Plutonium In Reactors Research Articles

Analysis of radwaste accumulation in various scenarios of NP development

Within the framework of “Proryv” project a radiation-equivalent approach to radioactive waste management is being envisioned with U and Pu recycling and MA transmutation. Successful industry-wide implementation of the design approaches should be planned in order to avoid considerable financial and radiological encumbrances caused by the NFC final stage for two-component nuclear power system (NPS) under formation on the basis of thermal and fast reactors. In order to ensure a successful industry-wide implementation of the approaches being developed, the back-end of the NFC should not constitute considerable a financial and radiological burden for the emerging two-component nuclear power system (NPS). This article addresses the problems concerning justification of radiological and technical-and-economic feasibility of MA partitioning and subsequent transmutation in FNR. The extent of MA accumulation as a result of TNR SNF reprocessing confirms the need for the introduction of MA partitioning technologies not only at all reprocessing plants planned for commissioning, but also at the plants now in operation. Based on available data, the study has shown that the implementation of the closed NFC with FNR contributes to significant reduction in the cost of disposal of radwaste compared to the scenario based exclusively on the development of VVER and open fuel cycle technologies. Recycling plutonium in fast reactors should be implemented in conjunction with MA to address environmental, non-proliferation and economic concerns of the back-end of advanced NFC. Within the scale of the future nuclear power system in Russia, an option such as this can only be realized on the basis of developing a FNR fleet.

Some economic aspects of reducing americium production in a two-component system of thermal and fast reactors

The article analyzes the economic aspects of reducing the production of americium during the transition from a single-component nuclear energy system (NES) based on thermal reactors in an open fuel cycle to a two-component system with thermal and fast reactors in a closed nuclear fuel cycle. Scenarios for the development of these systems in Russia up to the end of the century are modeled. Two methods are considered for reducing the production of americium in a two-component NES with fast sodium reactors. The first method, closing the fuel cycle for plutonium in BN reactors of SFR type, is based on the use of plutonium separated from spent nuclear fuel of thermal reactors with the shortest possible (according to technical specifications) time for MOX fuel preparation and use thus preventing the main part of plutonium-241 from decay into americium. The second way is transmutation of americium. The study was carried out by using the mathematical code CYCLE designed for modeling of the NES with closed nuclear fuel cycle (NFC). The technical and economic data used in the paper was taken from published studies of Russian specialists and materials of European Union specialists presented in the IAEA/INPRO SYNERGIES project. The results of the research show that the efficiency of closing the NFC by using plutonium from thermal reactors in MOX fuel of fast sodium reactors is comparable to the efficiency of the homogeneous transmutation considered in the paper. The combination of the americium accumulation prevention method and transmutation method might significantly reduce the rate of the americium accumulation in a nuclear energy system, but the estimated costs of the considered homogeneous transmutation can significantly worsen the economic performance of sodium fast reactors.

Burn-Up Analysis of Plutonium and Minor Actinide Recycling on Gas Cooled Fast Reactor (GCFR) Using SRAC COREBN

Research on the analysis of recycling, plutonium, and minor actinides in the Gas Cooled Fast Reactor (GFCR) using SRAC COREBN has been done. This research uses a fuel mixture of uranium, plutonium, and minor actinides. The analysis was conducted with computational simulation using the COREBN code, an additional code to SRAC. The purpose of this research is to find out the influence of the addition of plutonium and actinide on the composition of nuclear fuel at the end of reactor operation and the limitation of the value of the multiplication factor (keff at the end of the reactor burn up period). Found in this research is the final value, the multiplication factor (keff) of 1.19964, and the conversion ratio value of 0.766813 in the burn-up period of 1515 days as well as the maximum power density value of 125.85 watts/cm³, the relative power density value at the radius y of 1,230263 and radius x equal to 1.19737 the atomic density value experienced a change in the number of nuclides in the types of nuclides U235, U238, Pu239, Pu241, Np237, and Am243 at the end of the burn-up period.

