04/02775 Minimum critical mass and flat flux in a 2-groupmodel: Lewins, J. Annals of Nuclear Energy, 2004, 31, (5), 541–576
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hydraulic code 1Lo realize the coupling of neutronics and thermal- hydraulic. To generate the lattice cross-sections of instantaneous state, a refined parameterized cross-section method was used in FMPHWR. A modified time-averaged model was developed. It has been implemented in FMPHWR and tested with Qinshan Phase III CANDU reactor. The numerical results show that FMPHWR has superior computational efficiency and accuracy.
04/02768 Development of plasma stored energy feedback control and its application to high performance discharges on JT-60U Oikawa, T. et al. Fusion Engineering and Design, 2004, 70, (2), 175-183. The real-time feedback control of the plasma stored energy has been developed for control of the plasma MHD stability in the JT-60U tokamak. The plasma stored energy can be detected with high accuracy in real-time by a function parameterization method for various plasmas available in JT-60U, such as Ohmic plasmas, the L-mode, the H-mode, the high poloidal beta mode and the reversed shear mode over a wide range of the plasma parameters. By manipulating the neutral beam injection power, the plasma-stored energy has been successfully controlled along the pre-programmed reference waveform. Especially in the reversed shear mode, this feedback control scheme has improved the reproducibility of the formation of the internal transport barrier, and MHD activities could be suppressed keeping the normalized beta in a stable region. A DT equivalent fusion amplification gain of 0.5 was sustained for 0.8 s in a reversed shear plasma by employing this feedback control scheme.
04/02769 DIII-D tokamak control and neutral beam computer system upgrades Penaflor, B. G. et al. Fusion Engineering and Design, 2004, 71, (1-4), 5-9. This paper covers recent computer system upgrades made to the DII I - D tokamak control and neutral beam computer systems. The systems responsible for monitoring and controlling the DII I -D tokamak and injecting neutral beam power have recently come online with new computing hardware and software. The new hardware and software have provided a number of significant improvements over the previous Modcomp AEG VME and accessware based systems. These improve- ments include the incorporation of faster, less expensive, and more readily available computing hardware which have provided perform- ance increases of up to a factor 20 over the prior systems. A more modern graphical user interface with advanced plotting capabilities has improved feedback to users on the operating status of the tokamak and neutral beam systems. The elimination of aging and non supportable hardware and software has increased overall maintainability. The distinguishing characteristics of the new system include: (1) a PC based computer platform running the Redhat version of the Linux operating system; (2) a custom PCI CAMAC software driver developed by general atomics for the kinetic systems 2115 serial highway card; and (3) a custom developed supervisory control and data acquisition (SCADA) software package based on Kylix, an inexpensive interactive development environment (IDE) tool from borland corporation. This paper provides specific details of the upgraded computer systems.
0402770 Dynamic strain aging and grain size reduction effects on the 'fatigue resistance of SA533B3 steels Huang, J. Y. et al. Journal of Nuclear Materials, 2004, 324, (2-3), 140- 151. Constant amplitude axial fatigue tests were conducted on SA533B3 steels with four levels of sulfur content at room temperature and 300C. Fatigue life was significantly affected by the inclusions. Fractographic examination results suggested that inclusions near the specimen surface served as the crack initiation site for a majority of the fatigued specimens tested at room temperature and that for those tested at 300C, some cracks were identified to initiate around the inclusions even in the interior of the fatigued specimens. Under the cyclic deformation, the dynamic strain aging (DSA) prevailed in the first few cycles at 300C, then the interactions among dislocations became dominant. Carbide or nitride precipitation in SA533B3 steels was enhanced synergistically by thermal energy and mechanical stress. It was shown that the combined effects of DSA and grain size reduction are responsible fnr the better fatigue resistance of SA533B3 steel at 300C.
