Parallels between natural Oklo reaction zones and industrial reactors dynamics.

The first man-made nuclear reactor was developed by Fermi and collaborators at the University of Chicago and reached criticality in December 1942. This was the confirmation that men were able to use sustained fission reactions in order to produce energy. Following this success, nuclear reactors studies gave rise to several families of reactors corresponding to different orientations and technical choices. They are linked mainly to the choice of fuel (natural uranium, enriched uranium, plutonium, thorium), coolant (water, carbon dioxide, helium, sodium, ...) and moderator for slow neutron reactors (graphite, light water, heavy water). Out of all these choices, the pressurized water reactor (PWR) family is the closest to the Oklo natural reactors. Many intriguing similarities are observed and discussed in the present article. Our present-day understanding of the PWR operating conditions has been a great help for understanding the Oklo reactors.

Improved Measurement of the Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay.

Reactor neutrino experiments play a crucial role in advancing our knowledge of neutrinos. In this Letter, the evolution of the flux and spectrum as a function of the reactor isotopic content is reported in terms of the inverse-beta-decay yield at Daya Bay with 1958 days of data and improved systematic uncertainties. These measurements are compared with two signature model predictions: the Huber-Mueller model based on the conversion method and the SM2018 model based on the summation method. The measured average flux and spectrum, as well as the flux evolution with the ^{239}Pu isotopic fraction, are inconsistent with the predictions of the Huber-Mueller model. In contrast, the SM2018 model is shown to agree with the average flux and its evolution but fails to describe the energy spectrum. Altering the predicted inverse-beta-decay spectrum from ^{239}Pu fission does not improve the agreement with the measurement for either model. The models can be brought into better agreement with the measurements if either the predicted spectrum due to ^{235}U fission is changed or the predicted ^{235}U, ^{238}U, ^{239}Pu, and ^{241}Pu spectra are changed in equal measure.

Management and regulatory supervision of legacy sites and radioactively contaminated lands in Republic of Kazakhstan.

The safe management and disposal of radioactive waste (RW) arising from the nuclear legacy, as well as newly generated RW, are key problems. Their solution will have important implications for nuclear energy development, the introduction of other radiation technologies, and their public perception. In the framework of the cooperation between the Committee of Atomic and Energy Supervision and Control (CAESC) of the Ministry of Energy of the Republic of Kazakhstan and the Norwegian Radiation and Nuclear Safety Authority (DSA), work has been carried out to analyse the current state of nuclear and radiation safety in the Republic of Kazakhstan. The analysis was based on identifying gaps in national legislation and the assessment of corresponding threats in this area. Proposals for their elimination were developed, taking into account international experience and International Atomic Energy Agency recommendations. Analysis of the current situation in the Republic of Kazakhstan showed that at present the RWs are not properly regulated within an up-to-date regulatory framework. Currently, a list of key by-laws is being developed, which will support the provisions of a new law on RW management, and work is underway to adopt the already developed and drafted regulatory documents. Within the framework of the CAESC-DSA cooperation, the priority tasks established for 2021-2024 include the development of regulatory documents for the rehabilitation of uranium heritage sites, site selection for new nuclear facilities, and the management of nuclear materials for certain types of installations and manufactures. Practice has shown the need to use the advanced international experience and common approaches developed internationally, to develop and apply long-term and reliable solutions for the management of RW and nuclear legacy facilities and territories. The solution of these problems concerns not only scientists, technologists, and employers of the nuclear industry, but requires their cooperation with politicians, regulatory authorities, and the general population. The importance of sharing international experience to understand and solve these challenges is highlighted.

Determination of 90Sr activity concentration in black tea powder used as proficiency test sample by the modified liquid scintillation counting method.

Environmental monitoring is very important in nuclear facilities and its surroundings during the construction and operation. It requires many and different radioactivity measurements in order to radiation protection, radioactive waste management and environmental protection. An accurate and reliable determination of Strontium-90 in various environment matrices has been studied with many techniques such as co-precipitation, ion exchange, solvent extraction and extraction chromatography for environmental contamination monitoring purposes. In this study, the method used for dissolution and radiochemical separation to determine strontium-90 activity concentration in tea matrix used as proficiency test sample with liquid scintillation counter was investigated. As a result of these studies, the average mean activity concentration of strontium-90 in proficiency test material was determined as 155 ±â€¯16 Bq kg-1 dry matter (k = 2) by liquid scintillation efficiency tracing (CIEMAT/NIST) method for seven selected samples. This value was also confirmed by Joint Research Center-Geel with the CIEMAT/NIST method. The proficiency test results for consistency were checked by Dioxon's Q-test. Uncertainties arising from counting statistics, background, weighing, half-life, chemical recovery, efficiency and homogeneity were calculated.

