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NASA NTRS1987

In vivo synthesis of phosphatidylcholine in rat brain via the phospholipid methylation pathway

The in vivo synthesis of brain phosphatidylcholine (PC) by the methylation of phosphatidylethanolamine (PE) was examined. (H-3)methyl)methionine was infused i.c.v., by indwelling cannula, and brain samples were taken 0.5-18 h thereafter and assayed for (H-3)PC, as well as for its biosynthetic intermediates (H-3)phosphatidyl monomethylethanolamine ((H-3)PMME) and (H-3)phosphatidyl dimethylethanolamine ((H-3)PDME), and for (H-3)lysophosphatidylcholine ((H-3)LPC) and S-(H-3)adenosylmethionine ((H-3)SAM). Most of the (H-3)PC (79-94 percent) was present ipsilateral to the infusion site; indicating that the radioactivity in the (H-3)PC was primarily of intracerebral origin, and not taken up from the blood. Moreover, only very low levels of (H-3)PC were attained in brains of animals receiving (H-3)methionine i.p. and these levels were symmetrically distributed. (H-3)PMME and (H-3)PDME turned over with apparent half-lives of 2.2 h and 2.4 h. In contrast, the accumulation of brain (H-3)PC was biphasic, suggesting the existence of two pools, the more labile of which turned over rapidly (t(sub 1/2) = 5 h) and was formed for as long as (H-3)PMME and (H-3)PDME are present in the brain, and another, which was distinguishable only at 18 h after the (H-3)methionine infusion. (The latter pool may have been synthesized from (H-3)choline that was released via the hydrolysis of some of the brain (H-3)PC previously formed by the methylation of PE.) Subcellular fractionation of brain tissue obtained after in vivo labelling with (H-3)methionine revealed that mitochondrial PC had the highest specific radioactivity (dpm per micromol total lipid phosphorus), and myelin the least. These observations affirm that rat brain does synthesize PC in vivo by methylating PE, and the technique provides an experimental system which may be useful for examining the physiological regulation of this process.

Lakher, Michael↗

DOE OSTI2020

Feedback on Forty-year Long Clean-up Operations of a Contaminated Soil for Environmental Purpose - 20026

One of the CEA facilities in France is the place of miscellaneous mapping and clean-up operations since the 1980's. The final purpose of the conducted work deals with the environmental remediation of the building. In France, the absence of a regulatory framework for the management of sites polluted by radioactive substances, especially as the lack of release thresholds, has led to develop different approaches of soil radiological characterization. The geostatistical approach is part of them. Preliminary investigations are an unavoidable step prior to clearance and remediation. First, historical analysis enables to define site perimeters to investigate as well as to precisely locate expected contaminated parts. For radioactive pollutants, the media (water, soil, air) and transfer routes participate in the definition of the investigation perimeter and thus also need to be identified through a geological study. At the same time, functional analysis leads to choose the best measurement means which could provide useful information when combined with visual inspection. An equipment well adapted to easy-to-measure radionuclides as gamma emitters, such as gamma probe or on-site gamma spectrometry, is usually selected as a non-destructive assay. Coupled with spatial positions, processed on-site radiological data can be mapped. Data processing consists in mapping and kriging before a geostatistical analysis in order to identify the zones of interest to be targeted. Selected zones are not only linked to high level count rates but also to low level ones so that a pollution-free reference is known in the area. This data processing also enables to categorize waste types according to their origin and contamination levels. An expected volume of different waste then follows as well as the waste characterization equipments. In-depth investigations, e.g. core drillings and samplings, naturally ensue to aim at remediation optimization through the assessment of environmental impact. After clean-up operations, non-destructive assays as well as destructive samplings enable to check the effective clearance. Following this methodology in the present remediation work, results of the first non-destructive assay campaign highlighted a long-lived transuranic radionuclides contamination in the concrete flagstone. Successive non-destructive assays were then performed in the defined zones of interest. Count rates measured with a surface probe were first mapped. The best localization of destructive assays, i.e. core drillings at a 1 meter depth, came from this surface mapping followed by a geostatistical analysis. The concrete flagstone was totally removed. Then, the depth of soil to remove under the flagstone was optimized from the results of radiological activities measured in core drillings samples. A new surface mapping was then drawn after this partial soil excavation. The sand and demolition rubble samples, homogeneously constituted from the flagstone fragments, were analyzed by gamma spectrometry. Meanwhile, waste drums and other containers were also measured by gamma spectrometry and passive neutron measurement devices. All these results have brought miscellaneous pieces of information. In this context, major issues appear in terms of physical constraints such as the premises tininess as well as of soil sampling techniques and in terms of measurement performances assessment and the representativeness of homogeneous samples. The choice of radiological soil and waste characterization devices is described in the light of enhanced performances by keeping in mind radiation protection requirements and a willing to always optimize waste categories (very low level waste) and volume. The improvement of the characterization methods appears through this forty-year long work while sharing the feedback of encountered difficulties. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

