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Glioblastomas (GBM) are the most frequent and aggressive brain tumors due to their recurrence and resistance to current therapies. These characteristics are associated with the presence of glioma stem cells (GSCs), mainly identified by the detection of the membrane antigens CD133 and CD15. The main source of GSCs has been biopsies of tumors. However, alternatives are sought from cell lines because more homogeneous populations can be obtained with high yields. This chapter describes a method for the enrichment and characterization of GSCs from cell lines derived from human GBM by selective culture with serum-free neural stem cell medium and growth factors. The technique offers alternatives for the enrichment and characterization of GSCs, that could contribute to a better understanding of the biology of GBMs.

The radiocarbon (14C) content of irradiated graphite is the most important problem for the management of Spanish irradiated graphite (Vandellós I NPP) as L&ILW, due to this material exceeding the maximum14C inventory for the C.A. El Cabril repository. Therefore, the encapsulation of graphite in an impermeable matrix and making an appropriate waste form are indicated as potential management options to be studied. The conversion of the graphite to a long-term stable glass matrix, called IGM (impermeable graphite matrix), uses a long-term stable inorganic binder which additionally encloses the graphite pore system. The world’s first IGM samples made with irradiated graphite have been manufactured in CIEMAT facilities. The durability of the matrix is investigated in leaching experiments in deionized water and granitic bentonite water. The results show that ∼0.05% of14C is leached. A species of organic carbon was found as formate and oxalate (∼10–1mg/L). CO was detected as volatile specie in both media in the first leaching steps; for deionized water (∼3.101mg/L) and in granitic bentonite water (ranging 1.101–3.101mg/L). These low values demonstrated the durability of the IGM glass matrix for final disposal.

The understanding of the 14C behavior in waste packages could lead, in the Spanish context, to a revision of the management strategies for radioactive waste and a revaluation of the near surface repository devoted to the disposal of waste containing this radionuclide in high concentrations. To achieve this objective, and in the context of the EU project Carbon-14 Source Term (CAST), the authors of the work presented in this paper have performed leaching experiments with irradiated graphite considering two different scenarios. One, in which the leaching solution simulates some of the expected conditions in a repository where a granite/bentonite mixture has been used as backfill material, and the other, using deionized water as a high efficiency chemical removal agent and for comparison purposes. The analytical approach to measure the release rate and speciation of 14C from irradiated graphite samples in the aqueous and gaseous phase is also described. The main results obtained shows that, after 359 days of leaching, no 14C activity was detected above the detection limits, and only leaching rates regarding beta-gamma emitters were observed: 2×10–6 cm/day for 137Cs and 1×10–5 cm/day for 60Co in granite/bentonite water media and 4×10–6 cm/day for 137Cs in pure water.

It has been demonstrated that the human biomonitoring of susceptible populations is a valuable method for the identification of critical contaminants. Therefore, the purpose of this study was to assess the exposure profile for arsenic (As), lead (Pb), mercury (Hg), 1-hydroxypyrene (1-OHP), 1,1-bis(p-chlorophenyl)-2,2,2-trichloroethane (DDT), 1,1-dichloro-2,2-bis(p-chlorophenyl) ethylene (DDE), polybrominated diphenyl ethers (PBDEs), and polychlorinated biphenyls (PCBs) in children living in Ciudad Juarez, Chihuahua, Mexico (a major manufacturing center in Mexico). In 2012, we evaluated a total of 135 healthy children living in Ciudad Juarez since birth. The total PBDEs levels ranged from nondetectable (< LOD) to 215 ng/g lipid, with a mean total PBDEs level of 29.5 ± 53.0 ng/g lipid (geometric mean ± standard deviation). The mean total PCBs level in the study participants was 29.0 ± 10.5 ng/g lipid (range 4.50-50.0 ng/g lipid). The mean concentration of total DDT (DDT + DDE) was 11.9 ± 6.70 ng/g lipid (range 3.00-26.0 ng/g lipid). The mean 1-OHP levels was 1.2 ± 1.1 µmol/mol creatinine (range

A middle-aged man with cardiovascular risk factors, who suffered from a slight loss of unilateral vision for 6 months. After obtaining a diagnosis of diffuse choroiditis+papillitis within a probable masquerade syndrome, the patient was referred to outpatient Internal Medicine where this diagnosis was confirmed as an extended neoplastic disease.

