A new method for monitoring the redox potential of fuel salt based on the deposition of 95Nb on Hastelloy C276

2021 ◽  
Vol 109 (5) ◽  
pp. 357-365
Author(s):  
Zhiqiang Cheng ◽  
Zhongqi Zhao ◽  
Junxia Geng ◽  
Xiaohe Wang ◽  
Jifeng Hu ◽  
...  
Keyword(s):  
Redox Potential ◽  
Molten Salt ◽  
Specific Activity ◽  
Chemical Reduction ◽  
Quantitative Approach ◽  
Experimental Results ◽  
New Method ◽  
Molten Salt Reactor ◽  
Fuel Salt ◽  
Well Correlation

Abstract To develop the application of 95Nb as an indicator of redox potential for fuel salt in molten salt reactor (MSR), the specific activity of 95Nb in FLiBe salt and its deposition of 95Nb on Hastelloy C276 have been studied. Experimental results indicated that the amount of 95Nb deposited on Hastelloy C276 resulted from its chemical reduction exhibited a positive correlation with the decrease of 95Nb activity in FLiBe salt and the relative deposition coefficient of 95Nb to 103Ru appeared a well correlation with 95Nb activity in FLiBe salt. Both correlations implied that the measurement of 95Nb activity deposited on Hastelloy C276 specimen might provide a quantitative approach for monitoring the redox potential of fuel salt in MSR.

Distribution and behavior of fission product 95Nb in FLiBe salt

2021 ◽  
Vol 0 (0) ◽  
Author(s):  
Zhiqiang Cheng ◽  
Xiaohe Wang ◽  
Zhongqi Zhao ◽  
Junxia Geng ◽  
Jifeng Hu ◽  
...  
Keyword(s):  
Fission Product ◽  
Redox Potential ◽  
Molten Salt ◽  
Molten Salt Reactor ◽  
Lithium Metal ◽  
Site Monitoring ◽  
Salt Addition ◽  
Redox Indicator ◽  
Γ Ray ◽  
And Behavior

Abstract The 235,238UF4 was irradiated by photo-neutrons, distribution and behavior of the fission product 95Nb from irradiated 235,238UF4 in FLiBe salt were investigated by the measurement of its activity in the salt with the γ-ray spectroscopy. The experiments indicated that a part of 95Nb deposited on the surfaces of graphite and Hastelloy, as the moderator and the structural materials of molten salt reactor (MSR), respectively, and the majority of 95Nb maintained in molten salt. Addition of lithium metal made 95Nb in salt to be reduced and settled, leading to the decrease in its activity. Degree of the decrease was found to be correlated with niobium concentration. The experimental results supported the statement proposed early by ORNL, that 95Nb might be used as a redox indicator for MSR. Finally, the problem met with on-site monitoring for redox potential in MSR was pointed, and a possible protocol to resolve the problem was proposed.

On Voltage-Sensitive Managing the Redox-Potential of MSR Fuel Composition

2008 ◽  
Author(s):  
Pavel N. Alekseev ◽  
Alexander L. Shimkevich
Keyword(s):  
Redox Potential ◽  
Solid Electrolyte ◽  
Molten Salt ◽  
Galvanic Cell ◽  
Fuel Composition ◽  
Molten Salt Reactor ◽  
Exchange Processes ◽  
Cation Conductivity ◽  
The Given

The principles for optimal managing a composition of base solutions for the molten-salt reactor are formulated here for ensuring the given properties and exchange processes as a selective extracting of salt components. The correction of melt properties can be carried out by means of impurity additives parallel with the forced and controllable variation of reduction-oxidation (redox) potential of the non-stoichiometric salts. The accent is done on a possible application of the potentiometer for monitoring and managing of the properties of MSR fuel compositions. For this, one can use the precision methods of e.m.f and the coulomb-metric titration of sodium (lithium) in a galvanic cell upon the base of Na+(Li+)-β″-Al2O3 solid electrolyte with cation conductivity.