04102771 Integral data for most important materials, namely lithium, beryllium, lead and uranium in infinite medium per fusion neutron Yapici, H. and (gzceyhan, V. Energy Conversion and Management, 2004, 45, (9-10), 1443-1456. The physical behaviour of integral data in an infinite medium has been evaluated for incident fusion neutrons with the help of the 3-D Monte Carlo Code. In a fusion reactor blanket with finite dimension, the integral quantities will be more or less different from the infinite medium results, depending on the neutron leakage fraction. Design
05 Nuclear fuels (scientific, technical)
studies foresee reduction of the neutron leakage out of the blanket to very low levels in order to prevent nuclear heating in super-conducting fusion magnets and to keep all neutrons primarily in the coolant. The most important materials in fusion technology, namely tritium, beryllium, lead, and uranium have been investigated in an infinitive medium. The main purpose of this work is to calculate integral neutronic data for incident 14.1-MeV (D,T) fusion source neutrons in an infinite medium for the most important materials in fusion technology, namely lithium, beryllium, lead and uranium. The study concludes the calculation of the integral tritium breeding ratio, 239pu breeding rate, neutron multiplication ratio through (n, x) and fission reactions and that of the heat release in those mixtures that are composed, first when natural uranium and 238U, are mixed with natural lithium or 6Li for a volume fraction from 0% to i00%, and then, the variable uranium-lithium composition is mixed with Be or Pb for a volume fraction from 0% to 100%.
04102772 LWR spent fuel transmutation in a high power density fusion reactor Sahin, S. and lSlbeyli, M. Annals of Nuclear Energy, 2004, 31, (8), 871- 890. The prospect of light water reactor (LWR) spent fuel incineration in a high power density fusion reactor has been investigated. The neutron wall load is taken at 10 MW/m z and a refractory alloy (W-5Re) is used in the first wall. Neutron transport calculations are conducted over an operation period of 48 months on a simple experimental hybrid blanket in a cylindrical geometry with the help of the SCALE4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a Ss-P3 approximation. In the neutron rich environment, the tritium breeding ratio remains >1.05 so that the tritium self-sufficiency is maintained for the fusion reactor. The presence of fissionable nuclear waste fuel in the investigated blanket causes significant energy amplification. The energy multiplication factor is approximately 4 at startup and it increases steadily up to 5.55 during power plant operation so that even a modest fusion reactor can supply a significant quantity of electricity. In the course of nuclear waste incineration, most of the fissionable fuel is burnt in situ. In addition to that, excess fissile fuel production enhances the nuclear quality of the nuclear fuel. Starting with an initial cumulative fissile fuel enrichment (CFFE) value of the spent fuel of 2.172%, CFFE can reach 4% after an irradiation period of approximately 12 months. Then the spent fuel becomes suitable for a new recharge in an LWR as a regenerated fuel. Further residence in the fusion blanket continues to upgrade the nuclear waste so that after 48 months, CFFE can reach such a high level (9%) that it becomes qualified to be used in a new type of the advanced high temperature reactors for the Generation-IV.
04/02773 Master equation and Fokker-Planck methods for void nucleation and growth in irradiation swelling Surh, M. P. et al. Journal of Nuclear Materials, 2004, 325, (1), 44-52. A complete theory of void swelling in irradiated metals requires the treatment of defect cluster nucleation events, as well as subsequent growth of stable clusters. One difficulty is that small-voids evolve rapidly and reversibly, whereas the secular evolution of the overall system is extremely slow. Thus, rate theory models for the void size distribution entail a set of stiff, coupled equations. A combined Master equation and Fokker-Planck numerical approach is introduced to address this problem and permit large time-steps at late times. Calculations are stable in practice, easily converged, and computation- ally efficient to large doses over a wide range in temperatures. The results are encouraging compared to experiment and earlier, related calculations.
0402774 Microstructural development and radiation hardening of neutron irradiated Mo-Re alloys Nemoto, Y. et al. Journal of Nuclear Materials, 2004, 324, (1), 62-70. Stress-relieved specimens and recrystallized specimens of pure Mo and Mo-Re alloys with Re contents of 2, 4, 5, 10, 13 and 41 wt% were neutron irradiated up to 20 dpa at temperatures from 681 to 1072 K. On microstructural observation, sigma phase and chi phase precipitates were found in all irradiated Mo-Re alloys. Voids were observed in all irradiated specimens, and dislocation loops and dislocations were observed in the specimens that were irradiated at lower temperatures. On Vickers hardness testing, all of the irradiated specimens showed hardening. Especially Mo-41Re were drastically embrittled after irradiation at 874 K or below. From these results, the authors discuss about the relation between microstructure development and radiation hardening and embrittlement, and propose the optimum Re content and thermal treatment for Mo-Re alloys to be used under irradiation conditions.
