bluuebunny/crossref_metadata_2025 huggingface dataset:
Keeping the Clearance Levels of NPP Building Structures Validation by Measurement of the Nuclide-Specific Dose-Rate
Authors: Hummel Lothar
Published: April 2020

Clearance levels for building structures normally are defined surface-specific (Bq/cm2) by the projection of the specific activity onto the surface. This proceeding led to the actual release criteria for buildings in the EC, which were also used for the actual draft of the new German RPO. The practical ability of collimated in-situ spectrometry verifying the keeping of these revised nuclide-specific clearance levels and the corresponding radiological averaging area of 1 m2 will be demonstrated.

A completely different approach performing clearance measurements in building structures with poor remaining artificial activity was sketched in an earlier work, by performing a set of in-situ measurements without collimation. For this purpose, the spectrometer is calibrated for the measurement of the gamma dose rate at a very low level. The direct measurement of this value and its comparison with the fundamental, dose-based release criteria is a large advantage of this procedure. The sensitivity of a commonly used high-purity germanium detector is sufficient for this purpose with short measuring time; or in other words the detection limits are much smaller than 1 nSv/h.

Therefore this presentation is dealing also with a complete new concept for the release of building structures of NPP’s. It is not supported by the actual release criteria basing on conversion factors from effective dose to surface-specific activity. The concept shown here, does not need this conversion, the effective dose is evaluated directly. On the other hand, it might be a basis of a fast, simple procedure in the projects of the next decades. The basic theoretical work is finished. In future, a practical check in an actual decommissioning project has to be done and the experiences gained by this work have to be documented.
The Waste Isolation Pilot Plant: A Global Opportunity for Partnerships With a Purpose
Authors: Triay Inés R. , Matthews Mark L. , Eriksson Leif G. , Hansen Frank D.
Published: April 2020

On March 26, 1999, the United States (U.S.) Department of Energy (DOE) Carlsbad Field Office (CBFO) opened the Waste Isolation Pilot Plant (WIPP) in New Mexico, United States of America (USA), for safe deep geological disposal of up to 175,584 cubic meters (m3) of long-lived radioactive wastes/materials (LLRMs). Twenty-four years of intensive, iterative interactions with scientific, environmental, public, institutional, political, and regulatory interest groups resolved all regulatory and legal challenges involved in bringing a deep geological repository for LLRMs to adequate scientific, technical, institutional, political, and public acceptance and fruition.

International strategic partnerships and research and development (R&D) collaborations are cornerstones in both past and current strategies designed to timely, cost-effectively, and safely accomplish the CBFO mission. The primary objectives of the CBFO’s international programs are to:

1. Acquire information supporting the CBFO mission.

2. Present and share CBFO mission information, expertise, and facilities of potential interest and/or value to other radioactive waste management and disposal programs, including using the WIPP underground research laboratory (URL) for joint R&D and training.
Incineration of Radioactive Waste From NPP in Slovakia: 15 Years of Experiences
Authors: Chnapko Peter , Pisca Michal , Urbanec Marian
Published: April 2020

Paper deal with incineration of radioactive waste (RW) in Slovakia, description of VUJE incineration and also cementation unit is given. Furthermore management of incineration process, cementation of ash and some operating limits and results are mentioned.
Belgian Co-Operation With the Czech Republic, Hungary, the Slovak Republic and Slovenia
Authors: Schittekat Jacques , Volckaert Geert , De Valkeneer Michel
Published: April 2020

Since mid nineties, the Belgian Government has granted funding to Belgatom and SCK•CEN to initiate collaborations with four Eastern European countries in the field of nuclear waste disposal safety.

The covered matters are essentially the disposal of nuclear waste and the interim storage of spent fuel. This was a good opportunity for Belgatom and SCK•CEN to share their extensive expertise in the fields of geological disposal, site selection, performance assessment and spent fuel interim storage.

In the Czech Republic, the mission deals with assistance in the bidding process for an interim dry spent fuel storage facility for fuel originating from the Dukovany site.

The matters covered in the Slovak Republic are the interim storage of spent fuel and the disposal of high level waste.

In Hungary, the co-operation addresses spent fuel management, low-level and high-level waste management.

In Slovenia the co-operation included provision of expertise concerning LILW management and collaboration in the field of geological disposal.

The co-operation is since 2001 extended to Russia, focusing on low-level waste management.
The Management of Radioactive Concrete Arising From the Dismantling of a Pressurised Water Reactor: The R&D Project on the Recycling of Radioactive Concrete
Authors: Klein Michel , Huybrechts Noël , Rousseau Edmond
Published: April 2020

Activated concrete represents the greatest volume of radioactive materials produced during the dismantling operations of a PWR reactor. For heavy barytes concrete, 133Ba is the dominating γ nuclide whereas in the rebars, it is the 60Co isotope.

During the dismantling of the BR3 PWR reactor, we studied various aspects of the radioactive concrete issue:

• the characterization of the activation depth and its modelization;

• the efficiency of various demolition techniques and their application on real scale mock-ups;

• the active testing and use of various dismantling and demolition techniques among which the remote controlled jack hammer and the diamond cutting techniques were the most extensively used.

