Agency: European Commission | Branch: FP7 | Program: CP-FP | Phase: Fission-2011-2.3.1 | Award Amount: 3.06M | Year: 2012
In a nuclear power plant, a single metallic component may be fabricated from different materials. For example, RPV components are made from ferritic steel, whereas some of the connecting pipelines are fabricated from austenitic stainless steel. As a consequence different components often need to be connected by so-called dissimilar metal welds (DMW). Despite extensive research in Euratom Framework projects such as BIMET and ADIMEW, further work is needed to quantify the structural performance of ageing DMWs. The first purpose of this project shall be to gather relevant information from field experience: typical locations of DMWs in Western as well as Eastern LWRs will be identified and their characteristics considered, as well as applicable assessment methods. Micro-mechanical modelling of ductile failure processes will be used as an innovative technique to augment current numerical methods for structural integrity assessment of DMWs. The modelling will take account of ageing related phenomena and realistic stress distributions in the weld area and be supported by a comprehensive material test program. A procedure for measuring fracture toughness in DMWs will be developed. The work will also include an assessment of Leak Before Break (LBB) behaviour. Overall the project will serve to promote common understanding of structural integrity assessment of DMWs in existing and future NPPs of EU member states. This will be the technical basis towards the development of harmonised European codes and standards for multi-metal components, which is currently not available.
Alvarez Holston A.-M.,Studsvik Inc. |
Stjarnsater J.,Studsvik Inc.
Nuclear Engineering and Technology | Year: 2017
Delayed hydride cracking (DHC) was first observed in pressure tubes in Canadian CANDU reactors. In light water reactors, DHC was not observed until the late 1990s in high-burnup boiling water reactor (BWR) fuel cladding. In recent years, the focus on DHC has resurfaced in light of the increased interest in the cladding integrity during interim conditions. In principle, all spent fuel in the wet pools has sufficient hydrogen content for DHC to operate below 300°C. It is therefore of importance to establish the critical parameters for DHC to operate. This work studies the threshold stress intensity factor (KIH) to initiate DHC as a function of temperature in Zry-4 for temperatures between 227°C and 315°C. The experimental technique used in this study was the pin-loading testing technique. To determine the KIH, an unloading method was used where the load was successively reduced in a stepwise manner until no cracking was observed during 24 hours. The results showed that there was moderate temperature behavior at lower temperatures. Around 300°C, there was a sharp increase in KIH indicating the upper temperature limit for DHC. The value for KIH at 227°C was determined to be 2.6 ± 0.3 MPa √m. © 2017
Agency: European Commission | Branch: FP7 | Program: CP | Phase: Fission-2007-6.0.01;Fission-2007-1.1-01 | Award Amount: 12.29M | Year: 2008
Gas-Cooled Reactors (GCR), RBMK and some Material Test Reactors (MTR) make use of graphite as moderator of the fuel, structures of the core and/or thermal columns. During operation, the graphite and other carbonaceous materials like carbon brick, pyrocarbon and silicon carbide coatings are contaminated by fission products and neutron activation. These irradiated carbonaceous wastes are problematic due to their content of long-lived radioisotopes (e.g. Carbon14, Chlorine 36) and due to their large volumes. About 250000 t of i-carbon are existing, worldwide. Acceptable solutions have not yet been established to handle this kind of waste. This fact also represents a significant drawback for the market introduction of graphite-moderated reactors like Very/High-Temperature Reactors (V/HTR) as a promising Generation IV system candidate. Graphite moderated reactors represent the very first generation of nuclear reactors and therefore need to be decommissioned ahead of other reactor types which evolved later. Presently, accelerated decommissioning of GCR and RBMK and subsequent disposal of i-graphite is the preferred option for not leaving this waste as a legacy for future generations. The CARBOWASTE project aims at an integrated waste management approach for this kind of radioactive wastes which are mainly characterized as Intermediate Level Waste (ILW), due to the varying content of long-lived radioisotopes. Methodologies and databases will be developed for assessing different technology options like direct disposal in adopted waste containers, treatment & purification before disposal or even recycling i-carbonaceous material for reuse in the nuclear field. The feasibility of the associated processes will be experimentally investigated to deliver data for modeling the microstructure and localization of contaminants. This is of high importance to better understand the origin of the contamination and the release mechanisms during treatment and/or disposal.
