Fueled Core Research Articles

Divertor modelling and extrapolation to reactor conditions

Recent results of divertor modelling are summarized and some modelling results incorporated in the ITER-FEAT divertor design, such as the beneficial effect of V-shaped targets, the importance of high gas conductivity between the divertors, and the role of deep core fuelling in maintaining the plasma density, are discussed. Differences between carbon and seeded impurities are pointed out, and the beneficial effect of helium elastic scattering is demonstrated. A number of considerations for modelling reactor-like operation and additional requirements imposed on the code under these conditions are discussed.

Feasibility study of large MOX fueled FBR core and applicability of various coolants and fuels aimed at the Self-Consistent Nuclear Energy System

The potential of a large MOX fueled fast breeder reactor (FBR) is evaluated with regard to its ability to transmute radioactive nuclides and its safety when incorporated in the self-consistent nuclear energy system (SCNES). The FBR's annual production amounts of selected long -lived fission products (LLFPs), Se-79, Tc-99, Pd-107, I-129, Cs-135 and Sm-151, can be transmuted by using a radial blanket region and part of a lower axial blanket region without any significant impact on the reactor's nuclear and safety characteristics. The other LLFPs are confined in the system. The hazard index level of the LLFPs per one ton of spent fuel from the system after 1000 years is as small as that of a typical uranium ore. To realize self-controllability (passive safety), the proposed FBR core concept employs gas expansion modules and a sodium plenum above the core. To realize self-terminability, even if MOX fuel melting should cause a core compaction, re-criticality of the core can be avoided by a fuel dilution and relocation module. The results show the MOX fueled FBR core has potential applicability to the SCNES. The fundamental applicability of various coolants and fuels is evaluated based on neutron balance toward the final goal of the ideal SCNES. The results show that gas coolant has a potential for increasing the transmutation efficiency of LLFPs. And an improved SCNES with several conventional FBRs and a FP transmutation reactor is also studied.

Feasibility study on nitride fuel core and recycling system toward Self-Consistent Nuclear Energy System

The concept of a nuclear fuel recycle system with a nitride fueled FBR core has been investigated as a part of related studies towards the Self-Consistent Nuclear Energy System (SCNES). Nitride fuel has been given attention because of its relatively high fuel density and high thermal conductivity. To materialize the SCNES concept, it is important to adequately use the excess neutrons produced in the chain reaction. The high fuel density of the nitride fuel brings out more of the excess neutrons and has a higher potential to transmute the long-lived fission products (LLFP's). The high thermal conductivity, in addition, provides margin of fuel melting, and gives negative feedback due to the Doppler reactivity in unprotected loss of flow accidents. In this paper, we discuss good use of nitride fuel in the SCNES.

Elimination of recriticality potential for the Self-Consistent Nuclear Energy System

The ultimate safety goal of the Self-consistent Nuclear Energy System (SCNES) is to eliminate the recriticality-problem based on a simple safety logic. The principle of the elimination of the recriticality-problem is the Controlled Material Relocation (CMR) to establish the neutronic shutdown by removing the molten fuel to the out of core before a large scale pool formation which has potential of energetics driven by a super prompt criticality. The CMR concept should be reliable without significant impact on the core neutronic performance. As the typical core concepts to enhance this CMR characteristic, several design options are under consideration. They are fuel assemblies with inner duct structure (FAIDUS), fuel assemblies with hollow fuel pins in the axial blanket region (ABLE) for MOX fueled cores, and fuel assemblies without fuel pin bundle structure in the lower axial blanket region (ELAB) for the metallic fueled core. Based on the core design study and accident analyses, these CMR-oriented concepts have been found feasible without significant degradation of the neutronic performance In order to experimentally confirm the effectiveness of the CMR concept for the MOX fueled core, the EAGLE project has been started in 1998 by Japan Nuclear Cycle Development Institute (JNC) and The Japan Atomic Power Company (JAPC). The EAGLE project is the experimental program utilizing the out-of pile test facility and in-pile facility IGR of the National Nuclear Center of the Republic of Kazakhstan (NNC/RK).

A subchannel analysis code MATRA-LMR for wire wrapped LMR subassembly

In sodium cooled liquid metal reactors design limits are imposed on the maximum temperatures of the cladding and fuel pins. Thus an accurate prediction of the core coolant/fuel temperature distribution is essential to LMR core thermal hydraulic design. The detailed subchannel thermal hydraulic analysis code MATRA-LMR is being developed for LMFBR core design and analysis based on COBRA-IV-I and MATRA. The major modifications and improvements implemented in MATRA-LMR are as follows: sodium property calculation subprogram, sodium coolant heat transfer correlations, and most recent pressure drop correlations. To assess the development status of this code, benchmark calculations were performed with the ORNL 19 pin tests and EBR-II seven-assembly SLTHEN calculation results. The calculation results of MATRA-LMR were compared to the measurements and to the SABRE4 and SLTHEN code calculation results, respectively. Finally, the major technical results of the conceptual design for the KALIMER U-10%Zr binary alloy fueled core have been compared with the calculations of the MATRA-LMR, SABRE4 and SLTHEN codes.