NASA NTRS1974

Automated life-detection experiments for the Viking mission to Mars

As part of the Viking mission to Mars in 1975, an automated set of instruments is being built to test for the presence of metabolizing organisms on that planet. Three separate modules are combined in this instrument so that samples of the Martian surface can be subjected to a broad array of experimental conditions so as to measure biological activity. The first, the Pyrolytic Release Module, will expose surface samples to a mixture of C-14O and C-14O2 in the presence of Martian atmosphere and a light source that simulates the Martian visible spectrum. The assay system is designed to determine the extent of assimilation of CO or CO2 into organic compounds. The Gas Exchange Module will incubate surface samples in a humidified CO2 atmosphere. At specified times, portions of the incubation atmosphere will be analyzed by gas chromatography to detect the release or uptake of CO2 and several additional gases. The Label Release Module will incubate surface samples with a dilute aqueous solution of simple radioactive organic substrates in Martian atmosphere, and the gas phase will be monitored continuously for the release of labeled CO2.

Klein, H. P.↗

NASA NTRS1990

Techniques for determining total body water using deuterium oxide

The measurement of total body water (TBW) is fundamental to the study of body fluid changes consequent to microgravity exposure or treatment with microgravity countermeasures. Often, the use of radioactive isotopes is prohibited for safety or other reasons. It was selected and implemented for use by some Johnson Space Center (JCS) laboratories, which permitted serial measurements over a 14 day period which was accurate enough to serve as a criterion method for validating new techniques. These requirements resulted in the selection of deuterium oxide dilution as the method of choice for TBW measurement. The development of this technique at JSC is reviewed. The recommended dosage, body fluid sampling techniques, and deuterium assay options are described.

Bishop, Phillip A.↗

NASA NTRS2017

Measuring Surface Bulk Elemental Composition on Venus

The extreme surface environment (462 C, 93 bars pressure) of Venus makes subsurface measurements of its bulk elemental composition extremely challenging. Instruments landed on the surface of Venus must be enclosed in a pressure vessel. The high surface temperatures also require a thermal control system to keep the instrumentation temperatures within their operational range for as long as possible. Since Venus surface probes can currently operate for only a few hours, it is crucial that the lander instrumentation be able to make statistically significant measurements in a short time. An instrument is described that can achieve such a measurement over a volume of thousands of cubic centimeters of material by using high energy penetrating neutron and gamma radiation. The instrument consists of a Pulsed Neutron Generator (PNG) and a Gamma-Ray Spectrometer (GRS). The PNG emits isotropic pulses of 14.1 MeV neutrons that penetrate the pressure vessel walls, the dense atmosphere and the surface rock. The neutrons induce nuclear reactions in the rock to produce gamma rays with energies specific to the element and nuclear process involved. Thus the energies of the detected gamma rays identify the elements present and their intensities provide the abundance of each element. The GRS spectra are analyzed to determine the Venus elemental composition from the spectral signature of individual major, minor, and trace radioactive elements. As a test of such an instrument, a Schlumberger Litho Scanner oil well logging tool was used in a series of experiments at NASA's Goddard Space Flight Center. The Litho Scanner tool was mounted above large (1.8 m x 1.8 m x.9 m) granite and basalt monuments and made a series of one-hour elemental composition measurements in a planar geometry more similar to a planetary lander measurement. Initial analysis of the results shows good agreement with target elemental assays