Fourth-Order Adjoint Sensitivity and Uncertainty Analysis of an OECD/NEA Reactor Physics Benchmark: II. Computed Response Uncertainties

This work quantifies the impact of the most important 4th-order sensitivities of the leakage response of a polyethylene-reflected plutonium (PERP) reactor physics benchmark with respect to the benchmark’s 180 group-averaged microscopic total cross sections, on the expected value, variance and skewness of the benchmark’s leakage response. This work shows that, as the standard deviations of the cross sections increase, the contributions of the 4th-order sensitivities to the response’s expected value and variance become significantly larger than the corresponding contributions stemming from the 1st-, 2nd- and 3rd-order sensitivities. Considering a uniform 5% relative standard deviation for all microscopic total cross sections, the contributions from the 4th-order sensitivities to the expected value and variance of the PERP leakage response amount to 56% and 52%, respectively. Considering 10% uniform relative standard deviations for the microscopic total cross sections, the contributions from the 4th-order sensitivities to the expected value increase to nearly 90%. Consequently, if the computed value L(a) were considered to represent the actual expected value of the leakage response and the 4th-order sensitivities were neglected, the computed value would represent the actual expected value with an error of 3400%. Furthermore, uniform relative standard deviations of 5% and larger (10%) for the microscopic total cross sections cause the higher-order sensitivities to contribute increasingly higher amounts to the response standard deviation: the contributions stemming from the 4th-order sensitivities are larger than the contributions stemming from the 3rd-order sensitivities, which in turn are larger than those stemming from the 2nd-order sensitivities, which are themselves larger than the contributions stemming from the 1st-order sensitivities. This finding evidently underscores the need for computing sensitivities of order higher than first-order. The results obtained in this work also indicate that the 4th-order sensitivities produce a positive response skewness, causing the leakage response distribution to be skewed towards the positive direction from its expected value. Increasing the parameter standard deviations tends to decrease the value of the response skewness, causing the leakage response distribution to become more symmetrical about the mean value. The results presented in this work highlight the finding that the microscopic total cross section for hydrogen (H) in the lowest (“thermal”) energy group is the single most important parameter among the 180 microscopic total cross sections of the PERP benchmark, as it contributes most to the various response moments.

A highly precise micro-analytical XRF method for compositional characterization of fast breeder reactor fuels

Accurate and precise determination of plutonium and uranium in Fast Breeder Reactor (FBR) fuels is one of the most important steps in the chemical quality control of these fuels.

Utilization of Military Grade Plutonium in Generation-IV Gas Cooled Fast Reactor

The operation of fast spectrum concepts like Gas cooled Fast Reactor (GFR) requires fuel with high enrichment of the fissile isotope 239Pu. In this work the utilization of Weapon Grade Plutonium (WGrPu) in GFR is investigated. For that, the transport code MCNPX is employed to design two models of the GFR2400 core one is heterogeneous and the other is homogeneous. The designed models are used to perform a neutronic analysis of the use of WGrPu as a fissile material instead of the reference Reactor Grade Plutonium (RGrPu) vector. The proposed WGrPu fuel has shown some improved neutronic and safety parameters compared to the reference RGrPu fuel.

On the possibility to improve mixed uranium-plutonium fuel in fast reactors

Accumulation of plutonium extracted from the spent nuclear fuel (SNF) of light water reactors is one of the central problems in nuclear power. To reduce out-of-the-reactor Pu inventory, leading nuclear power countries (France, Japan) use plutonium in light water power reactors in the form of MOX fuel, with half of Pu fissioning in this fuel. The rest of Pu cannot be reused easily and efficiently in light water reactors because of the high content of even isotopes. Plutonium for which there are no potential consumers is accumulated. Unlike thermal reactors, fast reactors take plutonium of any isotopic composition. That makes it possible to improve plutonium isotopic composition and to reduce the fraction of even isotopes to the level that allows reuse of such plutonium in thermal reactors. The idea of changing the isotopic composition of Pu in fast reactors is well-known. The originality of the research lies in applying this idea to combine the fuel cycles of fast and thermal reactors. Pu isotopic composition can be improved by combining certain operational activities in order to supply fuel to thermal and fast reactors. Scientific and technological justification of the possibility will let Russian BN technologies and French MOX fuel technologies work in synergy with thermal reactors.