As alternative to the conditioning of the radioactive concrete as radioactive waste using the classical cement embedding strategy, we started an extensive R&D programme on the the recycling of the radioactive concrete.

The basic idea is to perform a pretreatment of the radioactive concrete so that it can be re-used as aggregates for the fabrication of “radioactive grout”. This grout is then used for the conditioning of metallic radioactive waste. We demonstrated that it is technically feasible to prepare a good quality grout using heavy radioactive concrete as raw material mixed with fresh cement.
Safety and Uncertainty Analyses for the Shallow-Land Disposal of Radioactive Waste With Uranium Decay Chain
Authors: Takdea Seiji , Kanno Mitsuhiro , Minase Naofumi , Kimura Hideo
Published: April 2020

Safety and uncertainty analyses for the shallow-land disposal of radioactive wastes with uranium decay chain were performed using the deterministic and probabilistic safety assessment models. The deterministic analyses show that the dose calculation in residence scenario is of great importance owing to the influence of daughters built up by uranium decay chain. The parameter uncertainties for the important pathways in residence scenario are estimated from the probabilistic analyses using the statistical methodology. The uncertainty analysis indicates that the influence of parameter uncertainty is the most remarkable in the estimation for the inhalation of radon gas with residence.
Leaching Study of the Behaviour of Spent Fuel and SIMFUEL Under Simulated Granitic Repository Conditions
Authors: Serrano J. A. , Quiñones J. , Cobos J. , Diaz Arocas P. , Rondinella V. V. , Glatz J. P. , Matzke Hj , Martinez A. , Esteban J. A.
Published: April 2020

The leaching behaviour of spent fuel is of importance for the concept of direct storage of spent fuel. The aim of this study was to study UO2 irradiated fuel under simulated granitic repository conditions. In parallel with these spent fuel tests, SIMFUEL leaching studies were also performed. Direct comparisons between spent fuel and its chemical analogues, SIMFUEL, are often difficult. On one hand, because of the differences existing between spent fuel and SIMFUEL. E.g., for irradiated fuel: different origin and burnup, presence of intense radiation fields, hence radiolysis effects, or formation of cracks and pores due to the volatile fission products, hence larger surface area. On the other hand, because of different experimental procedures used by different authors.

This work presents results of sequential leaching experiments in synthetic granite water in equilibrium with a cylinder of granite at room temperature in air using spent UO2 fuel and SIMFUEL. The experimental conditions and procedure for irradiated and non-irradiated materials were kept similar as much as possible. The specimens used were UO2 (43 MWd/kgU) and SIMFUEL (simulating a burnup of 30 MWd/kgU) as non-irradiated chemical analogue.

A thermodynamic study by means of geochemistry codes was also performed. Differences both in fractional release and in uranium concentration in the leachate were found. The highest fractional release of uranium was measured for UO2 spent fuel. Candidate solid phases calculated for controlling the uranium solubility were soddyite ((UO2)2(SiO4)·2H2O) in the case of spent fuel and haiweete (Ca(UO2)2(Si2O5)3·5H2O) for SIMFUEL. Further work is ongoing to characterise the surfaces of the leached fuel samples and to try to confirm the preliminary attempts to identify reprecipitated secondary phases.

Comparison of some fission product release between spent fuel and SIMFUEL was also performed.
Development of Radioactive Waste Management and Environmental Remediation Activities in Belgium
Authors: Dejonghe Paul
Published: April 2020

As of the 1950s the nuclear programme in Belgium was developed along three lines:

• nuclear power generation,

• nuclear fuel cycle and nuclear fuel fabrication,

• production and application of radioisotopes.
Belgium Programmes Related to the Themes of ICEM: International Programmes Participation
Authors: Gaube M.
Published: April 2020

Through their participation in the Belgium’s nuclear programme, Belgian companies acquired a valuable experience in radioactive waste management, nuclear installations decontamination and nuclear installations decommissioning.
Dry Storage of Spent Research Reactor Fuel in Castor BR3® Casks at Belgoprocess in Belgium
Authors: Braeckeveldt Marnix , Ooms Luc , Geenen Gustaaf
Published: April 2020

The BR3 reactor (10.5 MWe) at the Nuclear Research Center SCK•CEN was the first PWR plant installed in Europe and has been shut down in 1987. The BR3 reactor is from 1989 in a decommissioning stage and most of the spent fuel is presently still stored in the deactivation pool of the BR3 plant and has to be evacuated. The BR3 was used as a test-reactor for new fuel types and assemblies (Mixed Oxide (MOX) fuel, fuel rods containing burnable poison (Gd2O3) and other types of fuels). Some fuel rods, having undergone a destructive analysis, are stored in different laboratories at the SCK•CEN. In total, the BR3 spent fuel comprises the equivalent of almost 200 fuel assemblies corresponding to some 5000 fuel rods.

Beside the spent BR3 fuel, a limited number of spent fuel rods, with equivalent characteristics as the BR3 fuel but irradiated in research reactors outside Belgium and stored in other buildings at the SCK•CEN nuclear site, were added to the inventory of spent fuel to be evacuated.

