Core Transient Behavior

As described in the previous chapter, the MRX copes with anomaly including accidents by help of the engineered passive safety system The core is flooded by the water-filled containment and the decay heat is removed by the EDRS and the CWCS. To view the function and the transient behavior, the LOG A analysis using RELAP5/mod2[ll] and COBRA-IV[12] codes is presented here as followings. 

The typical duration of thermohydraulic transients is given by the time which the coolant needs to pass the core. This time being about 0.7 sec in the core average, all core transients with periods of a few seconds or more pose no problem in this regard. In particular, the behavior of the reactor due to ship motion of a 7 sec period should be described reasonably well by this model. Also the load following characteristics can be determined properly. 

Only the case of steady coinjection of surfactant solution and gas into a one-dimensional core initially filled with surfactant solution is addressed. Calculated transient foam displacement well represents both the measured wetting liquid saturations and pressure profiles with physically meaningful parameter values. It is predicted and experimentally verified that foam moves in a piston-like fashion through a linear porous medium presaturated with surfactant solution. Moreover, the proposed population-balance predicts the entire spectrum of unique steady foam-flow behavior in the capillary-pressure regime. 

The time course of selected ambient temperature and core temperature in the rat monitored by telemetry exemplifies the regulated hypothermic response induced by administration of chlorpyrifos (Fig. 6). When dosed with a control vehicle (corn oil), there was a transient decrease in selected temperature that reflects a heat dissipatory respon.se from the stress of handling and injection. When dosed with chlorpyrifos, selected ambient temperature decreased from 30 to 25 °C. The behavioral response to select a cooler temperature preceded a 2.5 C decrease in core temperature. At the nadir of the decrease in core temperature,. selected temperature increased rapidly,... 

Compared to the conventional cores the core melt-down behavior and the associated recriticality risk of burner cores are influenced by the specific burner properties, i.e. higher Pu- and MA-enrichment, lower fuel mass, no axial and radial blanket, etc. First calculations of the initiation phase of an oxide core were performed with the newest version of the SAS4A code and the unprotected loss-of-flow accident (ULOF) has been chosen as a representative initiator. The results show that energy releases as consequence of the initiation phase of ULOF transients in the CAPRA reference core design and their mechanical load potential remain rather small. Most probably consequences can be contained ithin the primary containment. 

Transient power behavior of a nuclear core is determined by a condition known as "reactivity." For a core operating at a steady power level, the various factors that affect reactivity are balanced so that the net reactivity is zero. If the net reactivity is positive, power level will increase and, conversely, decrease if reacfivify is negative. Power control of a PWR is based on balancing reactivity through the use of mechanical and... 

The complexity of the heat transfer processes in nuclear reactor core requires computer codes to handle both local and system-wide behavior under normal, transient, and accident conditions. The code models are assessed with experimental data to ensure that they were working properly. Some of the largest and most widely used codes in United States are as follows ... 

The appearance of non-volatile fission products or actinide isotopes in the coolant can indicate the presence of fuel rod defects with a direct contact between the fuel and liquid water. This can occur with large-sized defects, in particular in comparatively cold regions of the fuel rod at the vertical or horizontal periphery of the reactor core. However, any statement in this regard can only be based on radionuclides that are not present in the coolant as a remnant from preceding transients this means that in a PWR Cs or Cs are not appropriate indicators for such fuel rod failures. The requirements are in principle fulfilled by Np, which is a reliable indicator for defects with fuel-to-water contact, as are ruthenium and cerium isotopes, as well. However, because of the complex behavior of these radionuclides in the coolant (adsorption on suspended corrosion products and deposition on primary circuit surfaces), only qualitative assessments can be made, which means that a quantitative evaluation of the number of fuel rods showing... 

For this experiment, the reactor core had been equipped with a central fuel module containing 100 pre-pressurized fuel rods, the UO2 fuel of which had been enriched to 9.7% and which had been pre-irradiated to a bumup of about 450 MWd/t U. The transient phase started with the reactor scram and was terminated about 30 minutes later when the external temperatiure on the surface of the shroud of the central fuel module reached 1517 K at this time, the highest measured cladding temperature reached 2100 K. When reflooding of the reactor core with emergency coolant was started, a rapid temperature excursion occurred within the central fuel module which was caused by the enhanced metal—water reaction. The transient was followed by a post-transient period of 44 days during which the reactor core was cooled by recirculating coolant and the concentrations of fission products deposited in the primary coolant system as well as their behavior in the blowdown suppression tank were measured. 

This is a typical pressure decreasing transient. The maximum turbine control valve opening is assumed and it is 130% of the rated value. The cladding temperature is always below the initial temperature because the main steam flow rate and therefore the core coolant flow rate increase. A scram signal is released when the pressure reaches the low level 1 (24.0 MPa). A depressurization signal is released when the pressure reaches the low level 2 (23.5 MPa). After opening the ADS, the reactor behavior is similar to that shown in Fig. 6.7 [1]. 

Chapter 2 covers design and analysis of the core and fuel. It includes core and fuel design, coupled neutronic and thermal hydraulic core calculations, subchannel analysis, statistical thermal design methods, fuel rod design, and fuel rod behavior and integrity during transients. 

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