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1.
《Annals of Nuclear Energy》2002,29(3):271-286
To analyze the effect of an inhomogeneous mixture of an PuO2 powder on fission gas release in MOX fuel, a model has been developed using the assumption that gas release mechanism in Pu-rich particles is identical with that in UO2 fuel. A parametric study was performed to see the respective effect of the number density, size and fraction of Pu retained in the Pu-rich particles on gas release in MOX fuel. The model shows that, for the condition of all the other remaining parameters being fixed, more gas is released in a MOX fuel for lower number density of, smaller size of, and larger fraction of Pu retained in, the Pu-rich particles. However, there exists some condition or combination of parameters for which the effect of inhomogeneity on gas release is negligible depending on the characteristics of MOX fuel. Comparison with measured data for OCOM MOX fuel shows that the present model can predict the level of gas release in MOX fuel once the release mechanism in the Pu-rich particles is known. 相似文献
2.
Coated plutonia particle fuel has been proposed recently for use in radioisotope power systems and radioisotope heater units for a variety of space missions requiring power levels from milliwatts to tens or even hundreds of watts. The 238PuO2 fuel kernels are coated with a strong layer of ZrC designed to fully retain the helium gas generated by the radioactive decay of 238Pu. A recent investigation has concluded that helium retention in large-grain (200 μm) granular and polycrystalline fuel kernels is possible even at high-temperatures (>1700 K). Results of performance analysis showed that this fuel form could increase by 2.3–2.4 times the thermal power output of a light weight radioisotope heater unit. These figures are for a single-size (500 μm) particles compact, assuming 10% and 5% helium gas release respectively, and a fuel temperature of 1723 K, following 10 years of storage. A binary-size (300 and 1200 μm) particles compact increases the thermal power output of the RHU by an additional 15%. 相似文献
3.
Core Burnup Calculation of Plutonium Burning Inert-Matrix Fueled High Temperature Gas Cooled Reactor
《Journal of Nuclear Science and Technology》2013,50(6):853-861
For the efficient reduction of excess plutonium amount, Japan Atomic Energy Research Institute (JAERI, now Japan Atomic Energy Agency) has studied a concept of rock-like oxide (ROX) fuel as a kind of inert matrix fuel (IMF). In the JAERI study, ROX fuel is burnt in existing light water reactors (LWRs), while in this study, pebble bed type high temperature gas cooled reactor (HTGR) is studied, mainly because of its high neutron economy and softer neutron spectrum than LWRs. Here, PuO2-yttria stabilized zirconia (YSZ: (Zr,Y)O2-x) particles are dispersed in graphite matrix. In the ROX fueled LWR study, it was necessary to improve fuel temperature reactivity coefficients by adding such additives as 238U and Er. Here in HTGR, although the negative temperature coefficient is much larger than that in LWR without any improvements, temperature coefficient was improved as large as possible to the level of UO2 HTGR case by adding Er in the fuel. Burnup calculations on PuO2-YSZ fueled HTGR, by simulating the continuous refueling of fuel pebbles with the batch fuel loading, showed almost complete transmutation for 239Pu and more than 80% for the total plutonium. As the maximum power density of the fuel pebble obtained by the core burnup calculation was very large in comparison with the UO2 HTGR, the maximum temperature in YSZ fuel particle was also evaluated. Despite the low thermal conductivity of YSZ, the evaluated YSZ temperature was well below the melting point, thanks to the high thermal conductivity of graphite and small YSZ particle size. Here, the high power density in the Pu-YSZ fueled core might become a problem, but is possible to be reduced by adjusting the initial plutonium enrichment. 相似文献
4.
