Tokamak cooling systems and power conversion system options

DEMO will be a fusion power plant demonstrating the integration into the grid architecture of an electric utility grid. The design of the power conversion chain is of particular importance, as it must adequately account for the specifics of nuclear fusion on the generation side and ensure compatibility with the electric utility grid at all times. One of the special challenges the foreseen pulsed operation, which affects the operation of the entire heat transport chain. This requires a time-dependant analysis of different concept design approaches to ensure proof of reliable operation and efficiency to obtain nuclear licensing. Several architectures of Balance of Plant were conceived and developed during the DEMO Pre-Concept Design Phase in order to suit needs and constraints of the in-vessel systems, with particular regard to the different blanket concepts. At this early design stage, emphasis was given to the achievement of robust solutions for all essential Balance of Plant systems, which have chiefly to ensure feasible and flexible operation modes during the main DEMO operating phases – Pulse, Dwell and ramp-up/down – and to adsorb and compensate for potential fusion power fluctuations during plasma flat-top. Although some criticalities, requiring further design improvements were identified, these preliminary assessments showed that the investigated cooling system architectures have the capability to restore nominal conditions after any of the abovementioned cases and that the overall availability could meet the DEMO top-level requirements. This paper describes the results of the studies on the tokamak coolant and Power Conversion System (PCS) options and critically highlights the aspects that require further work

OECD/NEA PKL-4 benchmark activity. Code assessment of the relevant phenomena associated to a blind IBLOCA experiment

International audienceCode assessment and validation is one of the most relevant research lines in thermal hydraulics and best estimatecodes. During the last decades, the Nuclear Energy Agency (NEA) and the Organization for Economic Cooperation and Development (OECD) have sponsored dozens of experimental projects in this field. Most ofthem were compiled in the CSNI Code Validation Matrix in 1996. Several projects have been promoted in thenew century as the SETH, PKL, PKL-2, PKL-3 and PKL-4 at the PKL test facility. In 2017 a benchmark activity waslaunched within the framework of the OECD/NEA PKL-4 project with the aim of assessing the capabilities ofsystem codes to reproduce the relevant phenomena associated to the IBLOCA scenario. 16 participant organizations from 9 different countries simulated the i2.2 (run 3) experiment in semi-blind conditions. A large varietyof system codes were used in the activity: ATHLET, CATHARE, KORSAR, LOCUST, RELAP5, RELAPSCDASIM,SPACE and TRACE. This paper presents the main outcomes for the code assessment of such codes. The first partdescribes the main features of the experiment and the selection of the key phenomena for code validation. Inaddition, the paper intoduces a detailed description of each phenomena and the comparison between theexperimental data and the blind simulations of the participants. Finally, in the last part of the paper the mainsources of uncertainty associated to the codes and the modelling are listed as well as the code assessmentconclusions of the benchmark activity. In general, the results obtained by all participants showed a good performance and satisfactory agreement with experimental data, which increases the confidence in current TH codetechnologies. The overall quality of the contributions was partly a consequence of the excellent documentationand information provided by the PKL team

OECD/NEA PKL-4 benchmark activity. Code assessment of the relevant phenomena associated to a blind IBLOCA experiment

International audienceCode assessment and validation is one of the most relevant research lines in thermal hydraulics and best estimatecodes. During the last decades, the Nuclear Energy Agency (NEA) and the Organization for Economic Cooperation and Development (OECD) have sponsored dozens of experimental projects in this field. Most ofthem were compiled in the CSNI Code Validation Matrix in 1996. Several projects have been promoted in thenew century as the SETH, PKL, PKL-2, PKL-3 and PKL-4 at the PKL test facility. In 2017 a benchmark activity waslaunched within the framework of the OECD/NEA PKL-4 project with the aim of assessing the capabilities ofsystem codes to reproduce the relevant phenomena associated to the IBLOCA scenario. 16 participant organizations from 9 different countries simulated the i2.2 (run 3) experiment in semi-blind conditions. A large varietyof system codes were used in the activity: ATHLET, CATHARE, KORSAR, LOCUST, RELAP5, RELAPSCDASIM,SPACE and TRACE. This paper presents the main outcomes for the code assessment of such codes. The first partdescribes the main features of the experiment and the selection of the key phenomena for code validation. Inaddition, the paper intoduces a detailed description of each phenomena and the comparison between theexperimental data and the blind simulations of the participants. Finally, in the last part of the paper the mainsources of uncertainty associated to the codes and the modelling are listed as well as the code assessmentconclusions of the benchmark activity. In general, the results obtained by all participants showed a good performance and satisfactory agreement with experimental data, which increases the confidence in current TH codetechnologies. The overall quality of the contributions was partly a consequence of the excellent documentationand information provided by the PKL team

OECD/NEA PKL-4 benchmark activity. Code assessment of the relevant phenomena associated to a blind IBLOCA experiment

