atw 2015-01

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atw Vol. 60 (2015) | Issue 1 ı January 34 AMNT 2014 Paper presented at the RRFM 2014 facturing, construction, assembling and management. These phases will be carried out by national and international companies, and for these activities, a provision was made in the national budget, but not yet confirmed. Total project remaining time span is estimated in 5 years after contract signature and subject to availability of funds. 11. References | | [1] I. J. A. Perrotta, J. Obadia, “The RMB project development status”, on Proceedings of the 2011 International Conference on Research Reactors: Safe Management and Effective Utilization, held in Rabat, Morocco, 14-18 November 2011; International Atomic Energy Agency, Vienna, Austria (2012), available at: http:// www-pub.iaea.org/MTCD/Publications/ PDF/P1575_CD_web/datasets/abstracts/ C6Perrotta.html. | | [2] I. J. Obadia, J. A. Perrotta, “A sustainability analysis of the Brazilian Multipurpose Reactor Project”, on Transaction of 14 th International Topical Meeting on Research Reactor Fuel Management (RRFM-2010), held in Marrakesh, Morocco, 21-25 March 2010; European Nuclear Society, Brussels, Belgium (2010), ISBN 978-92-95064-10-2, available at: http://www.euronuclear.org/ meetings/rrfm2010/transactions/ RRFM2010-transactions-s6.pdf. | | [3] International Atomic Energy Agency, “Specific Considerations and Milestones for a Research Reactor Project”, Nuclear Energy Series NP-T-5.1, IAEA, Vienna, (2012), ISBN: 978–92–0–127610–0, Available at: - http://www-pub.iaea.org/MTCD/ publications/PDF/Pub1549_web.pdf. Authors José Augusto Perrotta and Adalberto Jose Soares Comissão Nacional de Energia Nuclear (CNEN) Avenida Prof. Lineu Prestes 2242 05508-000, Brazil 45 th Annual Meeting on Nuclear Technology: Key Topic | Reactor Operation, Safety – Report Part 3 The following reports summarise the presentations of the Technical Sessions “Reactor Operation, Safety: Radiation Protection”, “Competence, Innovation, Regulation: Fusion Technology” and “Competence, Innovation, Regulation: Education, Expert Knowledge, Knowledge Transfer” presented at the 45 th AMNT 2014, Frankfurt, 6 to 8 May 2014. The other Key Topics and Technical Sessions have been covered in previous issues of atw and will be covered in further issues of atw. Reactor Operation, Safety: Radiation Protection Angelika Bohnstedt Due to different circumstances the amount of presentations in the technical session “Radiation Protection” was at the Annual Meeting actually reduced to three lectures. But this gave the audience with about 23 to 27 participants the opportunity to have a lively discussion after each presentation, not only with the lecturer but also with other colleagues in the public. So the whole session was a fruitful exchange of interesting information and knowledge. The session was chaired by Dr. Angelika Bohnstedt, Karlsruhe Institute of Technology (KIT). The first presentation “Optimisation of Clearance Measurements According to DIN 25457 Taking Account of Type A and Type B Uncertainties” was hold by S. Thierfeld (co-author: S. Wörlen; both Brenk Systemplanung GmbH). In the beginning S. Thierfeld gave an overview of the DIN 25457, the widely applied standard for clearance measurements. He showed the evolvement from the fundamentals in 1993 via the Part 4 about contaminated and activated metal scrap, to the Part 6 of building rubbles and the latest Part 7 of the DIN about nuclear sites. And he emphasized that the primary aim is to get a reliable yes/no decision about the compliance with clearance levels. At the next step S. Thierfeld explained the incorporation of DIN ISO 11929, the standard for dealing with uncertainties in measurements, into DIN 25457. The consideration of Type A and Type B uncertainties for measurements and their calibrations was discussed. For different factors, influencing measurement and calibration, a conservative approach, taking only Type A uncertainties into account, and a realistic approach, combining Type A and Type B uncertainties, is possible. S. Thierfeld elucidated how to check step by step in the measurement and the calibration procedure which approach of uncertainty determination will be more reasonable for each respective factor. He concluded that finally a combination of all conservative and realistic approaches has to be done in a way to reach clearance measurements as precisely as necessary. At the end S. Thierfeld pointed out that the higher effort to reduce uncertainties will bring a decreased effort for decontamination work. The following presentation “Optimization of Handling Components and Large Scale Shielding Calculations with the Deterministic Code ATTILA” was given by S. Boehlke (co-author: M. Mielisch; both STEAG Energy Services GmbH), who started with the statement that in general shielding components are designed with conservative assumptions and