Various options such as reprocessing and intermediate storage awaiting final disposal were evaluated against criteria as available techniques, safety, waste production and overall costs. Finally the option of an AFR (away-from-reactor) intermediate dry storage of the BR3 and other spent fuel in seven CASTOR BR3® casks was adopted. As the SCK•CEN declared this spent fuel as radioactive waste, NIRAS/ONDRAF, the Belgian radioactive waste management agency became directly involved and the decision was taken to construct a small building at the Belgoprocess nuclear site for storing the CASTOR BR3® casks. Loading at the SCK•CEN followed by transport to Belgoprocess and storage is scheduled to take place at the end of 2001. The CASTOR BR3® cask weighing some 25 tonnes, consists of a monolithic body and has two independent lids with metal seals guaranteeing the long term leak-tightness of the cask. The CASTOR BR3® cask is designed for transport and the intermediate storage of at least 50 years.

Although a defect of the leaktightness of a CASTOR BR3® cask is very unlikely to occur, an intervention scenario had to be developed. As no pool is present at the Belgoprocess nuclear site to unload the fuel, an innovative procedure is developed that consists of transferring the basket, containing the spent fuel, into another CASTOR BR3® cask. This operation can be performed in the hot cell of the existing storage building for high level waste at the Belgoprocess site.
Modification and Testing of Advanced Hydraulic Cutting Tools for Use in an Arctic Environment
Authors: Stouky R. Jon , Griffith Andrew R. , Spargo Barry J. , Walsh Michael R. , Krumrine Paul H. , Butler Carey R.
Published: April 2020

An assessment was performed of Russian needs and capabilities related to decommissioning wastes from nuclear submarines. Hydraulically operated hand-held metal cutting tools can provide improved productivity and safety during the size and volume reduction of bulky metal scrap. Such tools are commercially available in the US, but not in Russia. Also, they have not been previously deployed in Arctic conditions. A system of metal cutting and spreading tools has been procured, modified and successfully tested at the Cold Regions Research and Engineering Laboratory (CRREL), and are now being deployed to Russia’s Northern Fleet.
Safety-Related Aspects due to the Chemotoxicity of Radioactive Waste to Be Disposed of
Authors: Brennecke Peter W.
Published: April 2020

According to the German disposal concept, all radioactive waste has to be emplaced in a repository constructed and operated in deep geological formations. To get a more detailed insight into site-specific safety-related issues, it is meaningful to perform additional investigations into the radiotoxicity and chemotoxicity. Thus, an evaluation including the hazard potential of organic and inorganic substances being the major constituents of waste packages to be disposed of becomes possible. In addition, possible releases via the water path must be investigated and assessed with regard to the safety of a repository during the post-closure phase, i.e. the radiological long-term effects (radionuclide-specific radiation exposures) as well as the possible pollution of near-surface groundwater by organic and inorganic substances. The latter will particularly take into account the principle of concern of section 34 “Reinhaltung” (Keeping Pure) of the Wasserhaushaltsgesetz (WHG – Water Resources Management Act).
Plasma Treatment of Problematic Waste
Authors: Deckers Jan , Vanbrabant Rik , Womack Ronald , Shuey Mark
Published: April 2020

Worldwide a great deal of the low and medium radioactive waste inventory is mixed with hazardous wastes and different non-combustibles. The path to treating these wastes historically has been to sort combustibles from non-combustibles and process them separately through incineration, supercompaction, cementation or other encapsulating technologies. Special attention has to be taken due to the presence of hazardous constituents. The cost and health physics exposure for sorting these types of mixed wastes and treating the separated streams in specialized infrastructure is not optimal and leaves a great potential for further optimization.

After several years of development, a commercially available high temperature treatment system has been developed and installed that treats heterogeneous low-level radioactive waste. High temperature plasma processing and unique torch design and operating features make it feasible to achieve a volume reduced, permanent, high integrity waste form while eliminating the personnel exposure and cost associated with sorting, characterizing and handling.

Plasma technology can also be used to recondition previous conditioned waste packages that don’t meet any longer the present acceptance criteria for final disposal. Plasma treatment can result in many cases in a substantial volume reduction, lowering the final disposal costs.

This paper covers the unique plasma centrifugal treatment principles and history. It also explains the roles of international partners that blend plasma, off gas treatment and nuclear expertise into one “best developed and available technology” (BDAT) for the treatment of problematic wastes.
Environmental Assessment Perspective of Decommissioning and Long-Term Management of Uranium Mine Tailings in Saskatchewan, Canada
Authors: Ross Malcolm K. , Hovdebo Don
Published: April 2020

Uranium was initially mined in Canada as a strategic mineral, primarily for export to the United States. Currently, uranium is produced for the global energy market and Saskatchewan is the sole producing province in Canada. Uranium development in Saskatchewan dates from 1953 and in 2000 accounted for 31% of global mine production.

In the 1990’s the Saskatchewan Government Environmental Assessment Branch reviewed a new generation of uranium mines with large reserves and extremely high average grades. Technically, the development of these mines has required the development of innovative technologies to manage the environmental and occupational health and safety issues associated with the mining of high-grade uranium ores. While the development of these innovative technologies posed a challenge to science and engineering, the potential environmental impacts and level of public concern associated with the development of the high grade uranium deposits equally challenged the ability of the Province of Saskatchewan’s environmental assessment process to evaluate the acceptability of the proposed mines.

