Pb-Bi Coolant Research Articles

Разработка усовершенствованной конструкции массообменного аппарата для реакторных установок с жидкометаллическим теплоносителем

This work was intended to develop an improved mass transfer apparatus for reactors with liquid-metal coolant. The prototypes were existing designs of mass transfer apparatuses developed by the Institute of Physics and Power Engineering (IPPE) JSC. The study yielded an improved design for mass transfer apparatus with adjustable coolant flow. The findings of this work made it possible to resolve the challenge of coolant flow control in the main duct of primary reactor circuits with Pb-Bi coolant.

Numerical analysis of melt migration and solidification behavior in LBR severe accident with MPS method

In Lead-based reactor (LBR) severe accident, the meltdown and migration inside the reactor core will lead to fuel fragment concentration, which may further cause re-criticality and even core disintegration. Accurately predicting the migration and solidification behavior of melt in LBR severe accidents is of prime importance for safety analysis of LBR. In this study, the Moving Particle Semi-implicit (MPS) method is validated and used to simulate the migration and solidification behavior. Two main surface tension models are validated and compared. Meanwhile, the MPS method is validated by the l-plate solidification test. Based on the improved MPS method, the migration and solidification behavior of melt in LBR severe accident was studied furthermore. In the Pb–Bi coolant, the melt flows upward due to density difference. The migration and solidification behavior are greatly affected by the surface tension and viscous resistance varying with enthalpy. The whole movement process can be divided into three stages depending on the change in velocity. The heat transfer of core melt is determined jointly by two heat transfer modes: flow heat transfer and solid conductivity. Generally, the research results indicate that the MPS method has unique advantage in studying the migration and solidification behavior in LBR severe accident.

Experimental analyses of bismuth sample reactivity worth at Kyoto University Critical Assembly

ABSTRACT Sample reactivity worth experiments are carried out by substituting aluminum (Al) plates for bismuth (Bi) ones at the Kyoto University Critical Assembly. At the beginning, uncertainty quantification of bismuth isotope is conducted by deterministic calculations with nuclear data library JENDL-4.0, with the use of experimental results of sample reactivity worth. Then, with the combined use of current (Bi) and previous (Pb) experimental results that demonstrate the comparative difference in the sensitivity and uncertainty of Bi and Pb isotopes, experimental results of cross-section uncertainties of Bi isotope are available for examination of neutron characteristics of Pb–Bi coolant material in the accelerator-driven system. From the experimental analyses, further uncertainty analyses by neutron transport calculations are needed for several reactions of Bi isotope, especially with the use of the covariance data of capture, elastic scattering and inelastic scattering reactions in another nuclear data library.

Choice of Materials for Sorption Treatment of Pb–Bi Coolant to Remove the Accumulated Radionuclides

Sorption and phase distribution of activation products (65Zn, 54Mn, 59Fe, 60Со) in systems consisting of Pb–Bi melt and a porous material were studied under laboratory conditions. Radionuclides of elements with minimal oxygen affinity (Ru, Te, and Sb) remain predominantly in the coolant melt and are weakly sorbed by the solid phase. Radionuclides with higher oxygen affinity are well sorbed by structural solid sorbents. The maximum of the 60Со sorption onto porous steel is reached at 853 K.

Sorption of Radionuclides from the Primary Circuits of Reactors with Pb–Bi Coolant onto the Surface of Structural Materials of the Reactor Core

Sorption and phase distribution of 54Mn, 59Fe, 60Со, 106Ru, 125Sb, 137Cs, 144Се, 154,155Eu, and 235,238U radionuclides in contact of the Pb–Bi melt with the steel surface were studied. Radionuclides of the elements with the minimal oxygen affinity (Ru, Te, and Sb) remain mainly in the coolant melt and are weakly sorbed onto the solid phase. The other fission product radionuclides having higher oxygen affinity are sorbed onto the surface of steel structural materials. The sorption increases at 350–400°С.

