Radiochemical aspects of production and processing of radiometals for preparation of metalloradiopharmaceuticals

Radiometals play an important role in nuclear medicine as involved in diagnostic or therapeutic agents. In the present work the radiochemical aspects of production and processing of very promising radiometals of the third group of the periodic table, namely radiogallium and radiolanthanides are investigated. The 68Ge/68Ga generator (68Ge, T½ = 270.8 d) provides a cyclotron-independent source of positron-emitting 68Ga (T½ = 68 min), which can be used for coordinative labelling. However, for labelling of biomolecules via bifunctional chelators, particularly if legal aspects of production of radiopharmaceuticals are considered, 68Ga(III) as eluted initially needs to be pre-concentrated and purified. The first experimental chapter describes a system for simple and efficient handling of the 68Ge/68Ga generator eluates with a cation-exchange micro-chromatography column as the main component. Chemical purification and volume concentration of 68Ga(III) are carried out in hydrochloric acid – acetone media. Finally, generator produced 68Ga(III) is obtained with an excellent radiochemical and chemical purity in a minimised volume in a form applicable directly for the synthesis of 68Ga-labelled radiopharmaceuticals. For labelling with 68Ga(III), somatostatin analogue DOTA-octreotides (DOTATOC, DOTANOC) are used. 68Ga-DOTATOC and 68Ga-DOTANOC were successfully used to diagnose human somatostatin receptor-expressing tumours with PET/CT. Additionally, the proposed method was adapted for purification and medical utilisation of the cyclotron produced SPECT gallium radionuclide 67Ga(III). Second experimental chapter discusses a diagnostic radiolanthanide 140Nd, produced by irradiation of macro amounts of natural CeO2 and Pr2O3 in natCe(3He,xn)140Nd and 141Pr(p,2n)140Nd nuclear reactions, respectively. With this produced and processed 140Nd an efficient 140Nd/140Pr radionuclide generator system has been developed and evaluated. The principle of radiochemical separation of the mother and daughter radiolanthanides is based on physical-chemical transitions (hot-atom effects) of 140Pr following the electron capture process of 140Nd. The mother radionuclide 140Nd(III) is quantitatively absorbed on a solid phase matrix in the chemical form of 140Nd-DOTA-conjugated complexes, while daughter nuclide 140Pr is generated in an ionic species. With a very high elution yield and satisfactory chemical and radiolytical stability the system could able to provide the short-lived positron-emitting radiolanthanide 140Pr for PET investigations. In the third experimental chapter, analogously to physical-chemical transitions after the radioactive decay of 140Nd in 140Pr-DOTA, the rapture of the chemical bond between a radiolanthanide and the DOTA ligand, after the thermal neutron capture reaction (Szilard-Chalmers effect) was evaluated for production of the relevant radiolanthanides with high specific activity at TRIGA II Mainz nuclear reactor. The physical-chemical model was developed and first quantitative data are presented. As an example, 166Ho could be produced with a specific activity higher than its limiting value for TRIGA II Mainz, namely about 2 GBq/mg versus 0.9 GBq/mg. While free 166Ho(III) is produced in situ, it is not forming a 166Ho-DOTA complex and therefore can be separated from the inactive 165Ho-DOTA material. The analysis of the experimental data shows that radionuclides with half-life T½ < 64 h can be produced on TRIGA II Mainz nuclear reactor, with specific activity higher than any available at irradiation of simple targets e.g. oxides.

Validation of the CFD code NEPTUNE for a full scale simulator for decay heat removal systems with in-pool heat exchangers