Unnatural Resources: Energy and Environmental Politics in Appalachia after the 1973 Oil Embargo by Michael Camp

Reviewed by: Unnatural Resources: Energy and Environmental Politics in Appalachia after the 1973 Oil Embargo by Michael Camp Merritt McKinney Unnatural Resources: Energy and Environmental Politics in Appalachia after the 1973 Oil Embargo. By Michael Camp. History of the Urban Environment. (Pittsburgh: University of Pittsburgh Press, 2019. Pp. viii, 192. $40.00, ISBN 978-0-8229-4571-0.) What a difference a few years and an energy crisis make. The late 1960s and early 1970s saw a surge in public support for regulations to protect the [End Page 940] environment. But the election of Ronald Reagan in 1980 led to a lessening of the federal government's environmental protection efforts. In Unnatural Resources: Energy and Environmental Politics in Appalachia after the 1973 Oil Embargo, Michael Camp identifies the 1973 oil embargo as an important turning point in the nation's changing attitudes about the proper balance between environmental quality and economic growth. Though he does not contend that the 1973 crisis was the sole reason for the change in environmental priorities, he does convincingly argue that concerns about inadequate energy supply—especially oil—had an important influence on the shift away from environmental protection. As he notes, "When energy supplies tighten, public concern for the environment can diminish" (p. 3). Though published as a volume in an influential environmental history series, this book is at heart a political history dealing with environmental issues. After an initial chapter providing an overview of national energy politics from the early 1970s to the Reagan presidency, each of the remaining chapters focuses on a narrower topic: the attempt to shift utility plants from oil to coal power to preserve the nation's oil supply; the controversy over the Clinch River breeder reactor in Tennessee; the impact of the Tellico Dam project on enforcement of the Environmental Protection Act; and the difficulties experienced by the Tennessee Valley Authority as it tried to adjust to the new energy politics. The analysis of each of these examples is mostly in the context of the emerging energy politics at the federal level, but the book points out that local interests often led to blurred ideological lines. For example, continued funding of the Clinch River plutonium reactor faced opposition from both environmentalists concerned about its environmental and health effects and fiscal conservatives worried about its cost. Though the project was eventually halted, funding for the reactor continued for more than a decade in part because many local leaders, backed by a powerful U.S. senator, touted its economic benefits to the surrounding area. The subtitle of the book, Energy and Environmental Politics in Appalachia after the 1973 Oil Embargo, is a bit misleading. Each of the case studies deals with events that took place in the Appalachian region—mostly East Tennessee—but the book is less a study of Appalachia than an examination of national energy politics with examples that happen to be from that region. There is little sense that the book's argument depends on distinct characteristics of Appalachia. Examples from other parts of the country would likely reveal a similar conflict of priorities and ideologies. Unnatural Resources is a useful book for scholars interested in the history of energy politics in the 1970s and 1980s as well as for those interested more broadly in how national debates on environmental regulation interacted with local interests. According to its author, "energy is a historical concept rooted in specific times and specific places" (p. 148). Through the study of several such times and places, this book aims to help explain the challenges President Jimmy Carter faced in achieving his energy goals. [End Page 941] Merritt McKinney Volunteer State Community College Copyright © 2020 Southern Historical Association

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.

Reactors as a Source of Antineutrinos: Effects of Fuel Loading and Burnup for Mixed-Oxide Fuels

In a conventional light water reactor loaded with a range of uranium and plutonium-based fuel mixtures, the variation in antineutrino production over the cycle reflects both the initial core fissile inventory and its evolution. Under the assumption of constant thermal power, we calculate the rate at which antineutrinos are emitted from variously fueled cores, and the evolution of that rate as measured by a representative ton-scale antineutrino detector. We find that antineutrino flux decreases with burnup for Low Enriched Uranium cores, increases for full mixed-oxide (MOX) cores, and does not appreciably change for cores with a MOX fraction of approximately 75%. Accounting for uncertainties in the fission yields, in the emitted antineutrino spectra, and the detector response function, we show that the difference in core-wide MOX fractions at least as small as 8% can be distinguished using a hypothesis test. The test compares the evolution of the antineutrino rate relative to an initial value over part or all of the cycle. The use of relative rates reduces the sensitivity of the test to an independent thermal power measurement, making the result more robust against possible countermeasures. This rate-only approach also offers the potential advantage of reducing the cost and complexity of the antineutrino detectors used to verify the diversion, compared to methods that depend on the use of the antineutrino spectrum. A possible application is the verification of the disposition of surplus plutonium in nuclear reactors.

