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This research is focused on using Thorium-Plutonium MOX fuel in the inner fuel pins of the CANDU fuel bundles for plutonium incineration and reduction of uranium demand and to reduce coolant void reactivity. The delayed neutron fraction and the power distribution amongst the fuel elements of the fuel bundle have been considered as main safety parameters. The 700 MWe Advanced CANDU Reactor (ACR-700) was selected as a case study. The inner eight UO 2 fuel pins of the ACR-700 fuel bundle are replaced by Thorium-Plutonium MOX fuel pins in the proposed design with 3% reactor grade PuO 2 . This amount represents 23.4 w/o of the fuel in the bundle. The outer two fuel rings (35 pins) enrichment is reduced from 2.1 w/o U-235 to 2 w/o U-235. The simulation using MCNP6 showed that about 27% reduction of uranium demand can be achieved. The proposed fuel bundle eliminate the use of burnable poisons in the central pin that was used for negative coolant void reactivity and more reduction in the coolant void reactivity was achieved (about 3.5 mk less than the reference fuel bundle). The power distribution throughout the fuel bundle is more flat in the proposed fuel bundle. Use of this fuel bundle reduces the delayed neutron fraction from 540 pcm in the reference case to 480 pcm in the proposed case.

Final disposition of spent nuclear fuel in a permanent repository is still on debate; meanwhile the stockpile of depleted fuel is increasing. Actinide recycling has the potential to reduce that stockpile. This study analyzes the potential to burn actinide from depleted fuel in a thermal reactor without reducing the nominal electricity generation of the reactor and meeting all the safety constraints. The study takes an equilibrium cycle of a Boiling Water Reactor (BWR) as the reference cycle and then it proposes the design of an equivalent fuel assembly in which 4 fuel rods are replaced by minor actinide-bearing rods, the process ends when the equilibrium cycle is met. The new BWR fuel assembly design, that is called MA-c, has the goal to maximize the actinide recycling while introducing the lowest perturbation possible to the control and safety systems of the reactor. MA-c is successful, according to our analysis. The performance of the core is analyzed using MA-c and it is compared against the UO 2 reference core, results show that both of these cores produce the same amount of energy for an 18-month cycle. The MA-c fuel assembly has a higher enrichment than the UO 2 reference fuel assembly to compensate for the use of four MA-bearing rods, however all the fuel pins inside the MA-c fuel assembly have enrichments below 5 w/o of U-235. The MA-c core met all the safety BWR constraints without modifying any of the control systems. Results for a BWR with a nominal power of 2027 MW t show a reduction of 135.30 kg HM of minor actinide per 18-month operating cycle when the whole core is composed of “once-through” MA-c fuel assemblies. The use of this new BWR fuel assembly design can be used effectively for the reduction of minor actinide without the need of control systems modification in the reactor to meet safety limits and without affecting the energy production.

The fuel element failure in an operating pressurized water reactor (PWR), including fuel element breaks, has an effect on the operation safety of PWR. In this paper, the RELAP5 model of the fuel element failure is established for the safety analysis. The RELAP5 time step sensitivity analyses for the element pre-break steady and post-break transient simulation are carried out. And the variations of main thermal-hydraulics parameters related to the fuel element break are quantitatively studied, which include the internal gap pressure and the maximum fuel pellet temperature as well as the releasement of noncondensables in the gap. It is found that (1) the results by the RELAP5 code is very sensitive to the time step in a volume system with the noncondensables, and the time step sensitivity analysis is necessary if the effective time step range is unknown, (2) the larger the break area is, the more quickly the gap pressure increases and the maximum pellet temperature reaches to the stable value, (3) when the gap pressure increases and reaches to the coolant pressure, at the break the liquid inflow from coolant to gap will be turned to the vapor outflow from gap to coolant, (4) during the failure transient, the gap thermal conductivity experiences a sharp decrease in the break instant, which results in the decrease of heat transferred to cladding and the sharp decrease of cladding temperature as well as the sharp increase of minimum departure from nucleate boiling ratio (MDNBR). These conclusions can provide the basic for the operation safety analysis of PWR during the fuel element failure.

Calculation of the primary circuit's coolant activation due to fission products (FPs) has been investigated for the eastern-type pressurized water reactor (VVER1000-V446). The reactor has been considered under normal full power operational condition for the first fuel cycle. Determination of the reactor coolant activity is based on time-dependent fission product core inventories. ORIGEN2.1 code has been used to determine the time-dependent fission product core inventories. The fission products activity in the primary coolant is calculated using a set of ordinary differential equations (ODEs) which governs the FPs concentration in the primary coolant. Results for 87 FPs have been calculated. The results of these calculations have been found to agree well with the corresponding available values found in the Final Safety Analysis Report (FSAR) of the Bushehr Nuclear Power Plant (BNPP).

