Prototype Fast Breeder Reactor Research Articles

Lessons Learned from Failures and Design Improvements for Large Diameter Seals

Large diameter seals are used in Prototype Fast Breeder Reactor (PFBR) to prevent the escape of radioactive argon cover gas through wide annular gaps. The design of these seals shall address different challenges than normal seals. The large tolerances needed on the width of annulus, achievable tolerances on rubber, clearances in the seal holder, etc., influence the seal performance. The paper discusses analytical and experimental work done to develop robust large diameter seals. The contours of the present study are based on the feedback from the testing and qualification of small size model seals. Various permutations of geometry for seal and its holder are numerically investigated. The study mainly highlighted the influence of seal holder on the seal performance. Various sensitivity studies done on selected seal and seal holder demonstrated robustness of the design solutions.Graphical

Evaluation of Reaction Rate Distribution for Shielding Region in the Prototype Fast Reactor Monju

The reliability and usefulness of the reaction rate distribution data measured in the prototype fast breeder reactor Monju were examined through a comparison with a calculation using JENDL-4.0, mainly focusing on shielding regions around the reactor core. The reaction rates of 238U (n,f) and 58Ni (n,p) sensitive to high-energy neutrons were all judged reliable. The calculation-to-experiment values were slightly worse in the shielding regions; however, those for the 58Ni (n,p) reaction rates were improved by employing JEFF-3.3 instead of JENDL-4.0. A different tendency was observed between the two reactions, probably due to the 238U (n,f) cross section in the energy range of around 700 eV. The reaction rates of 235U (n,f), 239Pu (n,f), 238U (n,γ), and 197Au (n,γ) sensitive to the lower-energy neutrons were mostly judged reliable. The data in the lower axial shielding region are less reliable but may be acceptable for the shielding calculation.

Evaluation of Reaction Rate Distribution Measurement in the Prototype Fast Reactor Monju

In the prototype fast breeder reactor Monju, reaction rate distributions of the fission reaction rates of 239Pu, 235U, and 238U and the capture reaction rate of 238U were measured using activation foils during its system startup test. The measurements in the core and radial blanket regions were evaluated in detail, and their reliability and usefulness as the validation data for fast reactor neutronics design methodologies were examined through a comparison with calculations. The reaction rate data measured in Monju were confirmed all reliable and useful as the validation data. The fission reactions of 239Pu, 235U, and 238U can be validated with an accuracy of a few percent in the core and blanket regions. The capture reaction of 238U in the core region also can be validated with a similar accuracy, whereas a precise calculation of the foil cross section is necessary to consider the resonance shielding effects of the surrounding fuel pins and a foil.

Evaluation of Specific Wear Coefficient of Sliding Contact Pairs through Tribological Experiments

In fuel handling machine of Prototype Fast Breeder Reactor, colomony-5 is extensively used for hardfacing of rubbing surfaces. To avoid the challenges associated with hardfacing of the components, it is proposed to eliminate hardfacing by using a forged component of a suitable material. 440C steel is a candidate material for this application. Hence, experiments were conducted to evaluate the tribological properties of 440C steel under various conditions seen during the operation. The tests were carried out using pin on disc tribometer by varying the load from 10-20 N and sliding speed from 5-30 mm/s. Friction and specific wear coefficients were evaluated from the experiments. Besides, Global Incremental Wear Model (GIWM) was used for prediction of wear behaviour of the steel.

Influence of Anions on the Corrosion Behavior of 9Cr-1Mo Ferritic Steel in Aqueous Media

Modified 9Cr-1Mo steel is used as steam generator material in prototype fast breeder reactors, owing to its good corrosion resistance, creep, and thermal conductivity characteristics. In the present study, the corrosion behavior of modified 9Cr-1Mo steel in aqueous environments containing various corrosive ions at potentiodynamic polarization conditions is reported. In chloride-exposed samples, randomly distributed dense shallow pits were seen, whereas mixed anions (, Cl−, and ) in fresh water resulted in deeper pits uniformly distributed on the entire surface. Laser Raman spectroscopy studies revealed the formation of Fe-oxides/hydroxides and Cr(VI) species, except in alkaline solution. A very thin film of only Fe-oxides (Fe3O4, γ-Fe2O3, and α-Fe2O3) was identified in an OH– ion-dominated solution. X-ray photoelectron spectroscopy results confirm the corroded surfaces comprised of Fe, Cr, and Mo-oxides of varying composition and enrichment of Mo-oxides in alkaline solution. Chloride ions present in the corroded layer influenced the pitting corrosion in neutral chloride and fresh water solutions. Field emission scanning electron microscopy images revealed abundant crystalline lepidocrocite with laminar morphology and doughnut-type magnetite together with pits on the samples exposed in neutral chloride and fresh water media.