During the assessment process a major technical, and public, issue was the decommissioning and long-term management of uranium tailings containing high levels of radionuclide and metal contaminants.

While technically decommissioning and reclamation are phases of mining that are considered at the end of mine life, scrutiny of these issues during the assessment process contributed significantly to the public and technical acceptability of the proposed mine developments.

The design, construction, operation, decommissioning and reclamation of uranium tailings management facilities for the proposed high-grade mines were subject to critical analysis during the technical and public review phases of the environmental assessment processes. Advances in tailings management design, incorporating innovative in-pit disposal methods capable of isolating decommissioned tailings from local groundwater regimes, presented a technical solution to concerns about long-term tailings containment after decommissioning. Public awareness and acceptance of the proposed mine developments was enhanced by the creation of an independent, public inquiry which ran concurrently with the existing federal and provincial assessment processes. The public inquiry was a critical factor, providing an independent forum where the technical acceptability of the proposed long-term tailings management methodologies was discussed. In retrospect the development of the new mines reflects the successful application of an assessment and review process in that the projects met the tests of technical and public acceptability in a process that was seen to be fair, timely, rigorous and public.
An Opportunity to Immobilize 1.6 MT or More of Weapons-Grade Plutonium at the Mayak and Krasnoyarsk-26 Sites
Authors: Jardine Leslie J. , Borisov Georg B. , Rovny Sergey I. , Kudinov Konstantin G. , Shvedov Alexander A.
Published: April 2020

The Mayak Production Association (PA Mayak), an industrial site in Russia, will be assigned multiple new plutonium disposition missions in order to implement the Agreement Between The Government Of The United States Of America And The Government Of Russian Federation Concerning The Management And Disposition Of Plutonium Designated As No Longer Required For Defense Purposes And Related Cooperation signed September 1, 2000, by Gore and Kasyanov, In addition, the mission of industrial-scale mixed-oxide (MOX) fabrication will be assigned to either the Mining Chemical Combine (MCC) industrial site at Krasnoyarsk-26 (K-26) or PA Mayak. Over the next decades, these new missions will generate radioactive wastes containing weapons-grade plutonium. The existing Mayak and K-26 onsite facilities and infrastructures cannot currently treat and immobilize these Pu-containing wastes for storage and disposal. However, the wastes generated under the Agreement must be properly immobilized, treated, and managed.

New waste treatment and immobilization missions at Mayak may include operating facilities for plutonium metal-to-oxide conversion processes, industrial-scale MOX fuel fabrication, BN-600 PAKET hybrid core MOX fuel fabrication, and a plutonium conversion demonstration process. The MCC K-26 site, if assigned the industrial-scale MOX fuel fabrication mission, would also need to add facilities to treat and immobilize the Pu-containing wastes.

This paper explores the approach and cost of treatment and immobilization facilities at both Mayak and K-26. The current work to date at Mayak and MCC K-26 indicates that the direct immobilization of 1.6 MT of weapons-grade plutonium is a viable and cost-effective alternative.
Radiochemical Characterization of Contaminants in Structural Materials of the AVR Reactor in Jülich
Authors: Lochny Michael , Bisplinghoff Bernhard
Published: April 2020

The complete decommissioning of the AVR reactor facility at Forschungszentrum Jülich GmbH (HTGR type) raised the question of the radioactive inventory built up during and since the operational phase. Therefore, a precise chemical and radiochemical analysis system was developed at the Institute for Safety Research and Reactor Technology at Forschungszentrum Jülich with the aim of identifying and quantifying 45 nuclides from both metallic and ceramic materials. Separation schemes and details of apparatus are given. Chromatograms are shown resulting from high-performance liquid chromatography which was applied for the fine separation of most β-emitters and of those nuclides which decay via electron capture.
Micro-Structural Studies of Leached SON-68-Type Glasses
Authors: Ollier N. , Panczer G. , Champagnon B. , Jollivet P.
Published: April 2020

Two types of borosilicate leached SON68-type glasses were studied, one doped with uranium and the other with rare-earth element (Nd, Eu).

Photoluminescence and cathodoluminescence properties of U doped samples have been correlated to microscopic features of the corroded glass.

Nuclear analysis, Electronic Microprobe and Scanning Electron Microscope investigations revealed the heterogeneous composition of the gels with differentiated phases. Enriched U phases (crystallised or not) and phosphorus precipitated phases in rare earth gel have been detected.
Molecular Dynamics Simulation of the Alpha-Recoil Nucleus Displacement Cascade in Zirconolite
Authors: Veiller L. , Crocombette JP. , Meis C. , Ghaleb D.
Published: April 2020

