Molten Salt Breeder Reactor Research Articles

Phase Diagram of Ternary RbF-PbF2-BiF3System as Target Salt in Accelerator Molten-Salt Breeder Reactor

Phase Diagram of Ternary RbF-PbF₂-BiF₃System as Target Salt in Accelerator Molten-Salt Breeder Reactor

Phase diagram of ternary RbF-PbF2-BiF3 system as target salt in accelerator molten-salt breeder reactor.

Click to increase image sizeClick to decrease image sizeKEYWORDS: accelerator breedermolten saltfertile materialfissile materialtargetblanketphase diagramrubidium fluoridelead fluorideternary systembismuth fluoride

溶解塩増殖炉への疑問に答えて : 理想化された核化学反応発電炉

溶解塩増殖炉への疑問に答えて : 理想化された核化学反応発電炉

Locked Rotor Accident Analysis in a Molten Salt Breeder Reactor

An analysis is made of locked rotor accidents in a molten salt breeder reactor (MSBR). The evaluation is performed using a point reactor model for the reactor power and a spatially lumped parameter model of primary system for fuel temperature. In a reactor with circulating fuel such as an MSBR, the reduction or the stoppage of fuel flow caused by locked rotor of fuel pumps will result not only in adding positive reactivity due to the decrease of the loss of delayed neutron precursors out of the core, but also in loosing heat sink. In this report, locked rotor accidents of one, two, three and four pumps out of four fuel salt pumps are evaluated. It is shown that the transients of the reactor system will be within the safety range in virtue of the excellent nuclear and thermal characteristics of the MSBR.

Alternative Breeder Reactor Technologies

The significance of employing breeder reactors to stretch the world resources of nuclear fuels is briefly discussed, and the various types of breeder concepts are described. General descriptions, advantages, and disadvantages of the liquid metal cooled fast breeder, gas cooled fast breeder, molten salt breeder, thermal breeders, and spectral-shift control reactors are presented. Aspects of safeguarding fissile material connected with breeder operation are examined. 31 references. (HDR)

Nuclear Safety Analysis of a Molten Salt Breeder Reactor

An analysis is made of the nuclear safety of a Molten Salt Breeder Reactor (MSBR) using one point kinetics model, with particular attention to the fact that the MSBR, operating on 233U-Th with circulating fuel, has a very small effective delayed neutron fraction, which is conducive to a very high power increase being induced by reactivity addition, particularly when applied sharply. The analysis proves that the transient phenomena of MSBR initiated by an abnormal reactivity addition do not exceed the bounds of structural safety, in virtue of the very favorable nuclear and thermal characteristics of the MSBR.

Reactivity Control Capability of Fuel-Salt Processing System in a Molten-Salt Breeder Reactor

Reactivity Control Capability of Fuel-Salt Processing System in a Molten-Salt Breeder Reactor

Locked Rotor Accident Analysis in a Molten Salt Breeder Reactor

An analysis is made of locked rotor accidents in a molten salt breeder reactor (MSBR). The evaluation is performed using a point reactor model for the reactor power and a spatially lumped parameter model of primary system for fuel temperature. In a reactor with circulating fuel such as an MSBR, the reduction or the stoppage of fuel flow caused by locked rotor of fuel pumps will result not only in adding positive reactivity due to the decrease of the loss of delayed neutron precursors out of the core, but also in loosing heat sink. In this report, locked rotor accidents of one, two, three and four pumps out of four fuel salt pumps are evaluated. It is shown that the transients of the reactor system will be within the safety range in virtue of the excellent nuclear and thermal characteristics of the MSBR.

Nuclear Safety Analysis of a Molten Salt Breeder Reactor

An analysis is made of the nuclear safety of a Molten Salt Breeder Reactor (MSBR) using one point kinetics model, with particular attention to the fact that the MSBR, operating on 233U-Th with circulating fuel, has a very small effective delayed neutron fraction, which is conducive to a very high power increase being induced by reactivity addition, particularly when applied sharply. The analysis proves that the transient phenomena of MSBR initiated by an abnormal reactivity addition do not exceed the bounds of structural safety, in virtue of the very favorable nuclear and thermal characteristics of the MSBR.

Reactivity Control Capability of Fuel-Salt Processing System in a Molten-Salt Breeder Reactor

An evaluation is made of the reactivity control capability of the fuel processing system (FPS) in a molten-salt breeder reactor. The principal functions required of the FPS are : (a) Isolation of 233Pa from regions of high neutron flux during its decay to 233U, and (b) the removal of fission products from the system. The FPS can very usefully serve also to control the primary system reactivity by appropriately utilizing its function of extracting uranium and reconstituting the fuel contained in the salt. The principles of operation are quite similar to the chemical shim control system currently installed in PWR's whereby the core reactivity, affected by changes in the moderator temperature, fuel burnup and transient Xe, is adjusted by regulating the concentration of boric acid introducted in the moderator as neutron absorber. The present study examines the capability of the FPS to follow transient Xe as in PWR's, and proves that the FPS should effectively serve as a system for adjusting not only long-term changes in reactivity but also short-term transient variations without any accompanying difficulties foreseen in operation.

Remarks on the thorium cycle

The use of thorium and neutrons to make U-233 would provide energy for many thousands of years. Thorium is more abundant than uranium and U-233 is the best fissile material for thermal neutron reactors. Four approaches to the use of thorium are worth developing: heavy water moderated reactors with conversion ratios greater than 0.9, such as modified CANDU with lower cost of separating D 2O and U-235; molten salt breeder reactors, from which fission products and excess fuel may be continuously removed; fusionfission hybrids that produce adequate tritium and excess neutrons for sustenance and U-233 production in a subcritical thorium U-233 blanket; and by fission-initiated-thermo nuclear explosions in cavities in salt beds one mile below the earth's surface yielding U-233 from the excess neutrons and thorium and decontaminated steam for power production.

