Plutonium Burning Research Articles

Neutronic evaluation of a small lead-cooled nuclear reactor as an actinides burner

This study employs the Monte Carlo N-Particle code, version 6 (MCNP6), to simulate ELECTRA and investigate its neutronic parameters and the transmutation of spent fuel over a 15-year burnup period, as referenced. The primary objective is to evaluate actinide burning in a low-power fast reactor. The study consists of three phases: 1) validating Model 1 for MCNP6 code with results obtained by SERPENT code from previous studies, 2) conducting neutronic analysis of three ELECTRA models (1, 2, and 3) with varying levels of detail, and 3) assessing burnup using the most detailed ELECTRA MCNP6 model (Models 3). The neutronic calculations obtained from the MCNP6 code align with previous studies. Notably, including all core components in the simulation yields significant disparities in criticality and neutron flux compared to a simpler model. Despite its small size and low power, ELECTRA, as a research reactor, exhibits essential characteristics for effectively burning plutonium from Light Water Reactor (LWR) spent fuel. Therefore, ELECTRA contributes to developing LFRs and supports the GIF’s research goals for reactor development.

An energy amplifier fluidized bed nuclear reactor concept

Abstract The concept of a fluidized bed nuclear reactor driven by an energy amplifier system is described. The reactor has promising characteristics of inherent safety and passive cooling. The reactor can easily operate with any desired spectrum in order to be a plutonium burner or have it operate with thorium fuel cycle.

Thorium-based CANDU qualification as plutonium burner

Abstract Converting the weapon grade Plutonium from the U.S.A., Russia, U.K. etc. to Mixed OXide fuel and using it in power reactors was seen as a feasible way to both dispose Plutonium and produce energy. Using Thorium-based fuels in CANDU has been investigated since early 1980’s, they were designed and tested in Canada as mixed ThO2 -UO 2 (both LEU and HEU) and mixed ThO2 -PuO 2, (both reactor- and weapons-grade) ([1]). In this respect, Thorium might also be seen as a valuable driver for weapon grade Plutonium annihilation. Our goal was to investigate ThO2 -PuO 2 MOX in the aim to propose a suitable fuel for the existing and future CANDU units in Romania. Both weapon grade and reactor grade Plutonium were considered as fissile drivers for Thorium. Since this is only an exploratory study, some key design parameters such as fuel pellet density and ThO2 /PuO 2 ratio were considered to span over a certain range imposed by MOX fuel fabrication technology and limited Plutonium availability. Eighteen fuel compositions were considered and cell calculations were performed for 37 and 43-element bundles using several computer codes.

Special issue on accelerator-driven system benchmarks at Kyoto University Critical Assembly

ABSTRACTThe accelerator-driven system (ADS) had been proposed for producing energy and transmuting minor actinide and long-lived fission products. ADS has attracted worldwide attention in recent years because of its superior safety characteristics and potential for burning plutonium and nuclear waste. At the Institute for Integrated Radiation and Nuclear Science, Kyoto University, a series of ADS experiments with 14 MeV neutrons was launched in fiscal year 2003 at the Kyoto University Critical Assembly (KUCA). Also, the high-energy neutrons generated by the interaction of 100 MeV protons with tungsten target was injected into the KUCA core on March 2009. The ADS experiments with 100 MeV protons obtained from the FFAG accelerator have been carried out to investigate the neutron characteristics of ADS. This special issue aims at concentrating on experimental analyses for the ADS benchmarks at KUCA on the basis of most recent advances in the development of computational methods, and contributing to academic progress for ADS research field in the future.

Study on plutonium burner high temperature gas-cooled reactor in Japan – Introduction scenario, reactor safety and fabrication tests of the 3S-TRISO fuel

Study on plutonium burner high temperature gas-cooled reactor in Japan – Introduction scenario, reactor safety and fabrication tests of the 3S-TRISO fuel

Self-shielding effect of double heterogeneity for plutonium burner HTGR design

The investigation on self-shielding effect of double heterogeneity for plutonium burner High Temperature Gas-cooled Reactor (HTGR) design has been performed. Plutonium burner HTGR designed in the previous study by using the advantage of double heterogeneity to control excess reactivity. In the present study, the mechanism of the self-shielding effect is elucidated by the analysis of burn-up calculation and reactivity decomposition based on exact perturbation theory. As a result, it is revealed that the characteristics of burn-up reactivity are determined by resonance cross section peak at 1 eV of Pu-240 due to the surface term of background cross section, this is, the characteristics of neutron leakage from fuel lump and collision to a moderator. Moreover, significant spectrum shift is caused during the burn-up period, and it enhances reactivity worth of Pu-239 and Pu-240 in EOL.

