Fission product simulants

UO2 specimens containing fission product simulants were subjected to heating similar to that used in the Sascha experiments discussed above. All tests were carried out at atmospheric pressure for 2 to 60 minutes at the test temperature different atmospheres (in particular different ratios of H2 H2O) have been used in different tests of the series. 

Computer simulation programs for process design optimization have been developed for the PUREX process utilizing these relationships (22). A subroutine has also been developed which describes the behavior of fission products (23). 

To evaluate fission product release in a reactor, it is necessary to supply the appropriate particle geometry, diffusion coefficients, and distribution coefficients. This is a formidable task. To approach this problem, postirradiation fission product release has been studied as a function of temperature. The results of these studies are complex and require considerable interpretation. The SLIDER code without a source term has proved to be of considerable value in this interpretation. Parametric studies have been made of the integrated release of fission products, initially wholly in the fueled region, as a function of the diffusion coefficients and the distribution coefficients. These studies have led to observations of critical features in describing integrated fission product releases. From experimental values associated with these critical features, it is possible to evaluate at least partially diffusion coefficients and distribution coefficients. These experimental values may then be put back into SLIDER with appropriate birth and decay rates to evaluate inreactor particle fission product releases. Figure 11 is a representation of SLIDER simulation of a simplified postirradiation fission product release experiment. Calculations have been made with the following pertinent input data ... 

Myasoedov, B.F. Kulyako, Y.M. Trofimov, T.I. Samsonov, M.D. Malikov D.A. Recovery of U and Pu from simulated spent nuclear fuel by adducts of organic reagents with HN03 followed by their separation from fission products by countercurrent chromatography, Plutonium Futures - The Science 2008, Dijon, France, July 7-11. 

Distribution ratios and transport were carried out on real HAW arising from dissolution of a mixed oxide of uranium and plutonium (MOX) fuel (burnup 34,650 MW d/tU), where uranium and plutonium have been previously extracted by TBP.86 The experiments were performed in the CARMEN hot cell of CEA Fontenay aux Roses with two dialkoxy-calix[4]arene-crown-6 derivatives (diisopropoxy and dini-trophenyl-octyloxy). High cesium distribution ratios were obtained (higher than 50) by contacting the HAW solution with diisopropoxy calix[4]arene-crown-6 (0.1 M in NPHE). Moreover, the high selectivity observed with the simulated waste was confirmed for most of the elements and radionuclides (actinides or fission products Eu, Sb, Ce, Mo, Zr, and Nd). The residual concentration or activity of elements, other than cesium, was less than 1% in the stripping solution, except for iron (2%) and ruthenium (8%) the extraction of these two cations, probably under a complexed... 

Element uptake from soil and transfer into the edible parts of plants have been addressed in several other studies. Soil-to-plant transfer factors in fruit and vegetables grown in various agricultural conditions have been determined for, for example, Pt [100], T1 [101], and various other metal contaminants [102], In a study on stable isotopes of fission product elements (Ce, Cs, Sr), an in vitro enzy-molysis method has been applied to investigate the solubilization of the analytes from fodder in a simulated ruminant digestion [103], The effect of inhibitors of fission product solubility was also considered and essential elements were determined simultaneously to evaluate potential nutrition problems for the animals from the use of such inhibitors. Selective leaching of individual classes of metal complexes with different ligands and sequential enzymolysis have been recently applied to estimate the potential bioavailability to humans of Cd and Pb in cocoa powder and related products [104]. 

In laboratory tests using simulated HLW solution spiked with fission product tracers, Am and Cm, the denitration step proved to be a sensitive process, but Am/Cm recoveries of ca. 90% in the aqueous supernate could be realized under optimized conditions. Decontamination factors (DF) > 1000 for Zr, Nb, Mo, and 100 for Ru and Fe were obtained in the precipitation step. The solvent extraction cycle gave > 98% recovery of Am/Cm and DF > 10 for rare earths, Sr and Cs. Appreciable decontamination was also obtained for Zr/Nb (DF = 20), Ru (50), U (650), Pu (250), Np (800) and Fe (420). The ion exchange cycle served mainly for Am-Cm concentration and for removal of DTPA and lactic acid based on tests with europium as a stand-in for trivalent actinides, concentration factors of about 50 could be expected under optimized conditions. 

Off-line Simulation with Recoiling Fission Products... 

To simulate recovery of uranium and thorium from irradiated 6 percent uranium, 94 percent thorium fuel from the first loading of Consolidated Edison Company s Indian Point 1 nuclear power plant, Oak Ridge National Laboratory [R3] made small-scale experiments on application of the acid Thorex process to fuel containing the appropriate amounts of uranium and thorium, with tracer quantities of the principal fission products. Spent uranium-thorium fuel from the Indian Point 1 plant was subsequently processed by Nuclear Fuel Services, Inc., at West Valley, New York, for recovery of uranium, but without separation of thorium from fission products. No account of this separation has been published. 

The fuel consists of a series of UN pellets that are total of 52 cm in length and 0.91 cm in diameter. These are capped by 5 cm of BeO to act as an axial reflector. There is a gas plenum on the top of the pin to allow for fission product gas build up. A layer of rhenium acts as a liner for the fuel pin to prevent interaction between the NblZr cladding and the UN pellets and also acts as a thermal neutron absorber in accident cases as discussed later. The NblZr end caps appear large in this Figure but their purpose is to simulate other components like the connectors that attach the fuel pin to a grid plate. 

During the same period, the range and sophistication of the applications developed. Representative new areas included the extensive range of simulation studies of defects in superionic conductors [20] in high-temperature superconductors (see, for example. Refs. [21, 22]) and the detailed studies of the complex defect chemistry arising from the creation of fission products in CUO2 nuclear fuels [23]. 

If pure water were substituted by 1 MHNO, as moderator, a reduction of Ak 0.04 was found for maximum k, . To simulate the content of nonvolatile fission products in the fuel, an equivalent gadolinium content was derived from ORIGEN data by weighing over thermal absorption cross sections. A remarkable reduction of k , occurs in this case only for unpoisoned lattices (Ak = -0.08 for k ,in ). For a Gd concentration of 3 g/litre in the moderator, the fission products in the fuel reduce k ,mair Onlv bv 0.01. 

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