boundary conditions which cover all possibly occurring situations. This can result in an overestimated shielding and the goal of an optimization procedure is to decrease on one hand the radiation level in accessible areas but on the other hand to decrease the amount of avoidable shielding material. S. Boehlke noted that for this optimization the calculation of the shielding geometry as well as the calculation of the dose rate distribution was done with the code ATTILA. He explained the different features of ATTILA, e.g. intuitive graphical user interface and the possibility to integrate simplified CAD geometries etc., and demonstrated in the following the use of ATTILA with 2 examples: a large scale dose rate mapping and the optimization of the shielding material of a handling machine for canisters of vitrified glass. For the large scale model (situation in a storage building) several aspects like superposition of all sources, the scattering of walls etc. and the scattering through openings was taken into account. As result S. Boehlke showed an overview about the shielding situation in the whole building. The second example was the calculation of the dose rate at the surface of a handling machine for canisters. Here S. Boehlke could demonstrate as consequence of the calculations a change in the design of the machine with the success that regions where the dose rate limit was exceeded before vanished and on the material site the reduction of used lead was about 30 % and the overall mass reduction of the machine was of about 10 %. AMNT 2014 Key Topic | Reactor Operation, Safety – Report Part 3
atw Vol. 60 (2015) | Issue 1 ı January In the third lecture “Shielding Factors of Newly Designed Ventilation Screws” H. Niegoth (co-authors: S. Boehlke, W. Stratmann and M. Bock; all STEAG Energy Services GmbH) presented the novel design of a shielding screw. He started with a short explanation that for a hot cell – to handle spent fuel and other sources of high radiation- there are two opposing requirements: on one hand sufficient ventilation is necessary but an effective radiation shielding has to be guaranteed on the other hand. Known construction solutions are Z-shaped channels, not able to penetrate a wall straight, and ventilation screws made of cast iron where the shielding is restricted to gammy radiation. Next H. Niegoth demonstrated a new approach for a screw design. This design bases upon a sandwich construction of several layers of winged discs where each individual element can vary in the numbers of wings, in the thickness and in the type of shielding material. H. Niegoth pointed out that because of different materials the shielding possibility is not restricted to gamma radiation but also neutron shielding can be achieved by using disc elements of polyethylene. He presented shielding calculations, performed with the ATTILA code, for screws with variations in the diameter, the length and the number of wings for each disc and in the number of turns per screw. In addition pressure loss coefficients have to be calculated for different screw parameters with CFD methods. At the end H. Niegoth summarized that the sandwich concept of the novel screw allows a flexible adaption of the shielding requirements with regard to the ventilation requirements for individual application cases. Competence, Innovation, Regulation: Fusion Technology Optimisation Steps in the ITER Design Thomas Mull ITER is the International Thermonuclear Experimental Reactor which is presently under construction at Cadarache in Southern France. This reactor is meant to demonstrate the feasibility of maintaining a burning fusion plasma – magnetically confined and producing a fusion power of 500 MW – in a steady state for several minutes and to investigate the possibility of breeding Tritium, which is needed for fuel, at a technical/industrial scale. The first presentation with the title “Effect of Diagnostic Apertures on Shut-Down Dose in ITER Upper Port Plug #18” was given by Arkady Serikov (Karlsruhe Institute of Technology, KIT). His co-authors were U. Fischer, B. Weinhorst (both KIT) and L. Bertalot, A. Suarez and S. Pak (all three: ITER Organisation). Shut-Down Dose Rates (SDDR) are a key condition for ITER maintainability. Dr. Serikov and his colleagues had focused on one of the port plugs, namely the Diagnostics Upper Port Plug (UPP) # 18. This port plug is hosting three diagnostic systems: vacuum ultra-violet (VUV) spectrometer, vertical neutron camera (VNC) and neutron activation system (NAS). The SDDRs are calculated in a multi-step process. The first step calculates the neutron and gamma flux in the plug volume during ITER operation, based on a prediction of the strength and spatial distribution of the power release of the ITER fusion plasma and shielding effects. The second step calculates the (space-resolved) activation of the materials in the plug, based on these radiation fluxes and assumptions for the preceding ITER operation time. A last step calculates the gamma dose rates due to these activated materials, based on a certain shutdown time before access (typically 106 s, i.e.: approx. 