Venus↗

DOE OSTI2022

Development of ultra-pure gadolinium sulfate for the Super-Kamiokande gadolinium project

This paper reports the development and detailed properties of about 13 metric tons of gadolinium sulfate octahydrate, Gd 2 (SO 4 ) 3 • 8H 2 O⁠, which has been dissolved into Super-Kamiokande (SK) in the summer of 2020. We evaluate the impact of radioactive impurities in Gd 2 (SO 4 ) 3 • 8H 2 O on diffuse supernova neutrino background searches and solar neutrino observation and confirm the need to reduce radioactive and fluorescent impurities by about three orders of magnitude from commercially available high-purity Gd 2 (SO 4 ) 3 • 8H 2 O⁠. In order to produce ultra-high-purity Gd 2 (SO 4 ) 3 • 8H 2 O, we have developed a method to remove impurities from gadolinium oxide, Gd 2 O 3 , consisting of acid dissolution, solvent extraction, and pH control processes, followed by a high-purity sulfation process. All of the produced ultra-high-purity Gd 2 (SO 4 ) 3 • 8H 2 O is assayed by inductively coupled plasma mass spectrometry and high-purity germanium detectors to evaluate its quality. Because of the long measurement time of high-purity germanium detectors, we have employed several underground laboratories for making parallel measurements including the Laboratorio Subterráneo de Canfranc in Spain, Boulby in the UK, and Kamioka in Japan. In the first half of production, the measured batch purities were found to be consistent with the specifications. However, in the latter half, the Gd 2 (SO 4 ) 3 • 8H 2 O contained one order of magnitude more 228Ra than the budgeted mean contamination. This was correlated with the corresponding characteristics of the raw material Gd 2 O 3 , in which an intrinsically large contamination was present. Based on their modest impact on SK physics, they were nevertheless introduced into the detector. To reduce 228 Ra for the next stage of gadolinium loading to SK, a new process has been successfully established.

79 ASTRONOMY AND ASTROPHYSICS↗

DOE OSTI2022

PET Imaging of the Neurotensin Targeting Peptide NOTA-NT-20.3 Using Cobalt-55, Copper-64 and Gallium-68

Introduction: Neurotensin receptor 1 (NTSR1) is an emerging target for imaging and therapy of many types of cancer. Nuclear imaging of NTSR1 allows for noninvasive assessment of the receptor levels of NTSR1 on the primary tumor, as well as potential metastases. This work focuses on a the neurotensin peptide analogue NT-20.3 conjugated to the chelator NOTA for radiolabeling for use in noninvasive positron emission tomography (PET). NOTA-NT-20.3 was radiolabeled with gallium-68, copper-64, and cobalt-55 to determine the effect that modification of the radiometal has on imaging and potential therapeutic properties of NOTA-NT-20.3. Methods: In vitro assays investigating cell uptake and subcellular localization of the radiolabeled peptides were performed using human colorectal adenocarcinoma HT29 cells. In vivo PET/CT imaging was used to determine the distribution and clearance of the peptide in mice bearing NTSR1 expressing HT29 tumors. Results: Cell uptake studies showed that the highest uptake was obtained with [ 55 Co] Co-NOTA-NT-20.3 (18.70 ± 1.30%ID/mg), followed by [ 64 Cu] Cu-NOTA-NT-20.3 (15.46 ± 0.91%ID/mg), and lastly [ 68 Ga] Ga-NOTA-NT-20.3 (10.94 ± 0.46%ID/mg) (p < 0.001). Subcellular distribution was similar across the three constructs, with the membranous fraction containing the highest amount of radioactivity. In vivo PET/CT imaging of the three constructs revealed similar distribution and tumor uptake at the 1 h imaging timepoint. Tumor uptake was receptor-specific and blockable by co-injection of non-radiolabeled NOTA-NT-20.3. SUV ratios of tumor to heart at the 24 h imaging timepoint show that [ 55 Co] Co-NOTA-NT-20.3 (20.28 ± 3.04) outperformed [ 64 Cu] Cu-NOTA-NT-20.3 (6.52 ± 1.97). In conclusion, our studies show that enhanced cell uptake and increasing tumor to blood ratios over time displayed the superiority of [ 55 Co] Co-NOTA-NT-20.3 over [ 68 Ga] Ga-NOTA-NT-20.3 and [ 64 Cu] Cu-NOTA-NT-20.3 for the targeting of NTSR1.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2023

