Liquid Metal Cooled Fast Breeder Reactors Research Articles

A chemical and radiological test [fast breeder reactor decommissioning

Decommissioning the the 250MW Prototype Fast Breeder Reactor (PFR) at Dounreay is the heart of the site restoration programme, but the reactor presents particular problems. In liquid-metal cooled fast breeder reactors (LMFBRs) the primary and secondary coolants are highly reactive alkali metals (sodium and potassium, NaK). The alkali metals react very strongly with water and in the presence of air liquid sodium can ignite, so it must first be converted to a stable chemical product that is then safely converted into effluent streams for final treatment. In 1995 UKAEA awarded NNC the contract to design, build, and commission a sodium disposal plant. The £17m process removes the sodium from the reactor vessel and other vessels at a rate of 2.5t per day and reacts in a special vessel with sodium hydroxide. The resulting solution is neutralised with hydrochloric acid to produce a saline solution, which is processed to remove the caesium content. The next problem is to deal with the residues, which is discussed in this paper.

Transient evolution of inter vessel gap pressure due to relative thermal expansion between two vessels

In a typical liquid metal cooled fast breeder reactor (LMFBR), a cylindrical sodium filled main vessel, which carries the internals such as reactor core, pumps, intermediate heat exchangers etc. is surrounded by another vessel called safety vessel. The inter vessel gap is filled with nitrogen. During a thermal transient in the pool sodium, because of the relative delay involved in the thermal diffusion between MV and SV, they are subjected to relative thermal expansion or contraction between them. This in turn results in pressurisation and depressurisation of inter vessel gap nitrogen respectively. In order to obtain the external pressurization for the buckling design of MV, transient thermal models for obtaining the evolutions of MV, SV and inter gap nitrogen temperatures and hence their relative thermal expansion and inter vessel gap pressure have been developed. This paper gives the details of the mathematical model, assumptions made in the calculation and the results of the analysis.

Fabrication and Thermal Conductivity of Boron Carbide/Copper Cermet

Studies on fabrication and thermal conductivity of B4C/Cu cermet were made to obtain high performance neutron absorber materials for Liquid Metal-cooled Fast Breeder Reactor (LMFBR). A mixed powder of B4C and Cu was mechanically blended at high speed thereby a coating layer of Cu was formed on the surface of B4C powder. Then the B4C powder with Cu coating was hot pressed at temperatures from 950 to 1,050°C to form a B4C cermet. A high density B4C/CU cermet with 70 vol% of B4C and relative density higher than 90% was successfully fabricated. In spite of the low volume fraction of Cu. the B4C/Cu cermet exhibited high thermal conductivity which originated from the existence of continuous metallic phase Cu in B4C/Cu cermet.

Study of rationalized safety design based on the seismic PSA for an LMFBR

Seismic PSA was carried out for a typical liquid metal cooled fast breeder reactor (LMFBR) in order to study the rationalized seismic design, maintaining and/or improving safety during seismic event. The seismic sequence quantification identifies the dominant structures, systems and components (SSCs) to the seismic core damage frequency (CDF). The sensitivity analyses by reducing or increasing the seismic capacity for SSCs are used to examine the optimized seismic design in view of safety and economical aspects. The LMFBR-specific risk-significant SSCs are reactor coolant boundary, decay heat removal coolant path and reactor control rod, which are different from those of light water reactors (LWRs). The electrical power supply system has a minor contribution to the seismic CDF. The sensitivity study shows that passive safety features of LMFBRs are important to maintain and/or enhance seismic capacity. The passive safety includes the decay heat removal capability via natural circulation and safety measures without depending on the support systems such as alternating current (AC) electrical power, for example. On the course of seismic sequence quantification, a methodology to evaluate the probability of seismic-induced multiple failure has been developed and applied to the decay heat removal function. The results suggest the multiplicity of the triply redundant system is to be considered for the significant components such as the decay heat removal path when one considers the difference in the seismic response.

History and perspective of fast breeder reactor development in Japan

Research and development on fast breeder reactors (FBRs) started in Japan in the mid-1960s, leading to the experimental FBR JOYO which went critical in 1977. It was followed by the prototype FBR MONJU which started operating in 1994. A 600 MWe loop-type demonstration liquid-metal-cooled FBR (LMFBR) is now under study, especially emphasizing its safety and economy. A demonstration plant could be in operation several years after 2010. Japan expects a significant fraction of world nuclear energy mix to be provided by FBRs by the end of the 21st century. This paper has detailed descriptions of the activities of the various Japanese organizations involved in FBR development, as well as program funding and policy considerations.