Gene Expansion and Positive Selection as Bacterial Adaptations to Oligotrophic Conditions.

We examined the genomic adaptations of prevalent bacterial taxa in a highly nutrient- and ion-depleted freshwater environment located in the secondary cooling water system of a nuclear research reactor. Using genome-centric metagenomics, we found that none of the prevalent bacterial taxa were related to typical freshwater bacterial lineages. We also did not identify strong signatures of genome streamlining, which has been shown to be one of the ecoevolutionary forces shaping the genome characteristics of bacterial taxa in nutrient-depleted environments. Instead, focusing on the dominant taxon, a novel Ramlibacter sp. which we propose to name Ramlibacter aquaticus, we detected extensive positive selection on genes involved in phosphorus and carbon scavenging pathways. These genes were involved in the high-affinity phosphate uptake and storage into polyphosphate granules, metabolism of nitrogen-rich organic matter, and carbon/energy storage into polyhydroxyalkanoate. In parallel, comparative genomics revealed a high number of paralogs and an accessory genome significantly enriched in environmental sensing pathways (i.e., chemotaxis and motility), suggesting extensive gene expansions in R. aquaticus The type strain of R. aquaticus (LMG 30558T) displayed optimal growth kinetics and productivity at low nutrient concentrations, as well as substantial cell size plasticity. Our findings with R. aquaticus LMG 30558T demonstrate that positive selection and gene expansions may represent successful adaptive strategies to oligotrophic environments that preserve high growth rates and cellular productivity.IMPORTANCE By combining a genome-centric metagenomic approach with a culture-based approach, we investigated the genomic adaptations of prevalent populations in an engineered oligotrophic freshwater system. We found evidence for widespread positive selection on genes involved in phosphorus and carbon scavenging pathways and for gene expansions in motility and environmental sensing to be important genomic adaptations of the abundant taxon in this system. In addition, microscopic and flow cytometric analysis of the first freshwater representative of this population (Ramlibacter aquaticus LMG 30558T) demonstrated phenotypic plasticity, possibly due to the metabolic versatility granted by its larger genome, to be a strategy to cope with nutrient limitation. Our study clearly demonstrates the need for the use of a broad set of genomic tools combined with culture-based physiological characterization assays to investigate and validate genomic adaptations.

Experience of on-site disposal of production uranium-graphite nuclear reactor.

The paper reported the experience gained in the course of decommissioning EI-2 Production Uranium-Graphite Nuclear Reactor. EI-2 was a production Uranium-Graphite Nuclear Reactor located on the Production and Demonstration Center for Uranium-Graphite Reactors JSC (PDC UGR JSC) site of Seversk City, Tomsk Region, Russia. EI-2 commenced its operation in 1958, and was shut down on December 28, 1990, having operated for the period of 33 years all together. The extra pure grade graphite for the moderator, water for the coolant, and uranium metal for the fuel were used in the reactor. During the operation nitrogen gas was passed through the graphite stack of the reactor. In the process of decommissioning the PDC UGR JSC site the cavities in the reactor space were filled with clay-based materials. A specific composite barrier material based on clays and minerals of Siberian Region was developed for the purpose. Numerical modeling demonstrated the developed clay composite would make efficient geological barriers preventing release of radionuclides into the environment.

High-Dose TL Dosimetry of Reactor Neutrons.

Radiation dose was measured with set of TL dosimeters during checkout of neutron radiation hardness of the ORTEC preamplifier type 142A in the experiment at the MARIA nuclear reactor at the National Centre for Nuclear Research (NCBJ), Otwock-Swierk, Poland. Different types of LiF-based TL detectors have been used for measurements in order to evaluate neutron and non-neutron components of the radiation field in the reactor channel during exposure and to check their relevancy for dose measurements in the reactor environment. For high-dose evaluation a new Ultra-High-Temperature Ratio (UHTR) method established for highly sensitive LiF:Mg,Cu,P detectors has been applied. Neutron fluence evaluated from TL measurements was in good agreement with one calculated using neutron flux data during the experiment.

Transmutation of All German Transuranium under Nuclear Phase Out Conditions - Is This Feasible from Neutronic Point of View?