Zirconolite (CaZrTi2O7) has been proposed as a crystalline ceramic host for the long-term disposal of actinides extracted from high-level nuclear waste (e.g. France) and from excess weapons-grade plutonium (e.g. USA). The disintegration of radionuclides induces modifications of the crystalline structure. During α-decay of actinides, localized cascades of displaced atoms occur primarily because of ballistic collisions in the material from the emitted α-recoil nuclei. Under α-decay irradiation, zirconolite undergoes a crystalline to amorphous transformation, which is associated to a volume expansion. We have focused our study on the understanding of radiation-induced structural changes at the atomic level in this ceramic. Molecular Dynamics (MD) has been used in the simulation of displacement cascades in zirconolite. Original Buckingham pair potentials have been established for zirconolite to characterize the two body short-range interactions between different ionic pairs. We present the potential parameters fitted to the structural equilibrium properties of the crystal. This fitting reproduces the characteristics of the cell parameters of zirconolite within 4% and gives reasonable values for the bulk modulus and the specific heat. The MD method is applied to determine the threshold displacement energies for the various sublattices. Finally, we have modelled the effects of displacement cascades in zirconolite, due to the α-recoil nuclei. For that purpose, two MD simulations of high recoil kinetic energies (2 and 6 keV) were performed. The preliminary results show that the complex matrix zirconolite tends to a structural disordering for high PKA energy values although a partial recrystallization step is observed during the energy dissipation.
Modelling of the Waste Form Behaviour in a Wet Near-Surface Repository Site Over Extended Time Periods
Authors: Ojovan Michael I. , Ojóvan Natalia V. , Startceva Irene V. , Golubeva Zoja I. , Barinov Alexander S.
Published: April 2020

A mathematical model was used to predict radionuclide release from bitumen and glass waste forms over extended time periods. To calculate some model parameters, we used experimental data derived from 12yr field tests with six borosilicate waste glass blocks (each ∼30 kg in weight) and a bitumen block (310 kg), containing real intermediate-level NPP operational waste (NaNO3, 86 wt.% of a dry salt content; 137Cs, 82% of the radioactive inventory). Specific radioactivities of the glass material containing 35 wt.% waste oxides were βtot(90Sr+90Y), 3.74×106 Bq/kg, and αtot(239Pu), 1.3×104Bq/kg. The bitumen block with ∼31 wt.% salt content and βtot(90Sr+90Y), 4.0·106 Bq/kg, and αtot(239Pu), 3.0×103 Bq/kg was manufactured on base of a hard bitumen BN-IV. Tests with the waste forms were performed under saturated conditions of an experimental near-surface repository with a free access of groundwater to the waste blocks through a covering of host loamy soil and backfill of coarse sand. The way used to quantify the amount of leached radioactivity was to measure the volume and radioactivity concentrations of contacting groundwater.

In the model, radionuclide release from the waste glass is assumed to be controlled by the processes of diffusion limited ion exchange and glass network dissolution. The mechanism of radionuclide release from the bitumen matrix is believed to remain the same throughout the long-term storage period, except for the initial stage when an enhanced leaching from the surface layer occurs. This long-term release is assumed to be controlled by diffusion of radionuclides through the bitumen matrix. So, identical formulae were applied to calculate the values of leached radioactivity fractions for two waste forms.

Radioactivity release curves were plotted for field data and calculation results. For both waste forms, there was good agreement between the modelled and available experimental data. According to the modelling results, fmax = 2.3×10−3% of the initial radioactivity will release from the waste glass into the environment within a proposed institutional control period of 300 years under conditions of the near-surface repository and in the absence of additional engineered barriers. For the bitumen block and the same 300-yr period, the total (maximum) leached radioactivity fraction will be fmax = 4.2×10−3%.

The main result of the modelling and experimental studies concerning the leaching behaviour of the bituminised and vitrified waste materials is that the fractional radioactivity release for two waste forms is on the same order of magnitude. Numerical release values per a unit of a surface area to volume ratio are also rather close for two waste forms (exposed surface area to volume ratio for the bitumen block is 2 to 4 times greater then for the glass).
Post-Closure Removal of Long-Lived Radioactive Waste From a Deep Geological Repository in Bedded Salt
Authors: Matthews Mark L. , Eriksson Leif G.
Published: April 2020

In recent years, retrievability (and various permutations of this term) has emerged around the world as a means to achieve and enhance public acceptance of deep geological disposal of long-lived radioactive wastes/materials (LLRMs). In this debate, it is often erroneously suggested that post-closure retrievability of the emplaced LLRMs cannot be accomplished in salt.

In October 1996, the U.S. Department of Energy (DOE) submitted the Waste Isolation Pilot Plant (WIPP) Compliance Certification Application (CCA) to the U.S. Environmental Protection Agency (EPA) for review and approval. The CCA included a feasibility analysis defining a five-phased approach to post-closure waste removal from the WIPP rock salt repository based on currently available equipment and technologies. The feasibility analysis addressed highly adverse workers’ safety and waste retrieval conditions, including:

1. Radioactivity.

2. Hazardous constituents.

3. Gas.

4. Brine.

5. Rock integrity (instability).

The concluding statement in the CCA was that “In no case, however, are the conditions expected to render removal impossible”. In May 1998, the EPA announced that WIPP complied with all applicable radioactive waste management and disposal regulations. This announcement was preceded by intense EPA and public scrutiny and oversight, which included successfully overcoming two legal challenges. Hence, the global application of the WIPP waste-removal feasibility analysis is:

LLRM emplaced in a rock salt repository can be removed during the post-closure period with currently available technologies!
Retrievability: A False Positive?!?
Authors: Eriksson Leif G.
Published: April 2020

Retrieval (and various permutations of this term) of long-lived radioactive wastes/materials (LLRMs) from deep geological repositories is a “concept” currently believed to increase public confidence in, and acceptance of, national LLRM-repository programs. If the “retrievability concept” provides the missing link to increased public acceptance, which is imperative in a democratic society for the siting and development of a deep LLRM repository, then the understanding and expectations of the terms used must be clearly defined, and the associated benefits and shortcomings clearly explained, to the public. To quote the Danish mathematician and philosopher Piet Hein:

Knowing what you knowest not is in a sense omniscience.