TRANSPORTATION CONSIDERATIONS FOR USED FUEL AND ADVANCED NUCLEAR ENERGY TECHNOLOGIES

This study focuses on requirements and challenges pertaining to advanced transportable nuclear energy systems, which are currently in development and feature varying degrees of mobility. These reactors mostly use high-assay low-enriched uranium (HALEU) fuel, and only limited refueling is required during their lifecycles. Due to the smaller source terms, higher burnup fuel, and potentially higher thermal efficiency involved, the waste inventory will be less than that generated by conventional light-water reactors. However, the relevant regulatory requirements must be met when transporting HALEU fuel (whether fresh or used) inside the reactor vessel—mainly the fuel material transportation, siting, and physical protection requirements outlined in US Nuclear Regulatory Commission (NRC) regulations 10 CFR 71 and 73. Spent nuclear fuel (SNF), as well as irradiated fuel (used during any interval of reactor operation), contains highly radioactive materials that generate heat and necessitate special precautions during transportation. This study addresses the challenges that this incurs, and outlines potential solutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

TRANSPORTATION CONSIDERATIONS FOR USED FUEL AND ADVANCED NUCLEAR ENERGY TECHNOLOGIES

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Data-driven background model for the CUORE experiment

Here, we present the model we developed to reconstruct the CUORE radioactive background based on the analysis of an experimental exposure of 1038.4 kg yr. The data reconstruction relies on a simultaneous Bayesian fit applied to energy spectra over a broad energy range. The high granularity of the CUORE detector, together with the large exposure and extended stable operations, allow for an in-depth exploration of both spatial and time dependence of backgrounds. We achieve high sensitivity to both bulk and surface activities of the materials of the setup, detecting levels as low as 10 nBq kg −1 and 0.1 nBq cm −2 , respectively. We compare the contamination levels we extract from the background model with prior radio-assay data, which informs future background risk mitigation strategies. The results of this background model play a crucial role in constructing the background budget for the CUPID experiment as it will exploit the same CUORE infrastructure.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

NASA NTRS1991

A miniaturized fibrinolytic assay for plasminogen activators

This report describes a micro-clot lysis assay (MCLA) for evaluating fibrinolytic activity of plasminogen activators (PA). Fibrin clots were formed in wells of microtiter plates. Lysis of the clots by PA, indicated by change in turbidity (optical density, OD), was monitored with a microplate reader at five minutes intervals. Log-log plots of PA dilution versus endpoint, the time at which the OD value was halfway between the maximum and minimum value for each well, were linear over a broad range of PA concentrations (2-200 International units/ml). The MCLA is a modification and miniaturization of well established fibrinolytic methods. The significant practical advantages of the MCLA are that it is a simple, relatively sensitive, non-radioactive, quantitative, kinetic, fibrinolytic micro-technique which can be automated.

NASA Discipline Cell Biology↗

NASA NTRS1987

Gibberellins and gravitropism in maize shoots: endogenous gibberellin-like substances and movement and metabolism of [3H]Gibberellin A20

