Plutonium Burning Research Articles

Burning of Reactor Grade Plutonium Mixed with Thorium in a Hybrid Reactor

Significant amount of plutonium have been discharged and accumulated from the conventional LWRs and CANDU reactors. Reducing this reactor grade (RG) plutonium is very important because it may be misused and/or released accidentally into the environment. Fusion-fission (hybrid) reactors have strong potential on burning plutonium effectively. This study presents the burning of RG plutonium mixed with thorium in a hybrid reactor for an operation period of 24 months. The effect of various fuel mixtures (98% ThO2 + 2% RG-PuO2, 94% ThO2 + 6% RG-PuO2 and 90% ThO2 + 10% RG-PuO2) and coolants (Flinabe, natural lithium and Li20Sn80) on the reactor’s performance was investigated. Numerical results showed that utilization of RG plutonium in the mixed fuel in such a hybrid reactor not only enhanced the reactor’s performance but also reduced its 239Pu content significantly.

Incineration of weapon grade plutoniumin a (DT) fusion driven hybrid reactor using various coolants

Abstract Reducing separated weapon grade (WG) plutonium is very crucial since this plutonium may be misused and/or released accidentally into the environment. Burning WG plutonium in nuclear reactors can help to solve this serious problem effectively. In this study, the incineration of WG plutonium mixed with thorium in a (DT) fusion driven high power density hybrid reactor was investigated for an operation period of 24 months. Furthermore, the effect of using different coolants on the burning of WG plutonium in the hybrid blanket was examined. Numerical calculations showed that WG plutonium can be denatured to reactor grade level within a very short time by using a high power density hybrid reactor. Moreover, addition of WG plutonium to thorium improved the neutronic performance of the reactor significantly.

Neutronic double heterogeneity effect in particle dispersed type inert matrix fuels

Rock-like oxide (ROX) fuel concept is studied in Japan for effective plutonium burning in light water reactors (LWRs). ROX is a heterogeneous fuel, where Pu containing yttria stabilized zirconia (YSZ) particles are dispersed in spinel matrix, and similar to the high temperature gas cooled reactor (HTR) fuel. The effect of such a ‘double’ heterogeneity (fuel, structure and coolant heterogeneity in reactor core, plus fuel heterogeneity) on HTR neutronic characteristics is important, while the effect was not taken into account in the ROX fueled LWR neutronics calculations. Here, this double heterogeneity effect is estimated for ROX fueled LWR, and compared with the Pu containing YSZ particle fueled HTR. As a result, the heterogeneity effect was negligible in the ROX–LWR system, while it is notable in YSZ–HTR system. The volume fraction of YSZ particle in the fuel region is one of the important parameter to cause the difference.

Use of Thorium in Light Water Reactors

Thorium-based fuels can be used to reduce concerns related to the proliferation potential and waste disposal of the conventional light water reactor (LWR) uranium fuel cycle. The main sources of proliferation potential and radiotoxicity are the plutonium and higher actinides generated during the burnup of standard LWR fuel. A significant reduction in the quantity and quality of the generated Pu can be achieved by replacing the 238U fertile component of conventional low-enriched uranium fuel by 232Th. Thorium can also be used as a way to manage the growth of plutonium stockpiles by burning plutonium, or achieving a net-zero transuranic production, sustainable recycle scenario. This paper summarizes some of the results of recent studies of the performance of thorium-based fuels.It is concluded that the use of heterogeneous U-Th fuel provides higher neutronic potential than a homogeneous fuel. However, in the former case, the uranium portion of the fuel operates at a higher power density, and care is needed to meet the thermal margins and address the higher-burnup implications. In macroheterogeneous designs, the U-Th fuel can yield reduced spent-fuel volume, toxicity, and decay heat. The main advantage of Pu-Th oxide over mixed oxide is better void reactivity behavior even for undermoderated designs, and increased burnup of Pu.