High temperature chemical reactions of ZrC powder or ZrC coating on alumina spheres with CeO2, UO2 and SrO were studied. CeO2 reacts with ZrC at 1473–2073 K to give Ce2O3 and Ce2Zr2O7. Virtually no reactions were observed in UO2-ZrC mixtures at 1673–1973 K. SrO reacts with ZrC above 1273 K to form SrZrO3, Sr vapor and, presumably, CO. The SrZrO3 formation on ZrC coated spheres contained in a bed of SrO powder was rapid at 1673 K. All particles suffered serious damages on the ZrC coatings by the reaction with SrO. Though the Sr concentration in the ZrC was below the detection limit, Sr was found to be distributed within the alumina kernel. 相似文献
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6.
Zafar Yasin 《Nuclear Engineering and Design》2010,240(10):2862-2865
A study of fuel burn-up and concentrations of uranium and plutonium isotopes for the three fuel cycles of a CANDU reactor are carried out in the present work. The infinite and effective multiplication factors are calculated as a function of fuel burn up for the natural UO2 fuel, 1.2% enriched UO2 fuel and for the 0.45% PuO2-UO2 fuel. The amount of 235U and 238U consumed and 239Pu, 240Pu and 241Pu produced in the three fuel cycles are also calculated and compared. 相似文献
7.
W.A. Stark 《Journal of Nuclear Materials》1978,73(2):169-179
To compare the relative effectiveness of ZrC vis-a-vis SiC as a fission product barrier in fuel structures for high temperature gas cooled reactor (HTGR) applications, a series of cesium infusion experiments on various ZrC powders, and ZrC coated graphite structures was performed to study the cesium solubility, diffusivity, and permeability of this coating material. The ZrC powder results yield a solubility of Cs in ZrC, S(ppm wt) = (1.7 × 10?6) exp[229 kJ/RT], over the temperature range 1485–1896 K. The diffusion coefficient of Cs in ZrC is 10?18–10?16 m2/s over a similar temperature interval. The activation energy of diffusion is estimated to be ≈ 50 kJ/mole.The results of experiments in which both SiC and ZrC coated graphite samples were exposed to cesium are more difficult to interpret. The results support the conclusions of the ZrC powder experiments that ZrC is comparable to SiC as a diffusion barrier to cesium. 相似文献
8.
Woong Ki Kim Author Vitae Young Woo Lee Moon Sung Cho 《Nuclear Engineering and Design》2011,241(9):3748-3752
The TRISO-coated fuel particle for the high temperature gas-cooled reactor (HTGR) is composed of a nuclear fuel kernel and outer coating layers. The coated particles are mixed with graphite matrix to make HTGR fuel element. Weight of fuel kernels contained in the element is one of the important items for evaluating the characteristics of fuel element, which is generally measured by chemical analysis or gamma-ray spectrometer. The chemical analysis is a destructive method, and gamma-ray spectrometer requires elaborate reference sample for the measurement. In this study, X-ray computed tomography (CT) is suggested to measure the weight of kernels in an element. The three-dimensional (3D) density information is acquired by the X-ray CT for a simulated compact including simulated TRISO-coated particles with ZrO2 kernels. The volume of kernels as well as the number of kernels in the simulated compact was calculated from the 3D density information. The weight of kernels in the simulated compact was calculated from the volume of kernels and the average density of kernels. It was also calculated from the number of kernels and the average weight of kernels for comparison. 相似文献
9.