[3H]Gibberellin A20 (GA20) of high specific radioactivity (49.9 gigabecquerel per millimole) was applied equilaterally in a ring of microdrops to the internodal pulvinus of shoots of 3-week-old gravistimulated and vertical normal maize (Zea mays L.), and to a pleiogravitropic (prostrate) maize mutant, lazy (la). All plants converted the [3H]GA20 to [3H]GA1- and [3H]GA29-like metabolites as well as to several metabolites with the partitioning and chromatographic behavior of glucosyl conjugates of [3H]GA1, [3H]GA29, and [3H]GA8. The tentative identification of these putative [3H]GA glucosyl conjugates was further supported by the release of the free [3H]GA moiety after cleavage with cellulase. Within 12 hours of the [3H]GA20 feed, there was a significantly higher proportion of total radioactivity in lower than in upper halves of internode and leaf sheath pulvini in gravistimulated normal maize. Further, there was a significantly higher proportion of putative free GA metabolites of [3H]GA20, especially [3H]GA1, in the lower halves of normal maize relative to upper halves. The differential localization of the metabolites between upper and lower halves was not apparent in the pleiogravitropic mutant, la. Endogenous GA-like substances were also examined in gravistimulated maize shoots. Forty-eight hours after gravistimulation of 3-week-old maize seedlings, endogenous free GA-like substances in upper and lower leaf sheath and internode pulvini halves were extracted, chromatographed, and bioassayed using the "Tanginbozu" dwarf rice microdrop assay. Lower halves contained consistently higher total levels of GA-like activity. The qualitative elution profile of GA-like substances differed consistently, upper halves containing principally a GA20-like substance and lower halves containing principally a GA20-like substance and lower halves containing mainly GA1-like and GA19-like substances. Gibberellins A1 (10 nanograms per gram) and A20 (5 nanograms per gram) were identified from these lower leaf sheath pulvini by capillary gas chromatography-selected ion monitoring. Results from all of these experiments are consistent with a role for GAs in the differential shoot growth that follows gravitropism, although the results do not eliminate the possibility that the redistribution of GAs results from the gravitropic response.

NASA Discipline Plant Biology↗

DOE OSTI2020

Initial Evaluation of Microreactor Disposition Options

The United States Department of Energy is supporting the U.S advanced reactor industry through funding, legislation and regulatory development to actively pursue several microreactor design concepts. The Idaho National Laboratory (INL) is strategically positioned to support demonstration of microreactor technology in the next three to five years. This report provides and initial evaluation of the disposition options for microreactor spent nuclear fuel (SNF) generated as part of the microreactor technology demonstration program. Currently available information constitutes the basis for the options identified and discussions thereof. In the absence of detailed microreactor design information, assumptions were made to facilitate the identification of disposition options Particulars pertaining to any component of an identified disposition pathway are naturally highly microreactor design specific and in general such details are not provided. The diverse nature of potential microreactor SNF is reflected in the diverse nature of DOE owned SNF stored at INL. Therefore, it is anticipated that DOE currently stores and manages fuels that can serve as analogs for most microreactor fuel concepts. As such disposition options for microreactor SNF are expected to be much the same as that for these existing fuels. Two generic microreactor concepts have been selected for the purposes of this options assessment. The selected reactor concepts are a tristructural isotropic (TRISO) fueled high temperature gas reactor concept and a sodium/potassium bonded heat pipe reactor with uranium oxide fuel. Both concepts are assumed to be using high-assay low enriched uranium (HALEU) as the initial fuel composition. Interim storage, treatment and neutralization, material recovery, packaging and extended dry storage options are identified these reactors. The disposition options include existing INL facilities and capabilities and new facilities and capabilities developed as part of the microreactor program or as part of DOE's overall strategy for the eventual transfer of all SNF at INL to a permanent repository.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Validation of Spent Fuel Dose Rate Calculations Using NDA Measurement Campaign Data

The predicted radiation gamma dose rate is a key parameter to determine if a spent nuclear fuel assembly is self-protected from misuse as well as an indicator for worker/inspector safety. Recent studies showed that using state-of-the-art techniques for dose rate calculations, including the one adopted for this study, the numerical calculations of the gamma dose rates are three times lower than those from the most commonly-cited reference values for spent LWR fuel used by domestic and international agencies. With only a few publically available experimental data points worldwide for this type of benchmark study, the gamma flux measurements from the Spent Fuel Non Destructive Assay (NDA) measurement campaign can be used to help validate the physics approach used in calculating gamma dose rates. Specifically, this data would validate the gamma flux calculation from spent fuel, which is a key component of these multi-physics techniques for dose rate assessments; dose rate measurements were not part of the NDA campaign. These data are primarily the passive gamma measurements of the spent nuclear fuel assemblies from the Swedish Nuclear Fuel and Waste Management Company (SKB).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Specification of FIPD Fission Gas Chemistry Data