In the present work, a multi physics simulation of an innovative safety system for light water nuclear reactor is performed, with the aim to increase the reliability of its main decay heat removal system. The system studied, denoted by the acronym PERSEO (in Pool Energy Removal System for Emergency Operation) is able to remove the decay power from the primary side of the light water nuclear reactor through a heat suppression pool. The experimental facility, located at SIET laboratories (PIACENZA), is an evolution of the Thermal Valve concept where the triggering valve is installed liquid side, on a line connecting two pools at the bottom. During the normal operation, the valve is closed, while in emergency conditions it opens, the heat exchanger is flooded with consequent heat transfer from the primary side to the pool side. In order to verify the correct system behavior during long term accidental transient, two main experimental PERSEO tests are analyzed. For this purpose, a coupling between the mono dimensional system code CATHARE, which reproduces the system scale behavior, with a three-dimensional CFD code NEPTUNE CFD, allowing a full investigation of the pools and the injector, is implemented. The coupling between the two codes is realized through the boundary conditions. In a first analysis, the facility is simulated by the system code CATHARE V2.5 to validate the results with the experimental data. The comparison of the numerical results obtained shows a different void distribution during the boiling conditions inside the heat suppression pool for the two cases of single nodalization and three volume nodalization scheme of the pool. Finaly, to improve the investigation capability of the void distribution inside the pool and the temperature stratification phenomena below the injector, a two and three dimensional CFD models with a simplified geometry of the system are adopted.

Nicht geträgerte Radioarsenisotope: ihre Herstellung, Abtrennung und Markierung von Proteinen

Das Element Arsen besitzt eine Reihe von Isotopen, die in nahezu trägerfreier Form (nca) produziert werden können und deshalb in der Radiopharmazie für die Diagnose oder Endoradiotherapie Verwendung finden können. Bei der Positronenemissionstomographie (PET) gibt es eine gewisse Lücke bei der Versorgung mit langlebigen Positronenemittern, die zur Untersuchung von langsamen physiologischen Prozessen wie z.B. der Biodistribution und Anreicherung von Antikörpern in Tumorgewebe eingesetzt werden können. Die beiden Arsenisotope 72As (T1/2 = 26 h, 88 % beta+) und 74As (T1/2 = 17,8 d, 29 % beta+) vereinen eine lange physikalische Halbwertszeit mit einer hohen Positronenemissionsrate und sind daher geeignete Kandidaten. Da das Verhalten von radioaktivem Arsen und seine Verwendung in der molekularen Bildgebung international relativ wenig bearbeitet sind, wurde die Radiochemie des Arsens von der Isotopenproduktion an Kernreaktor und Zyklotron, über die Entwicklung von Abtrennungsmethoden für Germanium und Arsen, bis hin zur Entwicklung einer soliden Markierungschemie für Antikörper weiterentwickelt. Die in dieser Arbeit bearbeiteten Felder sind: 1. Die Isotopenproduktion der relevanten Arsenisotope (72/74/77As) wurde an Kernreaktor und Zyklotron durch Bestrahlung von GeO2- und Germaniummetalltargets durchgeführt. Pro 6 h Bestrahlung von 100 mg Germanium konnten ca. 2 MBq 77As am TRIGA Reaktor in Mainz hergestellt werden. Am Zyklotron des DKFZ in Heidelberg konnten unter optimierten Bedingungen bei der Bestrahlung von Germaniummetall (EP = 15 Mev, 20 µA, 200 µAh) ca. 4 GBq 72As und ca. 400 MBq 74As produziert werden. 2. Die Entwicklung neuer Abtrennungsmethoden für nca 72/74/77As von makroskopischen Mengen Germanium wurde vorangetrieben. Für die Aufarbeitung von GeO2- und Germaniummetalltargets kamen insgesamt 8 verschiedene Methoden wie Festphasenextraktion, Flüssig-Flüssig-Extraktion, Destillation, Anionenaustauschchromatographie zum Einsatz. Die erzielten Ausbeuten lagen dabei zwischen 31 und 56 %. Es wurden Abtrennungsfaktoren des Germaniums zwischen 1000 und 1•10E6 erreicht. Alle erfolgreichen Abtrennungsmethoden lieferten *As(III) in 500 µl PBS-Puffer bei pH 7. Diese Form des Radioarsens ist für die Markierung von SH-modifizierten Molekülen, wie z.B. Antikörpern geeignet. 3. Die Entwicklung von Methoden zur Bestimmung des Oxidationszustandes von nca *As in organischem, neutralem wässrigen, oder stark sauren Medium mittels Radio-DC und Anionenaustauschchromatographie wurde durchgeführt und führte zu einem besseren Verständnis der Redoxchemie des nca *As. 4. SH-modifizierte Antikörper wurden mit 72/74/77As(III) markiert. Dabei wurden zwei Methoden (Modifizierung mit SATA und TCEP) miteinander verglichen. Während das *As(III) bei Verwendung von TCEP in Ausbeuten > 90 % mit dem Antikörper reagierte, wurde für SATA-modifizierte Antikörper in Abhängigkeit von der verwendeten Abtrennungsmethode eine breite Spanne von 0 % bis > 90 % beobachtet. 5. Es wurden Phantommessungen mit 18F, 72As und 74As am µ-PET-Scanner durchgeführt, um erste Aussagen über die zu erwartende Auflösung der Arsenisotope zu erhalten. Die Auflösung von 74As ist mit 18F vergleichbar, während die von 72As erkennbar schlechter ist.