Based on more than 20 years of involvement in the development of deep geological repositories for LLRMs in the United Sates of America and abroad, it is the preliminary conclusion of this author that retrievability conveys a false positive because it:

1. Does not add any appreciable post-closure physical safety to current and future generations and environments.

2. Could compromise (but does not necessarily have to) the long-term integrity of the disposal system.

3. Conveys the erroneous message that this generation does not have confidence in its ability to design, construct, decommission, and close a deep geological repository for LLRMs, in a manner that provides long-term containment and isolation of the emplaced LLRMs.
Start-Up of the ZWILAG Plasma Radwaste Treatment System
Authors: Wenger Jean-Pierre , Ineichen Robert , Vanbrabant Rik , Deckers Jan , Crouch James , Shuey Mark
Published: April 2020

ZWILAG (Zwischenlager Würenlingen AG, Switzerland) has installed a plasma system for the treatment and conditioning of low level radioactive wastes from nuclear power plants, industry, research and medical applications. In this plasma system radioactive wastes, composed off burnable and unburnable components, can be treated and conditioned in one single process without any pretreatment of waste drums. The glass product meets the final disposal requirements without any further conditioning. In this way the plasma system fulfills the general requirements for nuclear safety, minimal dose impact to the public and professional workers, industrial safety, low release limits for gaseous emissions, high volume reduction and disposable final waste forms.

This paper describes the start-up of the plant that was performed under the responsibility of ZWILAG together with the professional support of Belgoprocess, a consultant in radwaste processing, and of Retech Systems LLC, the supplier of the main components.

Some technical problems occurred during unit and integrated testing. The paper describes how these problems were solved and demonstrates the full operability of the system.
Decontamination of the Special Drainage System at NPP A-1
Authors: Rezbárik Ján , Majerský Dušan , Sekely Stanislav
Published: April 2020

The first Czecho-slovak nuclear power plant A-1 was shut down after an accident in 1977 and it is now under decommissioning. In spite of that, some technological systems in nuclear power plant must remain in operation. The planned service life of these systems has already been exceeded and their technical design is not suitable for present requirements. These operated technological systems must be repaired and reconstructed.

The special drainage system is one of the technological systems that must be in operation during the decommissioning process. After twenty-nine-year’-s of operation it became necessary to carry out the repairs, adaptations and replacements of some parts of this system with the aim to ensure more reliable operation and higher safety.

Radioactive sludge from the bottom of the collecting tank and drain pit was pumped out using an ejector and it was caught in drums. The fixed contamination on the inner surface of the collecting tank was removed by the chemical loop decontamination method using an external circulating decontamination device.

The surface of the drain pit was decontaminated with evenly spread decontamination gel that disrupted the fixed contamination. The contamination was then washed down using a water jet system with attached rotary brushes.

At NPP A-1, an immobilisation method for the treatment of radioactive sludge and solid particles based on the addition of silicates was successfully used by AllDeco for the first time. This sludge is often sticky with the high tendency to settle out the liquid and has a specific activity three orders of magnitude higher than are the limits for the Conditioning Centre (with high content of alpha radionuclides) and because it is not sound and effective to treat it by technologies generally used for another type of wastes. The immobilisation of the sludge was performed at room temperature. The resulting solid product has suitable properties for temporary storage. Procedures for the verification of the treatment of further sludge from NPP A-1 have been started.
Decontamination Method Data Base
Authors: Emets E. P. , Kolomeytsev G. Yu. , Polouektov P. P. , Shirokov V. V. , Yakovets A. N.
Published: April 2020

The Decontamination Method Data-Base (DMDb) is a versatile decision support information system. It is a computer based decision making tool which assists in the selection of best choice of metal surface decontamination technologies, procedures and agents. It is meant for educated and skilled experts in radiochemistry and for specialists in physics, chemistry, physical chemistry and other fields. The database has been developed to meet typical demands of users specialising in the decontamination of radiolagically contaminated materials and considers:

• detergent compositions for a specific contaminant;

• detergent compositions for a specific surface being decontaminated;

• detergent compositions for cleaning a specific surface of a specific contaminant;

• decontamination parameters in view of the composition of a detergent;

• corrosion effects in different conditions;

• etc.

In this connection, the major data entity of the database is a unit record describing a decontamination technology for a specific contaminated surface. The unit record contains information on the surface material, contamination, decontamination effectiveness, corrosion impacts and a literature source. The data presentation makes possible the realization of a record selection algorithm for a “lay” user to support the decontamination decision approaching the best one.

