Transition toward thorium fuel cycle in a molten salt reactor by using plutonium

The molten salt reactor(MSR), as one of the Generation Ⅳ advanced nuclear systems, has attracted a worldwide interest due to its excellent performances in safety, economics, sustainability, and proliferation resistance. The aim of this work is to provide and evaluate possible solutions to fissile 233 U production and further the fuel transition to thorium fuel cycle in a thermal MSR by using plutonium partitioned from light water reactors spent fuel. By using an in-house developed tool, a breeding and burning(BB) scenario is first introduced and analyzed from the aspects of the evolution of main nuclides, net 233 U production, spectrum shift, and temperature feedback coefficient. It can be concluded that such a Th/Pu to Th/~(233)U transition can be accomplished by employing a relatively fast fuel reprocessing with a cycle time less than 60 days. At the equilibrium state, the reactor can achieve a conversion ratio of about 0.996 for the 60-day reprocessing period(RP) case and about 1.047 for the 10-day RP case.The results also show that it is difficult to accomplish such a fuel transition with limited reprocessing(RP is 180 days),and the reactor operates as a converter and burns the plutonium with the help of thorium. Meanwhile, a prebreeding and burning(PBB) scenario is also analyzed briefly with respect to the net 233 U production and evolution of main nuclides. One can find that it is more efficient to produce 233 U under this scenario, resulting in a double time varying from about 1.96 years for the 10-day RP case to about 6.15 years for the 180-day RP case.     

Properties of phosphate glass waste forms containing fluorides from a molten salt reactor

Radioactive fluoride wastes are generated during the operation of molten salt reactors(MSRs) and reprocessing of their spent fuel.Immobilization of these wastes in borosilicate glass is not feasible because of the very low solubility of fluorides in this host.Alternative candidates are thus an active topic of research including phosphatebased glasses,crystalline ceramics,and hybrid glass-ceramic systems.In this study,mixed fluorides were employed as simulated MSRs waste and incorporated into sodium aluminophosphate glass to obtain phosphate-based waste form.These waste forms were characterized by X-ray diffraction,Raman spectroscopy,and scanning electron microscopy.Leaching tests were performed in deionized water using the product consistency test A method.This study demonstrates that up to 20 mol%of simulated radioactive waste can be introduced into the NaA1 P glass matrix,and the chemical durability is much better than that of borosilicate.The addition of Fe_2O_3 in the NaAlP glass matrix results in increases of the chemical durability at the expense of fluoride loading(to 6.4 mol%).Phosphate glass vitrification of radioactive waste containing fluorides is a potential method to treat and dispose of MSR wastes.     

小型模块化熔盐快堆燃料管理初步分析

由于燃料随熔盐流动的特性以及可以进行在线添料与处理的特点,液态燃料熔盐堆的燃耗分析与燃料管理和传统固态燃料反应堆有很大不同,需要针对液态燃料熔盐堆的特点重新开发燃耗分析与管理程序。本文针对液态燃料熔盐堆的熔盐流动特性以及在线添料与处理功能,基于MCNP5和ORIGEN2.1燃耗耦合程序,开发了适用于液态燃料熔盐堆的燃料管理程序,并应用于一种小型模块化熔盐快堆的燃料管理和分析,对比分析了5种不同运行方案以及分批在线添料情况下,运行30年期间keff的变化情况及重要核素的演化情况。计算结果表明,采用不断调整添料率的连续在线添料运行方案和固定批量添料的运行方案,都可以让小型模块化熔盐快堆维持运行在一个较小的keff波动范围之内。开发的燃料管理程序适用于液态燃料熔盐堆的研究,同时可以为液态燃料熔盐堆的设计及燃耗管理和分析提供有价值的参考。     

Analysis of heat source distribution in internal circulating surface heat transfer molten salt reactor