In this paper, a model for the actinide redistribution mechanism in mixed oxide fuel for a sodium fast reactor is presented and compared with measured data collected from post irradiation examination on fuel from the JOYO experimental reactor. The model considers that solid-state thermal diffusion and vapour transport can simultaneously contribute to the plutonium and americium radial profiles. The effect of fuel non-stoichiometry on actinide transport as well as on pore velocity is taken into account. The model is embedded into the TRANSURANUS fuel rod performance code and calculates the evolution of the Am and Pu concentrations as a function of the radial temperature profile. The calculated actinide distribution is in good agreement with the experimental data. The results confirm that under normal operation conditions with a decreasing fuel power during its life-time, redistribution via pore migration is extremely important only during the central hole formation at the beginning of life, whereas thermal diffusion represents the dominant effect for long-term irradiation.

The Cold Finger Apparatus (KuFA) is used to examine the combined effects of Depressurization and Loss of Forced Circulation accident scenarios on High Temperature Reactor fuel pebbles, constituted of a graphite matrix containing sub-mm-sized fuel particles with tri-isotropic coating. A fuel pebble is heated in He atmosphere for several hundred hours, mimicking accident temperature transients up to 1800 °C. Volatile fission products released from the fuel condense on a water cooled stainless steel plate (“Cold Finger”) positioned above the tested pebble in the furnace. Released fission gases are collected downstream in a charcoal trap cooled with liquid nitrogen and gamma monitored with a NaI detector. Exchanging plates frequently during the experiment and analyzing plate deposits by means of High Purity Germanium gamma spectroscopy allows a reconstruction of the fission product release as a function of time and temperature. In order to achieve a good quantification of the release, a careful calibration of the setup is mandatory. The analysis of condensation plates from recent KuFA tests shows that fission product release quantification is possible at high and low activity levels. An especially tailored collimator allows plate scanning with high spatial resolution, thus yielding information about the fission product distribution on the condensation plates.

The objective of this study is to develop an optimized BWR fuel assembly design for thorium–plutonium fuel. In this work, the optimization goal is to maximize the amount of energy that can be extracted from a certain amount of plutonium, while maintaining acceptable values of the neutronic safety parameters such as reactivity coefficients, shutdown margins and power distribution. The factors having the most significant influence on the neutronic properties are the hydrogen-to-heavy-metal ratio, the distribution of the moderator within the fuel assembly, the initial plutonium fraction in the fuel and the radial distribution of the plutonium in the fuel assembly. The study begins with an investigation of how these factors affect the plutonium requirements and the safety parameters. The gathered knowledge is then used to develop and evaluate a fuel assembly design. The main characteristics of this fuel design are improved Pu efficiency, very high fractional Pu burning and neutronic safety parameters compliant with current demands on UOX fuel.

In this paper the transmutation of light water reactors (LWR) spent fuel is analyzed. The system used for this study is the fusion-fission transmutation system (FFTS). It uses a high energy neutron source produced with deuterium-tritium fusion reactions, located in the center of the system, which is surrounded by a fission region composed of nuclear fuel where the fissions take place. In this study, the fuel of the fission region is obtained from the recycling of LWR spent fuel. The MCNPX Monte Carlo code was used to setup a model of the FFTS. Two fuel types were analyzed for the fissile region: the mixed oxide fuel (MOX), and the inert matrix fuel (IMF). Results show that in the case of the MOX fuel, an important Pu-239 breeding is achieved, which can be interesting from the point of view of maximal uranium utilization. On the contrary, in the case of the IMF fuel, high consumption of Pu-239 and Pu-241 is observed, which can be interesting from the point of view of non-proliferation issues. A combination of MOX and IMF fuels was also studied, which shows that the equilibrium of actinides production and consumption can be achieved. These results demonstrate the versatility of the fusion-fission hybrid systems for the transmutation of LWR spent fuel.

A fusion–fission hybrid reactor is proposed to achieve the energy gain of 3000 MW thermal power with self-sustaining tritium. The hybrid reactor is designed based on the plasma conditions and configurations of ITER, as well as the well-developed pressurized light water cooling technologies. For the sake of safety, the pressure tube bundles are employed to protect the first wall from the high pressure of coolant. The spent nuclear fuel discharged from 33GWD/tU Light Water Reactors (LWRs) and natural uranium oxide are taken as driver fuel for energy multiplication. According to thermo-mechanics calculation results, the first wall of 20 mm is safe. The radiation damage analysis indicates that the first wall has a lifetime of more than five years. Neutronics calculations show that the proposed hybrid reactor has high energy multiplication factor, tritium breeding ratio and power density; the fuel cannot reach the level of plutonium required for a nuclear weapon. Thermal-hydraulic analysis indicates that the temperatures of the fuel zone are well below the limited values and a large safety margin is provided.