Influence of seismic isolation on the earthquake response of internal components in a fast reactor

Nuclear energy is being considered as a frontrunner towards achieving the zero carbon goal. A sustainable use of nuclear energy is possible when a reactor design is deployed across regions with varying seismic hazard. Seismic isolation technology has been proven to reduce earthquake loads in the structures, and is being seen as a key towards the standardized reactor designs. This paper investigates the influence of seismic isolation on the response of a generic fast reactor (GFR) similar to the prototype fast breeder reactor (PFBR) currently under commissioning at Kalpakkam, India. This reactor comprises a head-supported main vessel that houses other reactor components (e.g., core, inner vessel). The major vessels of the reactor are partially filled with liquid sodium. In order to understand the influence of seismic isolation, simplified lumped mass stick models (LMSMs) for the GFR components were developed. These models were based on static analysis results obtained using corresponding finite element (FE) models. The component models were assembled to generate the LMSM and FE model for the GFR. The LMSM was verified through the comparison of dynamic properties obtained using the FE model, and through the comparison of response-history analysis results. These models considered fluid masses lumped at respective shells, thereby ignoring the fluid–structure interaction (FSI) effects. A representative system was analyzed with and without considering FSI effects. These effects did not influence acceleration response significantly. Fixed-base (FB) and isolated base (IB) configurations of the LMSM for the GFR were subjected to a series of response-history analyses. Seismic isolation led to a substantial reduction (e.g., up to 10 times) in the median floor accelerations, particularly for lower sliding friction and longer sliding period. The floor spectral ordinates for the IB configuration of GFR subjected to 1.00 g shaking were less than that for the FB configuration of GFR subjected to 0.25 g shaking at most periods, suggesting that the PFBR can be deployed at locations with seismic hazard much greater than that at Kalpakkam if suitable isolation systems are used.

Estimation Method for Residual Sodium Amount on Unloaded Dummy Fuel Assembly

This study has researched an estimation method for the amounts of residual sodium film and sodium lumps on dummy fuel pins in the Japanese prototype fast breeder reactor Monju by fundamental experiments and demonstration experiments. The residual sodium amounts on the pin surface were measured using three types of test specimens: (a) single pin, (b) 7-pin assembly, and (c) 169-pin assembly. The single pin and 7-pin assembly experiments revealed that the withdrawal speed of the pins and improvement of the sodium wetting drastically increased the residual sodium amounts. Furthermore, the 169-pin assembly experiments measured the practical amounts of the residual sodium in the Monju dummy fuel assembly and demonstrated sodium draining behavior through small gaps between the pins. The estimation method includes four models such as a viscosity flow model, Landau-Levich-Derjaguin (LLD) model, an empirical equation related to the Bretherton model, and a capillary force model in a tube. These calculation results were comparable to the residual sodium amounts obtained by the experiments. In the tests of improving sodium wetting, the amounts of residual sodium on the test specimen were close to 1.4 times larger than those of the thin sodium film estimated by the LLD model. The increased amount of residual sodium by improving the sodium wetting was explained by the ratio of the adhesion energy ( γ S O − γ S L ).

Numerical simulation of localization of a sub-assembly with failed fuel pins in the prototype fast breeder reactor

The early localization of a fuel subassembly with a failed (wet rupture) fuel pin is very important in reactors to limit the associated radiological and operational consequences. This requires a fast and reliable system for failure detection and their localization in the core. In the Prototype Fast Breeder Reactor, the system specially designed for this purpose is Failed Fuel Location Modules (FFLM) housed in the control plug region. It identifies a failed sub-assembly by detecting the presence of delayed neutrons in the sodium from a failed sub-assembly. During the commissioning phase of PFBR, it is mandatory to demonstrate the FFLM effectiveness. The paper highlights the engineering and physics design aspects of FFLM and the integrated simulation towards its function demonstration with a source assembly containing a perforated metallic fuel pin. This test pin mimics a MOX pin of 1 cm2 of geometrical defect area. At 10% power and 20% sodium flow rate, the counts rate in the BCCs of FFLM system range from 75 cps to 145 cps depending upon the position of DN source assembly. The model developed for the counts simulation is applicable to both metal and MOX pins with proper values of k-factor and escape coefficient.