The Molten Salt Reactor (MSR) is one of the Generation IV nuclear reactor concepts that were selected by the Generation IV International Forum in 2000. The concept is based on liquid fuel instead of solid fuel assemblies. Besides the advantages, there are several aspects of operation that can hinder the realization of this reactor concept. In this paper, the authors investigate the neutronics behaviour of a new sub-concept that offers solutions for many of the technical problems. The analysis was performed using the particle transport code MCNPX 2.7. The paper focuses on the short-term and steady state heat source distribution in the fuel salt and in the graphite moderator. Accordingly, neither burn-up effects nor reactivity transients are considered. The sensitivity of the effective multiplication factor on different geometrical and material parameters was studied. The results obtained indicate that the main region of heat deposition is in the internal and external channels of the graphite moderator. Only a few percent of the total heat power is released in the graphite moderator, where the gamma and neutron related heat deposition is on the same scale. The results also prove that the heat source distribution does not change drastically upon the actuation of the control rods.     

Development of a MCNP5 and ORIGEN2 based burnup code for molten salt reactor

The Molten Salt Reactor(MSR) is one of the six advanced reactor nuclear energy systems for further research and development selected by Generation IV International Forum(GIF),which is distinguished by its core in which the fuel is dissolved in molten fluoride salt.Because fuel flow in the primary loop,the depletion of MSR is different from that of solid-fuel reactors.In this paper,an MCNP5 and ORIGEN2 Coupled Burnup(MOCBurn) code for MSR is developed under the MATLAB platform.Some new methods and novel arrangements are used to make it suitable for fuel flow in the MSR.To consider the fuel convection and diffusion in the primary loop of MSR,fuel mixing calculation is carried out after each burnup time step.Modeling function for geometry with repeat structures is implicated for reactor analysis with complex structures.Calculation for a high-burnup reactor pin cell benchmark is performed using the MOCBurn code.Results of depletion study show that the MOCBurn code is suitable for the traditional solid-fuel reactors.A preliminary study of the fuel mixture effect in MSR is also carried out.     

Assembly-level analysis on temperature coefficient of reactivity in a graphite-moderated fuel salt reactor fueled with low-enriched uranium

To provide a reliable and comprehensive data reference for core geometry design of graphite-moderated and low-enriched uranium fueled molten salt reactors,the influences of geometric parameters on the temperature coefficient of reactivity(TCR)at an assembly level were characterized.A four-factor formula was introduced to explain how different reactivity coefficients behave in terms of the fuel salt volume fraction and assembly size.The results show that the fuel salt temperature coefficient(FSTC)is always negative owing to a more negative fuel salt density coefficient in the over-moderated region or a more negative Doppler coefficient in the under-moderated region.Depending on the fuel salt channel spacing,the graphite mod-erator temperature coefficient(MTC)can be negative or positive.Furthermore,an assembly with a smaller fuel salt channel spacing is more likely to exhibit a negative MTC.As the fuel salt volume fraction increases,the negative FSTC first weakens and then increases,owing to the fuel salt density effect gradually weakening from negative to positive feedback and then decreasing.Meanwhile,the MTC weakens as the thermal utilization coefficient caused by the graphite temperature effect deteriorates.Thus,the negative TCR first weakens and then strengthens,mainly because of the change in the fuel salt density coefficient.As the assembly size increases,the magnitude of the FSTC decreases monotonously owing to a monotonously weakened fuel salt Doppler coefficient,whereas the MTC changes from gradually weakened negative feedback to gradually enhanced positive feedback.Then,the negative TCR weakens.Therefore,to achieve a proper negative TCR,particularly a negative MTC,an assembly with a smaller fuel salt channel spacing in the under-moderated region is strongly recommended.     

Optimization of flux distribution in a molten-salt reactor with a 2-region core for plutonium burning

Molten-salt reactors (MSRs) are selected as one of the candidates of Generation IV reactor concepts. In GLOBAL2005 held in Tsukuba, Japan, one paper discussed the flattening of fast neutron flux in the core for a longer life of graphite moderator. In the paper a 3-region reactor concept was presented. The authors tried many cores changing configurations such as volume of each region and fractions of fuel salt in the regions or fuel compositions.

