Development of mathematical model for optimization of sodium leak collection tray

Sodium leaks and resultant fire containment play an important role in the safe operation of a fast breeder reactor. Leak collection tray (LCT) is a passive device which is used to collect the highly reactive liquid sodium in the case of an accidental leakage. The consequences of sodium fire are mitigated by oxygen starvation in the vessel which collects the liquid sodium after leakage. The current paper deals with the optimization of the LCT geometry based on the hydrodynamic characteristics of the leaked liquid sodium. Isothermal numerical simulations have been performed to understand the interfacial dynamics of the hot liquid sodium flow in the top tray part and the variation of sodium draining rate into the holdup vessel for various drainpipe diameters and leak rates. Since the numerical simulations involve very high computational effort, an equivalent semi-analytical sloshing/draining model has also been developed which emulates the flow process in the LCT. The predictions of transient mass distributions in the top part and in the holdup vessel for the semi-analytical model are in close match with the results obtained from the detailed numerical study. The results reveal critical geometric parameters at which the un-burnt sodium collected in the LCT will be maximum.     

The detection of sodium vapor bubble collapse in a liquid metal fast breeder reactor

Sodium boiling detection utilizing the sound pressure emanated during the collapse of a sodium vapor bubble in a subcooled media is discussed in terms of the sound characteristic, the reactor ambient noise background, transmission loss considerations and performance criteria. Data obtained in several loss of flow experiments on Fast Test Reactor Fuel Elements indicate that the collapse of the sodium vapor bubble depends on the presence of a subcooled structure or sodium. The collapse pressure pulse was observed in all cases to be on the order of a kPa, indicating a soft type of cavitational collapse. Spectral examination of the pulses indicates the response function of the test structure and geometry is important. The sodium boiling observed in these experiments was observed to occur at a low (<50°C) liquid superheat with the rate of occurrence of sodium vapor bubble collapse in the 3 to 30 Hz range. Reactor ambient noise data were found to be due to machinery induced vibrations, flow induced vibrations, and flow noise. These data were further found to be weakly stationary enhancing the possibility of acoustic surveillance of an operating Liquid Metal Fast Breeder Reactor. Based on these noise characteristics and extrapolating the noise measurements from the Fast Flux Test Facility Pump (FFTP), one would expect a signal to noise ratio of up to 20 dB in the absence of transmission loss. The requirement of a low false alarm probability is shown to necessitate post detection analysis of the collapse event sequence and the cross correlation with the second derivative of the neutronic boiling detection signal. Sodium boiling detection using the sounds emitted during sodium vapor bubble collapse are shown to be feasible but a need for in-reactor demonstration is necessary.     

Theoretical and experimental thermal-hydraulic analysis of a new sodium cooled fast breeder reactor core

This paper presents the experimental and theoretical results of the thermal-hydraulic design of a new fast breeder reactor core concept. The main feature of this concept is the omission of fuel element cans.The hydraulic function of these fuel element cans is substituted by a winding flow path through the radial blanket and a ring chamber without tubes.A computer code based on the quasi-continuum-theory and especially adapted to the features of the new core concept is developed for theoretical investigations. The pressure drop of the rod bundles is specified by a resistance tensor.The experimental investigations are realized in a test facility, where sodium is simulated by water. Pressures and velocities are measured.Theoretical and experimental results show good agreement. The aim of flattening of the coolant outlet temperature distribution can be reached with satisfying accuracy.     

Roughening characteristics and choices for the gas-cooled fast breeder reactor

The thermohydraulic performance of several types of rough surfaces proposed for use in the gas-cooled fast breeder reactor has been investigated experimentally at the Swiss Federal Institute for Reactor Research. Based on the tests, the most suitable roughness design has been defined. In addition to the thermohydraulic performance requirements, some other technological and operational criteria should be used for the final choice of roughness. There is not sufficient information on the different roughening methods to enable any decision to date, but when the new complex thermohydraulic performance criterion is considered, additional requirements become relatively more important.     