The construction of TRIGA-TRAP and direct high-precision Penning trap mass measurements on rare-earth elements and americium

Nuclear masses are an important quantity to study nuclear structure since they reflect the sum of all nucleonic interactions. Many experimental possibilities exist to precisely measure masses, out of which the Penning trap is the tool to reach the highest precision. Moreover, absolute mass measurements can be performed using carbon, the atomic-mass standard, as a reference. The new double-Penning trap mass spectrometer TRIGA-TRAP has been installed and commissioned within this thesis work, which is the very first experimental setup of this kind located at a nuclear reactor. New technical developments have been carried out such as a reliable non-resonant laser ablation ion source for the production of carbon cluster ions and are still continued, like a non-destructive ion detection technique for single-ion measurements. Neutron-rich fission products will be available by the reactor that are important for nuclear astrophysics, especially the r-process. Prior to the on-line coupling to the reactor, TRIGA-TRAP already performed off-line mass measurements on stable and long-lived isotopes and will continue this program. The main focus within this thesis was on certain rare-earth nuclides in the well-established region of deformation around N~90. Another field of interest are mass measurements on actinoids to test mass models and to provide direct links to the mass standard. Within this thesis, the mass of 241-Am could be measured directly for the first time.

Fast and accurate numerical solutions in some problems of particle and radiation transport: synthetic acceleration for the method of short characteristics, Doppler-broadened scattering kernel, remote sensing of the cryosphere

The aim of this work is to present various aspects of numerical simulation of particle and radiation transport for industrial and environmental protection applications, to enable the analysis of complex physical processes in a fast, reliable, and efficient way. In the first part we deal with speed-up of numerical simulation of neutron transport for nuclear reactor core analysis. The convergence properties of the source iteration scheme of the Method of Characteristics applied to be heterogeneous structured geometries has been enhanced by means of Boundary Projection Acceleration, enabling the study of 2D and 3D geometries with transport theory without spatial homogenization. The computational performances have been verified with the C5G7 2D and 3D benchmarks, showing a sensible reduction of iterations and CPU time. The second part is devoted to the study of temperature-dependent elastic scattering of neutrons for heavy isotopes near to the thermal zone. A numerical computation of the Doppler convolution of the elastic scattering kernel based on the gas model is presented, for a general energy dependent cross section and scattering law in the center of mass system. The range of integration has been optimized employing a numerical cutoff, allowing a faster numerical evaluation of the convolution integral. Legendre moments of the transfer kernel are subsequently obtained by direct quadrature and a numerical analysis of the convergence is presented. In the third part we focus our attention to remote sensing applications of radiative transfer employed to investigate the Earth's cryosphere. The photon transport equation is applied to simulate reflectivity of glaciers varying the age of the layer of snow or ice, its thickness, the presence or not other underlying layers, the degree of dust included in the snow, creating a framework able to decipher spectral signals collected by orbiting detectors.

Safety, Security And safeguards In GEN IV sodium fast reactors

This work presents first a study of the national and international laws in the fields of safety, security and safeguards. The international treaties and the recommendations issued by the IAEA as well as the national regulations in force in France, the United States and Italy are analyzed. As a result of this, a comparison among them is presented. Given the interest of the Japan Atomic Energy Agency for the aspects of criminal penalties and monetary, also the Japanese case is analyzed. The main part of this work was held at the JAEA in the field of proliferation resistance (PR) and physical protection (PP) of a GEN IV sodium fast reactor. For this purpose the design of the system is completed and the PR & PP methodology is applied to obtain data usable by designers for the improvement of the system itself. Due to the presence of sensitive data, not all the details can be disclosed. The reactor site of a hypothetical and commercial sodium-cooled fast neutron nuclear reactor system (SFR) is used as the target NES for the application of the methodology. The methodology is applied to all the PR and PP scenarios: diversion, misuse and breakout; theft and sabotage. The methodology is applied to the SFR to check if this system meets the target of PR and PP as described in the GIF goal; secondly, a comparison between the SFR and a LWR is performed to evaluate if and how it would be possible to improve the PR&PP of the SFR. The comparison is implemented according to the example development target: achieving PR&PP similar or superior to domestic and international ALWR. Three main actions were performed: implement the evaluation methodology; characterize the PR&PP for the nuclear energy system; identify recommendations for system designers through the comparison.