Fingerprinting commercial nuclear reactor forensics using the ORIGEN-S simulation code

The ability to determine the origin of a specific spent-fuel sample from a commercial nuclear reactor was studied using the Origen-S simulation code by calculating the plutonium and uranium isotopic concentration data for a range of nuclear power reactors. This range of reactors is based on a typical Westinghouse PWR fuel assembly with a fuel type of W17 X 17, having individual operating histories. Isotopic ratios of plutonium in nuclear reactors during the fuel-cycle period provide information on how the plutonium grows into the fuel as a function of burnup, as well as its attractiveness to proliferators. Using the results from the calculation of uranium and plutonium isotopic ratios, the origin of each spent-fuel assembly for a particular reactor can be predicted and documented for a future nuclear forensics reference database.

Variants of Regenerated Fissile Materials Usage in Thermal Reactors as the First Stage of Fuel Cycle Closing

At present, 240 000 t of spent nuclear fuel (SF) has been accumulated in the world. Its long-term storage should meet safety conditions and requires noticeable finances, which grow every year. Obviously, this situation cannot exist for a long time; in the end, it will require a final decision. At present, several variants of solution of the problem of SF management are considered. Since most of the operating reactors and those under construction are thermal reactors, it is reasonable to assume that the structure of the nuclear power industry in the near and medium-term future will be unchanged, and it will be necessary to utilize plutonium in thermal reactors. In this study, different strategies of SF management are compared: open fuel cycle with long-term SF storage, closed fuel cycle with MOX fuel usage in thermal reactors and subsequent long-term storage of SF from MOX fuel, and closed fuel cycle in thermal reactors with heterogeneous fuel arrangement. The concept of heterogeneous fuel arrangement is considered in detail. While in the case of traditional fuel it is necessary to reprocess the whole amount of spent fuel, in the case of heterogeneous arrangement, it is possible to separate plutonium and 238U in different fuel rods. In this case, it is possible to achieve nearly complete burning of fissile isotopes of plutonium in fuel rods loaded with plutonium. These fuel rods with burned plutonium can be buried after cooling without reprocessing. They would contain just several percent of initially loaded plutonium, mainly even isotopes. Fuel rods with 238U alone should be reprocessed in the usual way.

Radionuclide pollution inside the Fukushima Daiichi exclusion zone, part 2: Forensic search for the “Forgotten” contaminants Uranium-236 and plutonium

In the course of the Fukushima Daiichi nuclear accident (March 2011), large quantities of radionuclides, mainly of the volatile elements such as cesium and iodine were released to the environment. In small amounts non-volatile elements such as plutonium and uranium have also been released. For this study vegetation, litter and soil drill core samples were taken in 2013 and 2015 in the contaminated area in Fukushima prefecture. In 2015 the samples were taken as close to the sampling locations of the 2013 campaign as possible. To investigate the origin of plutonium and uranium the isotopic ratios 240Pu/239Pu and 236U/238U were measured using accelerator mass spectrometry. While the determined 240Pu/239Pu ratios indicate global fallout as the plutonium source for most samples, reactor plutonium of the reactors of Fukushima Daiichi is identified unambiguously in some vegetation and litter samples. This assumption is corroborated by the detection of 236U. None of the samples contained higher plutonium activity concentrations than in the rest of Japan, caused by global fallout. This holds even for the samples where the reactors were identified as the source from the isotope ratios. This and the strong localization, even for neighboring sampling sites, indicate that the overall small quantities of plutonium were most likely released in the form of particulate matter.

