No abstract available

No abstract available

No abstract available

[en] Description is given of a device for trapping the solid fission products carried by the coolant of a high temperature nuclear reactor, driven through the core, then through the reactor reflector through channels. This device is characterized in that it comprises stacks of balls or cylinders of an adsorbent substances, mounted in housings provided in the reflector. This device can adsorb 99% of the fission products carried by the coolant, without running the risk of re-cycling these products should be a depressurization occur

[en] Korea has developed a Helium Cooled Ceramic Reflector (HCCR) TBM to be tested in the ITER. It consists of two major loops, which are HCS (Helium Cooling System) and TES (Tritium Extraction System) (figure 1). Tritium is one of the most highly permeable molecule on earth, therefore tritium permeation takes place from TES to HCS in the TBM. Permeated tritium migrates along the system pipes and tritium release in port cell and port interspace area is one of important safety issues for human access. Tritium is radioactive molecule which is 12.3 years of half-life. It is difficult to find good property reference for the HCCR TBS and other fusion application. This paper summarizes degree of conservatism/margin of estimated tritium releases for HCCR-TBS. Degree of conservatism/margin of estimated tritium releases for HCCR-TBS has been assessed with respect to possible items. TES and CPS performance increase methods give big impact on entire HCCR TBS design, therefore, it is just possible design option at this level of detail. In this regard, 20 ~ 2000 of conservatism/margin could be suggested in total. Tritium permeation issue in fusion industry is critical and has huge uncertainty to get realistic release rate. It is the fact that permeability is highly depended on operational temperature and surface condition. Practically it is expected that tritium amount in human access area for fusion reactor is negligible or manageable, however, dynamic and multi-dimensional tritium release analysis could be a solution for detail estimation.

[en] Higher availability and reduced safety system requirements were seen as potential advantages of an MHTGR. An evaluation of both steam cycle/cogeneration and process heat MHTGRs was initiated in 1983. A key interim milestone for 1993 is the co-operative evaluation of the direct and indirect cycle version of the MHTGR-GT. The results of the studies will that the MHTGR will play a prominent role in the US nuclear power program. (DG)

[en] Helium-cooled Ceramic Breeder (HCCB) is one of the candidate Indian breeding blanket concepts for its DEMO. The design of manifolds is an important part in a breeding blanket design since it determines how helium is getting distributed in to different circuits for cooling different zones and components in blanket module. The flow scheme for HCCB blanket has been optimised based on its flow parameters. The inlet helium coolant of 300 C comes out at 360 C from FW and is then distributed into breeding unit zones, grid plates, top plate and bottom plate in parallel configuration. Hydraulic analysis of different circuits using ANSYS CFX has been performed to estimate the flow distribution from manifolds, velocities and pressure drop in different circuits. Heat transfer coefficients have also been evaluated using the obtained velocities in different cooling channels. The detailed design and analyses of the flow schemes in different circuits of blanket module and the purge helium flow have been discussed in this report. The thermo-mechanical design and analysis at using ANSYS mechanical has also been performed at design and operating conditions for p type and s type damages including fatigue and discussed in this report. (author)

[en] One of the main engineering performance goals of ITER is to test and validate the design concepts of the tritium breeding blankets relevant to a power producing reactor. The tests will focus on modules including a demonstration of the breeding capability that will lead to a tritium self sufficiency and extraction of heat suitable for an electricity generation. Korea has developed a Helium Cooled Molten Lithium (HCML) Test Blanket Module (TBM) and Helium Cooled Solid Breeder (HCSB) TBM to be tested in the ITER. Recently, solid type HCSB TBM was decided as a leading concept in National Fusion Committee and the other is developing as the breeding blanket for DEMO. The name of the solid type TBM was changed to a Helium Cooled Ceramic Reflector (HCCR) considering the unique concept of using the graphite reflector. In the present study, the overall design and its performance analysis were introduced according to the main components such as First Wall (FW), Breeding Zone (BZ), and Side Wall (SW)

[en] Variable speed AC electric drive are more suitable for totally submerged gas filtration in helium cooled nuclear reactors. A typical example is in the new design of the modular helium gas cooled nuclear reactors (MHR), in which the main gas circulator is a large induction motor rated between 5 MW and 6 MW of shaft power, and requires a coolant which is free of contamination. The filtration system proposed in this contribution comprises of a squirrel cage induction motor having a fabricated rotor construction and rated at 10 kW and running on variable speed up to 24,000 r/min. Variable speed is essential for optimisation of drive efficiency size and outout power. The drive performance is presented and a new method of estimating pump speed without using a shaft feedback element is suggested. (Author)

[en] The design study of tritium producing blanket systems has been carried out since the middle 1970's. The outlines of several designs in this study are introduced, and the design philosophies, the design bases, the design concepts, and the issues are discussed in this paper. These are discussed in the following 3 stages: Stage 1 contains the designs related to INTOR-Phase 0. The design of Li²O-breeder-He-coolant-pot type blanket is mainly discussed. Breeder in the pot vessel is directly cooled by high temperature He for utilizing generated heat and in-situ tritium recovery. Stage 2 contains the designs related to INTOR-Phase 1, which are based on the specifications standardized at INTOR workshop. The design of Li²O-breeder-water-coolant-tube-in-shell type blanket is mainly discussed. The blanket concept is Breeder Outside of Tube/Coolant Inside Tube. Advanced type blankets are discussed in Stage 3. The present design study of these blankets includes the improvements for the issues pointed out in Stage 1 and Stage 2. As an advanced design of He-cooled type blanket tube-in-shell type is studied for the simplicity of structure design and temperature control. This concept is Breeder and Coolant Inside Tube with moderator. The feasibility of He-cooled blanket is discussed and clarified here. The advanced design concept of water-cooled type blanket which performance is non-sensitive to operating conditions is also described. (author)