A disruptive approach to eliminating weapon-grade plutonium - Pu burning in a molten salt fast reactor.

The successful implementation of disarmament treaties of the last centuries has led to significant amounts of weapon-grade Plutonium which are currently stored in high security storage facilities. Disposing this Plutonium should be seen as ‘good housekeeping’ avoiding unnecessary costs and the hazards of storing this material indefinitely. In addition, the disarmament is only brought to a successful end when the Plutonium isn’t available for the production of new weapons anymore. We propose a disruptive approach for Plutonium disposition and demonstrate the feasibility in a neutronic study. Burning of weapon-grade Plutonium in a molten salt fast reactor is significantly more efficient than in the studied other reactors, while efficient process design has the potential to reduce the security concerns significantly. The proposed system could turn about 1.25 tons of weapon-grade Plutonium into electric energy worth £ 0.5 to 1 billion/year, depending on the electricity price while avoiding the hassle and eliminating the risk of high security Plutonium storage. In conclusion, burning of the weapon-grade Plutonium resulting from disarmament could be an economically very attractive approach to reduce the nuclear threat.

Treatment of Double Heterogeneity in the Resonance and Thermal Energy Regions in High-Temperature Reactors

The fuel in high-temperature reactors (HTRs) consists of a large number of tiny (a few hundred microns in size) TRISO-coated particles dispersed randomly in a graphite matrix. At the resonances of the major nuclide (238U or 232Th) of the fuel, the neutron mean free path in the fuel is often comparable to the kernel dimensions, and hence the dispersion of particles in the graphite matrix must be treated as a heterogeneous medium for obtaining self-shielded cross sections in the resonance (epithermal) groups. HTRs containing only plutonium as fuel (as may be the case in reactors designed for burning plutonium) have high concentrations of the isotopes 239Pu and 240Pu in the kernels. This fact together with their rather large resonance cross section in the thermal groups results in a very short neutron mean free path in the fuel that is comparable to the kernel dimensions. In such cases the fuel zone must be treated as a heterogeneous medium in the thermal energy region as well. However, the resonance treatment method in libraries such as the WIMS library does not cover the two large resonances of plutonium lying in the thermal region. Instead, a large number of groups are used to cover the details of cross-section variation in this region.The heterogeneity of the fuel region together with the heterogeneous distribution of fuel region, graphite moderator, and coolant is referred to as the double heterogeneity of HTRs. The paper describes work we carried out for addressing these problems. A new method for generating a random distribution of TRISO particles in the fuel zone of a pebble or fuel compact and Monte Carlo calculation of the Dancoff factor in the heterogeneous random medium, required for calculating the self-shielded resonance group cross sections, is presented. Dancoff factors obtained by our method are compared with values available in published literature on the subject. The paper also discusses a new methodology developed to solve the double-heterogeneity problem at the stage of the multigroup transport theory solution, which is particularly important in the thermal region occurring in high-content Pu-fueled HTRs. These features have been incorporated in the WIMS library–based lattice code BOXER3. An option for handling the spherical geometry of the lattice cell of a pebble bed reactor has been added in the code. Results of analysis of a number of HTR lattice cell benchmark problems are presented.