Thermodynamic analysis of Np–Zr–H, Am–Zr–H, Pu–Zr–H systems

High-level radioactive wastes generated after reprocessing spent nuclear fuels from nuclear reactors include long-lived and highly toxic radioactive nuclides. Recently, the transmutation method using actinide hydride targets in fast reactor has been proposed to reduce the amount of actinide in nuclear wastes. The burning of Pu using Pu–Zr–H fuel in light water reactor has also been proposed. Given the potential for high-temperature state during routine or extraordinary operating circumstances, accurate knowledge or reliable predictions of materials behaviour under a variety of conditions is required in order to develop safe and effective nuclear handling and remediation technologies. Semi-empirical Miedema and Griessen–Driessen models on the enthalpy of formation coupling with the available experimental entropy information were utilized to estimate the thermodynamic stability of the selected actinide–zirconium hydrides. The calculations showed that at high temperature, stable hydrides might be expected to form for the Np–Zr–H, Am–Zr–H and Pu–Zr–H systems at selected compositions, which can be effectively used for the MA transmutation or plutonium burning.

Advanced fuels for plutonium management in pressurized water reactors

Several fuel concepts are under investigation at CEA with the aim of manage plutonium inventories in pressurized water reactors. This options range from the use of mature technologies like MOX adapted in the case of MOX-EUS (enriched uranium support) and COmbustible Recyclage A ILot (CORAIL) assemblies to more innovative technologies using IMF like DUPLEX and advanced plutonium assembly (APA). The plutonium burning performances reported to the electrical production go from 7 to 60 kg (TW h) −1. More detailed analysis covering economic, sustainability, reliability and safety aspects and their integration in the whole fuel cycle would allow identifying the best candidate.

Analysis of the CIRANO Experiment for Plutonium Burning Fast Reactor using JNC’s Calculation Scheme

The French CIRANO experiments provide an integral database for plutonium burning fast reactor physics. A series of fuel substitution measurements parameterized by plutonium isotopic vector and enrichment was analyzed by using JNC’s standard calculation scheme with the evaluated nuclear data JENDL-3.2. A cross-section adjustment technique was employed for checking the consistency of the results and for deducing quantitative trends or errors in nuclear data, especially for the higher-mass plutonium 240Pu.

Categorisation and priorities for future research on inert matrix fuels: an extended synthesis of the panel discussions

There are three categories of basic fuel cycle needs, which are being addressed by the different types of inert matrix fuel (IMF) concepts currently under development. These are: plutonium burning in existing LWRs, plutonium burning in fast reactors and minor actinide transmutation - corresponding to three distinct timescales for perceived IMF implementation, viz. short, medium and long term, respectively. The current paper, based partly on the two panel discussions organised at the 6th IMF workshop, presents viewpoints and priorities for each of the three categories of IMF applications, both in terms of the fuel concepts to be pursued and the corresponding RD requirements.

Experiments and Analyses on Sodium Void Reactivity Worth in Uranium-Free Fast Reactor at FCA

A uranium-free fast reactor was simulated at FCA in order to examine the prediction accuracy for sodium void effect of plutonium burning fast reactors. Material sample worth for plutonium and B4C was also measured to compare its prediction accuracy with that for the sodium void reactivity worth. It was found that an axial distribution of plutonium sample worth and the central B4C sample worth for various kinds of 10B enrichment were precisely calculated by the conventional calculation method for fast reactors with 70-group structure. The sodium void reactivity worth was, however, poorly predicted especially for the non-leakage term. This discrepancy seems to be caused by the peculiar energy breakdown of the non-leakage term in the uranium-free fast reactor.

Studies on Plutonium Burning in the Prototype Fast Breeder Reactor Monju

Studies have been performed on plutonium burning in the Japanese prototype fast breeder reactor Monju. The main aims of these studies were to illustrate the plutonium-burning capabilities of fast reactors and to investigate the consequences of the related core design measures on the main core characteristics of Monju. Burner cores with diluting pins, with diluting subassemblies (also called diluents), and with an internal slice of inert material have been investigated; these require an increased average plutonium enrichment and thus offer an enhanced plutonium-burning rate. On the other hand, the consequences of the elimination of the radial and/or axial blanket have been investigated.Among the burner concepts, the B4C containing diluents have been found to be preferable because they cause the smallest maximum linear rating increase and offer the largest flexibility to adapt their reactivity via a modification of the B4C content. They also do not require a new fuel subassembly concept.For the case of the blanket elimination, the replacement of the blankets by steel reflectors has been found to be the best solution. The main consequence of the elimination of both blankets is the increase of the maximum linear rating by up to 11%. Whether this increase may lead to problems will depend on the actual linear power level of the core.