We investigated the other possibility of a 2-region core for the simplicity. Using one energy group neutron diffusion theory and considering extrapolation distance, the optimum selection of region wise neutron multiplication factors can be theoretically and easily obtained. In MSRs, there is no burnup distribution of the fuel. The region wise neutron multiplications can be obtained by adjusting the volume fraction of fuel in a cell with a given composition of the fuel salt. Using the theoretical results, the optimization of the actual core configuration was determined by a nuclear analysis code SRAC2002 with the nuclear data library of JENDL3.3.

In this paper, we considered MSRs using plutonium as a fissile material. Ordinary MSRs use uranium-233, which doesn't exist naturally, and utilizing plutonium is easier to startup.     

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(sm TMSRs), 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 sm TMSR, 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 sm TMSR, which is mainly affected by the molten salt volume fraction(VF) and the heavy nuclei concentration in the fuel salt(HN). The fourfactor 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 to12 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.     

Influence of ~(235)U 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 ‘‘a_(TM)'), the temperature coefficient of the fast fission factors(a_(TM)eeT) is positive, while those of the resonance escape probability(a_(TM)epT), the thermal reproduction factor(a_(TM)egT), the thermal utilization factor(a_(TM)ef T), and the total non-leakage probability(a_(TM)eKT) are negative. The results reveal that the magnitudes of a_(TM)eeT and a_(TM)epT for the MTC are similar. Thus, variations in the MTC with ~(235)U enrichment for different HM proportions are mainly dependent on a_(TM)egT, a_(TM)eKT, and a_(TM)ef T, 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.     

Project SPHINX spent hot fuel incinerator by neutron flux (the development of a new reactor concept with liquid fuel based on molten fluorides)

The SPHINX project is dealing with a solution of some principle problems of a very promising way of nuclear waste treatment, high level wastes from spent nuclear fuel in particular, by means of transmutation of radionuclides by use of a nuclear reactor with liquid fuel based on molten fluorides, which might be a subcritical system driven by a suitable neutron source. Its superiority lies also in the fact that it makes possible to utilize actinides contained, by others, in spent nuclear fuel and so to reach a positive energy effect.

The SPHINX project has been proposed by the consortium TRANSMUTATION being established by four leading nuclear research bodies in the Czech Republic (Nuclear Research Institute Rez plc, SKODA Nuclear Machinery plc in Pilsen, Nuclear Physics Institute of Academy of Sciences in Rez and Technical University in Praha) at the end of 1996 to which Technical University in Brno (specialized for a secondary circuit problems) has associated in the year 2000. The project has been supported by the Ministry of Industry and Trade of the Czech Republic, CEZ, a.s. (Czech Electricity Generating Company) and RAWRA (Radwaste Repository Authority).

The R&D program of the SPHINX project contains an experimental part, which serves for a verification of design inputs for designing a demonstration unit of a transmuter with liquid fuel based on molten fluorides. The current status of the experimental program performance has been focused upon the irradiation of samples of molten-salt systems as well as structural materials proposed for the blanket of the SPHINX transmuter in the field of high neutron flux of research reactors.

The main aims of this program called Irradiated Probes BLANKA are the following: (1) Experimental verification of long time behavior of transmuter blanket which contains molten fluoride salts as a fuel and coolant, (2) Validation of computational code system being developed for the computation of actinides concentration in long- term operation of the transmuter, and (3) Material research on behavior of materials in neutron and gamma fields, and materials interactions on high temperature conditions.

At present, two agreements on multinational cooperation in this field have been signed: One with European Commission and one with Russian Kurchatov Institute (joint experimental programs AMPULA containing fluorides of transuranium elements like Np, Pu, Am and Cm in irradiated samples and a joint development of the ISTAR code).     

Effects of molten material temperatures and coolant temperatures on vapor explosion

An observable experiment facility for low-temperature molten materials to be dropped into water was set up in this study to investigate the mechanism of the vapor explosion. The effect of the fuel and coolant interaction (FCI) on the vapor explosion during the severe accidents of a fission nuclear reactor has been studied. The experiment results showed that the molten material temperature has an important effect on the vapor explosion behavior and pressure. The increase of the coolant temperature would decrease the pressure of the vapor explosion.     

