Zirconia Inert Matrix for Plutonium Utilisation and Minor Actinides Disposition in Thermal Reactors

2006 ◽  
Vol 45 ◽  
pp. 1907-1914
Author(s):  
Claude Degueldre
Keyword(s):  
Solid Solution ◽  
Ion Irradiation ◽  
Near Field ◽  
Fission Products ◽  
Heavy Ion ◽  
Minor Actinide ◽  
Minor Actinides ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Inert Matrix

The toxicity of the UO2 spent fuel is dominated by plutonium and minor actinides (MA): Np, Am and Cm, after decay of the short live fission products. Zirconia ceramics containing Pu and MA in the form of an Inert Matrix Fuel (IMF) could be used to burn these actinides in Light Water Reactors. Optimisation of the fuel designs dictated by properties such as thermal, mechanical, chemical and physical must be performed with attention for their behaviour under irradiation. Zirconia must be stabilised by yttria to form a solid solution such as AnzYyPuxZr1-yO2-y where minor actinide oxides are also soluble. Burnable poison may be added if necessary such as Gd, Ho, Er, Eu or Np, Am them-self. These cubic solid solutions are stable under heavy ion irradiation. The retention of fission products in zirconia, under similar thermodynamic conditions, is a priori stronger, compared to UO2, the lattice parameter being larger for UO2 than for (Y,Zr)O2-x. (Er,Y,Pu,Zr)O2-x in which Pu contains 5% Am was successfully irradiated in the Proteus reactor at PSI, in the HFR facility, Petten as well as in the Halden Reactor. These irradiations make the Swiss scientists confident to irradiate such IMF in a commercial reactor that would allow later a commercial deployment of such a fuel for Pu and MA utilisation in a last cycle. The fuel forms namely pellet of solid solution, cercer or cermet fuel are discussed considering the once through strategy. For this strategy, low solubility of the inert matrix is required for geological disposal. As spent fuels these IMF’s are demanding materials from the solubility point of view, this parameter was studied in detail for a range of solutions corresponding to groundwater under near field conditions. Under these conditions the IMF solubility is 106 times smaller than glass, which makes the zirconia material very attractive for deep geological disposal. The desired objective would be to use IMF to produce energy in reactors, opting for an economical and ecological solution.

Zirconia Inert Matrix Fuel for Plutonium and Minor Actinides Management in Reactors and as an Ultimate Waste Form

2008 ◽  
Vol 1104 ◽  
Author(s):  
Claude Degueldre ◽  
Wolfgang Wiesenack
Keyword(s):  
Near Field ◽  
Fission Products ◽  
Minor Actinides ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Inert Matrix Fuel ◽  
Inert Matrix ◽  
Fuel Pellets ◽  
Reducing Conditions ◽  
Deep Geological Disposal

AbstractA plutonia stabilised zirconia doped with yttria and erbia has been selected as inert matrix fuel (IMF) at PSI. The results of experimental irradiation tests on yttria-stabilised zirconia doped with plutonia and erbia pellets in the Halden research reactor as well as a study of zirconia solubility are presented. Zirconia must be stabilised by yttria to form a solid solution such as MAz(Y,Er)yPuxZr1-yO2-ζ where minor actinides (MA) oxides are also soluble. (Er,Y,Pu,Zr)O2-ζ (with Pu containing 5% Am) was successfully prepared at PSI and irradiated in the Halden reactor. Emphasis is given on the zirconia-IMF properties under in-pile irradiation, on the fuel material centre temperatures and on the fission gas release. The retention of fission products in zirconia may be stronger at similar temperature, compared to UO2. The outstanding behaviour of plutonia-zirconia inert matrix fuel is compared to the classical (U,Pu)O2 fuels. The properties of the spent fuel pellets are presented focusing on the once through strategy. For this strategy, low solubility of the inert matrix is required for geological disposal. This parameter was studied in detail for a range of solutions corresponding to groundwater under near field conditions. Under these conditions the IMF solubility is about 109 times smaller than glass, several orders of magnitude lower than UO2 in oxidising conditions (Yucca Mountain) and comparable in reducing conditions, which makes the zirconia material very attractive for deep geological disposal. The behaviour of plutonia-zirconia inert matrix fuel is discussed within a burn and bury strategy.

scholarly journals In situ microstructural evolution in face-centered and body-centered cubic complex concentrated solid-solution alloys under heavy ion irradiation

2020 ◽  
Vol 198 ◽  
pp. 85-99 ◽  
Author(s):  
Calvin Parkin ◽  
Michael Moorehead ◽  
Mohamed Elbakhshwan ◽  
Jing Hu ◽  
Wei-Ying Chen ◽  
...  
Keyword(s):  
Solid Solution ◽  
Microstructural Evolution ◽  
Ion Irradiation ◽  
Heavy Ion ◽  
Body Centered Cubic ◽  
Heavy Ion Irradiation ◽  
Face Centered