Fast Burner Reactor Devoted to Minor Actinide Incineration

This study proposes a new fast reactor core concept dedicated to plutonium and minor actinide burning by transmutation. This core has a large power level of ~1500 MW(electric) favoring the economic aspect. To promote plutonium and minor actinide burning as much as possible, total suppression of 238U, which produces 239Pu by conversion, and large quantities of minor actinides in the core are desirable. Therefore, the 238U-free fuel is homogeneously mixed with a considerable quantity of minor actinides.From the safety point of view, both the Doppler effect and the coolant (sodium) void reactivity become less favorable in a 238U-free core. To preserve these two important safety parameters on an acceptable level, a hydrogenated moderator separated from the fuel and nuclides, such as W or 99Tc, is added to the core in the place of 238U. Tungsten and 99Tc have strong capture resonances at appropriate energies, and 99Tc itself is a long-lived fission product to be transmuted with profit.This core allows the achievement of a consumption rate of ~100 kg/TW(electric)·h of transuranic elements, ~70 kg/TW(electric)·h for plutonium (due to 238U suppression), and 30 to 35 kg/TW(electric)·h for minor actinides. In addition, ~14 kg/TW(electric)·h of 99Tc is destroyed when this element is present in the core (the initial loading of 99Tc is >4000 kg in the core).The activity of newly designed subassemblies has also been investigated in comparison to standard fast reactor subassemblies (neutron sources, decay heat, and gamma dose rate). Finally, a transmutation scenario involving pressurized water reactors and minor actinide-burning fast reactors has been studied to estimate the necessary proportion of burner reactors and the achievable radiotoxicity reduction with respect to a reference open cycle.

Experiments and Analyses on Sodium Void Reactivity Worth in Uranium-Free Fast Reactor at FCA.

A uranium-free fast reactor was simulated at FCA in order to examine the prediction accuracy for sodium void effect of plutonium burning fast reactors. Material sample worth for plutonium and B4C was also measured to compare its prediction accuracy with that for the sodium void reactivity worth. It was found that an axial distribution of plutonium sample worth and the central B4C sample worth for various kinds of 10B enrichment were precisely calculated by the conventional calculation method for fast reactors with 70-group structure. The sodium void reactivity worth was, however, poorly predicted especially for the non-leakage term. This discrepancy seems to be caused by the peculiar energy breakdown of the non-leakage term in the uranium-free fast reactor.

Development of Multiphase Ceramic-Ceramic Composites for Plutonium Burning

Faster disposition of large stock of plutonium available from both the dismantled weapons and accumulated stock from operation of the civilian nuclear power reactors has become a subject of international interest. Zirconia and/or alumina, magnesia based ceramic-ceramic composites (CCC) are considered as candidate matrix materials for plutonium burning on the basis of stable physical and chemical properties. The major advantage of such composite materials is that they consist of stable mineral and resemble ordinary rocks which are chemically and thermody- namically stable for geological periods even after irradiation to high burn-up. The spent fuel also remains in the form of stable minerals and, as such, suitable for direct disposal as high level waste in geological repositories. The present paper highlights the sintering and characterization of two such systems based on ZrO2-ThO2-CeO2 and Al2O3-MgO-ThO2-CeO2.

