Graphite-moderated Molten Salt Reactor Research Articles

Modeling Flow Blockage Accident in Graphite-Moderated Molten Salt Reactors

Fuel channel flow blockage is a postulated accident scenario in graphite-moderated molten salt reactors (MSRs), caused by debris partially obstructing the flow in a hole inside a graphite moderator brick. This obstruction produces an increase in the fuel salt bulk temperatures in the blocked channel, resulting in a significant fraction of the deposited fission power being conducted through graphite from the blocked channel to the surrounding unblocked channels. In the blocked channel, a combination of forced convection and natural convection takes place, the latter being dominant at very strong flow rate reductions. At full blockage (with the fragment completely obstructing the flow), because of the poor thermal conductivity of fuel salt, correct estimation of the natural convection impact on heat transfer between the channel bulk and the walls is fundamental to determine whether boiling conditions are reached in the blocked channel. The proposed assessing approach, developed in collaboration with ThorCon in the frame of the ThorCon reactor core design and optimization, is presented in this work. Results from a case study based on the molten salt breeder reactor (MSBR) design and operating conditions are presented and discussed. These analyses show that with respect to full blockage accidents in other reactor types, the consequences of full channel blockage accidents in graphite-moderated MSRs are milder. For graphite-moderated MSRs, the main mechanical limits during the accident arise in thermal stresses in the graphite brick. Graphite thermal and mechanical properties, namely, bulk modulus, thermal expansion coefficient, ultimate strength, and thermal conductivity, are severely impacted by neutron irradiation through the life of the graphite. All these effects are taken into account in the current analysis and are discussed in the present work.

Simulation-based methodology to assess the lattice defects creation as energy storing process

In nuclear industry, damages caused to in-core structural materials by high-energy neutrons have been studied for years. Recently, the possibility of storing energy through the creation of defects in crystalline materials via heavy particle bombardment was investigated. To evaluate the potential of the Wigner effect as energy-storing process, a graphite-moderated Molten Salt Reactor was adopted as a test case. Neutronics simulations performed with OpenMC Monte Carlo code allowed estimating the amount of energy released during fission events, the fraction that is deposited in graphite, and the fraction that has the potential to cause radiation damages. Molecular dynamics simulations were then performed to study the evolution of cascades triggered by high-energy neutrons. As a major outcome, a more accurate estimate of the radiation damage with respect to the traditional approaches was obtained, and the fraction of the deposited neutron energy that can be stored through long-lived defects was quantified.

Transition to thorium fuel cycle on a heavy water moderated molten salt reactor by using low enrichment uranium

Lacking fissile element (U-233) in nature is a key issue faced by Th-U fuel cycle, and existing available fuels should be taken as starting fuel to convert Th-232 to U-233 to realize transition to thorium fuel cycle. Low enrichment uranium (LEU, U-235 enrichment < 20%) is the most accessible fuel, and has been widely studied as starting fuel to realize transition to Th-U fuel cycle in graphite moderated molten salt reactors. Due to an inevitable introduction of large amount of fertile element U-238, however, this transition way has proven to be almost impossible. Heavy water moderated molten salt reactor (HWMSR) is an innovative molten salt reactor adopting heavy water as the moderator and heavy metal elements dissolved in fluoride salt as the fuel. A high neutron economy resulting from heavy water paves the way for the transition from LEU to thorium fuel cycle. The results obtained from this study indicated that transition can be achieved by using LEU with enrichment no lower than 5%. Raising enrichment would also enhance the breeding performance, but exert negative/positive influences on the attractiveness of uranium/plutonium contained in the fuel salt, decrease the heavy metal utilization efficiency and impose a small impact on the fuel cycle cost.

Ameliorating the positive temperature feedback coefficient for an MSR fueled with transuranic elements

Transuranic (TRU) elements can be used as an alluring fuel in a graphite-moderated Molten Salt Reactor (MSR) with liquid fuel and online reprocessing. However, the positive temperature feedback coefficient (TFC) stemming from the resolvable fission resonances of Pu-239 and Pu-241 at thermal neutron energies should be addressed. The aim of this study is to propose an effective solution to ameliorate the positive TFC, by adding Er-167 into the fuel salt or adjusting the minor actinides (MA) ratio in TRU. Er-167, with a strong capture resonance cross section of about 10,000 barns at 0.5 eV, is one of the most promising poisons. After adding Er-167 into the fuel salt with hectogram scale, the fuel TFC is decreased significantly to below −10 pcm/K and the total TFC to −2 pcm/K correspondingly. Similar to Er-167, some MA isotopes, especially Np-237 (at 0.5 eV) and Am-241 (at 0.3 eV, 0.6 eV and 1.1 eV) reveal significant capture resonance cross sections (>1000 barns). In this way, increasing the MA ratio in TRU is another available approach to obtain a negative TFC. The simulated results show that adjusting the MA ratio up to 50% can ameliorate the TFC efficiently. Combing the negative TFC and the salt solubility limit, only a narrow region of the fuel volume fraction lower than 25% can meet the two requirements, and the eligible region becomes smaller for a lower MA ratio.