International audienceCode assessment and validation is one of the most relevant research lines in thermal hydraulics and best estimatecodes. During the last decades, the Nuclear Energy Agency (NEA) and the Organization for Economic Cooperation and Development (OECD) have sponsored dozens of experimental projects in this field. Most ofthem were compiled in the CSNI Code Validation Matrix in 1996. Several projects have been promoted in thenew century as the SETH, PKL, PKL-2, PKL-3 and PKL-4 at the PKL test facility. In 2017 a benchmark activity waslaunched within the framework of the OECD/NEA PKL-4 project with the aim of assessing the capabilities ofsystem codes to reproduce the relevant phenomena associated to the IBLOCA scenario. 16 participant organizations from 9 different countries simulated the i2.2 (run 3) experiment in semi-blind conditions. A large varietyof system codes were used in the activity: ATHLET, CATHARE, KORSAR, LOCUST, RELAP5, RELAPSCDASIM,SPACE and TRACE. This paper presents the main outcomes for the code assessment of such codes. The first partdescribes the main features of the experiment and the selection of the key phenomena for code validation. Inaddition, the paper intoduces a detailed description of each phenomena and the comparison between theexperimental data and the blind simulations of the participants. Finally, in the last part of the paper the mainsources of uncertainty associated to the codes and the modelling are listed as well as the code assessmentconclusions of the benchmark activity. In general, the results obtained by all participants showed a good performance and satisfactory agreement with experimental data, which increases the confidence in current TH codetechnologies. The overall quality of the contributions was partly a consequence of the excellent documentationand information provided by the PKL team

Neutronics Benchmark of CEFR Start-Up Tests: temperature coefficient, sodium void worth, and swap reactivity

The China Institute of Atomic Energy (CIAE) proposed some of the China Experimental Fast Reactor (CEFR) neutronics start-up test data for the IAEA benchmark within the scope of the IAEA’s coordinated research activity. The coordinated research project (CRP) on “Neutronics Benchmark of CEFR Start-Up Tests” was launched in 2018. The benchmark aims to perform validation and verification (V&V) of the physical models and the neutronics simulation codes by comparing calculation results against collected experimental data. Twenty-nine participating research organizations finished performing independent blind calculations and refined their calculation results by referring to measurement data. The paper introduces the following three kinds of reactivity measurements in the CEFR start-up test and presents the results by participants: temperature coefficient, sodium void reactivity, and swap reactivity. First, for measuring temperature coefficients, ten sets of data were obtained by increasing and decreasing the temperature. The control rod position is changed for each temperature to maintain the reactor as critical. Second, sodium void reactivity is measured by replacing a fuel SA with vacuum-sealed SA and searching for the critical position of control rods. Third, for measuring the swap reactivity, fuel subassembly is replaced by stainless subassembly, and stainless subassembly is switched with one fuel subassembly. Swap reactivities are measured in two different ways, with more than two control rods moving to find the criticality of the core in the ‘Multiple Rods’ case and only one control rod moving in the ‘Single Rod’ case. All three reactivities are obtained by combining control rod worth for changed rod position and criticality difference. The comparison shows that uncertainty of calculations, modeling errors, and inaccurately determined control assembly worth make it challenging to calculate the temperature coefficient precisely. Meanwhile, the void worth and the swap reactivity results have similar trends and show good agreement with measurement

Maturation of critical technologies for the DEMO balance of plant systems

The Pre-Concept Design (PCD) of the Balance of Plant (BoP) systems of the EU-DEMO power plant is described in this paper for both breeding blanket (BB) concepts under assessment, namely the Water Cooled Lithium Lead (WCLL) BB and the Helium Cooled Pebble Bed (HCPB) BB. Moreover, the results of a preliminary evaluation of a number of BoP variants are discussed. This paper outlines the steps of the BoP design development, highlighting the project objectives and the strategy for their achievement under the very challenging requirements which include, among others, the intermittent nature of the DEMO plasma heat source. The main achievements during the PCD Phase will be reported together with the development plan for the Concept Design (CD) Phase to reach a mature (feasible) BoP concept for DEMO

Maturation of critical technologies for the DEMO balance of plant systems

The Pre-Concept Design (PCD) of the Balance of Plant (BoP) systems of the EU-DEMO power plant is described in this paper for both breeding blanket (BB) concepts under assessment, namely the Water Cooled Lithium Lead (WCLL) BB and the Helium Cooled Pebble Bed (HCPB) BB. Moreover, the results of a preliminary evaluation of a number of BoP variants are discussed. This paper outlines the steps of the BoP design development, highlighting the project objectives and the strategy for their achievement under the very challenging requirements which include, among others, the intermittent nature of the DEMO plasma heat source. The main achievements during the PCD Phase will be reported together with the development plan for the Concept Design (CD) Phase to reach a mature (feasible) BoP concept for DEMO.peer-reviewe

Verification and validation of neutronic codes using the start-up fuel load and criticality tests performed in the China Experimental Fast Reactor

Under the framework of coordinated research activities of the International Atomic Energy Agency (IAEA), the China Institute of Atomic Energy (CIAE) proposed a coordinated research project (CRP) to develop a benchmark based on the start-up tests of the China Experimental Fast Reactor (CEFR). 29 international organizations from 17 countries are participating in this CRP. Among the different physical start-up tests conducted in 2010 in the CEFR, the fuel loading and criticality experimental data is included. Before the start-up of the reactor, the core was preliminarily loaded with mock-up fuel sub-assemblies (SAs) in the active fuel positions. The reactor reached first criticality by replacing these mock-up SAs with real fuel SAs step by step. In a sub-critical extrapolation process, the number of fuel SAs to be loaded is determined by extrapolation of reciprocal of count rate and following safety requirements. As the reactor core approaches to criticality, the subcritical extrapolation ended and the next process is called super-critical extrapolation, which uses the control rods to reach criticality by period method. For the CEFR, the final clean-core criticality state was reached with 72 fuel SAs and the regulating control rod at the position of 70mm with a measured sodium temperature of 245°C. In the paper, the main results of the contributing international organizations for the fuel loading process in the blind and refined phase are summarized and compared with the experimental data. Additionally, code to code comparisons for normalized radial power are also presented. In general, results from all institutions show very good agreement while comparing with the experimental data. The results are divided in deterministic and stochastic codes and in each case, discussion and deep analysis is presented