Global modeling of pollutant transport and deposition from anthropogenic emission hot spots on global scale

Urban centers significantly contribute to anthropogenic air pollution, although they cover only a minor fraction of the Earth's land surface. Since the worldwide degree of urbanization is steadily increasing, the anthropogenic contribution to air pollution from urban centers is expected to become more substantial in future air quality assessments. The main objective of this thesis was to obtain a more profound insight in the dispersion and the deposition of aerosol particles from 46 individual major population centers (MPCs) as well as the regional and global influence on the atmospheric distribution of several aerosol types. For the first time, this was assessed in one model framework, for which the global model EMAC was applied with different representations of aerosol particles. First, in an approach with passive tracers and a setup in which the results depend only on the source location and the size and the solubility of the tracers, several metrics and a regional climate classification were used to quantify the major outflow pathways, both vertically and horizontally, and to compare the balance between pollution export away from and pollution build-up around the source points. Then in a more comprehensive approach, the anthropogenic emissions of key trace species were changed at the MPC locations to determine the cumulative impact of the MPC emissions on the atmospheric aerosol burdens of black carbon, particulate organic matter, sulfate, and nitrate. Ten different mono-modal passive aerosol tracers were continuously released at the same constant rate at each emission point. The results clearly showed that on average about five times more mass is advected quasi-horizontally at low levels than exported into the upper troposphere. The strength of the low-level export is mainly determined by the location of the source, while the vertical transport is mainly governed by the lifting potential and the solubility of the tracers. Similar to insoluble gas phase tracers, the low-level export of aerosol tracers is strongest at middle and high latitudes, while the regions of strongest vertical export differ between aerosol (temperate winter dry) and gas phase (tropics) tracers. The emitted mass fraction that is kept around MPCs is largest in regions where aerosol tracers have short lifetimes; this mass is also critical for assessing the impact on humans. However, the number of people who live in a strongly polluted region around urban centers depends more on the population density than on the size of the area which is affected by strong air pollution. Another major result was that fine aerosol particles (diameters smaller than 2.5 micrometer) from MPCs undergo substantial long-range transport, with about half of the emitted mass being deposited beyond 1000 km away from the source. In contrast to this diluted remote deposition, there are areas around the MPCs which experience high deposition rates, especially in regions which are frequently affected by heavy precipitation or are situated in poorly ventilated locations. Moreover, most MPC aerosol emissions are removed over land surfaces. In particular, forests experience more deposition from MPC pollutants than other land ecosystems. In addition, it was found that the generic treatment of aerosols has no substantial influence on the major conclusions drawn in this thesis. Moreover, in the more comprehensive approach, it was found that emissions of black carbon, particulate organic matter, sulfur dioxide, and nitrogen oxides from MPCs influence the atmospheric burden of various aerosol types very differently, with impacts generally being larger for secondary species, sulfate and nitrate, than for primary species, black carbon and particulate organic matter. While the changes in the burdens of sulfate, black carbon, and particulate organic matter show an almost linear response for changes in the emission strength, the formation of nitrate was found to be contingent upon many more factors, e.g., the abundance of sulfuric acid, than only upon the strength of the nitrogen oxide emissions. The generic tracer experiments were further extended to conduct the first risk assessment to obtain the cumulative risk of contamination from multiple nuclear reactor accidents on the global scale. For this, many factors had to be taken into account: the probability of major accidents, the cumulative deposition field of the radionuclide cesium-137, and a threshold value that defines contamination. By collecting the necessary data and after accounting for uncertainties, it was found that the risk is highest in western Europe, the eastern US, and in Japan, where on average contamination by major accidents is expected about every 50 years.