Ergodic divertor experiments on the route to steady state operation of Tore Supra

Ergodic divertor operation on Tore Supra is characterized by good performance in terms of divertor physics. Control of particle recirculation and impurity screening are related to the symmetry, both poloidally and toroidally, of the shell of open field lines and to its radial extent, Δr ≈ 0.16 m. Feedback control of the divertor plasma temperature has led to controlled radiative divertor experiments. In particular, good performance is obtained when the plasma is controlled to be at a temperature comparable to the energy involved in the atomic processes (15-20 eV). For standard discharges with 5 MW total power and ICRH heating, the low parallel energy flux ≈ 10 MW m-2 is reduced to ≈ 3 MW m-2 with nitrogen injection. This is achieved at a modest cost in core dilution, ΔZeff ≈ 0.3. Despite the large volume of open field lines ( ≈ 36%), the ergodic divertor does not reduce the possible current in the discharge since stable discharges are achieved with qsep ≈ 2. It is shown that the reorganization of the current profile in conjunction with a transport barrier in the electron temperature on the separatrix stabilizes the (2,1) tearing mode. Confinement follows the standard L mode confinement. In a few cases at high density and with no gas injection (wall fuelled discharges), `RI-like' modes are reported with a modest increase in confinement ( ≈ 40%). Despite the lack of core fuelling on Tore Supra, high densities during ICRH pulses can be achieved with Greenwald fractions fG ≈ 1. Compatibility with both ICRH and LH is demonstrated. In particular, long pulse operation with a flat-top in excess of 20 s is achieved with LHCD.

Mechanisms of Extending Operation Region in the HL-1M Tokamak

Stable operating region in the HL-1M tokamak has been extended by means of wall conditioning, core fuelling and current control techniques. The mechanisms of the extension are analyzed in this paper. Lithiumization diminishes the impurities and hydrogen recycling to the lowest level. After lithiumization a high density up to 7×1019 m-3 was obtained easily by strong gas puffing with ordinary ohmic discharge alone. More attractively we found that metal Li-coating exhibited the effects of wall stabilization. The low qa limit with higher density was extended by a factor of 1.5-2 in comparison with that for boronization, and 1.2 for siliconization. Siliconization not only extended stable operating region significantly by itself, but also provided a good target plasma for other experiments of raising density limit. Core fuelling schemes are favourable especially for siliconized wall with a higher level of medium-Z impurity (Z = 14). After siliconization the maximum density near to 1020 m-3 was achieved by a combination of supersonic molecule beam injection and multipellet injection. The new defined slope of Hugill limit illustrating more clearly the situation under low qa and high ne discharges was created to indicate the new region extended by combining Ip ramp-up with core fuelling. The slope with a large Murakami coefficient increased by a factor of 50-60%.

Comparison of the Burnup Characteristics and Radiotoxicity Hazards of Rock-like Oxide Fuel with Different Types of Additives

Burnup characteristics of rock-like oxide (PuO2-ZrO2: ROX) fueled LWR have been studied with an additive ThO2, UO2 or Er2O3. A ROX fueled LWR core has reactivity coefficients and burnup reactivity swing problems, which can be mitigated by these additives. From the burnup calculations of the ROX fueled LWR, it was found that the additive ThO2 causes a few percent decrease in Pu transmutation in ROX fuel, while the additive UO2 or Er2O3 decreases it largely. The production of 241 Am, which is a parent of long-lived 237Np, is smaller in ThO2 added fuel than that in UO2 or Er2O3 added fuel. However, the productions of 243Am and 244Cm are more in ThO2 added fuel than those in UO2 or Er2O3 added fuel. The long-lived fission products (LLFPs) production in ThO2 added fuel is nearly the same as that in UO2 added fuel, and smaller than that in Er2 O3 added fuel. As a result of these Pu transmutation and minor actinides (MAs) and LLFPs productions, ingestion radiotoxicity hazard index of ThO2 added ROX spent fuel is lower than that of UO2 or Er2O3 added ROX spent fuel, except at the 105 years cooling. From these calculations, ThO2 is recommended to be used as the additive in ROX fuel.

Evaluation on Vibrocompaction Fuel Core System. (Part 1). Effect of Amplitude, Size Ratio and Shape of Multi-Size Powders on Vibratory Compacting.