Conceptual Design of Passive Safety System for Lead-Bismuth Cooled Fast Reactor

This paper presents the results of the conceptual design of passive safety systems for reactor power 225 MWth using Pb-Bi coolant. Main purpose of this research is to design of heat removal system from the reactor wall. The heat from the reactor wall is removed by RVACS system using the natural circulation from the atmosphere around the reactor at steady state. The calculation is performed numerically using Newton-Raphson method. The analysis involves the heat transfer systems in a radiation, conduction and natural convection. Heat transfer calculations is performed on the elements of the reactor vessel, outer wall of guard vessel and the separator plate. The simulation results conclude that the conceptual design is able to remove heat 1.33% to 4.67% from the thermal reactor power. It’s can be hypothesized if the reactor had an accident, the system can still overcome the heat due to decay.

Pre-conceptual study of small modular PbBi-cooled nitride fuel reactor core characteristics

In this paper a pre-conceptual neutronics study on a small modular Pb-Bi cooled reactor with nitride fuel (SMoPN) is presented. The SMoPN is designed to meet the requirements for nuclear energy expansion in the next decades, by using the plutonium and thorium nitride fuel to increase the efficiency and performance of fuel. Based on the existing experiences of nuclear reactor, the primary design parameters were provided to match the design goals by the whole core three-dimensional calculation. The nitride fuel, stainless steel cladding and Pb-Bi coolant have a perfect compatibility with each other, as well as with excellent neutronics characteristics. High conversion ratios have been achieved to ensure 20 effective full power years (EFPYs) without refueling and shuffling. During the core lifetime, the burn-up swings slightly due to the perfect breeding capability. All of the important reactivity coefficients are negative to assure the SMoPN holding passive safety. The control system can provide enough shutdown margins in both normal and abnormal operational conditions. Therefore, the SMoPN concept satisfies completely the advanced design idea and the requirements of advanced nuclear reactor system.

Material development for India’s nuclear power programme

The area of materials research has registered a phenomenal growth in the recent years, assiduously accepting and assimilating ideas, concepts and analytical as well as experimental methodologies and techniques form almost all scientific disciplines, thereby demonstrating its remarkably multidisciplinary and interdisciplinary character. The focus of the materials programme of this centre is to provide materials, processes and processing solutions to the emerging needs of evolving indigenous nuclear energy systems by proactive research and development on a continuing basis. The initial stage of our activities was formulated around three stage Indian nuclear power programme. In stage I, material issues related to in-core materials with emphasis on development of fabrication routes of zirconium alloys for structural application were addressed. Subsequently the thrust areas were development and characterization of mixed oxide fuel, advanced zirconium alloys, structural steels, superalloys, neutron absorber materials based on boron carbides and borides, and shape memory alloys. The research was useful for in-service performance evaluation, safety assessment, residual life estimation and life extension of nuclear reactors built during stage I i.e., PHWRs and BWRs. It also included developments which would permit rapid expansion of nuclear power initially through fast breeder reactor based on mixed oxide fuel and later based on metallic fuels. For the 3rd stage, multi-layer coatings, graphite coolant tube, BeO, refractory metals and alloys, heat-treated zirconium alloys are being developed for CHTR, ADSS and AHWR. The materials being developed for fusion programme are low Z and high Z material for plasma facing application, Cu-alloys for heat sink, austenitic steels, RAFMS and ODS for structurals and NbTi, Nb3Sn and Nb3Al superconductors, lithium titanate, lithium silicate breeders, and Pb–Bi coolant. A brief overview of the materials research activities currently being pursued at Bhabha Atomic Research Centre is presented in this article.

