Blanket Region Research Articles

Evaluation of core nuclear analysis code system for LMR using measured physics parameters of BFS-73-1 critical assembly

As the first stage of the critical experiment plan for developing the Korea Advanced LIquid MEtal Reactor (KALIMER) core design, the basic neutronics characteristics of the BFS-73-1 critical assembly representing a uranium metal fueled benchmark core were investigated. The prediction capability of the core nuclear analysis code system, being developed as a standard analysis system in the KALIMER core design and analysis, has been evaluated against the BFS-73-1 experiment by comparing the calculated results to the measurements. The comparison shows that the effective multiplication factor was overpredicted by 0.3% in the C/E value. The 235U and 238U fission reaction rate distributions were underpredicted within a 2% discrepancy in the core region but appeared to be in poor agreement in the blanket region. The calculated values of spectral indices at the core center and the effective delayed neutron fraction, β eff, agreed with the measured ones within 2% and 1% discrepancies, respectively. The first order perturbation method predicted the sample reactivity worths for 235U, 238U, 239Pu and 10B within 7% but much larger discrepancies existed for the other materials. The calculated Doppler effect showed about an 8% overprediction.

Design and Operation of the APT Target/Blanket

The Accelerator Production of Tritium (APT) Target/Blanket (T/B) system is comprised of the T/B assembly and the attendant heat-removal systems. The T/B assembly produces tritium using a high-energy proton beam and a spallation neutron source. All systems reside within the T/B building, located at the end of a linear accelerator. During nominal operation, protons are accelerated to an energy of 1030MeV at a current of 100mA. The protons interact with tungsten and lead nuclei in the T/B assembly to produce neutrons through the process of nuclear spallation. Neutron capture in 3He gas produces tritium, which is removed on a semicontinuous basis. The T/B assembly is modular, to allow for replacement of spent components and minimization of waste. Inconel and Aluminum structural materials provide containment of spallation products, a coolant pressure boundary, and corrosion resistance. The operating environment is a combination of high-energy protons and neutrons. All structures operate at low temperature and are cooled with water. Structures directly in the proton beam, which endure the highest irradiation damage, are replaced annually. Materials in the blanket region are primarily in a low-neutron flux environment and potentially will last the life of the facility. Published by Elsevier Science Inc.

The Nonproliferative Light Water Thorium Reactor: A New Approach to Light Water Reactor Core Technology

The nonproliferative light water thorium technology, also known as RTF (Radkowsky thorium fuel), provides a new approach to light water reactor core design. An RTF core is completely nonproliferative for all practical purposes, provides major reductions in radwaste, reduces fuel cycle cost and consumption of natural uranium, does not require soluble boron control during operation, and is once-through (i.e., does not require reprocessing). The core is made up of multiple seed-blanket units with uranium-zirconium alloy in the seed regions and thorium oxide with ~10% uranium oxide in the blanket regions. A key advantage is that an RTF core has exactly the same control drives and support plates. An RTF core with plutonium substituted for uranium is also optimum for incinerating either weapons- or reactor-grade plutonium, burning at three times the rate obtainable with mixed oxide (MOX). Use of MOX also requires considerable core modifications and produces 60% new plutonium, while RTF core produces none.

Study on MA and FP transmutation in fast reactors

Feasibility studies have been performed to develop an optimized fast reactor core for reducing long-term radiotoxicity of nuclear waste by minor actinide(MA) and long-lived fission product(FP) transmutation, taking into consideration fuel cycle technology. Systematic parameter survey calculations were implemented to investigate the basic characteristics of MA and FP transmutation in a fast reactor core. The hybrid MA-loading method, where Np nuclide is dispersed uniformly in the core and target subassemblies containing Am, Cm and rare earth nuclides are loaded into the blanket region, has the potential to achieve the maximum transmutation of MA with no special fuel design considerations. The introduction of target subassemblies using duplex pellets - a moderator annulus surrounding a 99Tc core - has a great potential to transmute long-lived fission products in the radial blanket region of the fast reactor core.

