scholarly journals MAXIMISING DISCHARGE BURNUP IN AN OPEN CYCLE MOLTEN SALT REACTOR

2021 ◽  
Vol 247 ◽  
pp. 12002
Author(s):  
Charlie Constable ◽  
Ben Lindley ◽  
Geoff Parks
Keyword(s):  
Molten Salt ◽  
Fission Products ◽  
Molten Salt Reactor ◽  
The Core ◽  
Unit Cells ◽  
Internal Set ◽  
Open Cycle ◽  
Set Up ◽  
Work Done ◽  
Graphite Moderator

This paper discusses work done to find an estimate of the maximum achievable discharge burnup in an open cycle molten salt reactor (MSR). An in-development deterministic code (WIMS11) is used to create a model of a simple generic MSR, and the methodology employed is discussed. Some experimentation is done with regards to the internal set-up of the ‘unit cells’ within the core, which shows there is a strong link between this geometry and the achievable burnup. Work is done to quantify the effects of removing volatile fission products and implementing a two-batch refuelling scheme. Finally, an optimization process is carried out whereby the optimal proportion of graphite moderator within the core is found which balances power across the regions while maximising discharge burnup. Two fuels are tested, one which carries only 235U and 238U, and another which also carries 232Th. It is found that the maximum achievable discharge burnup is approximately 155 MWd/kg, which is considerably higher than modern PWRs, despite a lower enrichment and only two batches of fuel being used.

Primary Shielding Design for an Optimized Molten Salt Reactor

2013 ◽  
Author(s):  
Zhihong Zhang ◽  
Xiaobin Xia ◽  
Jianhua Wang ◽  
Changyuan Li
Keyword(s):  
Molten Salt ◽  
Dose Rate ◽  
Gamma Dose ◽  
Acceptable Level ◽  
Gamma Dose Rate ◽  
Molten Salt Reactor ◽  
Thermal Shield ◽  
Oak Ridge ◽  
The Core ◽  
Shielding Design

Molten salt reactor (MSR) system, a candidate of the Generation IV reactors, has inherent safety, on-line refueling and good neutron economy as typical advantages. An optimized MSR is developed by changing the size of fuel channel and the graphite-to-molten salt volume radio, based on the Molten-Salt Reactor Experiment (MSRE), which was originally developed at the Oak Ridge National Laboratory (ORNL). In this paper, shielding calculations for the optimized MSR are presented. The goal of this study is to determine the necessary shielding to decrease the neutron and gamma dose rate to the acceptable level according to national regulations. The operating temperature of the optimized MSR is designed in the range of 500 °C–700 °C, heat removal is also considered in the shielding design. The shielding calculations are carried out by using Monte Carlo method. The shielding system of the optimized MSR consists of 7 zones: the core, the core can, the reactor vessel, the thermal shield, the reactor cell containment, the shield tank and the concrete wall. The combinations of shielding materials in the thermal shield were evaluated. The thermal shield filled with carbon steel balls and circulating water gets an excellent shielding performance and heat removing effects. The neutron spectra and dose distributions, as well as the energy deposition over different shields have been analyzed. The total neutron dose rate outside the thermal shield is attenuated by a factor of about 104, and the gamma dose rate by a factor of about 103. These results show that the shielding design could low dose rate to an acceptable level outside the shielding and far below dose limit required.

A Review of Molten Salt Reactor Xenon Analysis Literature

2019 ◽  
Vol 6 (1) ◽  
Author(s):  
T. J. Price ◽  
O. Chvala
Keyword(s):  
Molten Salt ◽  
Void Fraction ◽  
Experimental Work ◽  
National Laboratory ◽  
A Priori ◽  
Molten Salt Reactor ◽  
Oak Ridge ◽  
Sensitive Dependence ◽  
Oak Ridge National Laboratory ◽  
Work Done

Abstract This paper presents a review of xenon analyses literature related to molten salt reactors (MSRs). A brief primer of reactor xenon theory is presented for fluid fueled reactors. A review of xenon analysis literature is presented for both the work done by the Oak Ridge National Laboratory, and the later work in academia. A review of experimental work is presented. The paper concludes with describing some of the difficulties in establishing a priori xenon models and includes a commentary on the sensitive dependence of the molten salt reactor xenon behavior on the circulating void fraction.