0402775 Minimum critical mass and flat flux in a 2-group model Lewins, J. Annals of Nuclear Energy, 2004, 31, (5), 541-576.
Fuel and Energy Abstracts November 2004 393
05 Nuclear fuels (scientific, technical)
Solutions for a flat thermal flux in a two-group diffusion model are found for a range of assumptions, since this is a requisite for both a minimum and a maximum fuel loading in the model. It is shown that a maximum fuel loading then arises when such a region, at unity infinite multiplication factor, is complemented by an outer core to bring the finite reactor critical. Correspondingly, when a core with a flat flux is surrounded by a reflector, the core fuel may be so distributed as to flatten the flux using the 'flux-trap' phenomenon and provide a minimum loading. Critical conditions and necessary fuel density are derived in finite and infinitely thick reflectors. A curiosity is the possibility of observing 'negative' reflector savings. Fuel savings are estimated for a range of models for the minimum loading compared to a correspondingly critical uniform core loading. In some circumstances a saving of up to 70% is indicated with economic and safety implications. It is shown how the reduction of the minimum loading solution from a two- to a one-group model retains the flat flux in the core but fails to satisfy the thermal boundary condition unless a reflector, with the flux-trap, is used so that without it, the two-group minimum fuel loading solution cannot fully transform to a minimum loading in one group. The full solution for a two-group model flat flux core with finite thickness reflector is given. It is shown that as the reflector is reduced it becomes necessary to increase the fuel density at the centre to a point where this exceeds the capability of the chosen fuel, thus providing a secondary criticality equation. The increasing steepness (negative slope) of the flux distribution is required to make the flux trap phenomenon possible in the reducing reflector region. The range in which there may be two reflector thicknesses leading to the same size core, but different fuel distributions observed by Williams, is determined. Constrained solutions that limit either the size of core or the maximum fuel density are considered, generalizing the original work of Goertzel to a practicable core design.
04/02776 Modified APEX reactor as a fusion breeder Sahin, S. and Ubeyli, M. Energy Conversion and Management, 2004, 45, (9-10), 1497-1512. An advanced fusion reactor project, called APEX, with improved effectiveness has been developed using a protective flowing liquid wall for tritium breeding and energy transfer. In the modified APEX concept, the flowing molten salt wall is composed of Flibe as the main constituent with increased mole fractions of heavy metal salt (ThF4 or UF4) for both fissile and fusile breeding purposes and to increase the energy multiplication. Neutron transport calculations are conducted with the help of the SCALE4.3 SYSTEM by solving the Boltzmann transport equation with the code XSDRNPM. By preserving a self sufficient tritium breeding ratio (TBR > 1.05) for a mole fraction up to 6% of ThF4 or 12% of UF4, the modified APEX reactor can produce u~to approximately 2800 kg of 33U/year or approximately 4950 kg of
Pu/year, assuming the same baseline fusion power production of 4000 MWth, as in the original APEX concept. With 6% ThF4 or 12% UF4 in the coolant, the total energy output will increase to 5560 MW~h or 8440 MWth, respectively. For a plant operation period of 30 full power years, the atomic displacement and helium production rates remain well below the presumable limits. The additional benefits of fissionable metal salt in the flowing liquid in a fusion reactor can be summarized as breeding of high quality fissile fuel for external reactors and increase of total plant power output.
0402777 Novel approach for the bulk synthesis of nanocrystalline yttria doped thoria powders via polymeric precursor routes Ganesan, R. et al. Journal of Nuclear Materials, 2004, 325, (2-3), 134- 140. Three different polymeric precursor routes namely, (i) amorphous citrate process, (ii) Pechini process and (iii) polyethylene glycol assisted process were compared for synthesizing nanocrystalline powders of 7.5 mol% yttria doped thoria (YDT). In each of the processes, parameters such as metal-to-fuel ratio or composition of fuel were varied and the effects were analysed. TG/DTA studies were conducted to identify the ignition temperatures of the precursors. Also, a novel experimental procedure with controlled combustion was devised for the preparation of powders based on the thermal analysis data. All the processes result in phase pure and nanocrystalline powders. The average crystallite size of the powders ranged between 9 and 18 nm. The powder samples were analysed for their carbon content and studied for their sinterability. Densities as high as 99% could be achieved by sintering the compacts of powders obtained from (i) amorphous citrate process with CA/M ratios 2....