Agency: European Commission | Branch: FP7 | Program: CP-FP | Phase: Fission-2011-1.1.1 | Award Amount: 4.74M | Year: 2012
The EURATOM FP7 Collaborative Project Fast / Instant Release of Safety Relevant Radionuclides from Spent Nuclear Fuel (CP FIRST-Nuclides) is established with the overall objective to provide for improved understanding of the fast / instantly released radionuclides from disposed high burn-up UO2 spent nuclear fuel. This issue is given a high priority in the SRA of the IGD-TP. The outcome of the project is relevant for all types of host rocks in Europe. European experimental facilities with specialised equipment for work with highly radioactive materials collaborate for improving the knowledge relevant for the period after loss of the disposed canister integrity. The project provides for experiments combined with modelling studies on integration of the different results as well as for up-scaling from experimental conditions to entire LWR fuel rods. Spent fuel materials are selected and characterized that have known initial enrichment, burn-up and irradiation histories. Experiments and modelling studies access the correlation between the fast release of fission gases and non-gaseous fission products. They also cover the chemical speciation of relevant fission/activation products and the retention of radionuclides in the rim and grain boundaries of the fuel. Complementary, existing data from previous investigations are evaluated. The 3 years project is implemented by a consortium with 10 Beneficiaries consisting of large Research Institutions and SMEs from 7 EURATOM Signatory States, and the EC Institute for Transuranium Elements. National Waste Management Organizations contribute to the project by participation in the End-User Group, by co-funding to Beneficiaries, and provide for knowledge and information.
Agency: European Commission | Branch: FP7 | Program: CP | Phase: Fission-2007-1.1-01 | Award Amount: 6.20M | Year: 2008
Main objectives of ReCosy are the sound understanding of redox phenomena controlling the long-term release/retention of radionuclides in nuclear waste disposal and providing tools to apply the results to Performance Assessment/Safety Case. Although redox is not a new geochemical problem, different questions are still not resolved and thus raised by implementers and scientists. From a top-down approach, the reliability of redox measurements for site characterization, redox disturbances by the near-field materials, changes induced by glaciation scenarios or the redox buffer capacity of host-rocks and the kinetics of response to redox perturbations are addressed. From a bottom-up approach, questions concerning the interpretation of mixed potentials, surface mediated reactions, redox states of actinides and long-lived fission products, the source term of spent nuclear fuel in the presence of corroding steel as well as the role of microbes and biofilms on the evolution of the redox state are tackled. Radionuclide redox transformations on minerals are decisive scenarios in the NEA FEP list and in the RETROCK project. In the large FP 6 IPs NF-PRO and FUNMIG, redox phenomena controlling the retention of radionuclides were addressed, although not systematically considered. The ReCosy concept is innovative in the scientific approach to the redox phenomena, including i) advanced analytical tools, ii) investigations of processes responsible for redox control iii) required data on redox controlling processes, and iv) response to disturbances in disposal systems. To this aim, the scientific-technical work program is structured along six RTD workpackages, covering near-field and far-field aspects as well as all relevant host-rocks considered in Europe. The 28 partners of ReCosy include the key European Research Institutes and Universities from 12 European countries, and Russia.
