توضیحات طراحی و عملکرد تجزیه و تحلیل از سیستم پوشش آزمون HCPB اروپا در شاهکار ITER

The helium cooled pebble bed (HCPB) blanket is one of the two European DEMO blanket concepts proposed for testing in international thermonuclear experimental reactor (ITER). The purpose of the tests is to validate the design principles and the operational feasibility for the demonstration blanket system. This includes the basic support functions like tritium extraction, helium cooling and heat transport, and helium purification. In addition, the basic properties and operating characteristics of the system's materials will be validated. Safety, reliability, maintenance and dismantling will be equally addressed. To assess these qualities, the ITER horizontal ports will be used to provide a relevant fusion plasma and the appropriate nuclear environment. At conclusion of the ITER feat EDA phase (July 2001), a revised design of the HCPB test blanket system have been completed to adapt the previous design (for ITER FDR) to the new operational conditions of ITER and to a new strategy for the blanket testing in this reactor. Design description, performance and safety analyses are presented in this paper.

This paper presents an overview on the work performed on the design and performance analyses of the helium cooled pebble bed (HCPB) test blanket system (TBS) during the international thermonuclear experimental reactor (ITER) feat EDA phase in the years 2000 and 2001. A first design of a HCPB TBS was completed in 1998 [1] and [2] for the previous version of ITER (ITER FDR). With the modification of the basic machine [3] was necessary to revise the previous design (DDD-98); the work performed by our team in collaboration with the ITER test blanket working group became immediately more than an adaptation to the new boundary conditions of ITER feat. The new operation parameters of the fusion reactor required a basic revision of the objectives and of the test strategy of the ITER blanket test programme [4]. This revision dictated the most important change introduced in the new design (DDD-01); the HCPB blanket testing programme for ITER-feat is now based on four test objects—the so called test blanket modules (TBMs)—that should be irradiated in the ITER equatorial port. Each TBM is devoted to a family of objectives; an electromagnetic (EM) TBM for measurements of eddy currents and electromagnetic forces in normal operation and during disruption should operate in the first 3 years during the H plasma. The neutronics and tritium production (NT) TBM should be irradiated in the next 3 years (D plasma and low-duty D–T plasma) with the aim to check the tritium breeding capability and to validate neutronic codes and data. The investigation of pebble bed behaviour, thermal creep, cyclic power operation, purged gas flow conditions will be carried out with a third module, namely the thermomechanics (TM) TBM (2 years of tests, low and high-duty D-T plasma). The plant integration (PI) TBM (2 years of test in high-duty D-T plasma) will be dedicated to Integration tests on thermo-mechanics, thermo-hydraulics, tritium and activation products handling and cooling system operation. An upgraded design description document (Status December 2001) of the HCPB TBM which will include all the work (test strategy, conceptual design, performance analyses including safety) performed up to end 2001, is under preparation [5].

The revision of the HCPB TBM design on respect to the operating conditions of ITER has been successfully completed. A new test strategy based on four TBMs has been proposed to reach the objectives of the blanket-testing programme. Consequently, the conceptual design of the testing system (TBMs and related auxiliary systems) has been revised and performance analyses have been carried out to determine the operating parameter of the system. A revised safety analysis shows that the new system is compatible to the safety requirement of ITER.