Analysis of Cell Configuration Effects on Nuclear Characteristics in a Lead-Bismuth Cooled Core for TRU Incineration

Sensitivity of the core characteristics to the fuel pin cell parameters change is analyzed for a lead-bismuth cooled reactor to incinerate transuranic nuclides. The pitch-to-diameter ratio is changed for a parametric study to investigate the effects of the coolant-to-fuel ratio. Not only the Zr-based fuel of TRU+Zr but also the Th-based fuel of TRU+Th+Zr is considered in order to investigate the sensitivity of nuclear characteristics of the fuel pin cell to neutron energy spectrum as well as effects of the fuel type on the core performance. For the sensitivity analyses, the neutron spectrum, the criticality performance parameters, and the non-fissile actinides destruction factor are evaluated. The obtained results clarify the unique property of nuclear characteristics of the fuel pin cell and give some useful information for design optimization of a lead-bismuth cooled reactor for TRU transmutation.     

Effects of Coolant Flow on Light Water Reactor Fuel Behaviors during Reactivity Initiated Accident

This paper describes the in-pile experimental results to study the influences of coolant flow on fuel behaviors under reactivity initiated accident (RIA) conditions performed in the Nuclear Safety Research Reactor (NSRR). A single PWR type test fuel rod was irradiated by a large neutron pulse in the NSRR to simulate a prompt power excursion of RIA's. The effects of coolant flow were studied at a coolant flow velocity of 0.3~1.8m/s and a coolant temperature of 20~90°C under the atmospheric pressure. It was found that the cooling conditions had considerable influences on fuel thermal behaviors under prompt heat-up. The increase of coolant flow velocity and subcooling enhanced heat transfer coefficient at cladding surface during film boiling, which resulted in large decrease of maximum cladding temperature and film boiling duration, and consequently in the increase of fuel failure threshold energy. The data tendencies are summarized and the influences of coolant flow are discussed with some computer analyses.     

Benchmark Analysis of Criticality Experiments in the TRIGA Mark II Using a Continuous Energy Monte Carlo Code MCNP

The criticality analysis of the TRIGA-II benchmark experiment at the Musashi Institute of Technology Research Reactor (MuITR, 100kW) was performed by the three-dimensional continuous-energy Monte Carlo code (MCNP4A). To minimize errors due to an inexact geometry model, all fresh fuels and control rods as well as vicinity of the core were precisely modeled. Effective multiplication factors (k_eff) in the initial core critical experiment and in the excess reactivity adjustment for the several fuel-loading patterns as well as the fuel element reactivity worth distributions were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated k_eff overestimated the experimental data by about 1.0Δk/k for both the initial core and the several fuel-loading arrangements (fuels or graphite elements were added only to the outer-ring), but the discrepancy increased to 1.8Δk/k for the some fuel-loading patterns (graphite elements were inserted into the inner-ring). The comparison result of the fuel element worth distribution showed above tendency. All in all, the agreement between the MCNP predictions and the experimentally determined values is good, which indicates that the Monte Carlo model is enough to simulate criticality of the TRIGA-II reactor.     

Benchmark Analysis of TRIGA Mark II Reactivity Experiment Using a Continuous Energy Monte Carlo Code MCNP

The benchmark analysis of reactivity experiments in the TRIGA-II core at the Musashi Institute of Technology Research Reactor (Musashi reactor, 100 kW) was performed by a three-dimensional continuous-energy Monte Carlo code MCNP4A. The reactivity worth and integral reactivity curves of the control rods as well as the reactivity worth distributions of fuel and graphite elements were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated values of integral reactivity curves of the control rods were in agreement with the experimental data obtained by the period method. The integral worth measured by the rod drop method was also consistent with the calculation. The calculated values of the fuel and the graphite element worth distributions were consistent with the measured ones within the statistical error estimates. These results showed that the exact core configuration including the control rod positions to reproduce the fission source distribution in the experiment must be introduced into the calculation core for obtaining the precise solution. It can be concluded that our simulation model of the TRIGA-II core is precise enough to reproduce the control rod worth, fuel and graphite elements reactivity worth distributions.