Aseismic study on the reactor vessel of a fast breeder reactor

This paper presents the results of a seismic study using an scale steel model and a scale plastic model which simulate the reactor vessel of a loop type Fast Breeder Reactor (FBR). The main purposes of this study are to confirm the structure/liquid interaction and the aseismic safety of the reactor vessel experimentally, and also to verify the validity of the seismic response analysis model of the prototype vessel.The characteristics of coupled vibration between the structure and liquid were clarified, and the approach of calculation model to aseismic design was worked out. And, the dip plate and other core internals were found to be effective in suppressing the liquid free surface oscillation.     

Status of gas-cooled fast breeder reactor programs

Most gas-cooled fast breeder reactor (GCFR) programs in Europe and the US are now coordinated and focused on a 300 MW(e) GCFR demonstration plant program. Except for venting and artificial surface roughening, GCFR fuel is similar to liquid metal fast breeder reactor (LMFBR) fuel and operates under nearly identical conditions. The primary helium system is integrated within a PCRV like all large gas-cooled thermal reactors, with three main loops and three auxiliary loops. Design and safety studies and various experiments, including heat transfer, irradiation, and critical experiments, indicate that most feasibility questions have been answered and a demonstration plant could be in operation within 12 years. This could be followed in the mid-1990s by a large-size GCFR with a doubling time of about 10 years fueled by (UO₂---PuO₂) and producing either ²³³U in thorium blankets as fuel for advanced converters or plutonium in depleted uranium blankets.     

Random thermal fatigue in fast breeder reactor: a narrow-band spectrum

Thermal fatigue crack growth in a fast breeder reactor is theoretically investigated with the aid of probabilistic fracture mechanics (PFM) under the conditions that (i) the temperature variation is a narrow-band stationary process and (ii) the crack grows owing only to the peak stress variation. First, a statistical property of residual life of the component with single crack is derived in an analytical form with the aid of an extended Markov approximation method, which is an efficient mathematical technique in PFM. Next, discussion is carried out on the generalization of the primitive model to the case with plural cracks, where a stress relaxation factor is introduced to express a stress intensity factor of each crack. Finally, a numerical example is shown to examine the quantitative behavior of the component's residual life, and sensitivity analysis is performed with respect to some model parameters.     

Safety features of the gas-cooled fast breeder reactor

The safety features of the gas-cooled fast breeder reactor (GCFR) are described in the context of the 300-MW(e) demonstration plant design. They are of two general types, inherent and design-related. The inherent features are principally associated with the helium coolant and the nuclear coefficients. Design-related features influencing safety include shutdown systems, residual heat removal systems, method of core support, and the prestressed concrete reactor vessel (PCRV). This paper discusses the safety-related aspects of each of these. Recently completed residual heat removal system reliability studies are also discussed. The probability of residual heat removal system failure in the GCFR is found to be lower than that described for light water reactors. The safety characteristics of larger plants are examined, and increases in size are found to improve GCFR safety margins.     

Reactivity insertion mechanisms in the gas-cooled fast breeder reactor

This report summarizes an analysis of reactivity insertion mechanisms in the gas-cooled fast breeder reactor (GCFR). Inherent reactivity feedback mechanisms are identified and their effects on reactor start-up, during normal operation, and on anticipated and postulated transients are analyzed. Potential sources of accidental reactivity insertions and the resulting transients are investigated, including potential reactivity effects due to cladding and fuel melting. All nuclear calculations are based on the ENDF-B, Version 3, cross-section file. It is concluded from these analyses that the GCFR is an inherently stable reactor during start-up and normal operation. Potential accidental reactivity insertions are mild, and in each case the reactor can be controlled with a substantial margin for fuel melting or cladding damage. In low-probability accident sequences which lead to core melting, there are potential fuel motion mechanisms which can mitigate reactivity effects and accident consequences.     

Effect of turbulent natural convection on sodium pool combustion in the steam generator building of a fast breeder reactor

A computational model is proposed to simulate sodium pool combustion considering the effect of turbulent natural convection in a vented enclosure of the steam generator building (SGB) of a fast breeder reactor. The model is validated by comparing the simulated results with the experimental results available in literature for sodium pool combustion in a CSTF vessel. After validation, the effects of vents and the location of the pool on the burning rate of sodium and the associated heat transfer to the walls are studied in an enclosure comparable in size to one floor of the steam generator building. In the presence of ventilation, the burning rate of sodium increases, but the total heat transferred to the walls of the enclosure is reduced. It is also found that the burning rate of sodium pool and the heat transfer to the walls of the enclosures vary significantly with the location of sodium pool.     