12 days). The parameters to be varied are choices for materials, presence or absence of filler materials and general geometry (where the real geometry is rather closely defined but for sake of simplicity there are calculations with a “homogenized” plug). Dr. Serikov and his colleagues found out that the SDDR directly behind the UPP under consideration could be reduced with respect to the current design by a factor of approx. 2 by proper choice of geometry and materials. Moreover they collected experience with radiation streaming effects inside the gaps which are surrounding the plugs inevitably and due to the apertures required by the spectrometers. This experience can be transferred to many other components. The following three contributions were all dealing with properties of tungsten which is foreseen as the plasma-facing material for the ITER divertor. The speakers (and main authors) of all these contributions were from Forschungszentrum Jülich GmbH (FZJ). Isabel Steudel presented her works on “Thermal Shock Behaviour of Tungsten Under Different Simulation Methods”. Her co-authors were A. Huber, J. Linke, G. Sergienko and M. Wirtz. Forged tungsten has a heterogeneous grain structure with grains flattened corresponding to the forging forces. I. Steudel had investigated the resistance of tungsten samples with different surface grain orientation to the impact of electron beams and Nd:YAG laser beams (with these beams simulating peak power impacts due to so-called edge-localized modes of the plasma, ELMs, which reach up to 1 GW/ m 2 and which are way higher than the “normal” irradiation due to a burning plasma which reaches “only” up to 20 MW/m 2 ). The experiments were starting from base temperatures of 20 °C and 400 °C, respectively. Mrs. Steudel explained that ELMs are unlikely to occur in operation states where the wall temperatures are below 200 °C. The results of these experiments show that there are different kinds of surface modifications and damages which are referable to two mechanisms. The transient heat loads and the related temperature increase leads to a local expansion that induces compressive stresses due to the cooler surrounding material. After the thermal shock event the material quickly cools down and compressive stresses are converted into tensile stresses. If the material is ductile these stresses can be compensated by plastic deformation. Otherwise cracks or crack networks occur. The damage behaviour strongly depends on the impacting power density and on the base temperature. A damage threshold was identified between 0.19 and 0.38 GW/m 2 for both base temperatures, for recrystallised material even between 0.38 and 0.76 GW/m 2 if starting at room temperature. An important conclusion from the experiments is that both simulation methods are capable to provide similar thermal loading conditions and that the damage patterns such as roughening and cracking are similar and show only minor deviations. Nathan Lemahieu reported on “Resistance of Tungsten with Yttrium Doping to ELM-Like Thermal Shocks”. His co-authors were J. Linke, G. Pintsuk, M. Wirtz (all three FZJ) as well as G. Van Oost (Ghent University, Belgium) and Z. Zhou (University of Science and Technology Beijing, China). The thermal shock resistance of spark plasma sintered tungsten grades, containing between 0.25 weight% and 1 weight% yttrium, was investigated under fusion relevant ELM-like loading conditions. The tungsten samples were cyclically tested at room temperature using a Nd:YAG laser beam and the electron beam facility JUDITH 1. Heat pulses with durations of 1 ms were applied to the samples leading to absorbed power densities between 0.37 GW/m 2 and 1.14 GW/m 2 . Furthermore, at a temperature of 400 °C, three samples were tested as well with the highest available power density of 1.14 GW/m 2 . The characterization of the samples before and after exposure shows that the thermal shock resistance improves with increasing yttrium content. However, in contrast to pure tungsten, for a base temperature of 400 °C there is still brittle behaviour for the yttrium-doped tungsten. These works clearly showed that tungsten-yttrium grades have benefits and should be considered as plasma facing materials (PFM). However, an observed rise in the ductile to brittle transition temperature (DBTT) could be a real draw-back. Therefore, future studies should include more tests at elevated base temperatures, higher than 400 °C to determine this DBTT and to exclude other drawbacks. A second approach to future research could include optimizing the yttrium content. 35 AMNT 2014 AMNT 2014 Key Topic | Reactor Operation, Safety – Report Part 3
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