Fueled Core Research Articles

An improved assembly homogenization approach for plate-type research reactor

We have developed an improved assembly homogenization approach for plate-type research reactor. Compared with the conventional cell and assembly homogenization way, three advanced homogenization methods are used in our improved approach, including the direct single and multi-assembly calculations, treatment of the assembly environmental effect by multi-assembly calculations and assembly classification, and the superhomogenization (SPH) method. The approach is carried out with DRAGON v4 code. The neutronic analysis for fresh fueled core of JRR-3M with UAlx–Al dispersion type fuel is carried out by diffusion code CITATION with 3-group homogenized cross sections which are generated by the improved approach. Results by Monte Carlo code are taken as references, and results by the conventional three-step method are given for comparison purposes. The calculation results show that the improved assembly homogenization approach is effective to improve the accuracy of keff eigenvalue, control rod worth and neutron flux distributions for whole core calculations. Especially, the control rod worth can be calculated accurately by the improved approach.

Safety evaluation of accident-tolerant FCM fueled core with SiC-coated zircalloy cladding for design-basis-accidents and beyond DBAs

The FCM fueled cores proposed as an accident tolerant concept is assessed against the design-basis-accident (DBA) and the beyond-DBA (BDBA) scenarios using MARS code. A thermal conductivity model of FCM fuel is incorporated in the MARS code to take into account the effects of irradiation and temperature that was recently measured by ORNL. Preliminary analyses regarding the initial stored energy and accident tolerant performance were carried out for the scoping of various cladding material candidates.A 16×16 FA with SiC-coated Zircalloy cladding was selected as the feasible conceptual design through a preliminary scoping analysis. For a selected design, safety analyses for DBA and BDBA scenarios were performed to demonstrate the accident tolerance of the FCM fueled core. A loss of flow accident (LOFA) scenario was selected for a departure-from-nucleate-boiling (DNB) evaluation, and large-break loss of coolant accident (LBLOCA) scenario for peak cladding temperature (PCT) margin evaluation. A control element assembly (CEA) ejection accident scenario was selected for peak fuel enthalpy and temperature. Moreover, a station blackout (SBO) and LBLOCA without a safety injection (SI) scenario were selected as a BDBA. It was demonstrated that the DBA safety margin of the FCM core is satisfied and the time for operator actions for BDBA s is evaluated.

Tungsten impurity transport experiments in Alcator C-Mod to address high priority research and development for ITERa)

Experiments in Alcator C-Mod tokamak plasmas in the Enhanced D-alpha H-mode regime with ITER-like mid-radius plasma density peaking and Ion Cyclotron Resonant heating, in which tungsten is introduced by the laser blow-off technique, have demonstrated that accumulation of tungsten in the central region of the plasma does not take place in these conditions. The measurements obtained are consistent with anomalous transport dominating tungsten transport except in the central region of the plasma where tungsten transport is neoclassical, as previously observed in other devices with dominant neutral beam injection heating, such as JET and ASDEX Upgrade. In contrast to such results, however, the measured scale lengths for plasma temperature and density in the central region of these Alcator C-Mod plasmas, with density profiles relatively flat in the core region due to the lack of core fuelling, are favourable to prevent inter and intra sawtooth tungsten accumulation in this region under dominance of neoclassical...

Study the neutronic analysis and burnup for BWR fueled with hydride fuel using MCNPX code

This work presents the neutronic characteristics of advanced boiling water reactor (BWR) which fueled with hydride fuel. MCNPX Computer code based on Mont Carlo Method, is used to design a three dimensional model for BWR fuel assembly in a typical operating temperature and pressure conditions. A test case was compared with a benchmark problem and good agreement was found. This model is used to determine attainable discharge burnup, pin-by-pin power distribution, axial power distribution and reactivity effect of voiding. It is found that hydride fuel bundle design can be simplified by reducing the volume of water in-between the fuel bundles. The increase in the number of fuel rods per given core volume enables an increase in the BWR power density relative to oxide fueled core design.