Applicability evaluation to a MOX fueled fast breeder reactor for a self-consistent nuclear energy system

The potential for a MOX fueled fast breeder reactor (FBR) is evaluated with regard to its ability to transmute radioactive nuclides and its safety when incorporated in a 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 two layer radial blanket region without a significant impact on core 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 10 2 years is as small as that of a typical uranium ore. Regarding self-controllability in the system's safety, the proposed FBR core concept has an inherent negative reactivity feedback with a gas expansion module, sodium plenum above the core and burnup reactivity compensation module. So sodium boiling and fuel melting will be avoided in anticipated transient without scram events. Regarding self-terminability, even if the MOX fuel melting should cause a core compaction process, re-criticality of the core can be avoided by a fuel dilution and relocation module.

Disruption problematics in segmented-blanket concepts

In tokamaks, the hostile operating environment originated by plasma disruption events requires that the first-wall/blanket/shield components sustain the large induced electromagnetic (EM) forces without significant structural deformation and within allowable material stresses. As a consequence, there is a need to improve the safety features of the segmented-blanket design concepts in order to satisfy the disruption problematics. The present paper describes recent investigations on internal blanket reinforcement systems needed in order to improve the first-wall/blanket/shield structural design for next-step and commercial fusion reactors. Particularly in the context of SEAFP and ITER activities, representative 3-D CAD models of the inboard and outboard blanket regions and the related magnetomechanical simulations are illustrated.

Local Texture and Electromigration in Fine Line Microelectronic Aluminum Metallization

ABSTRACTAluminum films 1 μm thick were deposited on oxidized silicon by sputtering and partially ionized beam evaporation to vary the crystallographic texture. These films were patterned into lines and subsequently annealed at 400 °C for 1 h. A strong correlation between the electromigration behavior and the blanket film texture (X-ray diffraction (XRD) / pole figures) has been reported previously for these films. In this work, an Electron Backscatter Diffraction (EBSD) a.k.a. Backscatter Kikuchi Diffraction (BKD) technique was employed using a scanning electron microscope (SEM) to interrogate individual grain orientations. BKD pole figures were acquired for lines ≥0.3 μm wide and for blanket (pad) regions. Identical, inverse pole figures were found for blanket films measured using both XRD and BKD (pads). Furthermore, the BKD (111) fiber texture shows a line width dependency, with narrow lines having a slightly improved texture over blanket (pad) regions. Local grain orientations were investigated near and within electromigration testing sites with characteristic void and hillock morphologies. The relationship of neighboring grain orientations to electromigration damage is shown.

Structural design aspects of the locking system needed for the next tokamak plasma-facing components against electromagnetic transients

A technical key issue during the design of the near-term Tokamak fusion devices (NET and ITER) is related to the structural problems of the needed attachments for the removable internal first-wall components (blanket segments and divertor plates) inside the vacuum vessel in the presence of major plasma disruption events. Both the understanding of the impact of the huge transient electromagnetic forces transmitted onto the vessel from the fastening/guiding support systems and the knowledge of the remote handling operations with related connection/disconnection of the first-wall component supports are involved. Among the various systems proposed as attaching locks, the “Twin-Belt” concept for the inboard blanket region is exposed. It has been developed at the TESLA Laboratory of the JRC-Ispra and has been retained by the NET/ITER Team as one of the options to be studied in detail. The use of such an intermediate toroidal, structurally continuous supporting system (belt bars) provides an independent system of attaching locks between the first-wall region and the vessel.

A new method for generating group constants for analysis of blanket parameters of HCLWR with Monte Carlo method

The preliminary analysis of natural uranium blanket region at the KUCA was carried out with continuous energy monte carlo method and group energy monte carlo method. From the results obtained, it is clear that group constants weighted with scalar neutron spectrum cannot predict the transitory effect of the forward biased on neutron angular distribution. It is suggested that the direction-dependent group constants have to be used to predict the reaction rate in the blanket region accurately.

