[en] IAEA Activities on NEA (non-electric applications): Support deployment of demonstration plant(s): - Establish an info-exchange forum (TMs, Workshops, CRPs,…); - Publications (Technical Reports, Journal Papers, Newsletter..); - Provide tools: DEEP, DE-TOP, HEEP, Toolkit; - Address issues of global concern: → Prospects of current nuclear reactors for NEA; → Prospects for cogeneration (including hybrid technologies); → Enhance Viability of NEA; → Establish close cooperation/collaboration on NEA.

[en] This report documents bounding functional and operating requirements (F and ORs)for the Next Generation Nuclear Plant (NGNP) Project to support selection of the nuclear system design and specification of the operating conditions and configuration of NGNP once the nuclear system design is selected. These requirements supplement the detailed F and ORs for NGNP developed in the FY07 NGNP Pre-conceptual design work

[en] An action plan presented here is the plan toward construction of a Demo reactor in Japan. A Task-force for development strategy of Tokamak Demo Reactors was established and has considered an action plan to construct the Demo in the 2040s. The action plan will lead works in a Joint Special Design Team for Fusion Demo established for design and R&D of the Demo reactor in Japan. Although the first version of action plan was reported in March 2016, a revised version, presented here, is being drafted in order to correspond to the new schedule of the ITER project agreed in 2016, because the ITER progress is one of critical issues in the action plan.

[en] The pulsed thermonuclear demonstration reactor (DEMO) features challenging operational conditions such as high neutron fluxes, high temperatures, and significant thermo-mechanical stresses. These conditions do not require only a selection of advanced structural materials, but also the development of reliable means to assemble the in-vessel components together; allowing thermal expansions, disassembly, and maintenance in attractive scenarios. Over the course of DEMO lifetime, the materials are subjected to embrittlement by neutron irradiation, swelling, considerable thermo-mechanical fatigue and creep. Traditional joining methods may be rarely used in the harsh fusion environment to assemble different components. In addition any proposed layout should cope with the limited space available inside the vacuum vessel (VV). The objective of this study is to review the proposed attachment systems (developed within the latest European DEMO Conceptual Study) for the vertical segmentation concept called 'multi module segments' (MMS). In order to find some place to house the attachments the blanket is cut respecting the Tritium Breeding Ratio limit for tritium self sufficiency. The conditions, neutronic and thermal, in which the attachments are supposed to operate, are calculated. The effects of pulsed operations have also been taken into account. The design of the attachments with the available structural materials with and without an active cooling system is analysed and a new concept for plug/unplug attachments is also suggested.

[en] Pre-operational tests of the Monju fast breeder reactor (FBR) by the Power Reactor and Nuclear Fuel Development Corporation (PNC) are now under way in Japan at the Siraki site located on the Tsuruga Peninsula of Fukui Prefecture, about 400 km west of Tokyo. Monju is a 280 MWe loop-type power reactor using plutonium-uranium mixed oxide as the fuel and liquid sodium as the coolant. The Monju plant incorporates the results of numerous research and development (R and D) activities, as well as the experience gained from the design, construction and operation of the experimental fast reactor Joyo at the Oarai Engineering Center (OEC) of PNC). The start of pre-operational tests follows the completion of construction on schedule in April 1991. The Monju reactor achieved first criticality on 5 April 1994, and nuclear heating tests were in progress during July 1995. (author)

[en] A fusion plant designed to demonstrate electricity production, known as DEMO, is likely to be very different from existing fusion devices and even from ITER. These differences will be found not only in the technology employed but also in physics areas including, but not confined to, the plasma behaviour. Here, some of the areas where DEMO is expected to be different are highlighted and explored. As a result of the differences between DEMO and existing devices, the fusion community needs to develop a new intuition when considering the design and performance of DEMO.

[en] Full text of publication follows. The HTR-PM demonstration plant is supported by the Chinese government within the framework of National Science and Technology Major Projects. INET is the designer of NSSS and the technology developer. Chinergy Co. Ltd is the contractor for EPC. HSNPC is the owner company for the demonstration plant, which is the joint venture between China Huaneng Group, China Nuclear Engineering Construction Group. HTR-PM project covers the design of the demonstration plant, the construction of the plant, operation of the plant, developing of key components as the first of a kind, test of the key components and system in full scale, development of the TRISO fuel element fabrication technology and fabrication plant, and research on new technology, and etc. The basic design of HTR-PM demonstration plant was finished in 2008, PSAR review was finished in 2009, and the construction started in 2012 because of the delay due to Fukushima accident. Now the construction goes well as well as the manufacturing of the components. The test of the components in full scale keeps on well, the performance of the fuel samples under irradiation behaves well, the installation of the components will be started soon. The HTR-PM demonstration plant is scheduled to be connected to the grid in the end of 2017. After the HTR-PM demonstration plant is completed, the commercial deployment is expected, based on the standardization of the HTR-PM design. (author)

[en] ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) is an integrated technology demonstrator designed to demonstrate the operability of the innovative choices enabling fast neutron reactor technology to meet the Generation IV criteria. ASTRID is a sodium-cooled fast reactor with an electricity generating power of 600 MWe. In order to meet the generation IV goals, ASTRID will incorporate the following decisive innovations: -) an improved core with a very low, even negative void coefficient; -) the possible installation of additional safety devices in the core. For example, passive anti-reactivity insertion devices are explored; -) more core instrumentation; -) an energy conversion system with modular steam generators, to limit the effects of a possible sodium-water reaction, or sodium-nitrogen exchangers; -) considerable thermal inertia combined with natural convection to deal with decay heat; -)elimination of major sodium fires by bunkerization and/or inert atmosphere in the premises; -) to take into account off-site hazards (earthquake, airplane crash,...) right from the design stage; -) a complete rethink of the reactor architecture in order to limit the risk of proliferation. ASTRID will also include systems for reducing the length of refueling outages and increasing the burn-up and the duration of the cycle. In-service inspection, maintenance and repair are also taken into account right from the start of the project. The ASTRID prototype should be operational by about 2023. (A.C.)

[en] Highlights: A few of the key highlights covered in this paper include: • The FNSF maintenance approach establishes guidance for future fusion plants. • The horizontal sector maintenance approach is preferred to the vertical approach. • Autonomous robotic maintenance will be a mainstream technology when FNSF is built. • Power Core maintenance times are estimated to yield an acceptable plant availability. • Hot Cell requirements and functionality are conceptualized for a fusion power plant. - Abstract: This paper addresses an approach to maintain and sustain the high-intensity Fusion Nuclear Science Facility (FNSF) experimental fusion facility that would enable it to have extended operations exceeding several exchanges of the primary power core components. The maintenance approach must be safe, quick, reliable, repeatable and precise. Moreover, this approach and equipment should be able to be extrapolated to the future fusion demonstration power plant (DEMO). A preliminary evaluation is presented for the underlying maintenance and safety requirements and the key design approaches that will define the FNSF and shape future high-power fusion facilities.

[en] A dissolver suitable for the dissolution of sheared nuclear fuel in which the fuel pieces undergoing dissolution are moved through the apparatus by pneumatic pulses has been developed by BNFL in conjunction with the All Russian State Scientific Institute for Inorganic Materials. The rate of transport of material through the dissolver is dependent upon amongst other things its mass and therefore separation of leached hulls from those containing fuel is achieved. The development of the dissolver is reviewed briefly and results relating to the transport of both simulated fuel pin and larger fuel assembly fragments are presented. (authors)