Optimum Ratio of Coverings Thickness to the Diameter of the Fuel Core of the Dispersive Nuclear Fuel

As a result of the conducted calculation experiments, optimum ratio of thickness of coverings to the diameter of the fuel core of the dispersive nuclear fuel was determined. It provides maximum duration of the fuel operating periods. New information about parameters of internal block effect and resonance neutrons capture in a nuclear fuel in the energy range up to 100eV is received. The research has been carried out with the support of the Ministry of Education and Science of the Russian Federation, agreement № 14.B37.21.0473, 2012, August, 3rd.

Oxide-Metal Core Is Possible Transition to the Metal Fuel Core for Fast Reactors of the BN-Type

The article deals with the initial requirements for metal fuel elements and commercial reactors using this type of elements: high smear density of fuel, high technological effectiveness of fuel, fuel column and fuel elements; interchangeability of metal fuel elements and oxide fuel elements; achievement of fuel breeding ratio of the core BRC ~1.0; prolongation of refuelling cycle up to Tc~11 months. The article presents content and results of examinations of radiation-thermal effects in U fuel column elements and UPu fuel column elements: irradiation of breeding and fuel elements in the BOR-60 and BN-350 reactors; research of macroeffects of radiation growth, radiation swelling, fuel-cladding chemical interaction. Applicability of the obtained results for developing advanced BN-reactors with heterogeneous cores of various types was discussed: with by FA heterogenization of the core (BFAH); intra FA heterogenization of the core (IFAH); intra fuel element heterogenization (IFEH).

Preliminary reactor physics assessment of the HTR module with 14% enriched UCO fuel

Preliminary reactor physics assessment of the HTR module with 14% enriched UCO fuel

Sawtooth period control strategies and designs for improved performance

The sawtooth instability is associated with the triggering of neo-classical tearing modes, core fuelling, α-confinement and the exhaust of thermal helium. Sawtooth control is therefore important for optimal reactor performance in ELMy H-modes. Control schemes for the sawtooth period have been published in the literature, but the systematic design of high-performance controllers (yielding accurate and fast convergent responses) has not been addressed. In this work, three control strategies for high-performance sawtooth control are presented using electron cyclotron current drive (ECCD). Both degrees of freedom of the ECCD actuator will be explored and combined with advanced controller designs. First, the ECCD deposition location is used as a control variable, for which a gain-scheduled feedback controller and static feedforward control is derived. Second, the use of the driven current as a control variable is explored, and a simple controller is designed based on the identified dynamics. In the third approach both control variables are joined in an overall controller design, which enables the combination of high-performance control of the sawtooth period and control of the gyrotron power. Time-domain simulations with a combined Kadomtsev–Porcelli sawtooth model show that each strategy obtains a better closed-loop performance than standard linear feedback techniques on merely the deposition location.

Ideal magnetohydrodynamic simulations of unmagnetized dense plasma jet injection into a hot strongly magnetized plasma

We present results from three-dimensional ideal magnetohydrodynamic simulations of unmagnetized dense plasma jet injection into a uniform hot strongly magnetized plasma, with the aim of providing insight into core fuelling of a tokamak with parameters relevant for ITER and National Spherical Torus Experiment (NSTX). Unmagnetized dense plasma jet injection is similar to compact toroid injection but with much higher plasma density and total mass, and consequently lower required injection velocity. Mass deposition of the jet into the background appears to be facilitated via magnetic reconnection along the jet's trailing edge. The penetration depth of the plasma jet into the background plasma is mostly dependent on the jet's initial kinetic energy, and a key requirement for spatially localized mass deposition is for the jet's slowing-down time to be less than the time for the perturbed background magnetic flux to relax due to magnetic reconnection. This work suggests that more accurate treatment of reconnection is needed to fully model this problem. Parameters for unmagnetized dense plasma jet injection are identified for localized core deposition as well as edge localized mode (ELM) pacing applications in ITER and NSTX-relevant regimes.