Target fuels for plutonium and minor actinide transmutation in pressurized water reactors

The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, containing approximately 95wt% uranium, 1wt% plutonium, and 4wt% fission products and transuranic elements. Fast reactors are a preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20years indicates a need for a nearer-term solution. This study considers a method for plutonium and minor actinide transmutation in existing light water reactors and evaluates a variety of transmutation fuels to provide a common basis for comparison and to determine if any single target fuel provides superior transmutation properties. A model developed using the NEWT module in the SCALE 6.1 code package provided performance data for the burnup of the target fuel rods in the present study. The target fuels (MOX, PuO2, Pu3Si2, PuN, PuUZrH, PuZrH, PuZrHTh, and PuZrO2) are evaluated over a 1400 Effective Full Power Days (EFPD) interval to ensure each assembly remained critical over the entire burnup period. The MOX (5wt% PuO2), Pu0.31ZrH1.6Th1.08, and PuZrO2MgO (8wt% Pu) fuels result in the highest rate of plutonium transmutation with the lowest rate of curium-244 production. This study selected eleven different burnable absorbers (B4C, CdO, Dy2O3, Er2O3, Eu2O3, Gd2O3, HfO2, In2O3, Lu2O3, Sm2O3, and TaC) for evaluation as spectral shift absorber coatings on the outside of the fuel pellets to determine if an absorber coating can improve the transmutation properties of the target fuels. The PuZrO2MgO (8wt% Pu) target fuel with a coating of Lu2O3 resulted in the highest rate of plutonium transmutation with the greatest reduction in curium-244 production.

Transitioning nuclear fuel cycles with uncertain fast reactor costs

This paper applies a novel decision making methodology to a case study involving choices leading to the transition from the current once-through light water reactor fuel cycle to one relying on continuous recycle of plutonium and minor actinides in fast reactors in the face of uncertain fast reactor capital costs. Unique to this work is a multi-stage treatment of a range of plausible trajectories for the evolution of fast reactor capital costs over time, characterized by first-of-a-kind penalties as well as time- and unit-based learning. The methodology explicitly incorporates uncertainties in key parameters into the decision-making process by constructing a stochastic model and embedding uncertainties as bifurcations in the decision tree. “Hedging” strategies are found by applying a choice criterion to select courses of action which mitigate “regrets”. These regrets are calculated by evaluating the performance of all possible transition strategies for every feasible outcome of the uncertain parameter. The hedging strategies are those that preserve the most flexibility for adjusting the fuel cycle strategy in response to new information as uncertainties are resolved.

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.

ZPPR Benchmarks for Large LMFBR Core Physics from JUPITER Cooperative Program Between United States and Japan

Under the International Reactor Physics Experiment Evaluation Project (IRPhEP) framework, in the cooperative JUPITER program between the United States and Japan, benchmarks are established to study large fast breeder reactor (FBR) core physics utilizing nine Zero Power Plutonium Reactor (ZPPR) critical experimental cores. These benchmarks cover a wide variety of core concepts including homogeneous and heterogeneous configurations, clean and engineering mock-up cores of 600- to 1000-MW(electric)–class sizes, and various core parameters such as criticality, reaction rate, and reactivity. Recently, detailed experimental information from original documents from Argonne National Laboratory has been scrutinized very carefully to establish the benchmark model and to evaluate quantitatively the experimental uncertainty. The benchmarks supply users with heterogeneous cell models and three-dimensional (3-D) core configurations, which are simplified to a degree that preserves the important physical features of the ZPPR cores such as plate heterogeneity, different drawer types, and 3-D core arrangement. Further, the benchmark handbook includes as-built information of the ZPPR cores as a complete set of electronic form; therefore, a user can develop his or her own benchmark model if necessary. The analysis of the benchmark with the deterministic or Monte Carlo method demonstrates its usefulness both for improving analytical methods and for validating nuclear data.