A Novel Molten Salt Reactor Concept to Implement the Multi-Step Time-Scheduled Transmutation Strategy

2002 ◽  
Author(s):  
Gyula Csom ◽  
Sandor Feher ◽  
Mate Szieberth
Keyword(s):  
Molten Salt ◽  
Liquid Fuel ◽  
Primary Circuit ◽  
Chemical Processing ◽  
Molten Salt Reactor ◽  
The Core ◽  
Continuous Feed ◽  
Fuel Mixtures ◽  
Continuous Chemical ◽  
Optimal Implementation

Nowadays the molten salt reactor (MSR) concept seems to revive as one of the most promising systems for the realization of transmutation. In the molten salt reactors and subcritical systems the fuel and material to be transmuted circulate dissolved in some molten salt. The main advantage of this reactor type is the possibility of the continuous feed and reprocessing of the fuel. In the present paper a novel molten salt reactor concept is introduced and its transmutational capabilities are studied. The goal is the development of a transmutational technique along with a device implementing it, which yield higher transmutational efficiencies than that of the known procedures and thus results in radioactive waste whose load on the environment is reduced both in magnitude and time length. The procedure is the multi-step time-scheduled transmutation, in which transformation is done in several consecutive steps of different neutron flux and spectrum. In the new MSR concept, named “multi-region” MSR (MRMSR), the primary circuit is made up of a few separate loops, in which salt-fuel mixtures of different compositions are circulated. The loop sections constituting the core region are only neutronically and thermally coupled. This new concept makes possible the utilization of the spatial dependence of spectrum as well as the advantageous features of liquid fuel such as the possibility of continuous chemical processing etc. In order to compare a “conventional” MSR and a proposed MRMSR in terms of efficiency, preliminary calculational results are shown. Further calculations in order to find the optimal implementation of this new concept and to emphasize its other advantageous features are going on.

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.

Parametric Lattice Study of a Graphite-Moderated Molten Salt Reactor

2014 ◽  
Author(s):  
Boris A. Hombourger ◽  
Jiři Křepel ◽  
Konstantin Mikityuk ◽  
Andreas Pautz
Keyword(s):  
Molten Salt ◽  
Unit Cell ◽  
Fuel Composition ◽  
Molten Salt Reactor ◽  
Closed Cycle ◽  
The Core ◽  
Channel Radius ◽  
Infinite Lattice

This article illustrates the influence of heterogeneity in an infinite lattice of a Molten Salt Reactor moderated by graphite. For a complete description of heterogeneity in a 2D lattice, two variables are needed; in this study the salt share in the unit cell and the channel radius are used. The equilibrium Thorium-based closed-cycle fuel composition is systematically derived for each chosen combination of points, and results such as kinf and the actinide vector composition are calculated. Results show that the heterogeneity effect can indeed be important for optimization of the core design of moderated molten salt reactors.

scholarly journals SIMULATION OF BREED AND BURN FUEL CYCLE OPERATION OF MOLTEN SALT REACTOR IN BATCH-WISE REFUELING MODE

2021 ◽  
Vol 247 ◽  
pp. 13003
Author(s):  
Valeria Raffuzzi ◽  
Jiri Krepel
Keyword(s):  
Molten Salt ◽  
Fuel Cycle ◽  
Fission Products ◽  
Molten Salt Reactor ◽  
Core Size ◽  
Spent Fuel ◽  
Fluid Core ◽  
Active Core ◽  
Breed And Burn ◽  
Gen Iv