The German government has decided for the nuclear phase out, but a decision on a strategy for the management of the highly radioactive waste is not defined yet. Partitioning and Transmutation (P&T) could be considered as a technological option for the management of highly radioactive waste, therefore a wide study has been conducted. In the study group objectives for P&T and the boundary conditions of the phase out have been discussed. The fulfillment of the given objectives is analyzed from neutronics point of view using simulations of a molten salt reactor with fast neutron spectrum. It is shown that the efficient transmutation of all existing transuranium isotopes would be possible from neutronic point of view in a time frame of about 60 years. For this task three reactors of a mostly new technology would have to be developed and a twofold life cycle consisting of a transmuter operation and a deep burn phase would be required. A basic insight for the optimization of the time duration of the deep burn phase is given. Further on, a detailed balance of different isotopic inventories is given to allow a deeper understanding of the processes during transmutation in the molten salt fast reactor. The effect of modeling and simulation is investigated based on three different modeling strategies and two different code versions.

The role of the health physicist in nuclear security.

Health physics is a recognized safety function in the holistic context of the protection of workers, members of the public, and the environment against the hazardous effects of ionizing radiation, often generically designated as radiation protection. The role of the health physicist as protector dates back to the Manhattan Project. Nuclear security is the prevention and detection of, and response to, criminal or intentional unauthorized acts involving or directed at nuclear material, other radioactive material, associated facilities, or associated activities. Its importance has become more visible and pronounced in the post 9/11 environment, and it has a shared purpose with health physics in the context of protection of workers, members of the public, and the environment. However, the duties and responsibilities of the health physicist in the nuclear security domain are neither clearly defined nor recognized, while a fundamental understanding of nuclear phenomena in general, nuclear or other radioactive material specifically, and the potential hazards related to them is required for threat assessment, protection, and risk management. Furthermore, given the unique skills and attributes of professional health physicists, it is argued that the role of the health physicist should encompass all aspects of nuclear security, ranging from input in the development to implementation and execution of an efficient and effective nuclear security regime. As such, health physicists should transcend their current typical role as consultants in nuclear security issues and become fully integrated and recognized experts in the nuclear security domain and decision making process. Issues regarding the security clearances of health physics personnel and the possibility of insider threats must be addressed in the same manner as for other trusted individuals; however, the net gain from recognizing and integrating health physics expertise in all levels of a nuclear security regime far outweighs any negative aspects. In fact, it can be argued that health physics is essential in achieving an integrated approach toward nuclear safety, security, and safeguards.

Feasibility studies towards future self-sufficient supply of the (99)Mo-(99m)Tc isotopes with Japanese accelerators.

In order to establish a self-sufficient supply of (99m)Tc, we studied feasibilities to produce its parent nucleus, (99)Mo, using Japanese accelerators. The daughter nucleus, (99m)Tc, is indispensable for medical diagnosis. (99)Mo has so far been imported from abroad, which is separated from fission products generated in nuclear reactors using enriched (235)U fuel. We investigated (99m)Tc production possibilities based on the following three scenarios: (1) (99)Mo production by the (n, 2n) reaction by spallation neutrons at the J-PARC injector, LINAC; (2) (99)Mo production by the (p, pn) reaction at Ep = 50-80 MeV proton at the RCNP cyclotron; (3) (99m)Tc direct production with a 20 MeV proton beam from the PET cyclotron. Among these three scenarios, scenario (1) is for a scheme on a global scale, scenario (2) works in a local area, and both cases take a long time for negotiations. Scenario (3) is attractive because we can use nearly 50 PET cyclotrons in Japan for (99m)Tc production. We here consider both the advantages and disadvantages among the three scenarios by taking account of the Japanese accelerator situation.

Further evaluations of the toxicity of irradiated advanced heavy water reactor fuels.