Development of Urinary Biomarkers for Internal Exposure by Cesium-137 Using a Metabolomics Approach in Mice

Cesium-137 is a fission product of uranium and plutonium in nuclear reactors and is released in large quantities during nuclear explosions or detonation of an improvised device containing this isotope. This environmentally persistent radionuclide undergoes radioactive decay with the emission of beta particles as well as gamma radiation. Exposure to (137)Cs at high doses can cause acute radiation sickness and increase risk for cancer and death. The serious health risks associated with (137)Cs exposure makes it critical to understand how it affects human metabolism and whether minimally invasive and easily accessible samples such as urine and serum can be used to triage patients in case of a nuclear disaster or a radiologic event. In this study, we have focused on establishing a time-dependent metabolomic profile for urine collected from mice injected with (137)CsCl. The samples were collected from control and exposed mice on days 2, 5, 20 and 30 after injection. The samples were then analyzed by ultra-performance liquid chromatography coupled to time-of-flight mass spectrometry (UPLC/TOFMS) and processed by an array of informatics and statistical tools. A total of 1,412 features were identified in ESI(+) and ESI(-) modes from which 200 were determined to contribute significantly to the separation of metabolomic profiles of controls from those of the different treatment time points. The results of this study highlight the ease of use of the UPLC/TOFMS platform in finding urinary biomarkers for (137)Cs exposure. Pathway analysis of the statistically significant metabolites suggests perturbations in several amino acid and fatty acid metabolism pathways. The results also indicate that (137)Cs exposure causes: similar changes in the urinary excretion levels of taurine and citrate as seen with external-beam gamma radiation; causes no attenuation in the levels of hexanoylglycine and N-acetylspermidine; and has unique effects on the levels of isovalerylglycine and tiglylglycine.

Antineutrino monitoring of burning mixed oxide plutonium fuels

This letter presents the physics and feasibility of reactor antineutrino monitoring to verify the burnup of plutonium loaded in the reactor as a Mixed Oxide (MOX) fuel. It examines the magnitude and temporal variation in the antineutrino signals expected for different MOX fuels, for the purposes of nuclear accountability and safeguards. The antineutrino signals from reactor-grade and weapons-grade MOX are shown to be distinct from those from burning low enriched uranium. Thus, antineutrino monitoring could be used to verify the destruction of plutonium in reactors, though verifying the grade of the plutonium being burned is found to be more challenging.

Utilization of TRISO fuel with reactor grade plutonium in CANDU reactors

Abstract Large quantities of plutonium have been accumulated in the nuclear waste of civilian LWRs and CANDU reactors. Reactor grade plutonium and heavy water moderator can give a good combination with respect to neutron economy. On the other hand, TRISO type fuel can withstand very high fuel burn-up levels. The paper investigates the prospects of utilization of TRISO fuel made of reactor grade plutonium in CANDU reactors. TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 68%. The fuel compacts conform to the dimensions of CANDU fuel compacts are inserted in rods with zircolay cladding. In the first phase of investigations, five new mixed fuel have been selected for CANDU reactors composed of 4% RG-PuO 2 + 96% ThO 2 ; 6% RG-PuO 2 + 94% ThO 2 ; 10% RG-PuO 2 + 90% ThO 2 ; 20% RG-PuO 2 + 80% ThO 2 ; 30% RG-PuO 2 + 70% ThO 2 . Initial reactor criticality ( k ∞,0 values) for the modes , , , and are calculated as 1.4294, 1.5035, 1.5678, 1.6249, and 1.6535, respectively. Corresponding operation lifetimes are ∼0.65, 1.1, 1.9, 3.5, and 4.8 years and with burn ups of ∼30 000, 60 000, 100 000, 200 000 and 290 000 MW d/tonne, respectively. The higher initial plutonium charge is the higher burn ups can be achieved. In the second phase, a graphical-numerical power flattening procedure has been applied with radially variable mixed fuel composition in the fuel bundle. Mixed fuel fractions leading to quasi-constant power production are found in the 1st, 2nd, 3rd and 4th row to be as 100% PuO 2 , 80/20% PuO 2 /ThO 2 , 60/40% PuO 2 /ThO 2 , and 40/60% PuO 2 /ThO 2 , respectively. Higher plutonium amount in the flattened case increases reactor operation lifetime to >8 years and the burn up to 580 000 MW d/tonne. Power flattening in the bundle leads to higher power plant factor and quasi-uniform fuel utilization, reduces thermal and material stresses, and avoids local thermal peaks. Extended burn-up grade implies drastic reduction of the nuclear waste material per unit energy output for final waste disposal.