Advanced Methods Comparisons of Reaction Rates in the Purdue Fast Breeder Blanket Facility

A review of worldwide results revealed that reaction rates in the blanket region are generally underpredicted with the discrepancy increasing with penetration; however, these results vary widely. Experiments in the large uniform Purdue Fast Breeder Blanket Facility (FBBF) blanket yield an accurate quantification of this discrepancy. Using standard production code methods (diffusion theory with 50 group cross sections), a consistent Calculated/Experimental (C/E) drop-off was observed for various reaction rates. A 50% increase in the calculated results at the outer edge of the blanket is necessary for agreement with experiments. The usefulness of refined group constant generation utilizing specialized weighting spectra and transport theory methods in correcting this discrepancy was analyzed. Refined group constants reduce the discrepancy to half that observed using the standard method. The surprising result was that transport methods had no effect on the blanket deviations; thus, transport theory considerations do not constitute or even contribute to an explanation of the blanket discrepancies. The residual blanket C/E drop-off (about half the standard drop-off) using advanced methods must be caused by some approximations which are applied in all current methods. 27 refs., 3 figs., 1 tab.

Chamber Design for the LIBRA Light Ion Beam Fusion Reactor

Scoping analysis was performed for LIBRA to determine the blanket design options that satisfy the tritium breeding and wall protection requirements. The blanket is made of banks of INPORT tubes of porous SiC and Li/sub 17/Pb/sub 83/ flowing in them. The reference design utilizes a 1.35 m thick blanket region backed by a 0.5 m thick HT-9 reflector. The lithium in the blanket and reflector is enriched to 90% /sup 6/Li. This design leads to the shortest possible beam propagation channel. The TBR is 1.5 and the end-of-life dpa in the chamber wall is 200 dpa. In this paper the authors describe the design and configuration of the chamber as well as the tritium fueling and the response to the target explosion.

Lithium corrosion/mass transfer test facility

Liquid lithium is a leading candidate as a tritium-breeding material for fusion reactors because of its good breeding and heat-transfer characteristics. The self-cooled blanket concept is particularly attractive because of the inherent simplicity associated with the use of lithium as both breeder and coolant. In addition to liquid metal magnetohydrodynamic issues (LMMHD), two major compatibility concerns arising from the use of lithium are corrosion/mass transfer and degradation of mechanical properties of the containment material. This paper describes a new experimental facility for investigating the corrosion/mass transfer/deposition mechanisms at characteristic velocities and system temperatures. This test facility, which is constructed of ferritic/martensitic steel (Fe-9Cr-1Mo), simulates conditions projected for a liquid metal heat transport system. The test facility will provide test velocities to 0.5 m/s, and temperatures to 550 °C, and system ΔTs up to 150 °C. The corrosion test region simulates the blanket region with a low inlet temperature and a higher (ΔT ~ 150 °C) outlet temperature. Test specimens in the deposition region provide a measure of the deposition processes that will occur in a heat exchanger as the liquid metal is cooled (ΔT ~ 150 °C). In addition to temperature and flow control, the facility provides for control and monitoring of impurities in the liquid.

An Improved Cell Calculation Method for Liquid-Metal Fast Breeder Reactor Blanket Analysis

Cell homogenization methods for the neutronics analysis of the blanket region of a plate-type liquid-metal fast breeder reactor critical assembly are investigated, and an improved method is proposed, which calculates cell-averaged cross sections so as to preserve groupwise reaction rates in each cell in a multidrawer cell model. The present homogenization method has been applied to the analysis of the heterogeneous fast critical assembly ZPPR-13. It was found that the keff difference between homogeneous and heterogeneous fast critical assemblies was reduced from 0.4 to 0.2% Δk/k by using the present method.

Measurement of relative power distribution in axially heterogeneous core by gamma-counting of each fuel plate.

In order to understand physics properties of axially heterogeneous LMFBR cores, relative power distribution was measured using a γ-counting method in zone-type partial mockup cores. By introducing an inner blanket at the core midplane, flat power distribution is achieved compared with that of conventional homogeneous cores. Axial power peaking factor is obtained from the measured power distribution in the cores having the inner blanket of different geometries. Investigation is also made on distortion of the power distribution caused by simulated B4C control rods having different 10B enrichments (20 and 55%), which are partially inserted into an upper half of the core. Discussions are made on degree of power distortion and its propagation distance. It is observed that the inner blanket mitigates this disturbance caused in the upper half of the core to a lower half of the core. The calculated distribution agrees quite well with the measured power distribution within experimental errors both in the inner blanket and core regions, while significant discrepancy is observed in an outer blanket region. The calculation predicts well the degree of the distortion of the power distribution caused by the simulated B4C control rods.

