Flow instabilities in boiling channels of pressure-tube-type reactor

The core of an advanced thermal reactor (ATR), which is a light-water-cooled, heavy-water-moderated, pressuretube-type reactor, consists of many parallel channels. For this reason the ATR was believed to exit from flow instability easily. Hence the flow instability conditions were investigated with the heat transfer loop (HTL) and safety experiment loop (SEL), which simulate the flow system of an ATR at full scale. The present data were compared with those for a boiling water reactor (BWR) system. The effect of the outlet pipe, which is not provided in a BWR, on the flow oscillation under low power conditions was also studied.     

The laboratory investigation of the experimental nuclear reactor VVR-S

Results are given in this paper of the laboratory investigation of the characteristics of the experimental reactor VVR-S undertaken with the aim of studying the neutron and physical parameters which are the most important for putting the reactor into operation and for its exploitation. As the result of the experimental work carried out, the critical mass and the maximum and operating fuel charges were found, the compensating capacity of the control and emergency rods was determined, the influence of the various factors (variation in the temperature of the water in the active zone, variations in the properties of the reflector, etc.) on the reactivity was studied, the distributions of neutron density with height and along the radius of the active zone were measured, and the operating time of the control rods was obtained.In conclusion the authors express their gratitude to T. N. Zubarev for discussion of the results and to O. I. Liubimtsev and I. V. Koptev for help in the work.     

中国实验快堆的安全特性

钠冷快堆因钠具有好的热物理特性而具有固有安全性,同时也因钠是活泼的碱金属,也难免会有钠的泄漏、钠火和钠水反应等工业事故.本文介绍了中国实验快堆利用钠冷快堆的固有安全性,装设了单靠自然循环和自然对流的事故余热导出系统等多项非能动安全系统及完善的能动安全系统,其安全性达到了第Ⅳ代先进核能系统的安全要求.对于大型快堆,因其保...     

Interpretation of the results of the CORA-33 dry core boiling water reactor test

All boiling water reactor (BWR) degraded core experiments performed prior to CORA-33 were conducted under ‘wet’ core degradation conditions, in which water remains within the core and continuous steaming feeds metal-steam oxidation reactions on the in-core metallic surfaces. However, one dominant set of accident scenarios would occur with reduced metal oxidation under ‘dry’ core degradation conditions and, prior to CORA-33, this set had been neglected experimentally. The CORA-33 experiment was designed specifically to address this dominant set of BWR ‘dry’ core severe accident scenarios and to resolve partially phenomenological uncertainties concerning the behavior of relocating metallic melts that drain into the lower regions of a ‘dry’ BWR core (the ex-reactor experiments at Sandia National Laboratories will further address these uncertainties). CORA-33 was conducted on 1 October 1992, in the CORA test facility at Karlsruhe. A review of the CORA-33 data indicates that the objectives were achieved; i.e. core degradation occurred at a core heat-up rate (characterized by the absence of any temperature escalation caused by oxidation) and a test section axial temperature profile (at incipient structural melting) that are prototypic of full-core nuclear power plant simulations under ‘dry’ core conditions. Simulations of the CORA-33 test at Oak Ridge National Laboratory (ORNL) have required the modification of existing control blade-canister materials interaction models to include the eutectic melting of the stainless steel-zircaloy interaction products and the heat of mixing of stainless steel and zircaloy. The timing and location of canister failure and melt intrusion into the fuel assembly appear to be adequately simulated by the ORNL models. This paper will present the results of the post-test analyses carried out at ORNL based on the experimental data and the post-test examination of the test bundle at Karlsruhe. The implications of these results with respect to degraded core modelling and the associated safety issues are also discussed.     

Core characteristics of fast reactor cycle with simple dry pyrochemical processing

Fast reactor core concept and core nuclear characteristics are studied for the application of the simple dry pyrochemical processing for fast reactor mixed oxide spent fuels, that is, the Compound Process Fuel Cycle, large FR core with half of loaded fuels are recycled by the simple dry pyrochemical processing. Results of the core nuclear analyses show that it is possible to recycle FR spent fuel once and to have 1.01 of breeding ratio without radial blanket region. The comparison is made among three kinds of recycle fuels, LWR UO₂ spent fuel, LWR MOX spent fuel, and FR spent fuel. The recycle fuels reach an equilibrium state after recycles regardless of their starting heavy metal compositions, and the recycled FR fuel has the lowest radio-activity and the same level of heat generation among the recycle fuels. Therefore, the compound process fuel cycle has flexibility to recycle both LWR spent fuel and FR spent fuel. The concept has a possibility of enhancement of nuclear non-proliferation and process simplification of fuel cycle.     

The BR-2 experimental fast neutron reactor

A brief survey is made of the design of the experimental fast neutron reactor and of its basic experimental and auxiliary equipment. The reactor was designed for physical experimentation with fast neutrons. The core is composed of plutonium rods; the lateral reflector is filled with depleted uranium. Heat is removed from the core by mercury and from the uranium reflector by air. The total rated power of the reactor is 150 kw of which about 100 kw is derived from the core.     

Neutron flux optimization in irradiation channels at NUR research reactor

Optimization of neutron fluxes in experimental channels is of great concern in research reactor utilization.The general approach used at the NUR research reactor for neutron flux optimization in irradiation channels is presented.The approach is essentially based upon a judicious optimization of the core configuration combined with the improvement of reflector characteristics.The method allowed to increase the thermal neutron flux for radioisotope production purposes by more than 800%. Increases of up to 60% are also observed in levels of useful fluxes available for neutron diffraction experiments (small angle neutron scattering (SANS), neutron reflectometry, etc.).Such improvements in the neutronic characteristics of the NUR reactor opened new perspectives in terms of its utilization. More particularly, it is now possible to produce at industrial scales major radio-isotopes for medicine and industry and to perform, for the first time, material testing experiments.The cost of the irradiations in the optimized configuration is generally small when compared to those performed in the old configuration and an average reduction factor of about of 10 is expected in the case of production of Molybdenum-99 (isotope required for the manufacturing of Technetium-99 medical kits).In addition to these important results, safety analysis studies showed that the more symmetrical nature of the core geometry leads to a more adequately balanced reactivity control system and contributes quite efficiently to the operational safety of the NUR reactor.Results of comparisons between calculations and measurements for a series of parameters of importance in reactor operation and safety showed good agreement.     

Aspects of 238Pu production in the experimental fast reactor JOYO

Experimental determination of ²³⁸Pu in ²³⁷Np samples irradiated in the experimental fast reactor JOYO was done as part of the demonstration of ²³⁸Pu production from ²³⁷Np in fast reactors within the framework of the protected Pu production project, which aims at reinforcement of proliferation resistance of Pu by increasing the ²³⁸Pu isotopic ratio. ²³⁸Pu production amount in the irradiated ²³⁷Np samples was determined by a radioanalytical technique. Aspects of ²³⁸Pu production were examined on the basis of the present radioanalysis. The ²³⁸Pu production amount depends on the neutron spectrum which can range from that of a typical fast reactor to a nearly epi-thermal spectrum. It is concluded that the fast reactor has not only high potential for use in protected Pu production, but also as an incinerator for excess Pu.