Steady Thermal Hydraulics Analysis for a Graphite-Moderated Channel Type Molten Salt Reactor

2008 ◽  
Author(s):  
Dalin Zhang ◽  
Suizheng Qiu
Keyword(s):  
Molten Salt ◽  
Flow Distribution ◽  
Axial Direction ◽  
Theoretical Models ◽  
Outer Wall ◽  
Molten Salt Reactor ◽  
Thermal Hydraulics ◽  
The Core ◽  
Fuel Salt ◽  
Channel Type

The Molten Salt Reactor (MSR) is one of the six GENIV systems capable of breading and burning. In this paper, a graphite-moderated channel type MSR was selected for conceptual research. For this MSR, a ternary system of 0.15LiF-0.58NaF-0.27BeF2 was proposed as the reactor fuel solvent, coolant and also moderator simultaneously with ca.1 mol% UF4 dissolving in it, which circulates through the whole primary loop accompanying fission reaction only in the core. 169 hexagonal graphite elements, each with a central fuel channel, are arranged in the core symmetrically by 30° angles. The theoretical models of the thermal hydraulics under steady condition are conducted in one-twelfth of the core and calculated by the numerical method. The DRAGON code is adopted to calculate the axial and radial power factors. The flow and heat transfer models in the fuel salt and graphite are founded basing on the fundamental mass, momentum and energy equations. The calculated results show the detailed mass flow distribution in the core; and the temperature of the fuel salt, inner and outer wall in the calculated elements along the axial direction are also obtained.

Fuel Salt for the Molten-Salt Reactor

Atomic Energy ◽  
2013 ◽  
Vol 115 (1) ◽  
pp. 5-10 ◽  
Author(s):  
L. I. Ponomarev ◽  
M. B. Seregin ◽  
A. P. Parshin ◽  
S. A. Mel’nikov ◽  
A. A. Mikhalichenko ◽  
...  
Keyword(s):  
Molten Salt ◽  
Molten Salt Reactor ◽  
Fuel Salt

Self-Breeding and Radiotoxicity Analysis on Thorium-Based Molten Salt Reactors

2013 ◽  
Author(s):  
Chun-yan Zou ◽  
Jin-gen Chen ◽  
Xiang-zhou Cai ◽  
Cheng-gang Yu ◽  
Da-zhen Jiang ◽  
...  
Keyword(s):  
Molten Salt ◽  
Nuclear Energy ◽  
Fuel Cycle ◽  
Molten Salt Reactor ◽  
Nuclear System ◽  
Energy Sustainability ◽  
Waste Production ◽  
Breeding Ratio ◽  
Fuel Salt ◽  
System Project

As one of the candidates in the Generation IV reactors program., the molten salt reactor (MSR) has the properties of online refueling and fuel salt reprocessing, MSR is especially attractive for the Thorium fuel cycle, which is very ideal for nuclear non-proliferation, radiotoxicity and nuclear energy sustainability. Therefore, the “Thorium-based Molten Salt Reactor (TMSR) nuclear system” project has been proposed as one of the “Strategic Priority Research Program” of Chinese Academy of Science (CAS). In this paper, we mainly investigated the influence on the breeding ratio and waste radiotoxicity with different reprocessing schemes. By considering the key parameters mentioned above, the aim is to choose an efficient reprocessing scheme for TMSR to reach self-breeding with Th/U fuel cycle and minimize the radioactive waste production of the molten salt.

Reactor Controllability of 3-Region-Core Molten Salt Reactor System: A Study on Load Following Capability

2006 ◽  
Author(s):  
Takahisa Yamamoto ◽  
Koshi Mitachi ◽  
Masatoshi Nishio
Keyword(s):  
Molten Salt ◽  
Solid Fuel ◽  
Molten Salt Reactor ◽  
Delayed Neutron ◽  
Reactor Kinetics ◽  
External Loop ◽  
Fuel Salt ◽  
Salt Flow ◽  
Point Reactor Kinetics ◽  
Graphite Moderator