Shear-bending buckling analyses of fast breeder reactor main vessels

The purpose of this paper is to describe the computation results and the knowledge of the buckling analysis strategy. Since fast breeder reactor main vessels are thin shell structures, plastic shear-bending buckling is one of the most important problems. To clarify the buckling behaviour, we carried out many tests and numerical calculations. Based on the experience of those buckling analyses, available elements, mesh division, modelling of shape imperfections etc. are described. These results show that the numerical analysis can be a useful tool for evaluating buckling phenomena.     

Optimization of material and production to develop fluoroelastomer inflatable seals for sodium cooled fast breeder reactor

The feasibility of producing thin-walled fluoroelastomer profiles under continuous, atmospheric-pressure vulcanization conditions in air has been demonstrated by successful manufacture of ∼2 m diameter test inflatable seals for the 500 MWe, Prototype Fast Breeder Reactor (PFBR) using a 50/50 blend formulation of Viton^® GBL-200S/600S based on advanced polymer architecture (APA). A commercial cold feed screw extruder with 90 mm diameter screw was used along with continuous cure by microwave (2.45 GHz) and hot air heating (190 °C) at a line speed of 1 m/min to produce the seals. The blend formulation promises significant improvement in the performance and safety of the seals. This article depicts the relevant characteristics of the original inflatable seal compound that was used as reference to achieve the objectives through synchronized optimization of material and production technologies. The production trials are outlined and the blend formulation used with minor factory modifications to produce the test seals is reported. Progressive refinements of the original, Viton^® A-401C based compound to the blend formulation is presented along with an assessment of potential performance gains. Possible uses of the reported formulation and production technique for other large diameter, high temperature seals of PFBR are indicated along with the envisaged activities en-route the production of perfected reactor inflatable seals.     

Thermal hydraulic investigations of intermediate heat exchanger in a pool-type fast breeder reactor

Intermediate heat exchanger (IHX) in a pool-type liquid metal cooled fast breeder reactor is an important heat exchanging component as it forms an intermediate boundary between the radioactive primary sodium in the pool and the non-radioactive secondary sodium in the steam generator (SG). The thermal loads during steady state and transient conditions impose thermal stresses on the heat exchanger tubes and on the shells which hold the tube bundle. Estimation of these thermal loads and achieving uniform temperature distribution in the tubes and shells by having uniform flow distributions are the major tasks of thermal hydraulic investigations of IHX. Through multi-dimensional thermal hydraulic investigations performed using commercially available computer codes such as PHOENICS, the flow and temperature distributions in the tubes and shells and in its secondary sodium inlet and outlet headers are obtained with and with out provisions of flow distribution devices. The effectiveness of these devices in achieving acceptably uniform flow and temperature distributions has been assessed and thermal loads on the tubes and shells for thermo mechanical analysis of the IHX have been defined. The predictions of the computational studies have been validated against simulated experiments.     

Heat transfer from a totally blocked fuel subassembly of a liquid metal fast breeder reactor: II. Numerical simulation

Fast breeder nuclear reactors used for power generation, have fuel subassemblies in the form of rod bundles enclosed inside tall hexagonal cavities. Each subassembly can be considered as a porous medium with internal heat generation. A three-dimensional analysis is carried out here to estimate the heat transfer due to natural convection, in such an anisotropic, partially heat generating porous medium, which corresponds to the typical case of blocked flow in a fuel subassembly inside the reactor core. Using the finite volume technique, the temperatures at various locations inside hexagonal cavity are obtained. The simulations by the three-dimensional code developed are compared with the results of experiments [Suresh, Ch.S.Y., Sateesh, G., Das, Sarit K., Venkateshan, S.P., Rajan, M., 2004. Heat transfer from a totally blocked fuel subassembly of a liquid metalfast breeder reactor. Part 1: Experimental investigation. Nucl. Eng. Design, present issue] conducted using liquid sodium as the heat transfer fluid. Further, the code is used to predict the maximum temperature in typical liquid metal fast breeder reactors to find the power level where the liquid sodium starts boiling. It helps to decide the power level for initiation of monitoring the temperature for the purpose of reactor control.     