The neutron economy and online refueling capability of heavy water moderated reactors enable them to use many different fuel types, such as low enriched uranium, plutonium mixed with uranium, or plutonium and/or U mixed with thorium, in addition to their traditional natural uranium fuel. However, the toxicity and radiological protection methods for fuels other than natural uranium are not well established. A previous paper by the current authors compared the composition and toxicity of irradiated natural uranium to that of three potential advanced heavy water fuels not containing plutonium, and this work uses the same method to compare irradiated natural uranium to three other fuels that do contain plutonium in their initial composition. All three of the new fuels are assumed to incorporate plutonium isotopes characteristic of those that would be recovered from light water reactor fuel via reprocessing. The first fuel investigated is a homogeneous thorium-plutonium fuel designed for a once-through fuel cycle without reprocessing. The second fuel is a heterogeneous thorium-plutonium-U bundle, with graded enrichments of U in different parts of a single fuel assembly. This fuel is assumed to be part of a recycling scenario in which U from previously irradiated fuel is recovered. The third fuel is one in which plutonium and Am are mixed with natural uranium. Each of these fuels, because of the presence of plutonium in the initial composition, is determined to be considerably more radiotoxic than is standard natural uranium. Canadian nuclear safety regulations require that techniques be available for the measurement of 1 mSv of committed effective dose after exposure to irradiated fuel. For natural uranium fuel, the isotope Pu is a significant contributor to the committed effective dose after exposure, and thermal ionization mass spectrometry is sensitive enough that the amount of Pu excreted in urine is sufficient to estimate internal doses, from all isotopes, as low as 1 mSv. In addition, if this method is extended so that Pu is also measured, then the combined amount of Pu and Pu is sufficiently high in the thorium-plutonium fuel that a committed effective dose of 1 mSv would be measurable. However, the fraction of Pu and Pu in the other two fuels is sufficiently low that a 1 mSv dose would remain below the detection limit using this technique. Thus new methods, such as fecal measurements of Pu (or other alpha emitters), will be required to measure exposure to these new fuels.

On the use of a molten salt fast reactor to apply an idealized transmutation scenario for the nuclear phase out.

In the view of transmutation of transuranium (TRU) elements, molten salt fast reactors (MSFRs) offer certain advantages compared to solid fuelled reactor types like sodium cooled fast reactors (SFRs). In the first part these advantages are discussed in comparison with the SFR technology, and the research challenges are analyzed. In the second part cycle studies for the MSFR are given for different configurations--a core with U-238 fertile, a fertile free core, and a core with Th-232 as fertile material. For all cases, the transmutation potential is determined and efficient transmutation performance for the case with thorium as a fertile material as well as for the fertile free case is demonstrated and the individual advantages are discussed. The time evolution of different important isotopes is analyzed. In the third part a strategy for the optimization of the transmutation efficiency is developed. The final aim is dictated by the phase out decision of the German government, which requests to put the focus on the determination of the maximal transmutation efficiency and on an as much as possible reduced leftover of transuranium elements at the end of the reactor life. This minimal leftover is achieved by a two step procedure of a first transmuter operation phase followed by a second deep burning phase. There the U-233, which is bred in the blanket of the core consisting of thorium containing salt, is used as feed. It is demonstrated, that transmutation rates up to more than 90% can be achieved for all transuranium isotopes, while the production of undesired high elements like californium is very limited. Additionally, the adaptations needed for the simulation of a MSFR, and the used tool HELIOS 1.10 is described.

Estimation of the respiratory tract burden resulting from a prolonged inhalation exposure to aerosols of DU, based on the U in a 24-h urine sample taken years after exposure.

A procedure is presented to estimate the respiratory tract burden from a prolonged inhalation exposure to particulate matter of depleted uranium, in cases where the rate of deposition is an unknown function. The precise range of possible values is identified. The calculations are based on the amount of depleted uranium measured in a single 24-h urine sample. In order to present an example, a simplified pharmacokinetical model is introduced. The results presented in this article are valid for any pharmacokinetical model represented by homogeneous linear differential equations with constant coefficients and non-zero initial values, and that clearly includes the International Commission on Radiological Protection model. In fact, they are applicable to any monitorable quantity measured over a short period of time, a monitorable quantity with a kinetic that can be described using a structurally similar system of differential equations to one describing these pharmacokinetical models.

First measurements of (236)U concentrations and (236)U/(239)Pu isotopic ratios in a Southern Hemisphere soil far from nuclear test or reactor sites.

The variation of the (236)U and (239)Pu concentrations as a function of depth has been studied in a soil profile at a site in the Southern Hemisphere well removed from nuclear weapon test sites. Total inventories of (236)U and (239)Pu as well as the (236)U/(239)Pu isotopic ratio were derived. For this investigation a soil core from an undisturbed forest area in the Herbert River catchment (17°30' - 19°S) which is located in north-eastern Queensland (Australia) was chosen. The chemical separation of U and Pu was carried out with a double column which has the advantage of the extraction of both elements from a relatively large soil sample (∼20 g) within a day. The samples were measured by Accelerator Mass Spectrometry using the 14UD pelletron accelerator at the Australian National University. The highest atom concentrations of both (236)U and (239)Pu were found at a depth of 2-3 cm. The (236)U/(239)Pu isotopic ratio in fallout at this site, as deduced from the ratio of the (236)U and (239)Pu inventories, is 0.085 ± 0.003 which is clearly lower than the Northern Hemisphere value of ∼0.2. The (236)U inventory of (8.4 ± 0.3) × 10(11) at/m(2) was more than an order of magnitude lower than values reported for the Northern Hemisphere. The (239)Pu activity concentrations are in excellent agreement with a previous study and the (239+240)Pu inventory was (13.85 ± 0.29) Bq/m(2). The weighted mean (240)Pu/(239)Pu isotopic ratio of 0.142 ± 0.005 is slightly lower than the value for global fallout, but our results are consistent with the average ratio of 0.173 ± 0.027 for the southern equatorial region (0-30°S).