ENTRAP comprises a pan-European cooperation of leading scientific institutions and regulatory bodies in the field of nuclear-waste characterization and its quality assurance for the safe disposal of radioactive waste. Here, the scope of this cooperation is presented and explained and links or interfaces for a potential collaboration with partners fulfilling tasks of IDG-TP are pursued.

The radioactive chlorine isotope, 36Cl, decays with a half-life of 3×105 years by emitting a beta particle (98 %) and by electron capture. The aim of this paper is to propose a radiochemical separation method of 36Cl from the other beta-gamma emitters present in low and medium radioactive wastes such as spent ion exchange resins and evaporator concentrates, that arise from Nuclear Power Plants and particularly in the wastes that come from decommissioning activities of graphite reactors, in order to provide data for 36Cl inventory calculations. The separation method proposed is based on an oxidation technique where chlorine is trapped by NaOH. 36Cl beta emissions are measured by liquid scintillation counting by the dual label technique in order to avoid the contamination produced by 14C which is also trapped by NaOH and which is the main contaminant present in graphite samples. The sensitivity of this method is sufficient to achieve the needed thresholds for the radiological characterization of the radioactive materials to which this method can be applied.

The zirconium isotope 93 Zr is a long-lived pure β -particle-emitting radionuclide, which is produced by nuclear fission and neutron activation of the stable isotope 92 Zr. This element is a constituent of the structural components of nuclear reactor vessels. A selective liquid–liquid extraction method for radiochemical separation of Zr, based on liquid–liquid extraction with 1-(2′-thenoyl)-3,3,3-trifluoroacetone in xylene and a subsequent stripping of 93 Zr by an aqueous acid solution, has been developed. The method was utilised to separate Zr from other pure β -particle and β – γ emitters in different kinds of samples. Decontamination factors higher than 99% for the pure β -particle and β – γ emitters and an overall chemical yield of 80% were obtained. The sensitivity of the method allows the determination of the isolated 93 Zr by liquid scintillation counting and the minimum detectable activity value obtained was 0.067 Bq over a counting period of 60 min.

The radioactive isotopes of calcium, 41Ca ( t 1/2 =1.03×10 5 yr ) and 45Ca ( t 1/2 =163 d ), are produced by neutron capture in the stable isotopes 40Ca and 44Ca, respectively. These radionuclides are present in the environment due to the reactions between the galactic cosmic rays and the earth's surface, and in nuclear power plants by the activation of the structural components of nuclear reactor vessels. The aim of this paper is to propose a radiochemical separation method of 41Ca and 45Ca from the other beta–gamma emitters present in radioactive materials, based on selective precipitation reactions. The activities were measured by liquid scintillation counting (LSC). The obtained decontamination factors are satisfactory for each radioactive component of the initial sample in that their activities in the final product were lower than the minimum detectable activity (MDA) due to the effectiveness of the radiochemical procedure. The sensitivity of the method allows the radiological characterization of 41Ca and 45Ca content in radioactive materials.

Qualification tests of coatings for inner walls in a Boiling Water Reactor (BWR) were performed under request for decontamination purposes. Several types of coatings were contaminated with fission products obtained from a UO 2 solution irradiated in a nuclear research reactor, afterwards a successive decontamination process according with standards based on γ activity was applied. In each decontamination stage, α and γ activity measurements were made on the probes. Results obtained for α activity using nuclear track detectors are presented as an alternative assessment method during the decontamination process.

A whole body counter determined the presence of radioactivities up to 21.8 kBq for 214Bi and up to 18.7 kBq for 214Pb attached to clothes of workers in a Nuclear Research Laboratory. A radon survey reveals that 80% of the monitoring areas have radon concentration values lower than 500 Bq m−3, while 10% of the sampling points with values bigger than 1 kBq m−3 correspond to the workers mentioned above. By exposing samples of 0.04 m2 clothes in a radon chamber, it was observed that radon decay products 214Bi and 214Pb were attached to them with an activity of 315–618 Bq per each kBq m−3 of Rn concentration additionally, fibres characterised with a lower electrostatics build up showed the lower attachment.