In order to increase homogeneity with a higher packing density of vibrocompaction fuel, we investigated the effect of size ratio and shape of multi-size powders on vibratory compacting, and the size segregation phenomenon which commonly occurs with vibratory compacting. Spherical glass beads and irregular glass particles were used in the present vibratory compacting. We propose a segregation model that expresses the relation between size segregation, size ratio and amplitude. Under the shaking condition of this study, our model well agreed with the experimental results. Before and after shaking, compacting behavior of the samples were observed by X-ray computed tomography (X-ray CT).

DT fusion power production in ELM free H modes in JET

Experiments in the ELM free hot ion H mode regime have been carried out in DT plasmas in JET. Initial experiments undertaken at constant neutral beam (NB) power (∼11 MW) demonstrated that core fuelling was dominated by wall/target recycling rather than NB fuelling and made it possible to arrange an optimum core DT mix by adjusting the DT mix in wall/target, gas and NB. High power experiments at 4.2 MA/3.6 T have successfully and reliably delivered fusion power (Pfus) up to 16.1 MW and plasma stored energy (Wdia) up to 17 MJ. The results are in good agreement with extrapolations, carried out with the TRANSP and JETTO codes, from similar deuterium discharges. Transiently, values of Qtot up to 0.95 ±0.17 were achieved, consistent with values of nDT(0)τE, diaTi (0) ≈ 8.7 × 1020 m-3·s·keV±20%. The ratio of fusion power to input power, Qin, is in excess of 0.6. There are indications of an isotope effect on the edge pressure pedestal, but no net dependence of global confinement on isotopic plasma composition has been found.

Optimization Study of Nuclear Design for a 150 MWe Fast Reactor with Uranium Metallic-Fuel

A 150 MWe uranium metallic fueled core design which uses only U—Zr binary fuel not in excess of 20wt% enrichment, is proposed in this study. The selected core design is based on extensive researches for the possible enrichment gains over various design options and in-core fuel management strategies. As a result, an economic and safe uranium metallic fuelled core design is established in the equilibrium cycle. Further optimization has been achieved by applying strategic loading schemes for initial and transition cycles to reach the target equilibrium cycle early. An iterative method has been developed to find initial core enrichment splits. With non-uniform feed enrichments at the initial core adopted, this iterative method shows the core can reach the uranium equilibrium cycle after just five reloads, keeping the feed enrichment unchanged from Cycle 2. Reactivity parameters are evaluated to assure the safety characteristics of this design at both BOC and EOC of the initial core and the equilibrium cycle core. The transition cycle analysis indicated that recycling of self-generated plutonium alone is not sufficient to make the designed reactor a pure plutonium equilibrium core even after 56 reloads.

Nuclear and thermal–hydraulic characteristics for an LMR core fueled with 20% enriched uranium metallic fuel

As a part of the core design development of KALIMER (150 MWe), the KALIMER core was initially designed with 20% enriched uranium metallic fuel. In this core design, the primary emphasis was given to realize the metallic fueled core design to meet the specific design requirements; 20% and below uranium enrichment and a minimum fuel cycle length of one year. The core was defined by a radially homogeneous core configuration incorporated with several passive design features to give inherent passive means of negative reactivity insertion. The core nuclear performance based on a once-through equilibrium fuel cycle scenario shows that the core has an average breeding ratio of 0.67 and maximum discharge burnup of 47.3 MWD/kg. When comparing with conventional plutonium metallic fueled cores of the same power level, the present uranium metallic fueled core has a lower power density due to its increased physical core size. The negative sodium void reactivity over the core shows a beneficial potential to assure inherent safety characteristics. The transition from the uranium startup to equilibrium cycle is feasible without any design change. Core nuclear performance characteristics in the present core design are attributed to the specific design requirements of enrichment restriction and fuel cycle length.

Particle control and behaviour of neutrals in the pumped W-shaped divertor of JT-60U

As a trial for an active particle control, simultaneous feedback control of the divertor neutral pressure and main plasma density by the dual locations of gas puff is performed in the W-shaped divertor of JT-60U. At a relatively low gas-puffing rate, the divertor neutral pressure and main density are well controlled as pre-programmed. Lowering the X-point height is effective for isolation of the main plasma from divertor neutral particles. Simultaneous control has some difficulty at a high gas-puffing rate due to an increase of the cross-term coupling with each gas puff. Particle flux from the neutral beam also makes it difficult to control divertor neutral pressure and density. The simultaneous controllability depends on the operational region of the divertor neutral pressure and the main density. The coupling of the divertor neutral pressure with gas puff becomes larger in the high flux region, so that core fuelling in order to minimise the coupling is necessary for an effective particle isolation.