Nuclear mechanism for TRU burning oxide fueled core in Advanced Recycling Reactor

This paper presents an approximation approach to predict the core characteristics based on parametric survey and an analysis of nuclear mechanism in a conceptual nuclear design for enhanced transuranics (TRU) burning mixed oxide fueled and sodium cooled fast reactor which can be realized in the near future. The design study of Advanced Recycling Reactor was conducted in the context of the program for the industry in Global Nuclear Energy Partnership initiatives, including a core in the first plant for demonstration and cores with enhanced TRU burning capability for the future plants. Both concepts for the first plant; low core height and large volume fraction of structure are deployed, seeking small TRU conversion ratio (CR) * and small void reactivity which are crucial in the design, but different approaches. In this paper, the TRU CR and the sodium void reactivity have been approximated with a single equation in these concepts, based on the theoretical formula related to the chain reaction in the reactor and the calculation results. Shortening the core height and increasing the structure volume fraction will enhance TRU enrichment through increased neutron leakage and capture, which will reduce a ratio of U-238 to sum of Pu-239 and Pu-241 so that TRU CR decreases from 0.78 to 0.57. A small ratio of sodium loss to plutonium fissile will be effective also in the reduction of positive reactivity effect by spectral hardening. On the other hand, when this ratio and geometrical buckling of flux reduce, negative contribution by the neutron leakage becomes small. Theses relations related to the positive void reactivity have been formularized by the approximation with few parameters within several percent respectively as well as the TRU CR, indicating that one of dominating parameters is the ratio of sodium loss to plutonium fissile in the void reactivity at large fast reactors. * = (1 − net loss of TRU/loss of heavy metal).

Postirradiation Testing of High Temperature Reactor Spherical Fuel Elements Under Accident Conditions

A new furnace for accident condition testing of spherical high temperature reactor fuel elements has been installed and now operates in the hot cells of the Institute for Transuranium Elements (ITU) Karlsruhe. The recent apparatus was constructed on the basis of a former development by Forschungszentrum Jülich (Schenk, Pitzer, and Nabielek, 1988, “Fission Product Release Profiles From Spherical HTR Fuel Elements at Accident Temperatures,” Jülich Report No. 2234), where it was named KüFA, the German acronym for cold finger apparatus. In a preceding publication (Kostecka, Ejton, de Weerd, and Toscano, 2004, “Post-Irradiation Testing of HTR-Fuel Elements Under Accident Conditions, Part 1 and 2,” Second International Topical Meeting on High Temperature Reactor Technology, Beijing, China) the general concept and details of the device were described. The present paper reports on the first operation under hot conditions, and the calibration of the fission gas measurement and of the efficiency of the cold finger, which is used to plate out solid fission products. Finally the results of fission product release and analysis of two heating tests on two fuel elements from the high temperature reactor K6 irradiation experiment (Nabielek, Conrad, Roellig, and Meyers, 1993, “Fuel Irradiation Experiments on HFR-K6 and HFR-B1 With Intermittent Water Vapour Injections,” Technical Committee Meeting on Response of Fuel, Fuel Elements and Gas Cooled Reactor Cores Under Accidental Air or Water Ingress Conditions, Beijing, China, Oct. 25–27) are presented and discussed.

Advanced Recycling Core Accommodating Oxide Fuel and Metal Fuel for Closed Fuel Cycle

This report presents a unique TRU burning core capable of accommodating oxide fuel and metal fuel and easy to change oxide core to metal core conforming to the design requirements. For the homogeneous oxide fueled core containing transuranics (TRU) fuel with 12% of the moderator pins, the results of calculation show the TRU conversion ratio (ratio of loss of TRU to loss of heavy metal) of 0.33 and the TRU burning capability (ratio of loss of TRU per electric generation) of 67 kg/TWeh. On the other hand, the calculations replacing from oxide fuel assemblies to metal fuel assemblies have indicated the TRU transmutation capability of 69 kg/TWeh with the TRU conversion ratio of 0.30. As the result of simulation calculations, three ordinary fuel exchanges transform the oxide equilibrium core to the full metal core by way of transitional cores, where the maximum linear heat rates are still equal to the metal equilibrium core or less. With this, the presented core concept is concluded that a full oxide core, a full metal core, mixed fueled cores can be materialized in the presented first unit of Advanced Recycling Reactor (ARR1).