A.M. Savchenko A.V. Vatulin E.M. Glagovsky I.I. Konovalov A.V. Morozov A.V. Kozlov S.A. Ershov V.A. Mishunin G.V. Kulakov V.I. Sorokin A.P. Simonov Z.N. Petrova V.V. Fedotov 《Journal of Nuclear Materials》2010,396(1):26-31
At A.A. Bochvar Institute a novel conception of IMF to burn civil and weapon’s grade Pu is currently accepted. It consists in the fact, that instead of using pelletized IMF, that features low serviceability and dust forming route of fuel element fabrication, the usage is made of dispersion type fuel element with aluminium or zirconium matrices.Dispersion fuels feature a high irradiation resistance and reliability; they can consequently reach high burnups and be serviceable under transient conditions.Three basic fuel element versions are under development in VNIINM for both thermal and fast reactors.The first version is a fuel element with a heterogeneous arrangement of fuel (PuO2 or YSZ granules) within an Al or Zr matrix. The second version of a fuel element has a heat conducting Al or Zr alloy matrix and an isolated arrangement of PuO2 in a fuel minielement more fully meets the ‘Rock Fuel’ requirements. According to the third version a porous meat of zirconium metallurgically bonded to a fuel cladding is formed through which a PuO2 powder is introduced. All the versions are technologically simple to fabricate and require minimal quantities of process operations related to treating MA and Pu. Preliminary in-pile tests of IMF prototypes are presented. 相似文献
10.
《Journal of Nuclear Science and Technology》2013,50(6):466-468
The crushing strength of thorium oxide and thorium-uranium mixed oxide fuel kernels for high-temperature gas-cooled reactors (HTGR) was estimated by application of the Hertz theory of contact to the failure load obtained in a simple crushing test. The strength data were interpreted assuming the Weibull distribution. The crushing strength of the ThO2 kernels ranged from 1.7 to 1.9 GPa. The strength was found to be closely related to the microstructure of the kernel and to be increased by the kernel properties: small grain size and smooth surface. Fracture would occur in a transgranular manner for the fine-grained kernels and in an intergranular manner for the coarse-grained. Flaws, which are located at or near surface of the boundary of the contact area, act as fracture origins. 相似文献
11.
Shuhei Miwa Isamu Sato Kosuke Tanaka Takashi Hirosawa Masahiko Osaka 《Journal of Nuclear Materials》2010,400(1):32-421
The melting behavior of MgO-based inert matrix fuels containing (Pu,Am)O2−x ((Pu,Am)O2−x-MgO fuels) was experimentally investigated. Heat-treatment tests were carried out at 2173 K, 2373 K and 2573 K each. The fuel melted at about 2573 K in the eutectic reaction of the Pu-Am-Mg-O system. The (Pu,Am)O2−x grains, MgO grains and pores grew with increasing temperature. In addition, Am-rich oxide phases were formed in the (Pu,Am)O2−x phase by heat-treatment at high temperatures. The melting behavior was compared with behaviors of PuO2−x-MgO and AmO2−x-MgO fuels. 相似文献
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13.
Franois Charollais Sophie Fonquernie Christophe Perrais Marc Perez Olivier Dugne Franois Cellier Grard Harbonnier Marie-Pierre Vitali 《Nuclear Engineering and Design》2006,236(5-6):534-2004
In the framework of the French V/HTR fuel development and qualification program, the Commissariat à l’Energie Atomique (CEA) and AREVA are conducting R&D projects covering the mastering of UO2 coated particle and fuel compact fabrication technology. To fulfill this task, a review of past knowledge, of existing technologies and a preliminary laboratory-scale work program have been conducted with the aim of retrieving the know-how on HTR coated particle and compact manufacture:
- • The different stages of UO2 kernel fabrication GSP process have been reviewed, reproduced and improved.
- • The experimental conditions for the chemical vapor deposition of coatings have been defined on dummy kernels and development of innovative characterization methods has been carried out.
- • Former CERCA compacting process has been reviewed and updated.
- • to recover and validate past knowledge,
- • to produce representative HTR TRISO fuel meeting industrial standards,
- • to permit the optimization of reference fabrication processes for kernels and coatings defined previously at a laboratory-scale and the investigation of alternative and innovative fuel design (UCO kernel, ZrC coating),
- • to test alternative compact process options and
- • to fabricate and characterize fuel required for irradiation and qualification purpose.
14.