Neutron Balance Analysis for Sustainability of Breed-and-Burn Reactors

One objective of the present work is to determine the minimum burnup (BU) required to sustain a breed-and-burn (B&B) mode of operation in a large 3000-MW(thermal) sodium-cooled fast reactor core fed with depleted uranium-based metallic fuel. Another objective is to assess the feasibility of using the fuel discharged at the minimum required BU for fabricating the starter of an additional B&B core without separation of actinides and most of the solid fission products. A melt-refining process is used to remove gaseous and volatile fission products and to replace the cladding when it reaches its 200 displacements per atom radiation damage limit. Additional objectives are to assess the validity of a simplified zero-dimensional (0-D) neutron balance analysis for determination of the minimum BU required and the maximum BU attainable in a B&B mode of operation and to apply this 0-D methodology to assess the feasibility of establishing a B&B mode of operation in fast reactor cores made of different combinations of fuels, coolants, and structural materials.It is found that the minimum BU required to sustain the B&B mode in the referenced depleted uranium-fueled B&B reactor is 19.4% FIMA. The number of excess neutrons that can be generated by the fuel discharged at 19.4% FIMA is found sufficient to establish the B&B mode in another B&B core. The net doubling time for starting new B&B reactors with fuel discharged from operating B&B reactors is 12.3 yr.The minimum BU required to sustain the B&B mode of operation in alternative core designs was found to be 29% FIMA when using Pb-Bi coolant with metallic uranium fuel and 40% FIMA when using nitride fuel with sodium coolant. The B&B mode of operation cannot be established using thorium fuel and liquid-metal coolant.The results derived from the neutron balance analysis strongly depend on the value of the estimated neutron leakage probability and the fraction of neutrons lost in the reactivity control systems. A neutron balance performed using a simplified 0-D core model, although not accurate due to, primarily, inaccurate spectra predictions, provides reasonable estimates of the minimum required and the maximum attainable BUs despite the fact that its k∞ evolution prediction is inaccurate. The 0-D approach can save much computational effort and time and is found to be useful for scoping analysis.

Safety design of Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR)

In Pb–Bi-cooled direct contact boiling water small fast reactor (PBWFR), steam is generated by direct contact of feedwater with primary Pb–Bi coolant above the core, and Pb–Bi coolant is circulated by steam lift pump in chimneys. Safety design has been developed to show safety features of PBWFR. Negative void reactivity is inserted even if whole of the core and upper plenum are voided hypothetically by steam intrusion from above. The control rod ejection due to coolant pressure is prevented using in-vessel type control rod driving mechanism. At coolant leak from reactor vessel and feedwater pipes, Pb–Bi coolant level in the reactor vessel required for decay heat removal is kept using closed guard vessel. Dual pipes for feedwater are employed to avoid leak of water. Although there is no concern of loss of flow accident due to primary pump trip, feedwater pump trip initiates loss of coolant flow (LOF). Injection of high pressure water slows down the flow coast down of feedwater at the LOF event. The unprotected loss of flow and heat sink (ATWS) has been evaluated, which shows that the fuel temperatures are kept lower than the safety limits.

Design study on reactor structure of Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR)

Pb–Bi-cooled direct contact boiling water small fast reactor (PBWFR) can produce steam by direct contact of feedwater with primary Pb–Bi coolant above the core, and circulate Pb–Bi coolant by means of buoyancy of steam bubbles. The PBWFR is capable of eliminating components of the cooling system such as primary pumps and steam generators, and thereby making the reactor system simple and compact. The specifications of the PBWFR are as follows: the fuel is Pu–U nitride; the core height is 75 cm; the core diameter is 278 cm; the average burnup is 80 GWd/t; the refueling interval is 10 years; the rated electric power is 150 MWe; the rated thermal power is 450 MWt; the core outlet/inlet temperatures are 460 °C/310 °C, respectively; and the operating steam pressure is 7 MPa. The reactor structure design has been formulated, where reactor vessel sizes are 4200 mm (ID) × 19,750 mm (H), the guard vessel is a closed type, the upper structure is made of chimneys, and the core support structure is hung up. An ultrasonic flow meter is installed inside the vessel. The seismic evaluation, design of refueling procedure and cost evaluation have been performed.

Numerical Analysis of Lead-Bismuth-Water Direct Contact Boiling Heat Transfer

Direct contact boiling heat transfer of sub-cooled water with lead-bismuth eutectic (Pb-Bi) was investigated for the evaluation of the performance of steam generation in direct contact of feed water with primary Pb-Bi coolant in upper plenum above the core in Pb-Bi-cooled direct contact boiling water fast reactor. An analytical two-fluid model was developed to estimate the heat transfer numerically. Numerical results were compared with experimental ones for verification of the model. The overall volumetric heat transfer coefficient was calculated from heat exchange rate in the chimney. It was confirmed that the calculated results agreed well with the experimental result.