Ideal magnetohydrodynamic simulations of low beta compact toroid injection into a hot strongly magnetized plasma

We present results from three-dimensional ideal magnetohydrodynamic simulations of low β compact toroid (CT) injection into a hot strongly magnetized plasma, with the aim of providing insight into CT fuelling of a tokamak with parameters relevant for the International Thermonuclear Experimental Reactor (ITER). A regime is identified in terms of CT injection speed and CT-to-background magnetic field ratio that appears promising for precise core fuelling. Shock-dominated regimes, which are probably unfavourable for tokamak fuelling, are also identified. The CT penetration depth is proportional to the CT injection speed and density. The entire CT evolution can be divided into three stages: (1) initial penetration, (2) compression in the direction of propagation and reconnection with the background magnetic field, and (3) coming to rest and spreading in the direction perpendicular to injection. Tilting of the CT is not observed due to the fast transit time of the CT across the background plasma.

High-density plasma with internal diffusion barrier in the Large Helical Device

An attractive high-density operational regime which is a so-called internal diffusion barrier (IDB) has been discovered in a helical divertor configuration on the Large Helical Device (LHD). The IDB is characterized by steep density gradients and the plasma profile is divided by the IDB into a high-density core plasma and a low density mantle plasma. The IDB enables the core plasma to access the high-density/high-pressure regime. The attainable central density exceeds 1 × 1021 m−3 and the central pressure reaches ≈1.5 times atmospheric pressure. Core pellet fuelling is absolutely essential for the IDB formation and it is reproducibly obtained by employing intensive multiple-pellet injection. In the IDB core plasma, the particle diffusion coefficient is kept at a considerably low level, 0.05 m2 s−1, in spite of high-density and steep-density gradients whereas an inward particle convection velocity is not observed.

Thermal hydraulic analysis for grid supported pressurized water reactor cores

This paper presents the methodology and results for thermal hydraulic analysis of grid supported pressurized water reactor cores using U(45% wt)–ZrH 1.6 hydride fuel in square arrays. The same methodology is applied to the design of UO 2 oxide fueled cores to provide a fair comparison of the achievable power between the two fuel types. Steady-state and transient design limits are considered. Steady-state limits include: fuel bundle pressure drop, departure from nucleate boiling ratio, fuel temperature (average for UO 2 and centerline/peak for U–ZrH 1.6), and fuel rod vibrations and wear. Transient limits are derived from consideration of the loss of flow and loss of coolant accidents, and an overpower transient. In general, the thermal hydraulic performance of U–ZrH 1.6 and UO 2 fuels is very similar. Slight power differences exist between the two fuel types for designs limited by rod vibrations and wear, because these limits are fuel dependent. Large power increases are achievable for both fuels when compared to the reference core power output of 3800 MW th. In general, these higher power designs have smaller rod diameters and larger pitch-to-diameter ratios than the reference core geometry. If the pressure drop across new core designs is limited to the pressure drop across the reference core, power increases of ∼400 MW th may be realized. If the primary coolant pumps and core internals could be designed to accommodate a core pressure drop equal to twice the reference core pressure drop, power increases of ∼1000 MW th may be feasible.