A study on transmutation of LLFPs using various types of HTGRs

In order to investigate the potential of high temperature gas-cooled reactors (HTGRs) for transmutation of long-lived fission products (LLFPs), numerical simulation of four types of HTGRs were carried out. In addition to the gas-turbine high temperature reactor system “GTHTR300”, which is the subject of our previous research, a small modular HTGR plant “HTR50S” and two types of plutonium burner HTGRs “Clean Burn with MA” and “Clean Burn without MA” were considered. The simulation results show that an early realization of LLFP transmutation using a compact HTGR may be possible since the HTR50S can transmute fair amount of LLFPs for its thermal output. The Clean Burn with MA can transmute a limited amount of LLFPs. However, an efficient LLFP transmutation using the Clean Burn without MA seems to be convincing as it is able to achieve very high burn-ups and produce LLFP transmutation more than GTHTR300. Based on these results, we propose utilization of variety of HTGRs for LLFP transmutation and storage.

On the Burning of Plutonium Originating from Light Water Reactor Use in a Fast Molten Salt Reactor—A Neutron Physical Study

An efficient burning of the plutonium produced during light water reactor (LWR) operation has the potential to significantly improve the sustainability indices of LWR operations. The work offers a comparison of the efficiency of Pu burning in different reactor configurations—a molten salt fast reactor, a LWR with mixed oxide (MOX) fuel, and a sodium cooled fast reactor. The calculations are performed using the HELIOS 2 code. All results are evaluated against the plutonium burning efficiency determined in the Consommation Accrue de Plutonium dans les Réacteurs à Neutrons RApides (CAPRA) project. The results are discussed with special view on the increased sustainability of LWR use in the case of successful avoidance of an accumulation of Pu which otherwise would have to be forwarded to a final disposal. A strategic discussion is given about the unavoidable plutonium production, the possibility to burn the plutonium to avoid a burden for the future generations which would have to be controlled.

Thoria based inert matrix fuel production via sol–gel process

There is an important concern that uranium reserves will not be sufficient for new nuclear facilities in the future. Thorium could offer a potential alternative fuel because it is three times more abundant than uranium. Thorium based nuclear fuels are important candidates for the future of nuclear fuels. The main interest is in considering thorium based fuels for burning plutonium in conventional reactors. Thoria based inert matrix fuels (IMF) are also candidates for burning plutonium and the transmutation of minor actinides. In this study, inert matrix fuel production was examined by external gelation process. Metal nitrate solution of each material (thorium nitrate, cerium nitrate, aluminum nitrate and magnesium nitrate) were mixed for the feed solution. The metal nitrate solution was 50% of total feed solution. After mixing the metal nitrate, solution polyvinyl alcohol (PVA, 10%) and tetrahydrofurfuryl alcohol (THFA, 40%) were added into mixed metal nitrate solution during stirring. PVA, THFA and metal nitrate solution were dropped through a nozzle into gelation medium ammonia solution. Then external gelation process was carried out. Cerium (Ce) was used to simulate Plutonium (Pu). Obtained gels from the external gelation process were aged, washed and dried. The powders were calcinated at 1273 K for 6 h. After calcination, the powders were pressed at 500 MPa with a manual press. Pellets were sintered at elevated temperatures (1473, 1573, 1673, 1773, 1873, and 1973 K) for 4 h. The densities of sintered pellets were measured by immersion method. Sintered pellets characteristics were determined by scanning electron microscopy, x-ray diffraction analysis and energy dispersive x-ray analysis and the results were discussed.

Plutonium Transmutation by a Gas Dynamic Trap Hybrid Reactor

A team at the Budker Institute of Nuclear Physics has been working for several years to develop the Gas Dynamic Trap Mirror Neutron Source (GDT-NS) for fusion materials irradiation. In 2010 they optimized the design for a transmutation mission forecasting a 16 meter DT plasma with a fusion power of 15 MW and neutrons preferentially emitted into blankets placed around the mirror turning points. While this remains to be demonstrated experimentally, it is intriguing to explore what could be done with a low fusion power neutron source.The GDT-NS team has previously modeled the burning of minor actinides. The work presented here builds on this by examining the burning of plutonium starting with transuranics recovered from spent nuclear fuel. It was found that a GDT plutonium burner with two blankets could eliminate nearly the plutonium produced in a single light water reactor core per full power year, 249 kg. By increasing the average blanket power with regular refueling, this quantity was increased to 381 kg per full power year. Next followed a preliminary overview of a GDT disposition blanket to meet US treaty commitments in burning surplus military plutonium.