The influence of dynamic strain ageing on stress response and strain-life relationship in low cycle fatigue of 316L(N) stainless steel

It has been recognized that the essential prerequisite for accurate prediction of low cycle fatigue life (LCF) at elevated temperatures is the determination of rate controlling time dependent damage process that influences the cyclic deformation and fracture behavior of alloys. At intermediate temperatures (0.25 < T/T{sub m} < 0.55, where T{sub m} is melting point in Kelvin), where the effects of time dependent processes such as creep and oxidation are found to be minimal, the drastic reduction in low cycle fatigue life observed with increasing temperature and decreasing strain rate in alloys such as type 04 stainless steel, Nimonic PE 16 superalloy and Haynes 188 superalloy has been ascribed primarily to the deleterious effects of dynamic strain ageing (DSA). These alloys displayed anomalous cyclic hardening and an increase of maximum cyclic stress with increasing temperature or decreasing strain rate due to DSA. The nitrogen alloyed 316L stainless steel is currently used for a variety of components in the liquid metal cooled fast breeder reactor (LMFBR). Although the temperature range of application of 316L(N) in LMFBRs encompass the domain in which DSA is active, to-date there has been little information available on the effects of DSA on cyclic behavior of 316L(N)more » stainless steel (SS). The present investigation is therefore focused on assessing the effects of DSA on LCF life and cyclic stress response of 316L(N) SS. The influence of DSA on cyclic strain-life relationship has also been examined in the regime where other time dependent mechanisms such as creep and oxidation are virtually absent.« less

Response Analysis of Electromagnetic Flowmeter

In liquid-metal-cooled fast breeder reactors (LMFBRs), electromagnetic flowmeters (EMFs) have been extensively used to measure the coolant flow rate. Because the coolant flow rate is one of the most important parameters, a high accuracy and a fast response are required for the flow rate measurement. However, it was thought that the response might become slower when the pipe diameter of the EMF was increased. Therefore, a quantitative evaluation of the response was needed. To evaluate the response time of EMFs, an equation of the transient response was derived based on the realistic approximation that the EMF pipe is made of nonconductive material. The response is expressed as a function of the reciprocal of the square of the pipe radius a and of the length L of the external magnetic field along the pipe axis. However, when the aspect ratio L/2a is larger than two, the length of the external magnetic field has an almost insignificant effect on the response, and the response time increases with increasing a{sup 2}. The transient response can be calculated with an uncertainty of less than a few percent. A first-order approximation of the derived equation is given by the first lag term with the timemore » constant of {mu}{sigma}a{sup 2}/3.83{sup 2} with permeability {mu} and conductivity {sigma} of the coolant. Even though the EMF has a diameter as large as 30 in., the response time is 45 ms and sufficiently fast compared with other sensors used in LMFBRs.« less

Benchmark exercise for multi-dimensional thermohydraulic analysis codes

A benchmark exercise for multi-dimsional thermohydraulic codes was carried out related to natural convection decay heat removal in liquid metal-cooled fast breeder reactors. A total of twelve computational methods were applied to the benchmark problem which simulated mixed forced and buoyancy driven penetration flow and thermal stratification phenomena. The applicability of turbulence models and higher order schemes of convection terms, was examined, and a combined method incorporating a higher order scheme and a turbulence model was found to be highly effective among the group of finite difference methods. The importance of turbulence models was also recognized for the finite element method. Development of a turbulence model applicable to the mixed convection flow regime was also discussed.

Experimental study on simulated molten jet-coolant interactions

Molten fuel-coolant interactions in a jet contact mode was studied with respect to the safety of liquid-metal-cooled fast breeder reactors (LMFBRs). From a series of molten Wood's metal (melting point 79 °C) jet-water interaction experiments, several distinct modes of interaction behaviors were observed for various combinations of initial temperature conditions of the two fluids. A semi-empirical model for a minimum film boiling temperature criterion was developed and used to reasonably explain the different interaction modes. It was concluded that energetic jet-water interactions are only possible under relatively narrow initial thermal conditions. Preliminary extrapolation of the present results to a UO2-sodium system suggests that mild interactions and coolable debris formation should be most likely in LMFBRs.

Attenuation of Temperature Fluctuations in Thermal Striping.