High temperature behavior of irradiated mixed nitride fuel during heating tests

Heating tests for U–Pu mixed nitride fuel irradiated in the experimental fast reactor, “JOYO” were performed for evaluation of fission gas release and nitride decomposition behavior. The gas release onset temperature of irradiated nitride fuel is greater than 1700 °C, which is higher than that of oxide fuel at the same heating rate. A high temperature transient may lead to a permanent expansion of the fuel, giving it a lower density than that of the Pb–Bi coolant. Fission gas release in irradiated nitride fuel does not tend to occur at similar temperatures to the decomposition of the fuel. Thus, in the scenarios of a core damage accident for a Pb–Bi cooled fast reactor, this decomposition behavior must be taken into consideration.

Experiment and Numerical Simulation of Bubble Behavior in Argon Gas Injection into Lead-Bismuth Pool

In a lead-bismuth alloy (45%Pb-55%Bi) cooled direct contact boiling water fast reactor (PBWFR), steam can be produced by direct contact of feed water with primary Pb-Bi coolant in the upper core plenum, and Pb-Bi coolant can be circulated by buoyancy forces of steam bubbles. As a basic study to investigate the two-phase flow characteristics in the chimneys of PBWFR, a two-dimensional two-phase flow was simulated by injecting argon gas into Pb-Bi pool in a rectangular vessel (400mm in length, 1500mm in height), and bubble behavior were investigated experimentally. Bubble sizes, bubble rising velocities and void fractions were measured using void probes. The experimental conditions are the atmospheric pressure and the flow rate of injection Ar gas is 10, 20, and 30 NL/min. The average of measured bubble rising velocity was about 0.6 m/s. The average chord length was about 7mm. An analysis was performed by two-dimensional and two-fluid model. The experimental results were compared with the analytical results to evaluate the validity of the analytical model. Although large diameter bubbles were observed in the experiment, the drag force model of lower value performed better for simulation of the experimental result.

Full-Height Thermal-Hydraulic Scaling Analysis of Pb-Bi-Cooled Fast Reactor PEACER

A liquid-metal–cooled fast reactor Proliferation-resistant Environment-friendly Accident-tolerant Continuable-energy Economical Reactor (PEACER) is being developed at Seoul National University with metallic-type U-Pu-Th fuel and lead-bismuth (Pb-Bi) coolant, employing the steam-generating Rankine cycle without an intermediate loop. The natural circulation potential is a key ingredient of the liquid-metal reactor design. PEACER is designed as a loop system, which increases the natural circulation capacity and reduces the maintenance cost. This work is concerned with design of a large-scale test loop Heavy Eutectic Loop Integrated Operation System (HELIOS) to simulate thermal-hydraulic behavior of Pb-Bi in the PEACER system following a single-phase one-dimensional flow loop model. The full-height loop is being constructed for Pb-Bi natural-circulation testing and operating procedure development. Electrically heated rods are used in the heating section, and a water-cooling method is adopted in the cooling section. HELIOS has the same elevation difference between thermal centers of the heat source and heat sink but reduced piping diameters.

Nuclide Composition Calculation for Pb-Bi Coolant of 80 MW Sub-Critical Reactor. Comparison of Different Data Libraries

Nuclide composition of Pb-Bi coolant of 80 MW sub-critical reactor has been investigated. Particle spectra for different parts of the reactor including spallation (target) module, the core and other units have been calculated. Special data files for nucleon induced reaction cross-sections including data for spallation reactions, fission product and heavy cluster yields at the energies of primary particles up to 1 GeV have been prepared. The special attention has been paid to the uncertainty of the results obtained, the influence of the choice of nuclear model parameters used for nuclear reaction cross-section calculations and to the comparison of the results obtained using different nuclear data libraries applied for nuclide composition calculations. The reactor configuration with the “window” for the proton beam and the “windowless” design were investigated.

Box Model of Radionuclide Dispersion and Radiation Risk Estimation for Population in Case of Radioactivity Release from Nuclear Submarine #601 Dumped in the Kara Sea

ABSTRACTA relatively simple model is suggested for radionuclide dispersion and assessment of additional population dose and radiation risk for radionuclide release from the nuclear submanne (NS) #601 with Pb-Bi coolant that was dumped in the Kara Sea.