Development of components for the gas-cooled fast breeder reactor program

The gas-cooled fast breeder reactor (GCFR) component development program is based on an extension of high temperature gas-cooled reactor (HTGR) component technology; therefore, the GCFR development program is addressed primarily to components which differ in design and requirements from HTGR components. The principal differences in primary system components are due to the increase in helium coolant pressure level, which benefits system size and efficiency in the GCFR, and differences in the reactor internals and fuel handling systems due to the use of the compact metal-clad core.The purpose of this paper is to present an overview of the principal component design differences between the GCFR and HTGR and the consequent influences of these differences on GCFR component development programs. Development program plans are discussed and include those for the prestressed concrete reactor vessel (PCRV), the main helium circulator and its supporting systems, the steam generators, the reactor thermal shielding, and the fuel handling system. Facility requirements to support these development programs are also discussed. Studies to date show that GCFR component development continues to appear to be incremental in nature, and the required tests are adaptations of related HTGR test programs.     

Static and stability analysis of the inner vessel of a fast breeder reactor

Liquid metal fast breeder nuclear reactors demand the usage of large sized thin shells for their reactor vessel components due to low operating pressure and high thermal load. Buckling is a very important aspect in the design of these vessels. In this article, analysis of the inner vessel of a typical 500 MWe fast breeder reactor is presented. Here, two different geometric configurations of the inner vessel are considered. One configuration is with the conical step joining the upper and the lower cylindrical portions, and the other is with the toroidal bottom joining the upper cylindrical part. The buckling strength of the vessel for both configurations are calculated and compared. Also, the effects of thermal load, initial geometric imperfection, geometric nonlinearity, etc. are investigated. The finite element method is used for analysis.     

Inservice inspection of the reactor block of sodium-cooled fast breeder reactors

By using sodium as coolant special boundary conditions result for the inservice inspection (ISI) of fast breeder reactors. For that reason in general it is not successful applying the methods and equipment proved for the 151 of light water reactors.This report presents inspection methods and equipment developed for the ISI of the reactor block of sodium cooled fast breeder reactors. The survey takes into account the state of the art as well as some R&D-work at home and abroad. Entering into particulars the methods and equipment used for leak monitoring, the inspection of the reactor vessel wall, the inspection' of reactor internals above and below the sodium level, monitoring of structure home noise and the measurement of the gap between the reactor vessel and the guard vessel are described.     

Design of an indirect-cycle fast breeder reactor cooled by supercritical steam

The neutronic feasibility of a large-sized FBR cooled by supercritical steam is assessed for finding the way to reduce the costs of FBR plants. A negative coolant void reactivity is realized without much deterioration in breeding capability, by our novel concept of inserting thin zirconium-hydride layers between the seed and blanket of the radially heterogeneous core. The gross electric power is 1040 MWe. The estimated equilibrium compound system doubling time is 16 years. The discharge burnup is 78.7 GWd/T and the refueling period is 1 year with a 77% load factor. Compared with the conventional steam cooled FBR, the pumping power fraction is reduced due to the high coolant density of supercritical water. Loeffler boilers for generating steam are not neccessary. The reactor system is simplified and similar to a PWR. The thermal efficiency is 39.6%, improved 15% relatively from PWR's. The pressure vessel is 32.5 cm thick.     

Hydraulic design of sparger arrangement for long guide tubes of a fast breeder reactor

In liquid metal cooled fast reactors, the core is submerged in sodium pool by ∼5 m below sodium free surface. This necessitates the control and shutdown of reactor be achieved by long overhanging mechanisms housed inside a control plug. These mechanisms are protected by porous guide tubes with a sparger type arrangement for the sodium flow through them. Comprehensive knowledge of flow distribution of sodium through these guide tubes is essential to assess the risks of flow induced vibration of thin thermowell tubes that pass close to these shroud tubes and entrainment of cover gas due to high free surface velocities. Three dimensional hydraulic analysis of single isolated shroud tube and integrated assembly of shroud tubes have been carried out using CFD tools to acquire this knowledge. The predictions of the CFD models have been validated against experimental predictions. These studies have provided important information regarding critical design parameters. Size of holes in the shroud tube, location of holes in the control plug shell and arrangement for breaking sodium jets emanating from shroud tubes have been optimized to reduce free surface velocity.