Plutonium recycling in hydride fueled PWR cores

The classic approach to the recycling of Pu in PWR is to use mixed U-oxide Pu-oxide (MOX) fuel. The mono-recycling of plutonium in PWR transmutes less than 30% of the loaded plutonium, providing only a limited reduction in the long-term radiotoxicity and in the inventory of TRU to be stored in the repository. The primary objective of this study is to assess the feasibility of plutonium recycling in PWR in the form of plutonium hydride, PuH 2, mixed with uranium and zirconium hydride, ZrH 1.6, referred to as PUZH, that is loaded uniformly in each fuel rod. The assessment is performed by comparing the performance of the PUZH fueled core to that of the MOX fueled core. Performance characteristics examined are transmutation effectiveness, proliferation resistance of the discharged fuel and fuel cycle economics. The PUZH loaded core is found superior to the MOX fueled core in terms of the transmutation effectiveness and proliferation resistance. For the reference cycle duration and reference fuel rod diameter and pitch, the percentage of the plutonium loaded that is transmuted in one recycle is 53% for PUZH versus 29% for MOX fuel. That is, the net amount of plutonium transmuted in the first recycle is 55% higher in cores using PUZH than in cores using MOX fuel. Relative to the discharged MOX, the discharged PUZH fuel has smaller fissile plutonium fraction – 45% versus 60%, 15% smaller minor actinides (MA) inventory and more than double spontaneous fission neutron source intensity and decay heat per gram of discharged TRU. Relative to the MOX fuel assembly, the radioactivity of the PUZH fuel assembly is 26% smaller and the decay heat and the neutron yield are only 3% larger. The net effect is that the handling of the discharged PUZH fuel assembly will be comparable in difficulty to that of the discharged MOX assembly while the proliferation resistance of the TRU of the discharged PUZH fuel is enhanced.

A Design Study of High Breeding Ratio Sodium Cooled Metal Fuel Core without Blanket Fuels

A Design Study of High Breeding Ratio Sodium Cooled Metal Fuel Core without Blanket Fuels

Internal transport barrier simulation with pellet injection in tokamak and helical reactor plasmas

In the future fusion reactor, the plasma density peaking is important for the increase in the fusion power gain. The density control and the internal transport barrier (ITB) formation due to the pellet injection have been simulated in tokamak and helical reactors using the toroidal transport linkage code TOTAL. Firstly, the pellet injection simulation is carried out including the neutral gas shielding model and the mass relocation model in the TOTAL code, and the effectiveness of the high field side (HFS) pellet injection is clarified. Secondly, the ITB simulation with the pellet injection is carried out with the confinement improvement model based on the E×B shear effects, and it is found that the deep pellet penetration is helpful for the ITB formation as well as the plasma core fuelling in the reversed shear tokamak reactor, but the deep pellet penetration is not effective in the helical reactor.

Internal Transport Barrier Formation and Pellet Injection Simulation in Helical and Tokamak Reactors

In the future fusion reactor, plasma density peaking is important for increase in the fusion power gain and for achievement of confinement improvement mode. Density control and internal transport barrier (ITB) formation due to pellet injection have been simulated in tokamak and helical reactors using the toroidal transport linkage code TOTAL. First, pellet injection simulation is carried out, including the neutral gas shielding model and the mass relocation model in the TOTAL code, and the effectiveness of high-field side (HFS) pellet injection is clarified. Second, ITB simulation with pellet injection is carried out with the confinement improvement model based on the E × B shear effects, and it is found that deep pellet penetration is helpful for ITB formation as well as plasma core fuelling in the reversed-shear tokamak and helical reactors.

Extended steady-state and high-beta regimes of net-current free heliotron plasmas in the Large Helical Device

The performance of net-current free heliotron plasmas has been developed by findings of innovative operational scenarios in conjunction with an upgrade of the heating power and the pumping/fuelling capability in the Large Helical Device (LHD). Consequently, the operational regime has been extended, in particular, with regard to high density, long pulse length and high beta. Diversified studies in LHD have elucidated the advantages of net-current free heliotron plasmas. In particular, an internal diffusion barrier (IDB) by a combination of efficient pumping of the local island divertor function and core fuelling by pellet injection has realized a super dense core as high as 5 × 1020 m−3, which stimulates an attractive super dense core reactor. Achievements of a volume averaged beta of 4.5% and a discharge duration of 54 min with a total input energy of 1.6 GJ (490 kW on average) are also highlighted. The progress of LHD experiments in these two years is overviewed by highlighting IDB, high β and long pulse.