The current FIPD library contains two main sets of fission gas chemistry data. The first set is data collected during the Integral Fast Reactor (IFR) program from 1984 to 1994, using a gas mass spectrometry system located in the Analytical Laboratory (AL) at Argonne National Laboratory-West (Argonne-West). Throughout this period, numerous fission gas release and chemistry datasets were gathered from a variety of metallic fuel pins. The fission gas was sampled by the Gas Assay, Sample and Recharge (GASR) System in the Hot Fuel Examination Facility (HFEF) and transferred to the AL to perform gas composition and isotopic abundance analysis. The second set is data collected after the IFR program. The fission gas samples were also collected by the GASR system at the HFEF, but analyzed using a similar gas mass spectrometer located in Pacific Northwest National Laboratory (PNNL). Many fuel pins irradiated in Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF) were measured, including the fuel pins for the MFF series of experiments, designed to qualify metal fuel for use as driver fuel in the FFTF and X496 experiment. For either set of data, fission gas was sampled with the gas sampling line in GASR using sample bottles after the capsule/element volume has been determined and the system is still full of radioactive gas. The sample bottles were then transferred to the sample packaging cylinder or an approved storage location pending transfer to the AL or prepared for shipment to another laboratory (such as PNNL) for analysis of the collected gas as directed on the GASR data form, other approved form. The receiving laboratories (AL or PNNL) required their Analytical Service Request form to be completed prior to sample transfer. Typical sample transfer processes were initiated at HFEF by the principal or process engineer. The laboratories performing the analyses (AL or PNNL) use the sample bottle numbers as well as a sample number produced by the respective laboratory. HFEF and the responsible experimenter tracked the sample using the analysis number, the gas bottle number, and the fuel pin number.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2025

Results from the fifth galaxy serpent exercise

Galaxy Serpent is an ongoing series of virtual, web-based international tabletop exercises designed to advance the application of National Nuclear Forensics Libraries (NNFLs) in investigations involving nuclear and other radioactive material found out of regulatory control. Here, this iteration emphasized interactions between scientific teams and mock investigative entities. Participants utilized their provided NNFLs to assess material consistency with a provided database of holdings, assign confidence levels, and identify key characteristics relevant to investigative queries. The exercise highlighted both challenges encountered and lessons learned, and advanced best practices for integrating a NNFL into nuclear forensics as part of an investigation.

Database↗

DOE OSTI2023

Trade Study of Nondestructive Assay Technologies

The purpose of this report is to summarize the capabilities of currently existing NDA technologies that can be applied to holdup measurements. Cataloguing the strengths and weaknesses of the existing state of the art is a necessary step towards setting goals for new technology development. The NDA technologies evaluated in this report include gamma imaging, non-imaging gamma detectors, and neutron coincidence counting. These technologies were chosen based on their potential for technological improvement or to be applied in different settings. Each technology will be analyzed on a set of evaluation criteria chosen to represent the most salient features of a detector system for holdup measurements. These criteria include sensitivity, energy resolution, angular resolution (for imaging technologies only), ease of use, and cost.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Requalification and Declaration of High-alpha Cemented TRU-waste from the Late 70's and Early 80's - 20499

KTE and Siemens launched a project to qualify a series of 391 waste drums and 115 concrete containers produced from 1979 to 1984. The waste is characterized by a high content of long-lived alpha-nuclides (TRU-waste - Pu, U, Am in the range of 1 E+11-1 E+12 Bq/ drum. This high radionuclide inventory requires a special packaging according to the waste acceptance criteria (WAC) of the repository Konrad [1] despite the fact, that the waste was treated manually. The raw waste originally was delivered in nitric acid solution and was solidified using ordinary Portland cement. The existing documentation comprises analyses, transfer documentation for the nuclear fuel nuclides and waste treatment documentation yielding information, which waste was treated together in a batch. Additionally all drums were subjected to non-destructive gamma-spectrometry assay. A small number of 7 drums were opened and sampled to support the existing declaration. For the entire series the waste products were attributed to defined batches using the accountancy data supported by the analyses and the gamma-spectrometry. Each batch was produced from a homogeneous stock of waste solution and can therefore be declared and documented together. The declaration concept is based on the gamma spectrometry results performed on the entire drums, which were corrected and extended for the uncertainty of the analyses. The declaration concept is verified by the regulator and its independent expert. The state of the project is in the documentation of the solidification campaigns followed by the planning of the container configuration. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