Evaluation of Control Rod Interaction Measurement in the Prototype Fast Reactor Monju

For evaluating control rod (CR) interaction, CR worth values with and without the insertion of interaction CRs were measured in six CR interaction patterns by the continuous period method and the compensation method in the prototype fast breeder reactor Monju. The measured CR worth was evaluated in detail by conducting corrections, such as the thermal contraction of CR drive shafts, and compared with calculations to examine the reliability. The calculation-to-experiment values (C/Es) of examined CR worths with and without the interaction were 0.96 to 0.98 ± 0.03. The values of the measured CR interactions were obtained from those examined CR worth measurements as a change in target CR worth by the insertion of interaction CRs: 5.0 ± 0.6% to 10.1 ± 0.1%. The C/Es of the interaction in those cases were 0.78 to 0.93 within 5% to 27% uncertainty. Through this study, the consistency and reliability of the measurements and the calculations of the CR interaction were confirmed, and the measurements were judged useful.

DESIGN AND DEVELOPMENT OF HMI IN SIMULATOR FOR FRESH FUEL HANDLING SYSTEM

KALBRSIM operator training simulator is used for training Prototype Fast Breeder Reactor (PFBR) nuclear power plant operators. In simulator, using virtual panel HMI which is an exact replica of actual hardware panels, all the essential fresh fuel operations can be done. Visualization power is also provided by 3D models which aid in quicker understanding of the complex process involved in Fresh Fuel Handling System (FFHS). Operation using Virtual Panel is an excellent tool for system developers to test the logic and process models of FFHS in KALBRSIM before integration with hardware panels. Also, operator training through virtual panel HMI can enable operator to monitor and control FFHS by running multiple screens at once at a central location. Teaching operators their effect of actions using 3D models is an add on with virtual panels which help the operator visualize the system accurately. This paper discuss the design and development of virtual panel HMI and 3D models of fresh fuel handling system used in full scope replica PFBR operator training simulator. Advantage of using HMI in simulator is also covered.

Effect of thickness on mechanical properties of modified 9Cr 1Mo steel welds made by narrow gap hot wire gas tungsten arc welding process

Hot wire Gas Tungsten Arc Welding (GTAW) process is selected for fabrication of thick section modified 9Cr 1Mo steel components of Steam Generators (SGs) for 500MWe Prototype Fast Breeder Reactor (PFBR) at Kalpakkam, India. During the welding procedure qualification for fabrication of SGs, the 30 mm and 62 mm thick modified 9Cr 1Mo steel welds were subjected to chemical analysis, visual examination, liquid penetrant examination, radiography examination, micro-hardness tests, longitudinal and transverse tensile tests (at ambient temperature and 525 °C), bend tests, drop weight tests and impact tests (at 0 °C and +18 °C). Literature reveals both positive and negative view on the welding of thick section steel components. Limited information is available in open literature on the effect of thickness on material properties of modified 9Cr 1Mo steel welds made by hot wire GTAW method. In view of the above, authors attempted to systematically investigate the influence of thickness on mechanical properties of modified 9Cr 1Mo steel welds made by hot wire GTAW process. In this paper, the chemical compositions, hardness, tensile strength, bending strength and impact properties of 30 mm and 62 mm thick welds produced by narrow gap hot wire GTAW process are characterized and compared. It is found that the mechanical properties of 30 mm and 62 mm thick welds are comparable except the impact properties at 0 °C. The impact strength at 0 °C for 62 mm thick weld was found inferior in comparison with 30 mm thick weld.

Evaluation of Sodium Radioactivity in the Primary System of the Prototype Fast Reactor Monju

In this study, the sodium radioactivity of 24Na and 22Na in the primary system measured in the prototype fast breeder reactor Monju was evaluated, and the reliability of the measurements and calculations was examined. The calculated-to-experiment values and their uncertainties for 24Na and 22Na radioactivities were 0.97 to 1.07 and 8.1% to 11.0% and 1.03 to 1.16 and 23.3% to 24.1%, respectively, using the JENDL-4.0 nuclear data library. The 22Na radioactivity calculated with ENDF/B-VIII.0 was larger by 40% than those calculated with JENDL-4.0 and JEFF-3.3 due to the 23Na(n,2n) cross-section discrepancy. The importance of the 22Na neutron capture effect is also confirmed herein for the accurate evaluation of the 22Na radioactivity. The experimental data were judged to be useful for validating the calculation method for improving the reliability of future designs of sodium-cooled fast reactors.