The analysis of the unit record subject and structure served as the basis for developing relational model “Tables – Relationships” and a tentative alternative of a user interface. Experimental records were entered to optimize users’ inquiries on the composition and corrosion effects of detergents under various conditions.
Ten Years’ Successful Operation of Nuclide Removal System in Loviisa NPP, Finland
Authors: Tusa Esko H. , Paavola Asko , Harjula Risto , Lehto Jukka
Published: April 2020

During the past ten years, selective removal of radionuclides from evaporator concentrates has been an essential part of waste management at the Loviisa NPP in Finland. Over 900 m3 of evaporator concentrates have been purified with 112 liters of highly selective ion exchange material. The volume reduction factor, comparing the volume of the original liquid to the volume of ion exchanger, is over 8,000. More importantly, comparing the volumes for final repository, the volume reduction factor for final repository is 1,550. This volume reduction, added to the avoidance of cementation of the original concentrates, results in very important economic savings.
Immobilization of Plutonium and Actinides in the Thorium Phosphate Diphosphate Ceramic
Authors: Dacheux N. , Thomas A. C. , Brandel V. , Genet M. , Le Coustumer P. , Podor R.
Published: April 2020

Thorium Phosphate-Diphosphate (TPD) can be considered as a potential matrix for the immobilization of actinides and especially for the plutonium coming from dismantled nuclear weapons. The capability of this solid to form solid solutions by substitution of thorium by tetravalent uranium, neptunium or plutonium is rather high. The synthesis of sintered pellets of thorium-uranium(IV) phosphate-diphosphate was successfully performed. Their density reached 95–99% of the value calculated from XRD data. The open and closed porosities were equal to 2–3% each. The resistance of the TPD and derivative solid solutions to aqueous corrosion are also very good. Leaching tests of powdered samples containing tetravalent plutonium or uranium showed that the normalized dissolution rate remained low (between 10−5 and 10−6 g/(m2.d)) even in acidic media. For tetravalent actinides, the saturation of the leachate seems to be controlled by the precipitation of the thorium phosphate-hydrogenphosphate hydrate which solubility product was found to be KS,0°* = 10−66.6 ± 1. This very low value allows a supplementary delay of the actinides migration in the hypothesis of their release in the leachate from the matrix.
Radiation Stability of Melted Titanate Waste Forms for Actinide Immobilization
Authors: Lian J. , Ewing R. C. , Yudintsev S. V. , Stefanovsky S. V.
Published: April 2020

Radiation stability of titanate ceramics suggested for actinide-bearing waste immobilization was studied. The major actinide (uranium) hosts in the samples prepared by melting followed by crystallization in a resistive furnace and in a cold crucible are phases with fluorite-related structure (zirconolite, pyrochlore, murataite) as well as brannerite. Critical amorphization doses at room temperature for these phases irradiated with 1 MeV Kr+ ions were (× 1018 ions/m2) ∼3, 1.8–2.4, 1.7–1.9, and 1.4 for zirconolite, pyrochlore, murataite, and brannerite, respectively. Murataite varieties with three-, five-, and eight-fold fluorite cells have the similar radiation stability. Recalculation of critical amorphization dose to dpa estimates a time required for murataite amorphization as 600–700 years at 10 wt.% 239Pu content. Similar values have been established previously for pyrochlore-structured titanates proposed for plutonium immobilization (because there will be also some commercial Pu issue).
Dissolution and Growth of Spinel Crystals in a High-Level Waste Glass
Authors: Hrma Pavel , Alton Jesse
Published: April 2020

A large fraction of HLW glass readily precipitates spinel. The presence of solid particles, including spinel crystals, is undesirable in a high-level waste (HLW) glass melter because the settling of solids can disrupt melter operation and shorten melter lifetime. Spinel formation in the melter can be reduced by lowering waste loading. When formulating HLW glass to maximize waste loading (thus minimizing the cost), the settling of insolubles must be considered.

The rate of nucleation, growth, and dissolution of spinel crystals in a molten HLW glass was measured as a function of temperature and the presence of nucleation agents, such as noble metals. The mass transfer coefficient was calculated using the Hixson-Crowell model and expressed as an Arrhenius function of temperature that was identical for both crystal growth and crystal dissolution. The Stokes law for hindered settling and the Hixson-Crowell equation for crystal growth and dissolution were chosen as a convenient representation of spinel behavior for mathematical models of HLW glass melters.

Levich’s analysis of the dissolution or growth of falling particles was used to estimate the effective diffusion coefficient (D) and the concentration-boundary-layer thickness (δ) around growing and dissolving crystals. The estimated values were in reasonable agreement with the measured concentration profile of Fe at a dissolving spinel crystal at 1200°C, determined with energy dispersive spectroscopy. The D value is comparable to that obtained by Borom and Pask for a sodium disilicate-magnetite couple. The calculated δ value agreed with the concentration distribution of Fe around falling crystals of spinel as imaged by optical microscopy.

