Kessler’s proposal that using only solely isotope fraction of plutonium.

Role of ADS in the back-end of the fuel cycle strategies and associated design activities: The case of Japan

Reduction of burden caused by radioactive waste management is one of the most critical issues for the sustainable utilization of nuclear power. The Partitioning and Transmutation (P&T) technology provides the possibility to reduce the amount of the radiotoxic inventory of the high-level radioactive waste (HLW) dramatically and to extend the repository capacity. The accelerator-driven system (ADS) is regarded as a powerful tool to effectively transmute minor actinides (MAs) in the “double-strata” fuel cycle strategy. The ADS has a potential to flexibly manage MA in the transient phase from light water reactors (LWRs) to fast breeder reactors (FBRs), and can co-exist with FBR symbiotically and complementarily to enhance the reliability and the safety of the commercial FBR cycle. The concept of ADS in JAEA is a lead–bismuth eutectic (LBE) cooled, tank-type subcritical reactor with the power of 800 MWth driven by a 30 MW superconducting LINAC. By such an ADS, 250 kg of MA can be transmuted annually, which corresponds to the amount of MA produced in 10 units of LWR with 1 GWe. The design study was performed mainly for the subcritical reactor and the spallation target with a beam window. In Japan, Atomic Energy Commission (AEC) has implemented the check and review (C&R) on P&T technology from 2008 to 2009. In the C&R, the benefit of P&T technology, the current status of the R&D, and the way forward to promote it were discussed.

Vibro-Packing Experiment of Non-Spherical Uranium Dioxide Particles with Spherical Metallic Uranium Particles

Japan Atomic Energy Agency has developed vibro-packed fuel as one of the candidates for commercial fast breeder reactor fuels. In this study, vibro-packing experiments were carried out to investigate particle movement during vibro-packing as well as particle distribution after the completion of vibro-packing. Non-spherical uranium dioxide particles and spherical metallic uranium particles were used to simulate mixed oxide particles and oxygen getter particles, respectively. These experiments revealed the following facts. It is important to prevent segregation by means of feeding each size of the fuel particles uniformly into a cladding tube in order to disperse oxygen getter particles uniformly. “Simultaneous feeding” with volumetric powder feeders is useful to realize the above.

Steam Generator Tubing Development for Commercial Fast Breeder Reactors

The development work to design, manufacture, and evaluate pre-stressed double-wall 2/one quarter/ Cr-1 Mo steel tubing for commercial fast breeder reactor steam generator application is discussed. The Westinghouse plan for qualifying tubing vendors to produce this tubing is described. The results achieved to date show that a long length pre-stressed double-wall tube is both feasible and commercially available. The evaluation included structural analysis and experimental measurement of the pre-stress within tubes, as well as dimensional, metallurgical, and interface wear tests of tube samples produced. This work is summarized and found to meet the steam generator design requirements. 10 refs.

Modified One-Mode Method for Fast-Reactor Diffusion Calculations

The modified one-mode method for fast-reactor neutron diffusion calculations was formulated by collapsing two- or three-energymode synthesis equations to an effective one-mode equation. The calculational procedure consists of solving an eigenvalue problem to determine the effective neutron multiplication factor and the first-mode expansion coefflcient, and solving inhomogeneous problems to determine the higher mode expansion coefficients. Therefore, the computer running time nearly equals that of the conventional one- group eigenvalue problem. The accuracy of this method was investigated by comparing the results obtained by a modified one-mode method with reference 26- group calculations, employing a one-dimensional radial model of a commercial fast breeder reactor. The discrepancies between the modifted one-mode method based on three-mode synthesis and the 26-group method are <0.1% in the effective multiplication factor, 5% in the control-rod reactivity and <2% in the power distribution. These results assure the applicability of this method to fast- reactor design studies. (auth)

Radiation-Induced Swelling of Stainless Steel

Significant swelling (1 to 10 percent due to small voids have been found in stainless steel when it is exposed to fast neutron doses less than expected in commercial fast breeder reactors. The main features of this new effect are: (i) the voids are formed by the precipitation of a small fraction of the radiation-produced vacancies; (ii) the voids form primarily in the temperature range 400 degrees to 600 degrees C (750 degrees to 1100 degrees F); and (iii) the volume increases with dose (fluence) at a rate between linear and parabolic. The limited temperature range of void formation can be explained, but the effects of fluence, microstructure, and composition are determined by a competition between several kinetic processes that are not well understood. This swelling does not affect the feasibility or safety of the breeder reactor,but will have a significant impact on the core design and economics of the breeder.Preliminary results indicate that one cannot eliminate the effect,but cold-working,heat treatment, or small changes in composition can reduce the swelling by a factor of 2 or more. Testing is hampered by the fact that several years in EBR-II are required to accumulate the fluence expected in demonstration plants. Heavyion accelerators,which allow damage rates corresponding to much higher fluxes than those found in EBR-II,hold great promise for short-term tests that will indicate the relative effect of the important variables.

高速増殖炉原型炉の設計の一覧

A survey has been attempted of the current design concepts of prototype fast breeder reactors in the world, and they are classified according to their characteristics. The survey was focused on the trend of a closer interrelationship between each prototype and experimental as well as large commercial fast breeder reactor.