Investigation on F/M material aspects of IRT-Sofia NCT channel.

The filter/moderator area of IRT-Sofia BNCT channel was investigated in this study in order to find a higher radiation resistant material as a suitable substitution for the Teflon(®). Two options - Al2O3 and graphite - were investigated. The results show, that both graphite and the Al2O3 can be successfully used as a filter/moderator material at IRT-Sofia. Initial evaluation of the in-phantom performance of the IRT-Sofia BNCT channel was made and merits similar to the best existing ones were found.

Extended set of activation monitors for NCT beam characterization and spectral conditions of the beam after reactor fuel conversion.

Since 2010 the LVR-15 reactor has been gradually converted from highly enriched fuel (36wt% (235)U) to low enriched fuel with the enrichment of 19.75wt% (235)U. Paper presents influence of the core pattern changes on the neutron characteristics of the epithermal beam. The determination of neutron spectrum free in the beam was done with a set of neutron activation monitors. After the reactor conversion the change in neutron spectrum is not provable as differences are in the range of measurement errors.

Toxicity of irradiated advanced heavy water reactor fuels.

The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels.

Uranium extraction from TRISO-coated fuel particles using supercritical CO2 containing tri-n-butyl phosphate.

High-temperature gas-cooled reactors (HTGRs) are advanced nuclear systems that will receive heavy use in the future. It is important to develop spent nuclear fuel reprocessing technologies for HTGR. A new method for recovering uranium from tristructural-isotropic (TRISO-) coated fuel particles with supercritical CO(2) containing tri-n-butyl phosphate (TBP) as a complexing agent was investigated. TRISO-coated fuel particles from HTGR fuel elements were first crushed to expose UO(2) pellet fuel kernels. The crushed TRISO-coated fuel particles were then treated under O(2) stream at 750°C, resulting in a mixture of U(3)O(8) powder and SiC shells. The conversion of U(3)O(8) into solid uranyl nitrate by its reaction with liquid N(2)O(4) in the presence of a small amount of water was carried out. Complete conversion was achieved after 60 min of reaction at 80°C, whereas the SiC shells were not converted by N(2)O(4). Uranyl nitrate in the converted mixture was extracted with supercritical CO(2) containing TBP. The cumulative extraction efficiency was above 98% after 20 min of online extraction at 50°C and 25 MPa, whereas the SiC shells were not extracted by TBP. The results suggest an attractive strategy for reprocessing spent nuclear fuel from HTGR to minimize the generation of secondary radioactive waste.

Preliminary radiological safety assessment for decommissioning of thoria dissolver of the ²³³U pilot plant, Trombay.

The thoria dissolver, used for separation of (233)U from reactor-irradiated thorium metal and thorium oxide rods, is no longer operational. It was decided to carry out assessment of the radiological status of the dissolver cell for planning of the future decommissioning/dismantling operations. The dissolver interiors are expected to be contaminated with the dissolution remains of irradiated thorium oxide rods in addition to some of the partially dissolved thoria pellets. Hence, (220)Rn, a daughter product of (228)Th is of major radiological concern. Airborne activity of thoron daughters (212)Pb (Th-B) and (212)Bi (Th-C) was estimated by air sampling followed by high-resolution gamma spectrometry of filter papers. By measuring the full-energy peaks counts in the energy windows of (212)Pb, (212)Bi and (208)Tl, concentrations of thoron progeny in the sampled air were estimated by applying the respective intrinsic peak efficiency factors and suitable correction factors for the equilibration effects of (212)Pb and (212)Bi in the filter paper during the delay between sampling and counting. Then the thoron working level (TWL) was evaluated using the International Commission on Radiological Protection (ICRP) methodology. Finally, the potential effective dose to the workers, due to inhalation of thoron and its progeny during dismantling operations was assessed by using dose conversion factors recommended by ICRP. Analysis of filter papers showed a maximum airborne thoron progeny concentration of 30 TWLs inside the dissolver.