Low-level technological radioactive wastes, in Spain, are commonly produced in research and medical centers. These wastes must be characterized before conditioning in order to determine their radioactive content for inventory purposes. A prototype has been designed for beta–gamma radiological characterization of standardized 25 l bags containing heterogeneous low-density technological radioactive wastes within the density range 0.05–0.6 g/cm 3 . The system consists of an iron shielding box with three NaI(Tl) and a silicon implanted detectors for gamma and gross beta activity determinations, respectively. The study of the measurement method, carried out with rotating scanning, included the optimization of the detection solid angle to minimize the uncertainties and the influence of the relative position of the radioactive material. Several materials and densities, in the range aforementioned, were considered to obtain the experimental attenuation factors, used for fitting a correction algorithm in function of density and γ-emission energy. The sensitivity of this method, calculated for the most frequent average density of this kind of waste (0.1 g/cm 3 ), is lower than 50 Bq/kg for the main β – γ emitters ( 137 Cs and 60 Co) and lower than 480 Bq/kg for gross beta activity.

This paper present a comparison between calculated and experimental absorbed dose values, for a typical product, irradiated in a 60 Co industrial irradiator, at ININ. Such industrial irradiator is two tiers and two layer systems, with overlapping product to source configuration; source activity was around 300 kCi. Calculated absorbed dose values were obtained by FUGI code, running in a PC microcomputer. This code is based in a point kernel over the source area, by means of Gauss quadrature technique, based in FUDGE-4A code. Build-up factors were estimated by Goldstein and Wilkins factors for punctual sources, using Taylor's approximation formula. Experimental dosimetry data of Petri dishes packages, with apparent density of 0.13 g/cm 3 data was performed using red acrylic dosimeters, during irradiation; required minimum dose for this material was 15 kGy. Absorbed dose data have shown good approximation, particulary a 15% desviation of results from FUGI code. However, this code can be used for rough calculation of irradiator parameters, before trying to use a more complex and time-consuming code, for detailed analysis.

A controlled radon chamber was used to test the detection performance of a portable active radon detector. The elapsed time to reach detection equilibrium with the radon atmosphere was three hours; the radon detector was sensitive to daily temperature changes. After using the radon detector three hours waiting time is required to place it in a new radon environment.

This paper presents a benchmark comparison between calculated and experimental absorbed dose values tor a typical product, in a 60 Co industrial irradiator, located at ININ, Mexico. The irradiator is a two levels, two layers system with overlapping product configuration with activity around 300kCi. Experimental values were obtanied from routine dosimetry, using red acrylic pellets. Typical product was Petri dishes packages, apparent density 0.13 g/cm 3 ; that product was chosen because uniform size, large quantity and low density. Minimum dose was fixed in 15 kGy. Calculated values were obtained from QAD-CGGP code. This code uses a point kernel technique, build-up factors fitting was done by geometrical progression and combinatorial geometry is used for system description. Main modifications for the code were related with source sumilation, using punctual sources instead of pencils and an energy and anisotropic emission spectrums were included. Results were, for maximum dose, calculated value (18.2 kGy) was 8% higher than experimental average value (16.8 kGy); for minimum dose, calculated value (13.8 kGy) was 3% higher than experimental average value (14.3 kGy).

Fluences data profile inside four beam tubes at a nuclear research reactor has been using activation foils together with CR39 plastic detectors covered with lithium borate converter, to obtain discrete and continuous neutron fluence measurements.

A portable radon detection system (α-Inin) has been designed and constructed for using it in adverse environmental conditions where humidity, temperature and chemical vaporous are present. The minimum integration time is in periods of 15 min during 41 days. A 12 V battery and a photovoltaic module allow the α-Inin autonomy in field measurements. Data is collected by means of a laptop computer where data processing and α-Inin programming are carried out. α-Inin performance was simultaneously tested in a controlled radon chamber, together with a commercial α-Meter.

Low- and medium-level nuclear wastes are generated mainly in the operation of nuclear power plants. The radiological characterisation of these wastes is required during their treatment and conditioning as well as during their storage as a final product in drums. Such requirements make it possible to know the radioactive content for radionuclide-inventory purposes in the final repository.

A case of post-traumatic temporomandibular joint ankylosis is described, and the technique of the treatment is detailed. Previous review of the literature was made in order to know the advantages of the treatment. There are several reports of children who have suffered facial injuries and fractures of the facial bones that became into ankylosis, characterized by the formation of new temporomandibular joint surfaces. The patient was a five years old girl who presented ankylosis of TMJ due to a car accident six months before. The symptom were: Limitation of motion, persistent close lock of mouth (2 mm) muscle pain that causes the patient to avoid eating foods, and the gradual loose of phonetic function. Condylectomy of the left mandibular condyle was performed to correct the mechanical problems associated to ankylosis. The patient had an uncomplicated postoperative course, and now are waiting the outcome of the treatment by a follow up study.