The FUBR-1B experiment - irradiation of lithium ceramics to high burnups under large temperature gradients

Solid breeder materials used for supplying the tritium for fueling fusion power reactors will be required to withstand a variety of severe environmental conditions such as irradiation damage, thermal stresses and chemical reactions while continuing to produce tritium and not interfering with other essential components in the complex blanket region. In the FUBR-1B experiment, several solid breeder candidates are being subjected to the hostile conditions (i.e., high burnup and large temperature gradients) foreseen in a fusion reactor's blanket. Materials with high lithium atom densities, such as Li 2 O, Li 8 ZrO 6 and Li 4 SiO 4 , and with irradiation stability such as LiAlO 2 and Li 2 ZrO 3 , are being irradiated to high burnup levels and under large temperature gradients.

The thermal conductivity of mixed beryllia/li ceramic in sphere-pac forms

Lithium-containing ceramic tritium-breeder materials have been envisaged to be deployed within the blanket region of a fusion reactor in several possible configurations. One of these is the sphere-pac configuration. For this configuration an important material parameter is its thermal conductivity ( K sp ). It is well known that K sp demonstrates rather complex behavior as a function of temperature, gas pressure, gas composition, particle size, and packing fraction. The interrelationship of these parameters has been satisfactorily accounted for with a hierarchical effective media theory (HEMT). For tritium self sufficiency, most lithium ceramic breeder materials would require the presence of a neutron-multiplier (e.g., Be or BeO). Here, the influence of configuration on K sp (i.e., how one put the different solid components together in the sphere-pac bed) becomes important. Using a generalized HEMT (i.e., a model with capability to describe systems with more than one solid material component), we have analyzed in detail the configurational dependence of K sp for sphere-pac beds composed of lithium ceramic/BeO microspheres. Substantial improvements in K sp can be achieved if a configuration of lithium ceramic spheres coated with BeO is chosen. Increases in K sp would lead to enhanced mechanical and thermal performance of the breeder materials.

Neutronic Parameters of a Cylindrical Hybrid Blanket Driven by a Simulated Line Source

A multidimensional neutronic analysis is carried out to determine the extent to which one dimensional neutron transport calculations can be applied to a fusion-fission (hybrid) experimental blanket in cylindrical geometry, driven by moveable (D,T) target to simulate a 14 MeV neutron line source. Length of the target trajectory has been chosen to be L = 20 and 100 cm by a blanket height of H = 120 cm. The study shows that, for the proposed blanket, one dimensional calculations will be satisfactory to interpret the experimental neutronic data over a blanket region of Z =1+-15 to 25 cm for a trajectory length of L = 120 cm. Whereas these calculations would be applicable over a very narrow strip of the blanket around the Z = 0 plane for a trajectory length of L = 20 cm.

Neutron streaming analysis for helical geometry system, heliotron-H fusion power reactor.

Nuclear analysis was carried out for the Heliotron-H reactor. Due to the helical geometry of the system, 14 MeV neutron streaming inherent in the reactor exists; some of source neutrons generated in the plasma region can see the coil shield surface in a direct line-of-sight manner and stream through blanket regions without collisions. With an appropriate coordinate transformation from the helical geometry, analysis on the neutron streaming was made possible by a method coupling the three-dimensional Monte Carlo code and two-dimensional discrete-ordinates transport code. For each radiation response parameter in the helical magnet, the streaming neutrons contribute 40–50%. This fact strongly suggests the need of taking account of the contribution from the streaming neutrons to provide an adequate shielding for the helical magnet.