Categorisation and priorities for future research on inert matrix fuels: An extended synthesis of the panel discussions

There are three categories of basic fuel cycle needs, which are being addressed by the different types of inert matrix fuel (IMF) concepts currently under development. These are: plutonium burning in existing LWRs, plutonium burning in fast reactors and minor actinide transmutation — corresponding to three distinct timescales for perceived IMF implementation, viz. short, medium and long term, respectively. The current paper, based partly on the two panel discussions organised at the 6 th IMF workshop, presents viewpoints and priorities for each of the three categories of IMF applications, both in terms of the fuel concepts to be pursued and the corresponding R&D requirements.

Inert matrix fuel for the utilisation of plutonium

Utilising fuel resources responsibly, reducing waste volume and emissions as well as conflict potentials within the international community (non-proliferation, energy demand) are among the principles for the judgment of sustainable development. Utilising and burning plutonium in a light water reactor has been shown to be feasible for the disposition of the large amount of excess plutonium produced in today's power reactors and resulting from the disarmament efforts of the super powers. With regard to material technology aspects, efforts have concentrated on the evaluation of fabrication feasibility and on the determination of the physicochemical properties of a single phase zirconium/erbium/plutonium oxide material stabilised as a cubic solution by yttrium for plutonium and minor actinide incineration and transmutation. Due to the absence of uranium as origin for plutonium build-up, such a nuclear fuel is called Inert Matrix Fuel. Irradiation testing with a dedicated experiment in the material test reactor in Halden, Norway, is underway within the framework of the OECD-Halden Reactor Project.

Inert matrix and thoria fuels for plutonium elimination

In our previous studies we analysed the plutonium burning performance of various kinds of fuel, based on mixing plutonium oxide with thorium oxide (TOX), or with inert matrix (IMF), or with inert matrix plus a limited addition (doping) of thorium oxide (TD-IMF). The present paper includes the neutronic analysis of a metal-based fuel variant and of fissile material recycling in TOX fuels. If the recycled fuel is topped with weapons grade plutonium (WG-Pu) as fissile material, it is possible to sustain indefinitely a closed cycle.

First In-Pile Experimental Results of High-Plutonium-Content Oxide Fuel for Plutonium Burning in Fast Reactors

An experimental program focused not only on the study of high-plutonium-content mixed-oxide fuels but also on more advanced ''Pu without U'' fuel concepts has been launched in the framework of the Consommation Accrue de Plutonium dans les RApides (CAPRA) project. First results of the in-pile and out-of-pile behavior of high-plutonium-content fuels with uranium, such as (U{sub 55%},Pu{sub 45%})O{sub 2}, and (U{sub 55%},Pu{sub 40%},Np{sub 5%})O{sub 2}, and also without uranium, such as (Pu{sub 44%},Ce{sub 56%})O{sub 2}, are now available. In particular, the Irradiation a FOrt Pu (IFOP) experiment in the SILOE reactor and the TRAnsmutation and Burning of ActiNides in Triox carrier (TRABANT1) experiments in the High Flux Reactor are presented and the results are analyzed: Up to a burnup of 1.5 at.%, destructive examinations of the IFOP pin have shown that the high-plutonium-content oxide fuel with a large central hole presents the usual global behavior (good pellet integrity, fuel microstructure). The TRABANT1 oxide fuel pin with a 40% Pu and 5% Np content demonstrates that a burnup of 9.5 at.% can be reached without failure by a high-plutonium-content fuel. However, the TRABANT1 pin 1 (oxide pin with 45% Pu), which had run under severe conditions, has failed at {approx}7 at.%more » burnup. Destructive examinations of these pins will give more evidence on the causes of the failure. The low-oxygen-to-metal fuel column of (Pu,Ce)O{sub 2-x} melted, thus confirming the poor conductivity of this fuel.« less

First In-Pile Experimental Results of High-Plutonium-Content Oxide Fuel for Plutonium Burning in Fast Reactors