Cui D.,Studsvik Inc. |
Cui D.,University of Stockholm |
Low J.,Studsvik Inc. |
Spahiu K.,Swedish Nuclear Fuel and Waste Management Company
Energy and Environmental Science | Year: 2011
The world's first spent nuclear fuel repository concept (Swedish KBS-3) is illustrated and the results of experiments on environmental behaviors of spent fuel and canister materials under a potential canister breaching at early stage of disposal are reported. In a deoxygenated synthetic groundwater (2 mM NaHCO3) under radiation (γ 0.9 Gy h-1), inventory fraction leaching rates of fission-products (137Cs, 90Sr and 99Tc) and actinides (238U, 237Np) from a spent fuel segment were found to be around 10-6 and 10-7 per day, respectively. A cast-iron canister surface was found to be able to immobilize 238U, 90Sr, 99Tc and 237Np dissolved from spent fuel, but a copper surface could not. In the presence of the oxidative species generated from water radiolysis, the corrosion rates of waste canister materials, copper and cast-iron were found to be 1 and 30 μm per year, respectively. The observation of insignificant dissolution of spent fuel in the leaching solution equilibrated with 0.1 atm H2 is explained by the reducing effects of H2 in the presence of fission-product alloy particles (Mo-Tc-Ru-Rh-Pd) as catalysts and dissolved Fe(ii) in groundwater. The coating effect of ferric precipitates on spent nuclear fuel dissolution is also discussed. © 2011 The Royal Society of Chemistry.
Rudstam G.,Studsvik Inc. |
Grapengiesser B.,Studsvik Inc.
Radiochimica Acta | Year: 2013
Die Ablagerungstemperaturen der Elemente Na, Co, Zn, Ge, As, Br, Rb, Ag, Cd, In, Sn, Sb, Te, J, Cs und Ba in elementarer Form bzw. als chemische Verbindungen gind bestimmt worden. In den meiaten Fällen fand die Ablagerung auf Quartz statt; es wurden jedoch auch einige Daten für andere Materialien ermittelt. © 2013, by Akademische Verlagsgesellschaft
Bahadir T.,Studsvik Inc.
5th Topical Meeting on Advances in Nuclear Fuel Management, ANFM 2015: Advances in Nuclear Fuel Management V | Year: 2015
The recent improvements implemented in Studsvik's next generation code package CMS5, with CASM05 and SIMULATE5, in modeling the PWR radial baffle/reflector is presented in this work. The shortcomings in the conventional approach of generating radial homogenized cross-sections and discontinuity factors from a ID fuel/reflector transport calculation have been eliminated by re-computing the reflector node cross-sections and discontinuity factors in real core geometry by using the submesh calculation model in SIMULATE5. The submesh constants for the radial reflectors are generated from either a ID fuel/reflector transport calculation or a multi-assembly core transport calculation. The effects of radial reflector modeling on core eigenvalue and assembly power predictions are demonstrated for the BEAVRS benchmark problem.
Kropaczek D.J.,Studsvik Inc.
Progress in Nuclear Energy | Year: 2011
COPERNICUS is the Studsvik code for performing nuclear fuel optimization over a multi-cycle planning horizon that provides for an implicit coupling between traditionally separate in-core and out-of-core fuel management decisions. These decisions include determination of: fresh fuel region size; sub-region enrichments and bundle designs; exposed fuel re-use; and core loading pattern. The COPERNICUS methodology is based on a parallel implementation of the Simulated Annealing optimization algorithm, modified by the technique of Mixing of States, that allows for deployment in a processor scalable environment. COPERNICUS utilizes the 3-D licensing grade code SIMULATE for evaluation of all core loading pattern constraints, such as those involving reactivity and thermal margin requirements. Results are presented for a transition cycle design that compares performance of multi-cycle optimization to successive, single cycle optimization with regard to reducing levelized fuel costs. © 2011 Elsevier Ltd. All rights reserved.
Studsvik Inc. | Date: 2015-01-30
Treatment of radioactive waste comprising organic compounds, and sulfur-containing compounds and/or halogen-containing compounds. An apparatus comprises a reaction vessel comprising a filter for carrying out thermal treatment of the waste and a thermal oxidizer. Utilizing co-reactants to reduce gas phase sulfur and halogen from treatment of wastes.