Reactivity feedback effects

One is the secondary- coolant reduction test by partial secondary loss of coolant flow in order to simulate the load change of the nuclear heat utilization system. This test will demonstrate that the both of negative reactivity feedback effect and the reactor power control system brings the reactor power safely to a stable level without a reactor scram, and that the temperature transient of the reactor core is slow in a decrease of the secondary coolant flow rate. The test will be perfonned at a rated operation and parallel-loaded operation mode. The maximum reactor power during the test will limit within 30 MW (100%). In this test, the rotation rate of the secondary helium circulator will be changed to simulate a temperature transient of the heat utilisation system in addition to cutting off the reactor-inlet temperature control system. This test will be performed under anticipated transients without reactor scram (ATWS). 

LMRs with oxide-fueled core Models modified and newly developed mto the code so far mclude models for reactivity feedback effects and pool thermal-hydraulics In order to venfy the logic of the models developed, and to assess the effectiveness of the inherent safety features based upon the negative reactivity feedbacks m achieving the safety design objectives of passive safety, a preliminary analysis of UTOP and ULOF/LOHS performance has been attempted... 

Thermal expansion of the sodium results in a net positive reactivity feedback. Thermal expansion results in fewer sodium atoms within and surrounding the core. The reduced density surrounding the core results in fewer neufrons being scattered back into the core and produces a small negative feedback effecf by increasing fhe leakage around the periphery. However, the dominant effect is the reduchon of collisions between neutrons and sodium atoms, which hardens the neutron energy spectrum and yields a net positive reactivity feedback effect. 

In the solution of the reactor kinetics equations for short term transients there are two prompt reactivity feedback effects. These are related to the uranium and coolant temperature coefficients 3tenon variation and graphite heating have too long a time constant to affect the short-time kinetics calclatlons ... 

The transient response of the reactor system is dependent on reactivity feedback effects, in particular, the moderator temperature coefficient and the Doppler power coefficient. These reactivity coefficients are discussed in subsection 6.3 of this PCSR. 

Experimental Validation. The following types of measurement have been used to evaluate the accuracy of Doppler effect calculations (a) the South-west Experimental Fast Oxide Reactor (SEFOR) was built and operated specifically to measure Doppler effects (or fast-acting fuel reactivity feedback effects with expansion effects minimised) (b) the dependence of reactivity on temperature in operating power reactors, such as PHENDC and SUPER-PHENDC (fi-om the non-linearity of the temperature coefficient, for example) (c) the ZEBRA 5 Doppler Loop experiments, in which a test zone was heated. Experiments were performed with and without sodium present (d) the temperature dependence of the reactivity worths of small samples oscillated at the centre of critical assemblies (e) the differences in reaction rates in samples irradiated at different temperatures and (f) temperature dependent thick sample transmission and self-indication measurements, which are usually analj ed together with the differential nuclear data to provide average resonance parameter data. The uncertainties in extrapolating fi-om these comparisons to the conditions in an operating power reactor must also be taken into account. 

Furthermore, there may be an overriding consideration because of the effect of relative material expansion on the instantaneous reactivity feedback effect, as explained below. 

The fuel thermo-mechanical behaviour under irradiation that determines both, fuel temperatures and associated reactivity feedbacks effects (fuel expansion and Doppler effects), fuel to clad heat transfers and interaction, as well as the consequent cladding thermo-mechanical behaviour and thermal expansion reactivity effects ... 

Additional developments in the SSC-K code include models for reactivity feedback effects for the metallic fuel, and the PSDRS. Also a two dimensional hot pool model has been build into SSC-K for analyzing the thermal stratification phenomenon in the hot pool. The control system model in SSC-K is flexible enough to handle any control system. For code maintenance and readability, SSC-K was converted to FORTRAN 90 free form and the use of standard FORTRAN 90 has enhanced code portability. 

On the other hand, the power generation rate is reduced much more than the core flow rate in the 50% core flow case. The reason of these power trends can be found in the reactivity feedback effects as shown in Figs 9 and 10. 

The presence of water rods reduces the density reactivity feedback effect due to the large time delay in the heat transfer to the water rods, and this affects the coupled neutronic and thermal-hydraulic stability. 

The interaction between thermal-hydraulics and neutronics may bring about instability in the Super LWR and this instability also needs to be considered in the Super LWR design. These two processes are coupled through the heat transfer from the fuel rod to the coolant and moderator and through the reactivity feedback effects due to changes in temperature and changes in density of the coolant and the moderator. 

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