An experimental program focused not only on the study of high-plutonium-content mixed-oxide fuels but also on more advanced “Pu without U” fuel concepts has been launched in the framework of the Consommation Accrue de Plutonium dans les RApides (CAPRA) project. First results of the in-pile and out-of-pile behavior of high-plutonium-content fuels with uranium, such as (U55%,Pu45%)O2, and (U55%,Pu40%,Np5%)O2, and also without uranium, such as (Pu44%,Ce56%)O2, are now available. In particular, the Irradiation à FOrt Pu (IFOP) experiment in the SILOE reactor and the TRAnsmutation and Burning of ActiNides in Triox carrier (TRABANT1) experiments in the High Flux Reactor are presented and the results are analyzed: Up to a burnup of 1.5 at.%, destructive examinations of the IFOP pin have shown that the high-plutonium-content oxide fuel with a large central hole presents the usual global behavior (good pellet integrity, fuel microstructure). The TRABANT1 oxide fuel pin with a 40% Pu and 5% Np content demonstrates that a burnup of 9.5 at.% can be reached without failure by a high-plutonium-content fuel. However, the TRABANT1 pin 1 (oxide pin with 45% Pu), which had run under severe conditions, has failed at ~7 at.% burnup. Destructive examinations of these pins will give more evidence on the causes of the failure. The low-oxygen-to-metal fuel column of (Pu,Ce)O2-x melted, thus confirming the poor conductivity of this fuel.

Neutronic analysis of U-free inert matrix and thoria fuels for plutonium disposition in pressurised water reactors

Inserting reactor-grade (RG) or weapons-grade (WG) plutonium in uranium-free matrices and burning it in light water reactors (LWRs) is an option gaining a wider consensus in the nuclear community. The main results of our neutronic studies performed in the last few years on this subject are reported. Our attention was mainly concentrated on two kinds of matrices: inert matrix in the form of calcia-stabilised zirconia, and thoria. Both materials are likely to exhibit excellent behaviour under irradiation (already demonstrated for thoria fuels) and high chemical stability. Direct disposal of spent fuel should be made feasible and attractive. A preliminary neutronic analysis was performed on these U-free fuels, imposing the constraint of maintaining the same assembly design and cycle length of a standard enriched-uranium fuel. In particular inert matrix fuel (IMF) showed a high plutonium burning capability, but associated with unacceptable feedback coefficients. Therefore, a whole IMF core results unfeasible, and only a partial core loading is possible. The solution then studied consists in replacing ≈21% of the pins of a standard enriched-U subassembly with IMF pins. Detailed assembly and core calculations were performed. A crucial aspect is the choice of a suitable burnable poison, which has to dampen the power peaks in the different fuel pin types without life penalisation. Among the considered poisons, a thin boron coating on the IMF pellets resulted the only effective one. Preliminary IMF pin cell calculations and the detailed ones gave similar results in terms of burnt plutonium fractions: 90% of fissile and 73% of total plutonium is burnt when RG plutonium is used. The main drawbacks of this fuel are the limited core loading capability and the lack of in-pile technological validation. In the case of Pu–Th fuels, pin cell calculations showed that increasing the plutonia content, decreasing the thoria content, and decreasing the pellet diameter are all possible ways to reach a longer fuel cycle and a higher percentage of burnt plutonium. Attained values for RG–Pu are ⩾80% and >60% for the fissile and total plutonium, respectively. The use of IMF is an effective solution to proliferation concerns, while some concerns remain for thoria fuels because of the production of 233U. This, however, can be eliminated by a small addition of 238U. Long time radiotoxicity is scarcely affected by these fuels with respect to conventional MOX.

Review of the Western European breeder programs

Since the beginning of the nuclear era, breeder reactors were recognized as the best way to make use of uranium resources. In Europe, the early development of the liquid metal fast breeder reactor (LMFBR) was initiated in England and in France in the mid-50s. Two demonstration plants, the PFR in Scotland and Phénix in France, started operation in the mid-70s. The 1200 MWe commercial size LMFBR Superphénix, built and operated by an international consortium from France, Germany, and Italy, first reached full power in 1986. Besides electricity production and the demonstration of breeding gain, it is scheduled to test burning of plutonium and other actinides in a fast flux. Present European developments are now focused on the design of a commercial European fast reactor (the 1500 MWe EFR), which could become economically competitive with LWRs when the price of uranium has increased several times.