The Effect of Minor Actinide Transmutation on Temperature Coefficient in PWR

2017 ◽  
Author(s):  
Wenxin Zhang ◽  
Haoyang Yu ◽  
Bin Liu ◽  
Jin Cai ◽  
Shuangshuang Cui
Keyword(s):  
Temperature Coefficient ◽  
Half Life ◽  
Nuclear Reactor ◽  
Minor Actinide ◽  
Minor Actinides ◽  
Spent Fuel ◽  
Radioactive Nuclides ◽  
Safety Parameter ◽  
Neutron Transmutation ◽  
Actinide Transmutation

Minor actinides in the spent fuel have strong radiotoxicity and very long half-life, the the properly dispose of spent fuel is indispensible to the development of nucler energy. Generally,we dispose the spent fuel by geological burying. But it can not compeletly solve the problem. Neutron transmutation is the only way to shorten the half-life of radioactive nuclides, under the irradiation of neutron MA nuclide will capture neutron or fission, and translate into the short lived nuclide or something valued nuclide. Reactivity temperature coefficient is an improtant safety parameter in nuclear reactor physics.In the reactor design, for the safely operation of reactor, reactivity temperature coefficient must be be negative. The introduction of MA in the PWR must have interference to the temperature coefficient. This paper mainly studied the influence of PWR transmutation minor actinide on the temperature coefficient.

The Effect of MA Transmutation in the PWR on Fuel Cycle

2017 ◽  
Author(s):  
Haoyang Yu ◽  
Bin Liu ◽  
Wenxin Zhang ◽  
Jin Cai
Keyword(s):  
Fuel Cycle ◽  
Minor Actinide ◽  
Minor Actinides ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water ◽  
Mcnp Program ◽  
And Control ◽  
Control Rods ◽  
Effective Multiplication

The minor actinides (MA) is important nuclides in the spent fuel which is bad for human ecological environment. Pressurized water reactor (PWR) is the main reactor type at commercial operation around world. It is important to find the appropriate loading patterns when introducing minor actinides to the PWR core. In this paper, we study the effect of MA transmutation in the PWR on fuel cycle. First, we use the MCNP program to simulate the model of PWR and the effective multiplication factor.Then,the MA is introduced into core in different ways and mass to simulate the effective multiplication factor. In conclusion,without considering chemical skim control and control rods, we change the thickness of the MA, until the keff closes to 1, We find that loading minor actinides to burnable poison rods for transmutation is an optimal minor actinide loading pattern.

In Situ Microstructural Evolution in Fcc and Bcc Complex Concentrated Solid-Solution Alloys Under Heavy Ion Irradiation

2020 ◽  
Author(s):  
Calvin Parkin ◽  
Michael Moorehead ◽  
Mohamed Elbakhshwan ◽  
Adrien Couet ◽  
Kumar Sridharan ◽  
...  
Keyword(s):  
Solid Solution ◽  
Microstructural Evolution ◽  
Ion Irradiation ◽  
Heavy Ion ◽  
Heavy Ion Irradiation

Influence of Container Base Material (Fe) on SIMFUEL Leaching Behavior

2000 ◽  
Vol 663 ◽  
Author(s):  
J. Quiñones ◽  
J.A. Serrano ◽  
P.P. Díaz ◽  
J.L. Rodríguez Almazán ◽  
J. Cobos ◽  
...  
Keyword(s):  
Experimental Data ◽  
Redox Potential ◽  
Uranium Concentration ◽  
Near Field ◽  
Base Material ◽  
Fission Products ◽  
Sharp Decrease ◽  
Spent Fuel ◽  
Leaching Behavior ◽  
Redox Sensitivity

ABSTRACTThe chemical stability of spent fuel will be greatly influenced by the redox potential of the near field. Presence of reductants such as iron is likely to be an important factor to maintain the original integrity of spent fuel. In this work experimental data about the influence of metallic iron (container base material) on SIMFUEL leaching behavior under simulated granite and saline repository conditions is presented. In the presence of iron uranium concentration undergoes a sharp decrease. This is much more noticeable in the experiments performed under initial oxic conditions. The effect of iron on simulated fission products of SIMFUEL is very important for the elements with high redox sensitivity such us molybdenum. On the contrary, strontium remains stable during the entire tests and it seems not be affected by changes in redox potential.