Irradiation and post-irradiation examination of uranium-free nitride fuel

Two identical Phénix-type 15-15Ti steel pinlets each containing a 70 mm Pu0.3Zr0.7N fuel stack in a 1-bar helium atmosphere have been irradiated in the HFR Petten at medium high linear power (46–47 kW/m at BOL) and an average cladding temperature of 505 °C. The pins were irradiated to a plutonium burn-up of 9.7% (88 MWd/kgHM) in 170 full power days. Both pins remained fully intact. Post-irradiation examination performed at NRG and PSI showed that the overall swelling rate of the fuel was 0.92 vol-%/%FIHMA. Fission gas release was 5–6%, while helium release was larger than 50%. No fuel restructuring was observed, and only mild cracking. EPMA measurements show a burn-up increase toward the pellet edge of up to 4 times. All investigated fission products except to some extent the noble metals were found to be evenly distributed over the matrix, indicating good solubility. Local formation of a secondary phase with high Pu content and hardly any Zr was observed. A general conclusion of this investigation is that ZrN is a suitable inert matrix for burning plutonium at high destruction rates.

Proposal of a plutonium burner system based on HTGR with high proliferation resistance

An innovative plutonium burner concept based on high temperature gas cooled reactor (HTGR) technology, “Clean Burn”, is proposed by Japan Atomic Energy Agency (JAEA). That is expected to be as an effective and safe method to consume surplus plutonium accumulated in Japan. A similar concept proposed by General Atomics (GA), Deep Burn, cannot be introduced to Japan because of its adopting highly enriched plutonium, which shall infringe on a Japanese nuclear nonproliferation policy according to Japan–US reprocessing negotiation. The Clean Burn concept can avoid this problem by employing an inert matrix fuel (IMF) and a tightly coupled fuel reprocessing and fabrication plants. Both features make it impossible to extract plutonium alone out of the fabrication process and its outcomes. As a result, the Clean Burn can use surplus plutonium as a fuel without mixing it with uranium matrix. Thus, surplus plutonium alone will be incinerated effectively, while generation of plutonium from the uranium matrix is avoided. High neutronic performance, i.e., achievement of burn-up of about 500 GWd/t and consumption ratio of plutonium-239 reaching to about 95%, is also assessed. Furthermore, reactivity defect caused by the inert matrix is found to be negligible. It is concluded that the Clean Burn concept is a useful option to incinerate plutonium with high proliferation resistance.

Utilization of rock-like oxide fuel in the phase-out scenario

Utilization of rock-like oxide (ROX) fuel in light water reactors for plutonium (Pu) burning was studied by nuclear material balance (NMB) analysis for a case of Japanese phase-out scenario under investigation after the Fukushima accident. For the analysis, the NMB code was developed with features of accurate burn-up calculation, flexible combination of reactors and fuels, and an ability to estimate waste and repository. Three scenario groups of once-through Pu burning in mixed oxide (MOX) fuel and in ROX fuel were analyzed. Using two full-MOX or full-ROX reactors the Pu amount is reduced to about one-half and the isotopic vector of Pu deteriorated for being used as a nuclear weapon, especially in terms of spontaneous fission neutron generation. Effects of ROX reactors are more significant than MOX reactors in terms of both reduction in the Pu amount and deterioration of the isotopic vector. Repository footprint and potential radiotoxicity are not reduced by the MOX and ROX reactors because the heat and toxicity of MOX and ROX spent fuels are considerably high.