Influence of 235U enrichment on the moderator temperature coefficient of reactivity in a graphite-moderated molten salt reactor

To optimize the temperature coefficient of reactivity (TCR) for a graphite-moderated and liquid-fueled molten salt reactor, the effects of fuel salt composition on the fuel salt temperature coefficient of reactivity (FSTC) were investigated in our earlier work. In this study, we aim to provide a more comprehensive analysis of the TCR by considering the effects of the graphite-moderator temperature coefficient of reactivity (MTC). The effects of $$^{235}$$U enrichment and heavy metal (HM) proportion in the salt mixture on the MTC are investigated from the perspective of the six-factor formula based on a full-core model. For the MTC (labeled “$$\alpha _{{\mathrm{TM}}}$$”), the temperature coefficient of the fast fission factors ($$\alpha _{{\mathrm{TM}}}(\varepsilon )$$) is positive, while those of the resonance escape probability ($$\alpha _{{\mathrm{TM}}}(p )$$), the thermal reproduction factor ($$\alpha _{{\mathrm{TM}}}(\eta )$$), the thermal utilization factor ($$\alpha _{{\mathrm{TM}}}(f )$$), and the total non-leakage probability ($$\alpha _{{\mathrm{TM}}}(\varLambda )$$) are negative. The results reveal that the magnitudes of $$\alpha _{{\mathrm{TM}}}(\varepsilon )$$ and $$\alpha _{{\mathrm{TM}}}(p )$$ for the MTC are similar. Thus, variations in the MTC with $$^{235}$$U enrichment for different HM proportions are mainly dependent on $$\alpha _{{\mathrm{TM}}}(\eta )$$, $$\alpha _{{\mathrm{TM}}}(\varLambda )$$, and $$\alpha _{{\mathrm{TM}}}(f )$$, but especially on the former two. To obtain a more negative MTC, a lower HM proportion and/or a lower $$^{235}$$U enrichment is recommended. Together with our previous studies on the FSTC, a relatively soft neutron spectrum could strengthen the TCR with a sufficiently negative MTC.

Research on the effect of the heavy nuclei amount on the temperature reactivity coefficient in a small modular molten salt reactor

Small modular thorium-based graphite-moderated molten salt reactors (smTMSRs), which combine the advantages of small modular reactors and molten salt reactors, are regarded as a wise development path to speed deployment time. In a smTMSR, low enriched uranium and thorium fuels are used in once-through mode, which makes a marked difference in their neutronic properties compared with the case when a conventional molten salt breeder reactor is used. This study investigated the temperature reactivity coefficient (TRC) in a smTMSR, which is mainly affected by the molten salt volume fraction (VF) and the heavy nuclei concentration in the fuel salt (HN). The four-factor formula method and the reaction rate method were used to indicate the reasons for the TRC change, including the fuel density effect, the fuel Doppler effect, and the graphite thermal scattering effect. The results indicate that only the fuel density has a positive effect on the TRC in the undermoderated region. Thermal scattering from both salt and graphite has a significant negative influence on the TRC in the overmoderated region. The maximal effective multiplication factor, which shows the highest fuel utilization, is located at 10% VF and 12 mol% HN and is still located in the negative TRC region. In addition, on increasing the heavy nuclei amount from 2 mol% HN to 12 mol% HN (VF = 10%), the total TRC undergoes an obvious change from − 11 to − 3 pcm/K, which implies that the change in the HN caused by the fuel feed online should be small to avoid potential trouble in the reactivity control scheme.

Numerical analysis on the dynamic behaviors of a graphite-moderated molten salt reactor based on MOREL2.0 code

Molten salt reactors (MSRs) are selected as one of six generation-IV nuclear power systems due to its remarkable advantages in terms of inherent safety, nuclear non-proliferation and economy. Liquid fuel makes the neutronics and thermal-hydraulics characteristics of MSRs different from those of conventional solid fuel reactors. In this study, transients perturbed by control rod ejection, local perturbation and over feeding U235 were analyzed for a graphite-moderated MSR based on a 3D coupled neutronics/thermal hydraulics code MOREL2.0. The effective delayed neutron fraction decreased with the rise of fuel mass flow. The core power changed more fiercely and even occurred supercritical phenomenon for larger fuel mass flow when the same reactivity was introduced. Local blockage introduced negative reactivity, and MSR maintained a safety state. While local overcooling resulted in positive reactivity introduction, and core power increased first then tended to be stable due to the negative temperature feedback. Furthermore, the response of core power at the transient of over feeding fissile isotopes was simulated. Those numerical results provide valuable information for the research and design of this new generation reactor.