The Molten Salt Reactor (MSR) is one of the most revolutionary Gen-IV reactors and it can be operated, especially with chloride salts, in the so-called breed and burn fuel cycle. In this type of fuel cycle the fissile isotopes from spent fuel do not need to be reprocessed, because the excess bred fuel covers the losses. The liquid phase of the MSR fuel assures its instant homogenization, and the reactor can be operated with batch-wise refueling thus reaching an equilibrium state. At the same time, the active core of the chloride fast MSR needs to be bulky to limit neutron leakage. In this study, the code Serpent 2 was coupled to the Python script BBP to simulate batch-wise operation of the breed and burn MSR fuel cycle. The script, previously developed for solid assemblies shuffling, was modified to simulate fuel homogenization after fertile material addition. Several fuel salts and fission products removal strategies were simulated and their impact was analyzed. Similarly, the influence of blanket volume was assessed in a two-fluid core layout. The results showed that the reactivity initially grows during the irradiation period and later decreases. The blanket has a large impact on the performance and it can be used to further increase the fuel burnup or to shrink the active core size. The breed and burn fuel cycle in MSR can reach high fuel utilization without fuel reprocessing and a multi-fluid layout can help to decrease the core size.

Numerical Investigation of Safety System Parameters in Molten Salt Reactor: Wall Effect on Freeze Valve Opening Time

2021 ◽  
Author(s):  
Muhammad Ilham ◽  
Indarta Kuncoro Aji ◽  
Tomio Okawa
Keyword(s):  
Molten Salt ◽  
Power Plants ◽  
Melting Process ◽  
Wall Effect ◽  
Fourth Generation ◽  
Reactor Wall ◽  
Molten Salt Reactor ◽  
Oak Ridge ◽  
Valve System ◽  
The Core

Abstract Molten Salt Reactor (MSRs) is one of the fourth generation Nuclear Power Plants with better capabilities and potentialities compared to previous generation, the enthusiasm for molten fuel reactor has been increasing around the world. MSRs has passive safety where if the core is overheating cause by accident event, the liquid salt fuel was required to be moved to the safety drain tank underneath the core vessel by gravity force. During this occasion, the freeze valve (FV) that formed in the pipe located between the core and drain tank must be melt out promptly to prevent the vessel to reach it is melting point. In this paper, we conduct on thermal analysis of the freeze valve at the solidification and melting process based on finite elements methods. The enthalpy-porosity method adopted by ANSYS Fluent was used to simulated the designed system at specified condition. The Oak Ridge National Laboratory of Molten Salt Reactor Experiment Freeze valve system was used as a references for parameters investigation. Using pipe wall thickness of 5 mm, 10 mm, and 15 mm to examined the wall effect to thermal properties of the designed freeze valve. The wall pipe for FV systems material was also investigate in order to examine its effect to the opening time. Further, the temperature distributions of the valve system were obtained and analyzed. It was found that the wall effect has significant impact to the solidification and melting process.

Breeding Properties Study on High Power Thorium Molten Salt Reactor

2017 ◽  
Author(s):  
Yafen Liu ◽  
Rui Guo ◽  
Xiangzhou Cai ◽  
Rui Yan ◽  
Yang Zou ◽  
...  
Keyword(s):  
Molten Salt ◽  
High Power ◽  
Liquid Fuel ◽  
Molten Salt Reactor ◽  
The Core ◽  
Optimized Model ◽  
Core Conditions ◽  
Thorium Molten Salt Reactor ◽  
Reactivity Insertion ◽  
Two Fluid

The Molten Salt Reactor (MSR), the only one using liquid fuel in the six types ‘Generation IV’ reactors, is very different from reactors in operation now and has initiated very extensive interests all over the world. This paper is primarily aimed at investigating the breeding characteristics of high power (1000 MWe) Thorium Molten Salt Reactor (TMSR) based on the two-fluid Molten Salt Breeder Reactor (MSBR) with superior breeding performance. We explored the optimized structure to be a thorium based molten salt breeder reactor with different core conditions and different post-processing programs, and finally got the breeding ratio of 1.065 in our TMSR model. At last we analyzed the transient security of our optimized model with results show that the temperature coefficient of core is −3 pcm/K and a 2000 pcm reactivity insertion can be successfully absorbed by the core if the insertion time is more than or equal to 5 seconds and the core behaves safely.

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.

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