Thermal striping is random temperature fluctuation generated by the mixing of flow streams at different temperatures. In a liquid-metal-cooled fast breeder reactor (LMFBR), it occurs in the upper plenum as a result of the mixing of sodium coolant jets from fuel assemblies and control rods. It is necessary to consider the temperature cycling in the context of fatigue on the component. A simplified model has been developed to analyze the attenuation of temperature fluctuations from the flow streams to the structure surface. The analytical results were compared with experiments that simulated the upper core structure and impinging jets from fuel sub-assemblies and control rod channels in an LMFBR. Sodium and water were used as simulated fluids to examine the differences in their physical properties. The surface attenuation ratio decreased with the jet flow velocity. The ratio of surface temperature fluctuation to fluid temperature fluctuation is 30–80% in water tests and is 20–50% in sodium tests. These expe...

Two-Phase Flow in Wire-Packed Bed.

For the establishment of liquid metal cooled fast breeder reactors (LMFBRs) as commercial reactors, one of the most important tasks is to develop highly reliable heat transfer tubes for steam generators (SGs) used in LMFBRs because their failure could lead to a potentially explosive sodium/water reaction. The authors proposed a new concept of a wire-packed duplex tube in order to improve stress concentration after its bending, thermal stress and so on against the contact type duplex tubes and knitted wire-packed duplex tubes which had been proposed. In this paper, two-phase flow in wire-packed beds, which simulate the wire-packed layer of the proposed duplex tubes, is experimentally investigated using air and water as test fluids in order to obtain a fundamental data base for a water-leakage detection method using absolute pressure and differential pressure measurements. The measured pressure drops are analyzed by using the following analytical models; (1) homogeneous flow model, (2) channeling model and ...

Living PSA program: LIPSAS development for safety management of an LMFBR plant

Abstract During construction and subsequent operation of a nuclear power plant, many changes occur in components, systems and operating procedures, which continuously modify the configuration of the power plant. The unique ability of PSA techniques to effect of possible changes to baseline system configurations has been noted. A living PSA program can assess and manage safety-related operations and plant changes by adequately reproducing plant models and structured databases corresponding to the changes in system configuration. A living PSA system, LIPSAS, has been developed for the Japanese prototype liquid metal-cooled fast-breeder reactor (LMFBR), Monju, which is in the preoperation functional test stage. In order to utilize the LIPSAS as a risk management tool, equations for the schematic time history of the plant risk level and the relative risk criteria have been developed. Experience with LIPSAS shows that this program is a prospective tool to support decisions that affect plant safety, although a continuing and significant resource commitment of the operations staff at the site is still required.

Second generation columnar grain nickel-base superalloy: Cetel, A. D. and Duhl, D. N. Proc. Conf. Superalloys 1992, Champion, Pennsylvania, USA, 20–24 Sept. 1992 (The Minerals, Metals & Materials Society, 420 Commonwealth Dr, Warrendale, PA 15086, USA) 1992 pp 287–296

The design of various components of Liquid Metal Cooled Fast Breeder Reactors (LMFBR) requires knowledge of the natural convection heat transfer conditions. A review has therefore been made of natural convection heat transfer for low Prandtl number fluids: this includes both theoretical and experimental studies. The importance of the surface orientation is emphasized with reference to vertical, inclined, and horizontal surfaces.The most recent measurements made by the authors are compared with the analytical predictions and measurements made in other low Prandtl number fluids.

Prediction method for thermal stratification in a reactor vessel

The prediction method for thermal stratification phenomena in a fast breeder reactor is described. The focus of attention is placed on the applicability of water test results to predict thermal stratification phenomena in a real plant. The basic feature of thermal stratification was examined in a cylindrical plenum, using water and sodium as test fluids. The similitude relationship between a small-scale test and a real plant is discussed in order to understand the experimental results. The scale-model experiments for LMFBRs (liquid metal-cooled fast breeder reactors) were also performed to see the effects of a reactor configuration and reactor-trip operation condition. Then the magnitudes of the temperature gradient and the ascending speed of stratified interface in the hot plenum of LMFBRs were predicted, based on the results of the water scale-model.

Limit analysis for flat ends in vessels: Kieswetter, N.Tech. Mech. 1989 10, (1), 36–43 (in German)

Low-carbon, nitrogen-controlled 316 stainless steel called 316FR was developed and is regarded as a principal candidate for a main structural material of liquid metal-cooled fast breeder reactor plants in Japan. To develop a creep-fatigue evaluation method suitable for this steel, a number of uniaxial creep-fatigue tests have been conducted for three products of this steel. Long-term data up to about 35,000 h were obtained and applicability of failure life prediction methods was studied based upon their results. Cruciform shaped specimens were also tested under biaxial loading conditions to examine the effect of stress multiaxiality on failure life under creep-fatigue condition.