Possible scenarios for the transition to thorium fuel cycle in molten salt reactor by using enriched uranium

Molten Salt Reactor (MSR) with Th-233U fuel cycle attracts more and more attention with its excellent performance such as desirable breeding capacity, low waste production and high inherent safety. Considering the fact that there is no available 233U in the nature, it is necessary to analyze the fuel transition from enriched 235U/Th to 233U/Th and then give a flexible transition scenario for a graphite-moderated MSR. By employing an in-house developed tool which is based on SCALE6.1, two scenarios, a Breeding and Burning (B&B) scenario and a Pre-breeding scenario, are studied. The evolution of the inventories of main nuclides, net 233U production and isothermal temperature coefficient are presented and discussed in the B&B scenario. It is found that the fuel transition can be achieved smoothly by using enriched uranium with greater than 40% concentration of 235U. The fuel transition can still be accomplished with 20% enriched uranium but takes a long double time of about 79 years. Meanwhile, we perform an analysis of the Pre-breeding scenario and conclude that it is efficient to produce 233U and the double time ranges from 2.07 years for the 10-day reprocessing to 10.7 years for the 180-day reprocessing. A comparison of these two scenarios is conducted, which indicates that the B&B scenario is more favorable than the Pre-breeding scenario from the aspect of resource utilization efficiency. Finally, a combined three-stage program for developing Th-based MSRs is proposed.

Lattice optimization for graphite moderated molten salt reactors using low-enriched uranium fuel

The family of Generation-IV reactor concepts comprises of multiple promising designs, one of which is the molten salt reactor. These reactors have traditionally been chosen for the possible use of the thorium-uranium fuel cycle and used 7LiF-BeF2 carrier salt. This particular salt choice however presents several challenges due to the cost of highly depleted 7Li isotope for the carrier salt, tritium production, and beryllium toxicity. Additionally, lack of developed and accepted safeguards methodology for thorium fuel cycle presents a barrier. While none of these issues are insurmountable, alternatives are worth investigating. The purpose of this paper is to analyze the more cost effective and regulatory amenable fuel salt choices by using low-enriched uranium fuel in the form of UF4. Several eutectic mixtures are examined that avoid the use of 7Li and Be while maintaining a melting point low enough to be compatible with standard structural materials. The optimal conditions for hexagonal lattice arrangements using nuclear graphite moderation are discussed for multiple fuel salt choices. The aim of this study is to present options of using simpler molten salt reactor alternatives focused on thermal single-fluid low-enriched uranium converter concepts.

Parametric Lattice Study of a Graphite-Moderated Molten Salt Reactor

Molten salt reactors (MSRs) are promising advanced nuclear reactors for closure of the fuel cycle. This paper discusses the core design of graphite-moderated MSRs, thanks to a parametric study of the fuel and moderator lattice. The study is conducted at equilibrium of the thorium-uranium fuel cycle for several fuel channel radius and moderator block size combinations. The equilibrium composition for each studied configuration is derived with the help of an in-house MATLAB code, EQL0D, which uses the Serpent 2 Monte Carlo neutronics code for the calculation of reaction rates. The results include excess reactivity at equilibrium, mirroring the breeding gain, and the actinide vector composition for each configuration. Moreover, the occurence of an optimum of the excess reactivity per percent uranium-233 was observed. The investigations showed that it is systematically seen at an interchannel distance equal to the neutron slowing-down length in graphite for each configuration and does not depend on the salt channel radius beyond a certain size, which is given by the thermal fission rate reaching the levels of the fast fission rate. In this way, an exotic energy and spatial distribution of the neutrons are attained. The investigations highlight the potential attractiveness, from a neutronics/fuel cycle point of view, of both large fuel channels and moderators with a shorter neutron slowing-down length.