The Amplifier - v. 7, no. 3

In this issue...Nuclear Reactor Testing Station, Arco, Idaho, Wesley club, Northern Pacific Railroad, Newman Club, geology, Russell Barthell, Main Hall, Thanksgiving, Montana Power

Thermochromatography study of volatile polonium species in various gas atmospheres

Phenomena related to the volatilization of polonium and its compounds are critical issues for the safety assessment of the innovative lead–bismuth cooled type of nuclear reactor or accelerator driven systems. The formation and volatilization of different species of polonium and their interaction with fused silica was studied by thermochromatography using carrier gases with varied redox potential. The obtained results show that under inert and reducing conditions in the absence of moisture, elemental polonium is formed. Polonium compounds more volatile than elemental polonium can be formed if traces of moisture are present in both inert and reducing carrier gas. The use of dried oxygen as carrier gas leads to the formation of polonium oxides, which are less volatile than elemental polonium. It was also found that the volatility of polonium oxides increases with increasing oxidation state. In the presence of moisture in an oxidizing carrier gas, species are formed that are more volatile than the oxides and less volatile than the elemental polonium. Considering the redox potential of the carrier gas those species are likely oxyhydroxides.

A design-phase PSA of anuclear-powered hydrogen plant

A probabilistic safety assessment (PSA) is being developed for a steam-methane reforming hydrogenproduction plant linked to a high-temperature gas-cooled nuclear reactor (HTGR). This work is based on the Japan Atomic Energy Research Institute's (JAERI) High Temperature Engineering Test Reactor (HTTR) prototype in Japan. The objective of this paper is to show how the PSA can be used for improving the design of the coupled plants. A simplified HAZOP study was performed to identify initiating events, based on existing studies. The results of the PSA show that the average frequency of an accident at this complex that could affect the population is 7 × 10−8 year−1 which is divided into the various end states. The dominant sequences are those that result in a methane explosion and occur with a frequency of 6.5 × 10−8 year−1, while the other sequences are much less frequent. The health risk presents itself if there are people in the vicinity who could be affected by the explosion. This analysis also demonstrates that an accident in one of the plants has little effect on the other. This is true given the design base distance between the plants, the fact that the reactor is underground, as well as other safety characteristics of the HTGR.

A proposed parameterization of interface discontinuity factors depending on neighborhood for pin-by-pin diffusion computations for LWR

There exists an interest in performing full core pin-by-pin computations for present nuclear reactors. In such type of problems the use of a transport approximation like the diffusion equation requires the introduction of correction parameters. Interface discontinuity factors can improve the diffusion solution to nearly reproduce a transport solution. Nevertheless, calculating accurate pin-by-pin IDF requires the knowledge of the heterogeneous neutron flux distribution, which depends on the boundary conditions of the pin-cell as well as the local variables along the nuclear reactor operation. As a consequence, it is impractical to compute them for each possible configuration. An alternative to generate accurate pin-by-pin interface discontinuity factors is to calculate reference values using zero-net-current boundary conditions and to synthesize afterwards their dependencies on the main neighborhood variables. In such way the factors can be accurately computed during fine-mesh diffusion calculations by correcting the reference values as a function of the actual environment of the pin-cell in the core. In this paper we propose a parameterization of the pin-by-pin interface discontinuity factors allowing the implementation of a cross sections library able to treat the neighborhood effect. First results are presented for typical PWR configurations.

Estudio de la funcionalización de secciones eficaces de celda en multigrupos para cálculos best-estimate 3D pin-by-pin de núcleos PWR

Uno de los objetivos del Proyecto europeo NURISP (NUclear Reactor Integrated Platform) del 7º Programa Marco es avanzar en la simulación de reactores de agua ligera mediante el acoplamiento de códigos best-estimate que profundicen en la física de núcleo, termohidráulica bifásica y comportamiento del combustible [1]. Uno de estos códigos es COBAYA3, código de difusión 3D en multigrupos pin-by-pin desarrollado en la UPM, que requiere de librerías de secciones eficaces homogeneizadas a nivel de la barrita de combustible.