Fusion reactor design with ceramics

Ceramics are required to serve in a conventional role as electrical and thermal insulators and dielectrics in fusion power reactors. In addition, certain ceramic materials can play a unique role in fusion power reactors by virtue of their very low induced radioactivity from transmutation products produced by fusion neutron capture. The aspects of safety, long-term radioactive waste management, and personnel access for maintenance and repair can all be significantly improved by applying the low-activation ceramics to the first wall and blanket regions of a fusion reactor. This application imposes tensile, compressive, and shear structural loads and thermal stresses on the materials, and it is primarily in support of tensile stresses where problems in ceramic design lie. Silicon carbide, carbon, and graphite materials are three primary candidate structural ceramics. Electrical insulators and radio frequency electromagnetic wave windows commonly employ ceramics, such as Al 2O 3, MgO, SiO 2, Si 3N 4 and glasses. Material properties characteristic of the radiation damaged state must be used. The structural failure modes of low-ductility ceramics are by immediate fracture when a critical stress is applied or by slow crack growth eventually propagating to fracture. Both failure modes follow a statistical distribution with a finite probability of failure with low applied loads. Proof testing, however, can reduce this probability of failure to zero below some threshold stress by eliminating the weaker components, thus easing design problems. Design studies have been performed to develop conceptual designs of fusion power reactor components using low-activation ceramic and metallic materials. These components include limiters, first walls, blanket modules, shields, superconducting magnets, diagnostic instrumentation, electrical insulation, and radio frequency windows. Present day ceramics can fulfill all the functional requirements of these components without undue performance penalties. Improved ceramic materials, both monolithic and fiber composites, are being developed at a rapid pace, and these can readily be applied to improved fusion designs. It thus appears possible to design a fusion reactor using only low-activation ceramic materials, principally structural ceramics, in the high-neutron flux zones of the reactor. Presently operating fusion plasma devices employ graphite for limiters and armor, and ceramics for electrical insulators, providing a base for continued utilization when power-producing devices are built. The ultimate potential of fusion as an environmentally benign energy source with a high degree of safety and public acceptance is optimally achieved through the use of low-activation structural ceramics.

Fusion reactor design with ceramics

Ceramics are required to serve in a conventional role as electrical and thermal insulators and dielectrics in fusion power reactors. In addition, certain ceramic materials can play a unique role in fusion power reactors by virtue of their very low induced radioactivity from transmutation products produced by fusion neutron capture. The aspects of safety, long-term radioactive waste management, and personnel access for maintenance and repair can all be significantly improved by applying the low-activation ceramics to the first wall and blanket regions of a fusion reactor. This application imposes tensile, compressive, and shear structural loads and thermal stresses on the materials, and it is primarily in support of tensile stresses where problems in ceramic design lie. Silicon carbide, carbon, and graphite materials are three primary candidate structural ceramics. Electrical insulators and radio frequency electromagnetic wave windows commonly employ ceramics, such as Al 2 O 3 , MgO, SiO 2 , Si 3 N 4 and glasses. Material properties characteristic of the radiation damaged state must be used. The structural failure modes of low-ductility ceramics are by immediate fracture when a critical stress is applied or by slow crack growth eventually propagating to fracture. Both failure modes follow a statistical distribution with a finite probability of failure with low applied loads. Proof testing, however, can reduce this probability of failure to zero below some threshold stress by eliminating the weaker components, thus easing design problems. Design studies have been performed to develop conceptual designs of fusion power reactor components using low-activation ceramic and metallic materials. These components include limiters, first walls, blanket modules, shields, superconducting magnets, diagnostic instrumentation, electrical insulation, and radio frequency windows. Present day ceramics can fulfill all the functional requirements of these components without undue performance penalties. Improved ceramic materials, both monolithic and fiber composites, are being developed at a rapid pace, and these can readily be applied to improved fusion designs. It thus appears possible to design a fusion reactor using only low-activation ceramic materials, principally structural ceramics, in the high-neutron flux zones of the reactor. Presently operating fusion plasma devices employ graphite for limiters and armor, and ceramics for electrical insulators, providing a base for continued utilization when power-producing devices are built. The ultimate potential of fusion as an environmentally benign energy source with a high degree of safety and public acceptance is optimally achieved through the use of low-activation structural ceramics.