Low-cycle fatigue analysis of notched members: Zhang, D. and Sandor, B. I.ASTM J. Test. Eval. July 1989 17, (4), 212–217

Low-carbon, nitrogen-controlled 316 stainless steel called 316FR was developed and is regarded as a principal candidate for a main structural material of liquid metal-cooled fast breeder reactor plants in Japan. To develop a creep-fatigue evaluation method suitable for this steel, a number of uniaxial creep-fatigue tests have been conducted for three products of this steel. Long-term data up to about 35,000 h were obtained and applicability of failure life prediction methods was studied based upon their results. Cruciform shaped specimens were also tested under biaxial loading conditions to examine the effect of stress multiaxiality on failure life under creep-fatigue condition.

Diffusion Calculation Model for Streaming Effects in Low Density Channels in LMFBRs

A study is made of the use of modified diffusion theory to calculate the negative reactivity worth of recently proposed flow activated reactor shutdown devices, in which sodium is voided from purely sodium filled channels located at the boundary between the core and the radial blanket in a Liquid Metal cooled Fast Breeder Reactor (LMFBR). Three-dimensional diffusion theory calculations using various definitions of modified diffusion coefficients in the channels are compared with each other and with three-dimensional transport theory computations. While normal diffusion coefficient is found to be inadequate for these reactivity worth calculations, the various modified diffusion coefficients appear adequate for the considered case. Modelling method of the small thickness of steel wall and interstitial sodium layer of the modules are also studied.

Diffusion calculation model for streaming effects in low density channels in LMFBRs.

A study is made of the use of modified diffusion theory to calculate the negative reactivity worth of recently proposed flow activated reactor shutdown devices, in which sodium is voided from purely sodium filled channels located at the boundary between the core and the radial blanket in a Liquid Metal cooled Fast Breeder Reactor (LMFBR). Three-dimensional diffusion theory calculations using various definitions of modified diffusion coefficients in the channels are compared with each other and with three-dimensional transport theory computations. While normal diffusion coefficient is found to be inadequate for these reactivity worth calculations, the various modified diffusion coefficients appear adequate for the considered case. Modelling method of the small thickness of steel wall and interstitial sodium layer of the modules are also studied.

Freezing of a thermite melt injected into an annular channel experiments and recalculations

The analysis of hypothetical accidents of liquid metal cooled fast breeder reactors (LMFBR) shows that the reactor core can under certain circumtances undergo melting. The movement and freezing behavior of the molten core material are of primary importance for the subsequent course of the accident. To investigate the characteristics of freezing and crust formation of a melt penetrating into cold structure, a series of out-of-pile experiments was performed using a simple cylindrical geometry. In the experiments the molten fuel was simulated by a thermite melt that was formed by the reaction of iron oxide and aluminum and was forced into a freezing channel. The movement of the melt was observed by X-ray cinematography and on-line instrumentalation. Post-test examination supplied information about the distribution of the relocated material and the formation of the crusts. The tests were then analyzed with the multicomponent computer code BUCOGEL. The test results show that the freezing behavior of the thermite melt is comparable to that of a uranium oxide/stainless steel mixture as used in in-pile tests conducted in similar geometry. The results of the thermite tests are therefore regarded as suitable to validate the relocation and freezing modules of computer codes describing whole-core accidents.

有限要素法によるタンク型高速増殖炉上部プレナム

The pool-type LMFBR (liquid-metal cooled fast breeder reactor) is characterized by the unpredictably complex flow because of the complicated arrangement of structural components immersed in the liquid sodium. Such flow behavior must be fully understood for the sophisticated design of the LMFBR. Furthermore, the thermal-hydraulic feature of the pool-type LMFBR is one of the most important key issues from the viewpoint of reactor safety. The present study is devoted to the numerical analyses of the thermal-hydraulic characteristics of the pool-type LMFBR.The finite element method (FEM) is chosen for the spatial approximation, taking an advantage of the FEM's flexibility for the approximation of the complicated geometry. Considering the fact that the efficient computation is essential for the FE analysis of complicated three-dimensional flow, an efficient FE scheme for the incompressible viscous flow is proposed in the present paper. Focus of attention is placed on the time integration algorithm and some techniques for the reduction of storage demand. The proposed computational scheme is applied to the three-dimensional thermal-hydraulic problem in the upper plenum of the pool-type LMFBR to demonstrate the feasibility of the present FE strategy.