Steps Ahead in the Few-Group CRoss-Section Library Generation at the Pin Level

There exists an interest in performing pin-by-pin calculations coupled with thermal hydraulics so as to improve the accuracy of nuclear reactor analysis. In the framework of the EU NURISP project, INRNE and UPM have generated an experimental version of a few group diffusion cross sections library with discontinuity factors intended for VVER analysis at the pin level with the COBAYA3 code. The transport code APOLLO2 was used to perform the branching calculations. As a first proof of principle the library was created for fresh fuel and covers almost the full parameter space of steady state and transient conditions. The main objective is to test the calculation schemes and post-processing procedures, including multi-pin branching calculations. Two library options are being studied: one based on linear table interpolation and another one using a functional fitting of the cross sections. The libraries generated with APOLLO2 have been tested with the pin-by-pin diffusion model in COBAYA3 including discontinuity factors; first comparing 2D results against the APOLLO2 reference solutions and afterwards using the libraries to compute a 3D assembly problem coupled with a simplified thermal-hydraulic model.

Assessment of CTF boiling transition and critical heat flux modeling capabilities using the OECD/NRC BFBT and PSBT benchmark databases

The need to refine models for best-estimate calculations, based on good-quality experimental data, has been expressed in many recent meetings in the field of nuclear applications. The modeling needs arising in this respect should not be limited to the currently available macroscopic methods but should be extended to next-generation analysis techniques that focus on more microscopic processes. One of the most valuable databases identified for the thermalhydraulics modeling was developed by the Nuclear Power Engineering Corporation (NUPEC), Japan. From 1987 to 1995, NUPEC performed steady-state and transient critical power and departure from nucleate boiling (DNB) test series based on the equivalent full-size mock-ups. Considering the reliability not only of the measured data, but also other relevant parameters such as the system pressure, inlet sub-cooling and rod surface temperature, these test series supplied the first substantial database for the development of truly mechanistic and consistent models for boiling transition and critical heat flux. Over the last few years the Pennsylvania State University (PSU) under the sponsorship of the U.S. Nuclear Regulatory Commission (NRC) has prepared, organized, conducted and summarized the OECD/NRC Full-size Fine-mesh Bundle Tests (BFBT) Benchmark. The international benchmark activities have been conducted in cooperation with the Nuclear Energy Agency/Organization for Economic Co-operation and Development (NEA/OECD) and Japan Nuclear Energy Safety (JNES) organization, Japan. Consequently, the JNES has made available the Boiling Water Reactor (BWR) NUPEC database for the purposes of the benchmark. Based on the success of the OECD/NRC BFBT benchmark the JNES has decided to release also the data based on the NUPEC Pressurized Water Reactor (PWR) subchannel and bundle tests for another follow-up international benchmark entitled OECD/NRC PWR Subchannel and Bundle Tests (PSBT) benchmark. This paper presents an application of the joint Penn State University/Technical University of Madrid (UPM) version of the well-known subchannel code COBRA-TF, namely CTF, to the critical power and departure from nucleate boiling (DNB) exercises of the OECD/NRC BFBT and PSBT benchmarks

Neighborhood-corrected interface discontinuity factors for multi-group pin-by-pin diffusion calculations for LWR

Performing three-dimensional pin-by-pin full core calculations based on an improved solution of the multi-group diffusion equation is an affordable option nowadays to compute accurate local safety parameters for light water reactors. Since a transport approximation is solved, appropriate correction factors, such as interface discontinuity factors, are required to nearly reproduce the fully heterogeneous transport solution. Calculating exact pin-by-pin discontinuity factors requires the knowledge of the heterogeneous neutron flux distribution, which depends on the boundary conditions of the pin-cell as well as the local variables along the nuclear reactor operation. As a consequence, it is impractical to compute them for each possible configuration; however, inaccurate correction factors are one major source of error in core analysis when using multi-group diffusion theory. An alternative to generate accurate pin-by-pin interface discontinuity factors is to build a functional-fitting that allows incorporating the environment dependence in the computed values. This paper suggests a methodology to consider the neighborhood effect based on the Analytic Coarse-Mesh Finite Difference method for the multi-group diffusion equation. It has been applied to both definitions of interface discontinuity factors, the one based on the Generalized Equivalence Theory and the one based on Black-Box Homogenization, and for different few energy groups structures. Conclusions are drawn over the optimal functional-fitting and demonstrative results are obtained with the multi-group pin-by-pin diffusion code COBAYA3 for representative PWR configurations.