This methodology, in conjunction with mathematical modeling, provides a basis for developing optimized technology and glass formulation for HLW vitrification.
The Effect of Temperature and Composition on Spinel Concentration and Crystal Size in High-Level Waste Glass
Authors: Mika Martin , Patek Milan , Maixner Jaroslav , Randakova Simona , Hrma Pavel
Published: April 2020

High-level radioactive wastes can be safely immobilized in alkali-aluminoborosilicate glass. To reduce the cost of the vitrification process, the waste loading should be maximized. This can be done by optimizing the process using mathematical modeling. The main objective of our work was to determine one of the necessary inputs for the mathematical model, which is the effect of temperature and composition on the concentration of spinel crystals and their size. We prepared six glasses with a different content of Li+, Na+, Mg2+, Ni2+, Cr3+, and SiIV and studied the effect of composition on the temperature dependence of spinel equilibrium concentration in glass by X-ray powder diffraction. The size of crystals was determined using optical microscopy. It was found that the temperature effect on spinel concentration significantly increased as the content of Ni2+ or Mg2+ in glass increased and slightly decreased as the content of Cr3+ increased and Li+ and Na+ content decreased. Both Ni2+ and Cr3+ acted as nucleating agents, producing a huge number of tiny spinel crystals (∼2 μm). In particular, Ni2+ seems to very significantly facilitate spinel crystallization.
Radionuclide Identification and Quantification in the Waste Streams of the Belgian Nuclear Research Centre SCK•CEN
Authors: Bruggeman M. , Cornelissen R. , Maris P.
Published: April 2020

The radiological characterisation of nuclear wastes of a research centre is difficult seen the many different processes that generate waste. Since these wastes may contain radionuclides relevant for the disposal option, the nuclide content and activity have to be known. Considering the fact that some wastes are generated only in minor quantities, complex approaches, involving sampling and successive analysis are generally not justified. Basic physical models can generally be applied to estimate activity ratios from which the radionuclide inventory can be determined by non-destructive assay on waste packages. This article discusses waste streams at the Belgian Nuclear Research Centre SCK•CEN and explains how nuclide inventories and activity are determined. The physical models, used to derive activity ratios and other simple approaches are discussed.
Vitrification decoupling from α-relaxation in a metallic glass
Authors: Monnier Xavier , Cangialosi Daniele , Ruta Beatrice , Busch Ralf , Gallino Isabella
Published: April 2020

For a metallic glass, vitrification kinetics takes place at temperatures lower than expected from the α-relaxation.
A molten carbonate shell modified perovskite redox catalyst for anaerobic oxidative dehydrogenation of ethane
Authors: Gao Yunfei , Wang Xijun , Liu Junchen , Huang Chuande , Zhao Kun , Zhao Zengli , Wang Xiaodong , Li Fanxing
Published: April 2020

Molten carbonate leads to a 10-fold ethylene yield increase by facilitating oxygen transport while blocking nonselective sites.
Transcriptome Analysis Reveals IsiA-Regulatory Mechanisms Underlying Iron Depletion and Oxidative-Stress Acclimation in
Synechocystis
sp. Strain PCC 6803
Authors: Cheng Yarui , Zhang Tianyuan , Wang Li , Chen Wenli
Published: April 2020

This study analyzed the impact of
isiA
deletion on the transcriptomic profile of
Synechocystis
. The
isiA
gene encodes an iron stress-induced chlorophyll-binding protein, which is significantly induced under iron starvation. The deletion of
isiA
affects the expression of various genes that are involved in photosynthesis and ABC transporters. WGCNA revealed three functional modules in which the blue module was correlated with oxidative stress. We further demonstrated that the
isi
operon contained the following five genes:
isiA
,
isiB
,
isiC
,
ssl0461
, and
dfp
by cotranscriptional PCR. Three sRNAs were identified that were related to oxidative stress. This study enhances our knowledge of IsiA-regulatory mechanisms under iron deficiency and oxidative stress.
Covert Cross-Feeding Revealed by Genome-Wide Analysis of Fitness Determinants in a Synthetic Bacterial Mutualism
Authors: LaSarre Breah , Deutschbauer Adam M. , Love Crystal E. , McKinlay James B.
Published: April 2020

Microbial communities impact life on Earth in profound ways, including driving global nutrient cycles and influencing human health and disease. These community functions depend on the interactions that resident microbes have with the environment and each other. Thus, identifying genes that influence these interactions will aid the management of natural communities and the use of microbial consortia as biotechnology. Here, we identified genes that influenced
Escherichia coli
fitness during cooperative growth with a mutualistic partner,
Rhodopseudomonas palustris
. Although this mutualism centers on the bidirectional exchange of essential carbon and nitrogen,
E. coli
fitness was positively and negatively affected by genes involved in diverse cellular processes. Furthermore, we discovered an unexpected purine cross-feeding interaction. These results contribute knowledge on the genetic foundation of a microbial cross-feeding interaction and highlight that unanticipated interactions can occur even within engineered microbial communities.
The Gut Microbiota of Laying Hens and Its Manipulation with Prebiotics and Probiotics To Enhance Gut Health and Food Safety
Authors: Khan Samiullah , Moore Robert J. , Stanley Dragana , Chousalkar Kapil K.
Published: April 2020

The microbiota plays a vital role in maintaining gut health and influences the overall performance of chickens. Most gut microbiota-related studies have been performed in broilers, which have different microbial communities compared to those of layers. The normal gut microbiota of laying chickens is dominated by
Proteobacteria
,
Firmicutes
,
Bacteroidetes
,
Fusobacteria
, and
Actinobacteria
at the phylum level. The composition of the gut microbiota changes with chicken age, genotype, and production system.