[en] Fault suppression capability (FSC) test with DC short downstream DC reactor of international thermonuclear experimental reactor (ITER) poloidal field (PF) converter is performed to verify the thermal stability characteristics of the converter bridge arms. In this paper, firstly, DC short downstream DC reactor of ITER PF converter is analyzed to get the fault current. Then, the appropriate test scheme is designed and the feasibility of the test scheme is verified by simulation based on test platforms in Institute of Plasma Physics, Chinese Academy of Science (ASIPP). Finally, the effectiveness of the designed test scheme results are verified by the test platforms in ASIPP.

[en] Objectives of the GIF-RDTF: - Identify essential large experimental infrastructure needed in support of GEN IV systems R&D activities in terms of feasibility/performance as well as demonstration/deployment; - Facilitate R&D collaboration across GEN IV systems; - Promote utilization of experimental facilities for collaborative R&D activities among GIF partners; - Facilitate GIF partners' access to the various R&D facilities in the GIF member countries.

[en] A design is described of a device for the suspension, transport and securing of control rods of experimental and training nuclear reactors. On the body of the control rod there is at least one claw and one recess on the support beam. A revolving key with a lever passes through the body of the control rod. On the face of the control rod there are suspension pins, a removable transport suspension with a hollow and arms of which one has a contour identical to that of the lever in the unblocked position. On the bottom part of the pin joint in the secured position of the control rod is the bush of the pin joint. The constant position of the control rod is secured by a quide in the fuel assembly and a notch in the claw in which engages the pin fitted in the beam. Technical drawings are included. The described device increases safety and shortens handling time with control rods during an experiment. (J.B.)

[en] A simplified design is described of a fuel assembly for experimental reactors and critical assemblies. The design eliminates the assembly head and transfers its function to the outer can. Thus, access is facilitated to the individual fuel elements while transportability of the assembly is maintained. The outer assembly can is extended by a part which houses bores for mounting and handling. Access to the bores is possible by adjusting the relative orientation of assemblies in the reactor core or by different length of the extended parts of the can. Direct acces to fuel elements makes it possible to plan experiments without transporting the assembly between the reactor and the mounting stand. Handling time in the mounting stand is reduced and the risk is removed of damage during assembly head mounting and dismantling. (Pu)

No abstract available

[en] This paper discusses salient aspects of severe-accident-related recriticality modeling and analysis in the Advanced Neutron Source (ANS) reactor. The development of an analytical capability using the KEN05A-SCALE system is described including evaluation of suitable nuclear cross-section sets to account for the effects of system geometry, mixture temperature, material dispersion and other thermal-hydraulic conditions. Benchmarking and validation efforts conducted with KEN05-SCALE and other neutronic codes against critical experiment data are described. Potential deviations and biases resulting from use of the 16-group Hansen-Roach library are shown. A comprehensive test matrix of calculations to evaluate the threat of a criticality event in the ANS is described. Strong dependencies on geometry, material constituents, and thermal-hydraulic conditions are described. The introduction of designed mitigative features are described

[en] The design is described of the vessels and components of a multi-pool experimental reactor working with a variable reactor level. The reactor is composed of separated vessels which are placed in shafts in the external shielding block whose upper surface forms the technological foundation. Horizontal openings and vertical channels are formed in the block walls. One of the vessels is fitted with experimental channels, fuel elements and control elements. The shielding bushings of the tangential channel with the shielding plug and of the radial channel with a shielding closure reach as far as the outer shielding block. A common shaft for the reactor vessel and experimental vessel is formed in this block. The radioactive fuel element storage room is in the bottom part of the experimental vessel. (Z.S.). 2 figs

[en] A possible approach to choosing the power level for an experimental reactor, based on correlation between the reactor power and the volume of the information obtained is discussed. Calculation results for the experimental reactor with the 1x10¹⁶ neutron/cm²xS flux density are presented to illustrate the approach proposed. The main conclusions are the following: with the reactor power increase in the range up to approximately 400 MW the total information volume increases in an approximate proportion to the cube of the power increase, and specific information and the rate of information obtained increases in an approximate proportion to the square of the power increase. With the power increase above approximately 400 MW the growth of the total information volume slows down. In a large reactor the same level of neutron flux density can be attained at significantly lower power intensity of the core, that facilitates the problem of fuel element construction for a high-flux reactor. The great power of the experimental reactor permits to use fuel of a relatively low inrichment. It means that a rather high fuel conversion coefficient (approximately 0.5) can be obtained that to a certain extent facilitates the problem solution of reactivity compensation in the experimental reactor

No abstract available

No abstract available