Implementation of optimal oil flushing methodology for nuclear steam turbo generator oil circuits

The first turning gear operation is regarded as a significant milestone in the overall activities of Turbine Generator system commissioning. To accomplish this critical goal, the lube oil circuit must be clean and free of all type of contaminants. The purpose of this paper is to elaborate on the various flushing tactics used in the Prototype Fast Breeder Reactor 500 MW Turbo Generator (TG) oil systems, to understand the factors that influence oil circuit flushing, the difficulties encountered during flushing, and to optimise the flushing process of TG oil circuits. Also A simple method was used in this paper to determine the minimum amount of oil required to begin flushing and the make-up requirements of lube oil at various stages of flushing. The percentage of oil passing through the cooler and cooler bypass line during the thermal shocking process was also determined using an energy balance approach. Based on TG oil circuit commissioning experience, an optimised methodology for completing the TG oil circuits flushing exercise is also proposed. This study demonstrates that the cooling and heating times of lube oil can be controlled during oil flushing, resulting in an increase in the number of thermal shocks per day and a significant reduction in flushing time.Article HighlightsThis paper is the result of flushing experiences with oil circuits of a 500 MW Nuclear power plant. This can be used to minimize oil quantity and the time flushing thereby lot of man hours and cost of flushing could be saved during flushing of turbine oil circuits.The minimum oil required for turbine oil circuit flushing is not readily available. The method outlined here can be used to determine the minimum amount of oil required for each stage of oil circuit flushing.This paper describes an energy balance methodology that can be used to increase the number of thermal shocks per day during flushing, thereby significantly reducing flushing time.

Application of external flood probabilistic safety assessment methodology to Prototype Fast Breeder Reactor

ABSTRACT This paper presents the methodology adopted, accident sequence analysis and salient results from the external flood probabilistic safety analysis exercise for Prototype Fast Breeder Reactor. The major steps performed are the probabilistic hazard estimation, fragility computation, system reliability modelling and risk quantification. The hazard analysis has been performed for rainfall, storm surge and tsunami as typically associated with a coastal site, from which correlation function relating flood level to the annual exceedance frequency is estimated. The fragility analysis of different safety systems is estimated by considering submergence mode of failure using step fragility functions. The secondary effects of flood such as debris impact, load falling and flood-induced sodium fires are not considered. The human reliability analysis of manual operations necessary for plant safety is done for the control room as well as onsite actions. The challenges encountered during the analysis, salient results and uncertainties involved are presented with inferences. Abbreviations: AHX: air heat exchanger; CDF: core damage frequency; EFPSA: external flood probabilistic safety assessment; EL: elevation levels; EOP: emergency operation procedure; ET: event tree; FT: fault tree; HRA: human reliability analysis; IE: initiating events; NPP: nuclear power plant; OGDHRS: operation grade decay heat removal system; PFBR: prototype fast breeder reactor; POT: peak over threshold; PSA: probabilistic safety assessment; PTHA: probabilistic tsunami hazard assessment; RY: reactor year; RCB: reactor containment building; RP: return period; SGDHRS: safety grade decay heat removal system; SSC: structures, systems and components

Computational Analysis of Thermal Striping in Primary Sodium System of Liquid Metal Fast Breeder Reactor Using Finite Volume Method

Thermal striping is associated with random fluctuations of temperature that occur at the nonisothermal jet stream interface or across thermally stratified fluid layers due to the high heat transfer coefficient of liquid sodium flow. The temperature fluctuations in the jet mixing or stratified layer regions are transmitted to the adjoining structures after minimal attenuation in a Liquid Metal Fast Breeder Reactor (LMFBR). In turn, the adjoining structure may experience high cycle fatigue and catastrophic failure caused by crack propagation. Investigations have been carried out in detail numerically, and frequency and amplitude of temperature fluctuations in 500-MW(electric) pool-type fast reactor [Prototype Fast Breeder Reactor (PFBR)] structures for practical applications have been observed. The investigations consist of numerical simulations at two levels. First, a published benchmark experiment is analyzed, and then, a suitable computational fluid dynamics (CFD) model is identified for simulating the thermal striping phenomenon numerically. After that, detailed flow and temperature fluctuations are predicted in the reactor structures by analysis carried out based on the CFD model. The values of the temperature fluctuations predicted are found to be within acceptable limits, as required by structural mechanics considerations in the study.