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The oxide chemistry of niobium in molten LiFBeF 2 mixtures

The oxides Nb 2O 5, NiNb 2O 6, NiO and BeO were equilibrated in various combinations with LiFBeF 2 melts containing Nb(V) and NiF 2 over the temperature range 500–750°C. The concentration of Nb(V) in the fluoride phase at equilibrium with NiNb 2O 6 and NiO increased as the square root of the NiF 2 concentration, indicating the formation of dissolved NbO 2F by the reaction ( 1 2 ) NiNb 2 O 6(c) + ( 1 2 ) NiF 2(d) ⇌ NbO 2 F(d) + NiO(c) . Changes in the NbO 2F concentration with melt composition suggest that NbO 2F and BeF 2 have similar acidities and hence that NbO 2F 3 2− is the principal species of Nb(V) present. Formation of the oxy-ion NbO 2 + explains the unexpected stability of Nb(V) in molten fluorides over a wide range of oxide concentration and redox potential. The significance of this to the behavior of fission-product niobium in the fuel of a molten salt breeder reactor is discussed. Formation free energies of dissolved NbO 2F and crystalline NiNb 2O 6 are calculated from the present results and available thermochemical data.

NEEDS: Nuclear Energy Electrical Demand Simulation

A computer model is presented called NEEDS — an acro nym for Nuclear Energy Electrical Demand Simulation- which can simuZate the nuclear sector of the United States electric power economy. The code can forecast the impact of various nuclear-reactor growth patterns over the next fifty years in terms of economics, resource utilization, and the requirements for a supporting fuel-cycle industry. NEEDS was developed for the Office of Energy Research and Development Policy of the National Science Foundation as a tool to help evaluate alternative fission reactor develop ment strategies involving advanced reactor concepts, including advanced high-temperature gas-cooled reac tor (HTGR) systems, the gas-cooled fast reactor (GCFR), the molten-salt breeder reactor (MSBR), the Zight-raater breeder reactor (LWBR), and the heavy- water reactor (HWR) . The objectives of the study were to calculate the impacts, benefits, and require ments of various nuclear growth patterns and to determine the sensitivity of these measures to input parameters which are dominating, uncertain, or sub ject to wide variations. The NEEDS model is currently being used to answer such questions as: What benefits would the United States realize by developing the MSBR, the GCFR, or the LWBR as alternatives to the LWR/HTGR/LMFBR power economy? What are the potential impacts on enrich ment requirements and uranium demand of changing per cent tails assay? How does a power economy based upon a thorium-U233 fuel cycle compare with a pluto nium economy? The model has been used to relate the research-and-development requirements for developing an advanced reactor concept to the availability of the reactor system and its impact on the nuclear economy. In summary, the NEEDS model was developed to study questions of national nuclear policy using a simple modeling approach and minimal input data requirements. The model can forecast the impact of nuclear power development under selected national policy options and potential growth scenarios.

Removal of Iodide from LiF-BeF2 Melts by HF-H2 Sparging—An Application to Iodine Removal from Molten Salt Breeder Reactor Fuel

The results of experiments in which iodine, dissolved as I- in LiF-BeF2 melts, was stripped as HI by sparging with HF-H2 mixtures have indicated that it may be possible to use such treatment to rem...

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Viscoelastic analysis of irradiated graphite with variable creep coefficient

This paper provides a refined analysis of a previous study concerning a method of stress analysis for irradiated graphite which may be used for molten salt breeder reactor (MSBR) core design. The present method includes the effect of a variable creep coefficient which is caused by the non-uniform temperature distribution. To facilitate a simple formulation, it is assumed that the temperature dependence of the elastic response of the material is approximated to be inversely proportional to the creep rate. It is shown that the problem reduces to the solution of several associated (fictitious) elastic problems which have a common elastic modulus inversely proportional to the creep rate of the irradiated graphite. Numerical examples in the previous paper were recalculated on the basis of the present theory. It shows, for large dose values, some small differences compared with the previous calculation. This method is expected to be more effective for higher dose values and more sensitive creep coefficients.

The chemistry and thermodynamics of molten salt reactor fuels

Abstract A fairly detailed knowledge of the chemistry and thermodynamics of molten salt breeder reactor fuels, typically LiF-BeF2ThF 4-UF4 ( 72-16-12- mol % ), has been gained from measurements of heterogeneous equilibria involving various gases (e.g. HF, H2O) and solids (usually oxides). This information combined with the thermochemical data available for the pure compounds has given formation free energies for the fual components. Their entropies can be estimated from the known entropies of the corresponding solid oxides and the charge on the cation. Their activity coefficients depend primarily on the mol % of LiF in the salt mixture. The solubility of oxides generally decreases with the square of the cation charge divided by the radius. Thus the actinide dioxides and, especially, Pa2O2 (or an addition compound of it) are of low solubility. Pourbaix diagrams are presented which summarize the conditions of redox potential and oxide concentration wherein the fuel may be contained without significant corrosion or precipitation reactions, or conditions wherein selective oxide precipitation may be carried out for purposes of fuel reprocessing.

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溶融塩炉開発の現状と将来

溶融塩炉開発の現状と将来