A. F. Grachev A. A. Maershin O. V. Skiba V. A. Tsykanov A. V. Bychkov M. V. Kormilitsyn Yu. S. Sokolovskii 《Atomic Energy》2004,96(5):320-326
A pyroelectrochemical process for reprocessing spent fuel and fabricating granular oxides UO2, PuO2 or (U, Pu)O2 from chloride melts has been developed at the Scientific-Research Institute of Nuclear Reactors for a prospective nuclear fuel cycle. The basic equipment has been developed. The basic results of a comprehensive study of fuel elements with vibrationally compacted (U, Pu)O2 fuel for fast reactors are presented. The performance of the reactors remains high up to 30% burnup in standard BOR-60 reactor fuel assemblies and 32% burnup in experimental fuel elements. An assessment is made of the effectiveness of the pyroelectrochemical methods and vibrational compaction technology for plutonium utilization. 相似文献
15.
D. Gryaznov S. Rashkeev E. Heifets 《Nuclear instruments & methods in physics research. Section B, Beam interactions with materials and atoms》2010,268(19):3090-3094
UO2 and (U, Pu)O2 solid solutions (the so-called MOX) nowadays are used as commercial nuclear fuels in many countries. One of the safety issues during the storage of these fuels is related to their self-irradiation that produces and accumulates point defects and helium therein.We present density functional theory (DFT) calculations for UO2, PuO2 and MOX containing He atoms in octahedral interstitial positions. In particular, we calculated basic MOX properties and He incorporation energies as functions of Pu concentration within the spin-polarized, generalized gradient approximation (GGA) DFT calculations. We also included the on-site electron correlation corrections using the Hubbard model (in the framework of the so-called DFT + U approach). We found that PuO2 remains semiconducting with He in the octahedral position while UO2 requires a specific lattice distortion. Both materials reveal a positive energy for He incorporation, which, therefore, is an exothermic process. The He incorporation energy increases with the Pu concentration in the MOX fuel. 相似文献
16.
《Journal of Nuclear Science and Technology》2013,50(10):1259-1267
The effect of Pu-rich agglomerates in U-Pu mixed oxide (MOX) fuel on a lattice calculation has been demonstrated. The Pu-rich agglomerate parameters are defined based on the measurement data of MIMAS-MOX and the focus is on the highly enriched MOX fuel in accordance with increased burnup resulting in a higher volume fraction of the Pu-rich agglomerates. The lattice calculations with a heterogeneous fuel model and a homogeneous fuel model are performed simulating the PWR 17 × 17 fuel assembly. The heterogeneous model individually treats the Pu-rich agglomerate and U-Pu matrix, whereas the homogeneous model homogenizes the compositions within the fuel pellet. A continuous-energy Monte Carlo burnup code, MVP-BURN, is used for burnup calculations up to 70 GWd/t. A statistical geometry model is applied in modeling a large number of Pu-rich agglomerates assuming that they are distributed randomly within the MOX fuel pellet. The calculated nuclear characteristics include k-inf, Pu isotopic compositions, power density and burnup of the Pu-rich agglomerates, as well as the pellet-averaged Pu compositions as a function of burnup. It is shown that the effect of Pu-rich agglomerates on the lattice calculation is negligibly small. 相似文献
17.
A.M. Savchenko I.I. Konovalov A.V. Vatulin E.M. Glagovsky 《Journal of Nuclear Materials》2009,385(1):148-152
New type of metal base fuel element is suggested for fast reactors. Basic approach to fuel element development - separated operations of fabricating uranium meat fuel element and introducing into it Pu or MA dioxides powder, that results in minimizing dust forming operations in fuel element fabrication. According to new fuel element design a framework fuel element having a porous uranium alloy meat is filled with standard PuO2 powder of <50 μm fractions prepared by pyrochemical or other methods. In this way a high uranium content fuel meat metallurgically bonded to cladding forms a heat conducting framework, pores of which contain PuO2 powder. Framework fuel element having porous meat is fabricated by capillary impregnation method with the use of Zr eutectic matrix alloys, which provides metallurgical bond between fuel and cladding and protects it from interaction. As compared to MOX fuel the new one features high thermal conductivity, higher uranium content, hence, high conversion ratio does not interact with fuel cladding and is more environmentally clean. Its principle advantage is a simple production process that is easily realized remotely, feasibility of involving high background Pu and MA isotopes into closed nuclear fuel cycle at the minimal influence on environment. 相似文献
18.