Selection of smooth motion profile for a tube locator module of an inspection device

<span lang="EN-US">The Prototype Fast Breeder Reactor steam generators inspection system has seven modules. In this, tube locator module is a planar serial two-link robotic arm, which is used to place the eddy current probe above the steam generators tube hole in the tube sheet region. The trajectory planning of the two-link robotic arm is one of the important tasks, so the peak velocity, peak acceleration, peak jerk of various motion profiles for a given distance has to be selected properly for smooth motion and to avoid actuator saturation. The fifth-order polynomial gives lower acceleration and velocity than the jerk-limited S-curve. In this paper, the comparison of peak values of kinematic variables (velocity, acceleration, and jerk) for different motion profiles has been presented.</span>

Subset simulation for optimal sensors positioning based on value of information

Greedy and non-greedy optimization methods have been proposed for maximizing the Value of Information (VoI) for equipment health monitoring by optimal sensors positioning. These methods provide good solutions, but still with limitations and challenges: greedy optimization does not guarantee to find the optimal solution, due to the non-submodularity of the VoI; non-greedy optimization does not suffer from the non-submodularity of the VoI but requires computationally expensive and tedious simulations to find the optimal solution. In this work, the Subset Simulation (SS) method is originally proposed to address these limitations and challenges. A real case study is considered concerning the condition monitoring of a Steam Generator (SG) of a Prototype Fast Breeder Reactor (PFBR). Results show that SS, even if initialized with a small number of Monte Carlo samples, is capable of finding the optimal set of sensors positions in a very short computational time and is insensitive to the non-submodularity of VoI.

Approach towards qualification of TCP/IP network components of PFBR

Distributed control system architecture is adopted for I&C systems of Prototype Fast Breeder Reactor, where the geographically distributed control systems are connected to centralized servers & display stations via switched Ethernet networks. TCP/IP communication plays a significant role in the successful operations of this architecture. The communication tasks at control nodes are taken care by TCP/IP offload modules; local area switched network is realized using layer-2/3 switches, which are finally connected to network interfaces of centralized servers & display stations. Safety, security, reliability, and fault tolerance of control systems used for safety-related applications of nuclear power plants is ensured by indigenous design and qualification as per guidelines laid down by regulatory authorities. In the case of commercially available components, appropriate suitability analysis is required for getting the operation clearances from regulatory authorities. This paper details the proposed approach for the suitability analysis of TCP/IP communication nodes, including control systems at the field, network switches, and servers/display stations. Development of test platform using commercially available tools and diagnostics software engineered for control nodes/display stations are described. Each TCP link behavior with impaired packets and multiple traffic loads is described, followed by benchmarking of the network switch's routing characteristics and security features.

Influence of long term Sodium Exposure on the Corrosion and Tensile Properties of AISI Type 316LN stainless steel and modified 9Cr-1Mo steel

AISI Type 316LN stainless steel (SS) and modified 9Cr-1Mo (P91) steel are used in the intermediate heat exchanger and steam generator (SG), respectively of prototype fast breeder reactor (PFBR). This paper presents the effect of long-term sodium exposure (50000 h) on the corrosion and tensile properties of 316LN SS and P91 steel. The results showed that the prolonged sodium exposure of 316LN SS can lead to selective leaching of alloying elements (depletion of Cr, Ni, and Mo), enrichment of Fe, formation of ferrite layer and carburized zones with increased hardness at the surface (272 HV), and a reduction in ductility by nearly 40 %. Double-loop electrochemical potentiokinetic reactivation (DLEPR) results of sodium exposed 316LN SS showed a 25% higher degree of sensitization (DOS) as compared to thermally aged one. The microstructural examination of sodium exposed 316LN SS revealed a ditch structure near the surface and carbide precipitates along the grain boundaries indicating the diffusion of carbon through grain boundaries and surface carburization. XRD studies of sodium exposed 316LN SS revealed predominant phases of ferrite and Cr23C6 phases, which further corroborated the results of DLEPR, microstructure, and hardness data. The sodium exposed P91 steel showed no significant microstructural, microchemical changes and tensile properties after sodium exposure. The carbon depth profiles carried out using SIMS in sodium exposed 316LN SS and P91 steel showed a surface carbon concentration of 0.85 and 0.32 wt.%, respectively. Fractography of sodium exposed 316LN SS showed intergranular fractures whereas the fractographs of P91steel showed fibrous dimple and cleavage type of fractures indicating the growth of carbides. The present results confirm that P91 steel can sustain the reactor life in flowing sodium containing 25 ppm carbon and less than 2 ppm oxygen.

Evaluation of Fuel Reactivity Worth Measurement in the Prototype Fast Reactor Monju

In the prototype fast breeder reactor Monju, fuel reactivity worth was measured at six positions as the reactivity corresponding to the differences of critical control rod positions between cores with and without a dummy fuel subassembly. In this paper, the measurements are evaluated in detail, and their reliability and usefulness as validation data for fast reactor neutronics design methodologies are examined through a comparison with calculations. Calculated-to-experiment values and their uncertainty of fuel reactivity worth were 0.97 to 1.02 and 4% to 6%. Through this study, the consistency and reliability of the measurements and calculations were confirmed.