Prospects for charge-exchange-recombination-based measurements on International Thermonuclear Experimental Reactor using a helium diagnostic neutral beam

Several important measurements in the International Thermonuclear Experimental Reactor (ITER) diagnostic mission, including the primary one of core helium ash density, are expected to be addressed using active spectroscopic techniques. These methods rely on the use of a dedicated diagnostic neutral beam (DNB) which has been optimized for the dual requirements of beam penetration and charge exchange cross section. For hydrogenic beams, this results in an optimal beam energy of ∼100 keV/amu. Signal-to-noise estimates using realistic geometries and the existing ITER profile and equilibrium data have confirmed the stringent requirements on beam quality and intensity to satisfy the stated ITER measurement precisions. In this article we consider the use of a neutral helium DNB for making active spectroscopic measurements on ITER, since helium beams offer better penetration in dense plasma for a given energy, and the prospects for given source performance may also be improved. Drawbacks include the more difficult absolute calibration of the beam density profile as well as the fundamental problem of uniquely identifying the source (fusion-based ash, beam core fuelling, or edge DNB neutralizer/source efflux) of the observed He charge-exchange recombination line in order to unambiguously characterize core helium buildup and confinement on ITER.

Optimization Study of Nuclear Design for a 150 MWe Fast Reactor with Uranium Metallic-Fuel.

A 150 MWe uranium metallic fueled core design which uses only U—Zr binary fuel not in excess of 20wt% enrichment, is proposed in this study. The selected core design is based on extensive researches for the possible enrichment gains over various design options and in-core fuel management strategies. As a result, an economic and safe uranium metallic fuelled core design is established in the equilibrium cycle. Further optimization has been achieved by applying strategic loading schemes for initial and transition cycles to reach the target equilibrium cycle early. An iterative method has been developed to find initial core enrichment splits. With non-uniform feed enrichments at the initial core adopted, this iterative method shows the core can reach the uranium equilibrium cycle after just five reloads, keeping the feed enrichment unchanged from Cycle 2. Reactivity parameters are evaluated to assure the safety characteristics of this design at both BOC and EOC of the initial core and the equilibrium cycle core. The transition cycle analysis indicated that recycling of self-generated plutonium alone is not sufficient to make the designed reactor a pure plutonium equilibrium core even after 56 reloads.

ＭＯＸ燃料炉物理試験と解析の現状 プルトニウムの有効利用に向けて

ＭＯＸ燃料炉物理試験と解析の現状 プルトニウムの有効利用に向けて

96/03844 Viscosity of coal-derived liquids. 1. A group contribution method for pure model compound liquids

96/03844 Viscosity of coal-derived liquids. 1. A group contribution method for pure model compound liquids

Core fuelling to produce peaked density profiles in large tokamaks

Peaking the density profile increases the usable bootstrap current and the average fusion power density; this could reduce the current drive power and increase the net output of power producing tokamaks. The use of neutral beams and pellet injection to produce peaked density profiles is assessed. It is shown that with radially 'hollow' diffusivity profiles (and no particle pinch) moderately peaked density profiles can be produced by particle source profiles that are peaked off-axis. The fuelling penetration requirements can therefore be relaxed and this greatly improves the feasibility of generating peaked density profiles in large tokamaks. In particular, neutral beam fuelling does not require megavolt particle energies. Even with beam voltages of approximately 200 keV, however, exceptionally good particle confinement is needed to achieve net electrical power generation. The required ratio of particle to thermal diffusivities is an order of magnitude outside the range reported for tokamaks. In a system with no power production requirement (e.g., neutron sources) neutral beam fuelling should be capable of producing peaked density profiles in devices as large as ITER. Fuelling systems with low energy cost per particle-such as cryogenic pellet injection-must be used in power producing tokamaks when tau p approximately tau E. Simulations with pellet injection speeds of 7 km/s show that the peaking factor, ne0/(ne), approaches 2

Performance of Metal and Oxide Fuel Cores during Accidents in Large Liquid-Metal-Cooled Reactors

This paper reports on a cooperative effort among European and U.S. analysts, which is an assessment of the comparative safety performance of metal and oxide fuels during accidents in a 3500-MW (thermal), pool-type, liquid-metal-cooled reactor (LMR) is performed. The study focuses on three accident initiators with failure to scram: the unprotected loss-of-flow (ULOF), the unprotected transient overpower, and the unprotected loss-of-heat-sink (ULOHS). Core designs with a similar power output that have been previously analyzed in Europe under ULOF accident conditions are also included in this comparison. Emphasis is placed on identification of design features that provide passive, self-limiting responses to postulated accident conditions and quantification of relative safety margins. The analyses show that in ULOF and ULOHS sequences, metal-fueled LMRs with pool-type primary systems provide larger temperature margins to coolant boiling than do oxide-fueled reactors of the same design.

Fuel Management and Core Design Code Systems for Pressurized Water Reactor Neutronic Calculations

A package of connected code systems for the neutronic calculations relevant in fuel management and core design of pressurized water reactors (PWRs) has been developed and applied for validation to ...