Shohei Ueta Jun AiharaKazuhiro Sawa Atsushi YasudaMasaki Honda Noboru Furihata 《Progress in Nuclear Energy》2011,53(7):788-793
This paper describes experiences and present status of research and development works for the high temperature gas-cooled reactor (HTGR) fuel in Japan. Recently, Very High Temperature Reactor (VHTR) is evaluated highly worldwide, and is a principal candidate for the Generation IV reactor systems. In Japan, HTGR fuel fabrication technologies have been developed through the High Temperature Engineering Test Reactor (HTTR) project in Japan Atomic Energy Agency since 1960’s. In total about 2 tons of uranium of the HTTR fuel has been fabricated successfully and its excellent quality has been confirmed through the long-term high temperature operation. Based on the HTTR fuel technologies, SiC TRISO fuel has been newly developed for burnup extension targeted VHTR. For ZrC-TRISO coated fuel as an advanced fuel designs, R&Ds for fabrication and inspection have been carried out in JAEA. The irradiation with the Japanese uniform stoichiometric ZrC coating has been completed in the cooperation with Oak Ridge National Laboratory of the United States. 相似文献
19.
《Journal of Nuclear Science and Technology》2013,50(8):465-468
Several kinds of coated fuel particles, with their coating either intact or artificially cracked, were heated out-of-pile in such manner as to create a sharp temperature gradient across the particles (60°120°C per particle), at temperatures from 1,500° to 1,950°C. The purpose was to obtain information on the displacement of the kernel material relative to the coating. To examine this amoeba effect, the particles were observed, after heating, by both ceramography and ×-ray radiography. The results revealed that: (1) In the case of UO2 kernel with artificially impaired coating, their kernels were found to move more readily toward the crack, regardless of the temperature gradient, as compared with UC2. (2) The amoeba effect is observed even in out-of-pile heating on intact coated particles with UO2kernel which moves down the temperature gradient. This UO2 movement was given a new explanation based on the evaporation and subsequent condensation of the UO2 within the particle, when the coating is intact. (3) In case of UC2 kernel, which moves up the temperature gradient, the sealing-in of the kernel by the intact coatings appears to assume a controlling factor, and the occurrence of evaporation is negligible. 相似文献
20.
R.E. Bullock 《Nuclear Engineering and Design》1980,61(3):331-345
Designs have been developed for coated ThO2 fuel particles to be used in a hybrid fusion-fission system that could be operated without reprocessing. The fresh fertile fuel particle would first be cycled through the blanket of a fusion reactor to breed 233U, which would then be ‘burned’ in a thermal fission reactor. The depleted fuel would then be refreshed in a second pass through the fusion reactor, and the process above repeated as many times as feasible. Designs of coated particles for up to three cycles through the hybrid system of reactors have been developed. The outer structural layer for these particles is made from vapor-deposited silicon carbide, because of its remarkable dimensional stability under fast neutron irradiation, and an inner layer of porous pyrocarbon is used to accommodate the buildup of gaseous reaction products inside the particle. The production of gaseous emission products from the interaction of high-energy fusion neutrons with coating materials and with the oxygen in the kernel contributes significantly to pressure vessel stresses in these coatings, whereas gaseous fission products alone dominate in conventional thermal reactors. The most stringent design for the three-cycle particle is identical in fuel loading to the reference fertile particle for an HTGR, which would constitute an ideal hybrid partner for the fusion reactor. Consideration is also given to coated-particle designs for the containment of the bred tritium used to fuel the D-T fusion reactor. 相似文献