The use of thorium in pressurized water reactor fuel assemblies is investigated in this paper. The novelty of the reported work is to study a fuel design primarily intended to control the excess of reactivity at beginning of life, and flatten the intra-assembly power distribution rather than converting fertile Th-232 into fissile U-233. The fuel assembly is a traditional 17 x 17 pressurized water reactor fuel design. The majority of the fuel pins contain a mixture of uranium and thorium oxides, while a few fuel pins contain a mixture between uranium and gadolinium oxides. The calculation were performed by two-dimensional transport calculations with the Studsvik Scandpower CASMO-4E code in order to determine the main neutronic properties of the new fuel design, compared with the traditional uranium-based fuel assembly containing gadolinium used as reference. The majority of the neutronic properties of the uranium-thorium-based fuel assembly were similar to the reference fuel assembly. The Doppler and the moderator temperature coefficients of reactivity were found to be appreciably more negative in the uranium-thorium-based design, but still within acceptable limits. One advantage of this new uranium-thorium-based design is a reduction of the pin peak power at beginning of life, because of smaller amount of gadolinium being used. This is important from an operational and safety viewpoint, since the margin to departure from nucleate boiling becomes larger. Consequently, this new type of thorium-based fuel assembly shows advantageous properties for use in power-uprated cores.

MTR fuel elements burned-up inside the core of nuclear research reactors are stored worldwide mainly under the water of storage pools. When cladding breach is present in one or more fuel plates of such elements, radioactive fission products are released into the storage pool water. This work proposes a model to describe the release mechanism considering the diffusion of nuclides of a radioactive fission product either through a postulated small cylindrical breach or directly from a large circular hole in the cladding. In each case, an analytical expression is obtained for the activity released into the water as a function of the total storage time of a breached fuel plate. Regarding sipping tests already performed at the IEA-R1 research reactor on breached MTR fuel elements, the proposed model correlates successfully the specific activity of 137Cs, measured as a function of time, with the evaluated size of the cladding breach.

Gas and Vapor Core Reactors (G/VCR) are externally reflected and moderated nuclear energy systems fueled by stable uranium compound in gaseous or vapor phase. In G/VCR systems the functions of fuel and coolant are combined and the reactor outlet temperature is not constrained by solid fuel-cladding temperature limitations. G/VCRs can potentially provide the highest reactor and cycle temperature among all existing or proposed fission reactor designs. Furthermore, G/VCR systems feature a low inventory and fully integrated fuel cycle with exceptional sustainability and safety characteristics. With respect to fuel utilization, there is practically no fuel burn-up limit for gas core reactors due to continuous recycling of the fuel. Owing to flexibility in nuclear design characteristics of cavity reactors, a wide range of conversion ratio from almost solely a burner to a breeder is achievable. The continuous recycling of fuel in G/VCR systems allows for continuous burning and transmutation of actinides without removing and reprocessing of the fuel. The only waste product at the backend of the gas core reactors' fuel cycle is fission fragments that are continuously separated from the fuel. As a result the G/VCR systems do not require spent fuel storage or reprocessing. G/VCR systems also feature outstanding proliferation resistance characteristics and minimum vulnerability to external threats. Even for comparable spectral characteristic, gas core reactors produce fissile plutonium two orders of magnitude less than Light Water Reactors (LWRs). In addition, the continuous transmutation and burning of actinides further reduces the quality of the fissile plutonium inventory. The low fuel inventory (about two orders of magnitude lower than LWRs for the same power generation level) combined with continuous burning of actinides, significantly reduces the need for emergency planning and the vulnerability to external threats. Low fuel inventory, low fuel heat content, and online separation of fission fragments are among the key constituent safety features of G/VCR systems.

New concept of a passive-safety reactor “KAMADO” has a negligible possibility of core melting and flexibility of total reactor power. The reactor core of KAMADO consists of fuel elements of graphite blocks, which have UO2 fuel rods and cooling water holes. These fuel elements are located in a reactor water pool of atmospheric pressure (1 atm) and low temperature (< 60°C). In case of LOCA, decay heat from fuel rods is removed by conduction heat transfer to the reactor water pool. Since the cooling water does not contact a fuel rod directly, core design has much flexibility without considering dry-out limitation and Minimum Critical Power Ratio (MCPR). Additionally an effective use of spent fuel is expected.