LIFE hybrid reactor as reactor grade plutonium burner

The early version of the conceptual modified design of the Laser Inertial Confinement Fusion Fission Energy (LIFE) engine consists of a spherical fusion chamber of 5m diameter, surrounded by a multi-layered blanket. The first wall is made of 2cm thick ODS and followed by a Li17Pb83 zone (2cm), acting as neutron multiplier, tritium breeding and front coolant zone. It is separated by an ODS layer (2cm) from the FLIBE molten salt zone (50cm), containing fissionable fuel. A 3rd ODS layer (2cm) separates the molten salt zone on the right side from the graphite reflector (30cm).Calculations have been conducted for a constant fusion driver power of 500MWth in S8-P3 approximation using 238-neutron groups. Reactor grade (RG) plutonium carbide fuel in form of TRISO particles with volume fractions of 2%, 3%, 4%, 5% and 6% have been dispersed homogenously in the FLIBE coolant.Tritium breeding ratio (TBR) values per incident fusion neutron for the above cited cases start with TBR=1.35, 1.52, 1.73, 2.02 and 2.47, respectively. With the depletion of fissionable RG-Pu isotopes, TBR decreases gradually. At startup, higher fissionable fuel content in the molten salt leads to higher blanket energy multiplication, namely M0=3.8, 5.5, 7.7, 10.8 and 15.4 with 2%, 3%, 4%, 5% and 6% TRISO volume fraction, respectively. Calculations have led to very high burn up values (>400,000 MD.D/MT). TRISO particles can withstand such high burn ups. Such high burn ups would lead to drastic reduction of final nuclear waste per unit energy production.

The GDT Based Neutron Source as a Driver in a Sub-Critical Burner of Radioactive Wastes

Transmutation of long-lived radioactive nuclear waste, including plutonium, minor actinides and fission products, represents a highly important problem of fission reactor technology and is presently studied worldwide in large-scale. Sub-critical systems seem to be a promising option for efficiently burning plutonium and minor actinides provided a sufficiently high-intense neutron source is available. For a number of years the Budker Institute of Nuclear Physics (Russia) in collaboration with the Russian and European organizations developed the project of a 14 MeV neutron source for fusion material irradiation and other applications. The projected plasma type neutron source is based on the Gas Dynamic Trap (GDT) which is a special magnetic mirror system for the plasma confinement. This poster presents different version of the GDT-based neutron source for hybrid fusion-fission sub-critical system for the transmutation of the long-live radioactive waste in spent nuclear fuel.

A neutronic investigation on a helium cooled hybrid reactor using nitride fuels containing reactor grade plutonium

There has been a significant amount of reactor grade (RG) plutonium accumulated from the conventional nuclear reactors’ spent fuel. Destruction or reducing this RG plutonium is very important to prevent its misuse and/or release accidentally into the environment. Using very energetic fusion neutrons in fusion–fission (hybrid) reactors can burn the RG plutonium effectively. This study presents the burning of the mixed fuel containing RG plutonium and uranium in a helium cooled hybrid reactor for an operation period of 24 months. Effect of various fuel mixtures and tritium breeders on the neutronic performance of the reactor as well as the burning of the RG plutonium was investigated. Calculations were carried out on an experimental hybrid blanket with the aid of SCALE4.3 by solving the Boltzmann transport equation with the XSDRNPM code. Numerical results showed that increasing RG plutonium content in the fuel increased the burning of plutonium and fusion energy multiplication substantially. The tritium breeder having the lowest lithium atomic density allowed the highest fission of the fuel in the blanket.

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.

Comparative study of minor actinide transmutation in sodium and lead-cooled fast reactor cores

This paper shows that lead-cooled and sodium-cooled fast reactors (LFRs and SFRs) can preferentially consume minor actinides without burning plutonium, both in homogeneous and in heterogeneous mode. The former approach consists of admixing about 5% of minor actinides (MAs) into uranium–plutonium fuels in the core and using a limited number of thermalising pins consisting of UZrH 1.6. These are needed to keep the negative Doppler feedback larger than the positive coolant reactivity coefficient. Our Monte Carlo burn-up calculations showed that a 600 MW e LFR self-breeder without blankets can burn an average of around 67 kg annually of MAs with a reactivity swing of only about −0.7$ per year. The reactivity swing of a corresponding 600 MW e SFR is more than three times larger due to the poorer breeding and half the critical mass in comparison to the LFR. However, when axial and radial blankets loaded with 10% MAs are added, the SFR burns 25% more MAs (131 kg/yr) and breeds 30% more Pu (150 kg/yr) than an equally sized LFR. When only the blankets are loaded with MAs, the SFR breeds 30% more Pu (198 kg/yr) and still burns about 60 kg a year of MAs. However, in terms of severe accident behaviour, the LFR, with its superior natural coolant circulation and larger heat capacity, has definite advantages.