Metastable phase formation by ion mixing of Nb–Al multilayers

1989 ◽  
Vol 4 (6) ◽  
pp. 1385-1392 ◽  
Author(s):  
K. Pampus ◽  
K. Dyrbye ◽  
B. Torp ◽  
R. Bormann
Keyword(s):  
Solid Solution ◽  
Amorphous Phase ◽  
Phase Formation ◽  
Ion Irradiation ◽  
Metastable Phase ◽  
Heavy Ion ◽  
Free Energies ◽  
Transmission Electron ◽  
Radiation Enhanced Diffusion ◽  
Metastable Phase Formation

The structure of Nb–Al thin films after ion mixing was studied for compositions from 20 to 85 at. % Al as a function of temperature in the range between 40 and 620 K. The phase formation was determined by transmission electron microscopy. At lower temperatures, only supersaturated bcc-solid solution, NbAl, and amorphous phase were found throughout the studied composition range. Besides these phases irradiation at temperatures above 470 K causes the formation of a metastable crystalline compound at an overall composition close to Nb25Al75, and for T = 623 K the equilibrium compound NbAl3 is formed. The other intermetallic phases Nb2Al and Nb3Al have not been observed at any irradiation temperature. Calculations of the Gibbs free energies of the various phases are presented, and the reliability of extrapolations to regions of metastability with respect to temperature and composition is commented on. The phase formation during heavy-ion irradiation is discussed in the context of the calculated free energies and kinetic constraints. For temperatures above 300 K, the attainment of a metastable phase equilibrium between the bcc solid solution and the amorphous phase is proposed due to the influence of radiation enhanced diffusion.

scholarly journals Different Radiation Tolerances of Ultrafine-Grained Zirconia–Magnesia Composite Ceramics with Different Grain Sizes

Materials ◽  
2019 ◽  
Vol 12 (17) ◽  
pp. 2649
Author(s):  
Wenjing Qin ◽  
Mengqing Hong ◽  
Yongqiang Wang ◽  
Jun Tang ◽  
Guangxu Cai ◽  
...  
Keyword(s):  
Grain Size ◽  
Ion Irradiation ◽  
Composite Ceramic ◽  
Heavy Ion ◽  
Composite Ceramics ◽  
Displacement Damage ◽  
Inert Matrix ◽  
Grain Sizes ◽  
Ultrafine Grained ◽  
Nuclear Systems

Developing high-radiation-tolerant inert matrix fuel (IMF) with a long lifetime is important for advanced fission nuclear systems. In this work, we combined zirconia (ZrO2) with magnesia (MgO) to form ultrafine-grained ZrO2–MgO composite ceramics. On the one hand, the formation of phase interfaces can stabilize the structure of ZrO2 as well as inhibiting excessive coarsening of grains. On the other hand, the grain refinement of the composite ceramics can increase the defect sinks. Two kinds of composite ceramics with different grain sizes were prepared by spark plasma sintering (SPS), and their radiation damage behaviors were evaluated by helium (He) and xenon (Xe) ion irradiation. It was found that these dual-phase composite ceramics had better radiation tolerance than the pure yttria-stabilized ZrO2 (YSZ) and MgO. Regarding He+ ion irradiation with low displacement damage, the ZrO2–MgO composite ceramic with smaller grain size had a better ability to manage He bubbles than the composite ceramic with larger grain size. However, the ZrO2–MgO composite ceramic with a larger grain size could withstand higher displacement damage in the phase transformation under heavy ion irradiation. Therefore, the balance in managing He bubbles and phase stability should be considered in choosing suitable grain sizes.

Spent Fuel Leaching under Anoxic Conditions and the Effect of Canister Materials

2003 ◽  
Vol 807 ◽  
Author(s):  
Daqing Cui ◽  
Jeanett Low ◽  
Max Lundström ◽  
Kastriot Spahiu
Keyword(s):  
Cast Iron ◽  
Aqueous Phase ◽  
Atmospheric Oxygen ◽  
Near Field ◽  
Fission Products ◽  
Spent Fuel ◽  
Metal Foils ◽  
Oxygen Contamination ◽  
Leaching Experiment ◽  
Fuel Pin

ABSTRACTThe results of a spent fuel leaching experiment in which a fuel pin (17.7 g) was contacted with 380 mL of a 10 mM NaCl, 2 mM NaHCO3 solution by taking special care to minimize atmospheric oxygen contamination are presented. During the first 287 days, the fractions of inventory in the aqueous phase per day (f/d) increased nearly constantly for all nuclides (except for 100Mo), but were higher for fission products f/d(137Cs)=1.210−6, f/d(99Tc)=1.1·10−6 and f/d(90Sr)= 6.7 · 10−7 than for actinides: f/d (238U) =1.0 · 10−7, f/d(237Np)= 2.6 · 10−7 and f/d(239Pu) = 5.1 ·10−9. After adding iron, cast iron and copper foils (of ∼30 mm2 size), the concentrations of 238U, 237Np and 99Tc decreased by 80%, 97% and 88% to relatively stable levels (500ppb, 0.2 ppb and 0.6 ppb respectively). 239Pu concentrations increased from a level around 0.05 ppb to PuO2 solubility level, 0.5 ppb, and stabilized. The leaching process for 137Cs, 100Mo and 90Sr seems not to be influenced by the addition of metal foils. The observations in the present work contribute to an improved understanding of the behavior of spent fuel under near field repository conditions.

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