The Molten Salt Reactor (MSR) systems are liquid-fueled reactors that can be used for actinide burning, production of electricity, production of hydrogen, and production of ssile fuels (breeding). Thorium (Th) and uranium-233 (233U) are fertile and ssile of the MSR systems, and dissolved in a high-temperature molten fluoride salt (fuel salt) with a very high boiling temperature (up to 1650K), that is both the reactor nuclear fuel and the coolant. The MSR system is one of the six advanced reactor concepts identified by the Generation IV International Forum (GIF) as a candidate for cooperative development [1]. In the MSR system, fuel salt flows through a fuel duct constructed around a reactor core and fuel channel of a graphite moderator accompanied by fission reaction and heat generation, and flows out to an external-loop system consisted of a heat exchanger and a circulation pump. Due to the motion of fuel salt, delayed neutron precursors that are one of the source of neutron production make to change their position between the ssion reaction and neutron emission events and decay even occur in the external loop system. Hence the reactivity and effective delayed neutron precursor fraction of the MSR system are lower than those of solid fuel reactor systems such as Boiling Water Reactors (BWRs) and Pressurised Water Reactor (PWRs). Since all of the presently operating nuclear power reactors utilize solid fuel, little attention had been paid to the MSR analysis of the reactivity loss and reactor characteristics change caused by the fuel salt circulation. Sides et al. [2] and Shimazu et al. [3] developed MSR analytical models based on the point reactor kinetics model to consider the effect of fuel salt flow. Their models represented a reactor as having six zones for fuel salt and three zones for the graphite moderator. Since their models employed the point reactor kinetics model and the rough temperature approximation, their results were not sufficiently accurate to consider the effect of fuel salt flow.

Numerical Research on Steady Coupling of Neutronics and Thermal-Hydraulics for a Molten Salt Reactor

2008 ◽  
Author(s):  
Dalin Zhang ◽  
Changliang Liu ◽  
Libo Qian ◽  
Guanghui Su ◽  
Suizheng Qiu
Keyword(s):  
Thermal Neutron ◽  
Molten Salt ◽  
Molten Salt Reactor ◽  
Thermal Hydraulics ◽  
Steady Condition ◽  
Delayed Neutron ◽  
The Core ◽  
Production Of Hydrogen ◽  
Fuel Salt ◽  
Salt Flow

The Molten Salt Reactor (MSR), which is one of the ‘Generation IV’ concepts, can be used for production of electricity, actinide burning, production of hydrogen, and production of fissile fuels. In this paper, a single-liquid-fueled MSR was selected for conceptual research. For this MSR, a ternary system of 15%LiF-58%NaF-27%BeF2 was proposed as the reactor fuel solvent, coolant and also moderator with ca. 1 mol% UF4 dissolving in it, which circulates through the whole primary loop accompanying fission reaction only in the core. The fuel salt flow makes the MSR different from the conventional reactors using solid fissile materials, and makes the neutronics and thermal-hydraulic coupled strongly, which plays the important role in the research of reactor safety analysis. Therefore, it’s necessary to study the coupling of neutronics and thermal-hydraulic. The theoretical models of neutronics and thermal-hydraulics under steady condition were conducted and calculated by numerical method in this paper. The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six-group delayed neutron precursors considering flow effect. The thermal-hydraulic model was founded on the base of the fundamental conservation laws: the mass, momentum and energy conservation equations. These two models were coupled through the temperature and heat source. The spatial discretization of the above models is based on the finite volume method (FVM), and the thermal-hydraulic equations are computed by SIMPLER algorithm with domain extension method on the staggered grid system. The distribution of neutron fluxes, the distribution of the temperature and velocity and the distribution of the delayed neutron precursors in the core were obtained. The numerical calculated results show that, the fuel salt flow has little effect to the distribution of fast and thermal neutron fluxes and effective multiplication factor; however, it affects the distribution of the delayed neutron precursors significantly, especially long-lived one. In addition, it could be found that the delayed neutron precursors influence the neutronics slightly under the steady condition, and the flow could remove the heat generated by the neutron reactions easily to ensure the reactor safe. The obtained results serve some valuable information for the research and design of this new generation reactor.

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