Optimization of temperature coefficient and breeding ratio for a graphite-moderated molten salt reactor

Molten salt reactor (MSR) has fascinating features: inherent safety, no fuel fabrication, online fuel reprocessing, etc. However, the graphite moderated MSR may present positive feedback coefficient which has severe implications for the transient behavior during operation. In this paper, the feedback coefficient and the breeding ratio are optimized based on the fuel-to-graphite ratio variation for a thorium based MSR (TMSR). A certain thermal core with negative feedback coefficient and relative high initial breeding ratio is chosen for the reprocessing scheme analysis. The breeding performances for the TMSR under different online fuel reprocessing efficiencies and frequencies are evaluated and compared with other MSR concepts. The results indicate that the thermal TMSR can get a breeding ratio greater than 1.0 with appropriate reprocessing scheme. The low fissile inventory in thermal TMSR leads to a short doubling time and low transuranic (TRU) inventory. The lifetime of graphite used for the TMSR is also discussed.

New breeding gain definitions and their application to the optimization of a molten salt reactor design

The molten salt reactor (MSR) is an attractive breeder reactor. A graphite-moderated MSR can reach breeding because of the online salt processing and refueling. These features give difficulties when the breeding gain (BG) of the MSR is evaluated. The inventory of the core and external stockpiles have to be treated separately in order to quantify the breeding performance of the reactor. In this paper, an improved BG definition is given and it is compared with definitions used earlier. The improved definition was used in an optimization study of the graphite – salt lattice of the core. The aim of the optimization is a passively safe, self-breeder reactor. The fuel channel diameter, graphite volume and thorium concentration were varied while the temperature feedback coefficient of the core, BG – as defined in the paper – and the lifetime of the graphite were calculated. There is a small range of lattices which provide both negative temperature feedback and breeding. Furthermore, breeding is possible only at low power densities in case of the salt processing efficiencies set in this study. In this range of power the lifetime of the graphite is between 12 and 20 years.

An approach for the modelling and the analysis of the MSR thermo-hydrodynamic behaviour

In the last years, there has been a rapid growth of research and development activities on high temperature molten salts for nuclear and non-nuclear applications. The study of their heat transfer characteristics is a key issue in the current development of the molten salt reactor (MSR) that is one of the innovative nuclear reactors proposed by the Generation IV International Forum. The MSR is a sort of circulating fuel reactor (CFR), which adopts a molten halide salt mixture containing the fissile material and playing the distinctive role of both fuel and coolant. In a thermal-neutron-spectrum MSR, the reactor core is composed by a graphite matrix (neutron moderator), through which the liquid nuclear fuel flows and leads to a strong and intrinsic coupling between thermo-hydrodynamics and neutronics. This peculiar feature requires a suitable and qualified multi-physics simulation environment for a proper description of the system (fuel/coolant+graphite) behaviour. With reference to such complex and non-linear system, the present work is intended to give two different but linked contributions, in the perspective of a multi-physics modelling able to accurately describe the synergy of the involved different physical phenomena (e.g., by means of COMSOL Multiphysics ®, 2008 software). (I) On one hand, this work is aimed at establishing a useful validation framework for the assessment of computational fluid-dynamics (CFD) analyses of liquids with internal heat generation. For this purpose, an analytic approach for both fluid velocity and temperature fields in a circular pipe surrounded by a solid region has been developed. It is an extension of the approach elaborated for pipe flow in ( Di Marcello, V., Cammi, A., Luzzi, L., 2010. A generalized approach to heat transfer in pipe flow with internal heat generation. Chemical Engineering Science 65, 1301−1310), in order to take into account the heat conduction in the solid domain (represented by the graphite matrix in the specific case of interest). (II) On the other hand, the paper presents a preliminary investigation of the thermal-hydraulic behaviour occurring in a typical graphite-moderated MSR core channel. The developed analytic approach has been conveniently applied to this case study: (i) by testing the capabilities of COMSOL Multiphysics ® (2008) to evaluate the heat transfer characteristics and the hydrodynamic behaviour of such system; and (ii) by investigating the applicability of correlations for the Nusselt number to fluids with internal heat generation. For a deeper insight into the numerical solutions provided by COMSOL, a code-to-code comparison has been also carried out, adopting a dedicated CFD finite volume software (i.e., FLUENT ®). As a result, a satisfactory agreement has been found between the analytic solution and the numerical computations provided by COMSOL and FLUENT for a wide range of Reynolds numbers. As concerns the Nusselt number evaluation, the correlations usually adopted for molten salts have been proved as unsuitable for the analysed system, whereas the correlation advanced in the above mentioned paper (that explicitly takes account of the internal heat source) appears more appropriate. The results of the present work are thought to be useful because they provide: (i) a “generalized” analytic approach to the heat transfer that is applicable in a more general context; (ii) an insight into the heat transfer characteristics of the considered graphite-moderated MSR core channel; and (iii) a preliminary assessment of computations that is essential in view of the adoption of multi-physics tools (like COMSOL) for more complex and representative simulations of the dynamic behaviour of the molten salt reactor, and more in general of other kinds of CFRs.