[en] Full text: In this paper, total decay heat values generated in a proto-type fast reactor are estimated. These values are compared with those of certain fast reactors. Simple analytical fits are also obtained for these values which can serve as a handy and convenient tool in engineering design studies. These decay heat values taken as their ratio to the nominal operating power are, in general, applicable to any typical plutonium based fast reactor and are useful inputs to the design of decay-heat removal systems

[en] This paper describes an arrangement for removal of decay power of a reactor core

[en] MONJU has three decay heat removal systems. The intermediate heat exchanger of the decay heat removal system is incorporated within the main IHX shell, and the heat from the secondary system is rejected to the air. Forced circulation is adopted for both primary and secondary coolant, though natural circulation capability is designed into the plant itself. Feasibility of rejecting the decay heat through steam plant is also being studied. In this paper, MONJU's decay heat removal system design, operational procedures, and the considerations behind the concept will be presented. (author)

[en] The upward heat removal capability in the 300 MW(e) GCFR demonstration plant following a hypothetical core meltdown accident has been investigated. It was assumed that no forced helium circulation is available, and a delay of 30 minutes has been assumed before the loop isolation valves in the Core Auxiliary Cooling System (CACS) are opened. Following opening of the isolation valves, a natural convection helium flow is established which provides heat transport from the melt surface to the water cooled CACS heat exchangers. The analysis showed that no melting of internal components would occur for system pressures greater than 6 atm and that no melting in the radial blanket assemblies would occur for system pressures greater than 14 atm

[en] With all security requirements of the reactor operator met and the proper function of the apparatus demonstrated in a test run, the operating permission was granted. With dummy samples the time constant of the calorimeter was determined to be 2 s. The time constant of the adiabatic jacket was found to be 1 s, the accuracy of its temperature control being +- 2 x 10^-3 K. The sensitivity of the calorimeter reaches 10^-4 W compared with expected powers of 200 - 50 m W. (orig./RW)

[en] Since fusion power and neutron fluence of the compact ITER has been reduced, thermal transients due to decay heat of the tokamak components such as in-vessel components and vacuum vessel (VV) were numerically analyzed for the purpose of identifying the necessity of cooling function as the safety measure. The result shows that the maximum VV temperature remains around 500 deg. C even under the extremely hypothetical conditions, assuming that all the coolants in the VV and in-vessel components are lost instantaneously. In addition, the maximum temperature appears after about 100 days and hence reducing the temperature rise can be practically achieved during such a long grace period. As a whole, it has been clarified that the decay heat removal can be passively achieved by only radiation without any active cooling measures. This paper describes the analysis results on thermal transient due to decay heat, including sensitivity study on the effect of heat connection and removal characteristics on the temperature rise

[en] This paper presents the recent analyses of Phénix End of Life tests performed with the most advanced modelling developments, and a global assessment of the campaign. The following items are specifically touched upon: core decay heat removal measurement and associated uncertainties, natural convection in reactor vessel using coupled system and CFD models, and neutronic/mechanical behaviour of the core during flowering test. Finally, it appears that the Phénix decay heat measurement test can be added to the validation basis of DARWIN code associated to JEFF3.1.1. The discrepancies with experimental data are weak (under +/- 8%) and globally covered by uncertainties. Concerning the primary natural convection test, the use of a CFD/system coupled calculation allows to model very accurately the sodium temperatures during the transient. The behaviour of the core during static flowering test can be well estimated by using HARMONIE code for mechanical aspects and TRIPOLI for neutronics. A new test is scheduled in Phénix to assess the core mechanics code in case of dynamic load. (author)

[en] The device is designed to remove the afterheat of a e.g. gas-cooled reactor if the steam turbine of the steam power plant including a water-steam separating vessel is shut down. For this purpose already existing components are used to a large extent. E.g. a heat exchanger, which may be cooled externally, is connected to the outlet for the condensed water of a start-up flash tank of the plant. It is also possible to use the heat exchanger as a preheater, whose connections for the turbine steam may selectively be connected to coolant pipes. (DG)

[en] The paper summarizes experimental data of all power losing in the HFETR, and compares the experimental results with the calculated values (the transient flow and temperature distribution in the fuel subassembly after all power losing). The good agreement between the experimental results and the calculated values (with a departure each other of less than 4%) shows that the HFETR is in safety during all power losing, and that theoretical analysis done by the authors is correct

[en] This specification establishes the requirements to design the SAFETY WATER TESTS to be constructed in the Hydraulic Test Facility (HTF) at the GE San Jose site. The test is an 1/8th scale model of a large loop type breeder reactor or a 1/14th scale model of a large pool type breeder reactor and uses water as the test fluid. It simulates a breeder reactor system with a 0.5 MW heated core with an upper and a lower plenum, a primary loop with 300 gpm flow rate and four auxiliary cooling systems (DRACS) that are to be immersed in the upper plenum and connected to the inlet plenum through a check valve