The BREST fast reactor with nitride fuel and lead coolant is being developed as a reactor of new generation, which has to meet a set of requirements placed upon innovative reactors, namely efficient use of fuel resources, nuclear, radiation and environmental safety, proliferation resistance, radwaste treatment and economic efficiency. Mixed uranium-plutonium mononitride fuel composition allows supporting in BREST reactor CBR≈1. It is not required to separate plutonium to produce “fresh” fuel. Coarse recovered fuel purification of fission products is allowed (residual content of FPs may be in the range of 10−2 – 10−3 of their content in the irradiated fuel). High activity of the regenerated fuel caused by minor actinides is a radiation barrier against fuel thefts. The fuel cycle of the BREST-type reactors “burns” uranium-238, which must be added to the fuel during reprocessing. Plutonium is not extracted during reprocessing being a part of fuel composition, thus exhibiting an important nonproliferation feature. The radiation equivalence between natural uranium consumed by the BREST NPP closed system and long-lived high-level radwaste is provided by actinides (U, Pu, Am) transmutation in the fuel and long-lived products (I, Tc) transmutation in the blanket. The high-level waste must be stored for approximately 200 years to reduce its activity by the factor of about 1000. The design of the building and the entire set of the fuel cycle equipment has been completed for the demonstration BREST-OD-300 reactor, which includes all main features of the BREST-type reactor on-site closed fuel cycle.

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.

The use of a high-density low-enriched uranium fuel in low-power reactors was investigated. These reactors can be run on this fuel without changing either dimensions of the fuel meat or the cladding. The new reactor is safer in the accidental conditions but it is not in the normal operation. The power of the reactor is higher by about 19.5%. The initial excess reactivity is reduced to 3.5950 instead of 3.980 mk. Fifteen new dummy rods are added to the core of the reactor substituting as many as fuel rods. No changes to the cladding material are required.

In both the reactor portion and the fuel-cycle portion of the Integral Fast Reactor (IFR), handling, treatment and disposal of wastes are simpler than in current fuel cycles. The vast majority (> 99.9%) of the very-long-lived radioactive TRU elements are not sent to the repository; rather, they are recycled. High-level waste volume from the IFR process (called “the pyroprocess”) is lower than that from either the direct disposal of spent fuel or from conventional PUREX-type reprocessing. The quantity of low-level waste is very low. In the pyroprocess, the actinides are recovered and separated from the bulk of the fission products by an electrorefining step wherein the actinides are electrotransported from chopped fuel elements and deposited at cathodes. The volatile fission products xenon, krypton, and tritium are collected for long-term storage and decay. Zirconium and the “noble metal” fission products (those that are less easily oxidized than zirconium) remain in the anode compartment, to be removed with the fuel cladding fragments and made into a metal waste form. The remaining fission products collect in the salt as chlorides. A process has been developed to periodically remove the contaminated salt from the electrorefiner, separate most of the fission products, and return the purified salt in a form that is ready for continuing use. To clean up the electrorefiner salt, the fission products are removed by ion exchange onto a column of Zeolite A. After the purification step, the column material and the contained fission products are converted to a mineral waste form for disposal. The processes and equipment for waste isolation and conversion to suitable disposal forms are described in this paper.

Once-through burning process has been proposed for the disposition of excess plutonium. A new stable fuel material of multi-phases is fabricated based on conventional MOX fuel technologies. After irradiation in LWR, the spent fuels would be geologically stable, and become high level radioactive waste (HLW) without further processing. From the chemical properties and crystal structures, two oxide systems have been proposed: PuO 2 -ThO 2 - Al 2 O 3 -MgO and PuO 2 -ZrO 2 (Y,GD)-Al 2 O 3 -MgO systems. The experimental study has been made to examine the phase relations of the fuel materials and the distribution of fission products. From the burnup calculation study, it was estimated more than 80% of plutonium is transmuted after irradiation and the quality of plutonium becomes very poor. For the zirconia type fuel (PuO 2 ZrO 2 (Y,Gd)-Al 2 O 3 ), the void and Doppler reactivities were calculated to be very small. The two modified fuel systems are therefore proposed to avoid these small reactivities, i.e. the zirconia-thoria fuel system (PuO 2 -(Zr,Th)O 2 -Al 2 O 3 -MgO) and zirconia fuel system with W or Yb additive (PuO 2 -ZrO 2 (Yb)-Al 2 O 3 -MgO-(W).

The fast reactor designs for transuranium and long-lived fission product confining equilibrium fuel cycle are proposed and analyzed. Lead-cooled fast reactor concept is proved to be a candidate for equilibrium phase of Self-Consistent Nuclear Energy System.

A fuel cycle concept for Self-Consistent Nuclear Energy System which releases no fission products to the environment is proposed. Material and energy balances are shown for this system.