Reactor decay heat calculation

Because reactor unwatering considerably worsens decay heat pick-up from Spent Fuel Assemblies (SFAs), it should be performed after some period of fuel hold up in non-unwatered reactor. Many calculations were performed taking into account both factual and possible-limiting state of cores depending on fuel bumup. Two options were examined ... 

The PWR.ECP code begins by employing a chemical speciation model to calculate the composition of the coolant and the pH at closely spaced points around the primary coolant circuit, including the pressurizer, RWCU, and RHRS loops (Fig. 39). Note that the purpose of the RHRS is to remove decay heat from the core upon shutdown of the reactor. Accordingly, this system is isolated from the primary coolant circuit during normal... 

Estimated maximum values of the ratio G of fission-product decay-heat rate, with neutron absorption in fission products considered, to the decay-heat rate in the absence of neutron absorption in fission products are given in Table 2.13 [A2]. The data are calculated for U- U fuel irradiated for 4 years in a light-water reactor. For cooling times of <10 s, the... 

The calculated elemental composition, radioactivity, and decay-heat rate for discharge fuel are shown in Table 8.7 for the uranium-fueled PWR (cf. Fig. 3.31), in Table 8.8 for the liquid-metal fast-breeder reactor (LMFBR) (cf. Fig. 3.34), and in Table 8.9 for the uranium-thorium-fueled HTGR (cf. Fig. 3.33). These quantities, expressed per unit mass of discharge fuel, are useful in the design of reprocessing operations. For the purpose of comparison, all quantities are calculated for 150 days of postirradiation cooling. 

ISO (1992) Nuclear energy - Light water reactors Calculation of the decay heat power in nuclear fuels, ISO 10645. 

The K-, L-, and P-Reactor Disassmbly Basins house irradiated fizd and target assemblies, each emitting varying amounts of decay heat. As shown in the K-Reactor in Cold Standby PHA and P-and L-Reactor Disassembly Basins PHA (Ref 8-17, 8-51), the irra ated fiid assemblies currently in the K-, L-, and P-R u tor Disassembly Basins have relativdy low decay heat since the assemblies have been in the basins for an tended period of time or were not irradiated for a long period of time. Conservative calculations performed in the PHAs (Ref 8-17,8-51) show that the... 

Klapdor, H.V. and Metzinger, J., Predictions of the Decay Heat of Nuclear Reactors by Microscopic Beta Decay Calculations., Proc. Int. Conf. on Nuclear Data for Science and Technology - Mito, Japan (1988). 

ABSTRACT Technological advancements in area of sensor-based online maintenance systems have made the possibility of repairing some failed safety support systems of Nuclear Power Plants (NPP) such as electrical supply, I C systems, ventilation systems. However, the possibility of repair during accident situation is yet to be included into PSA level-1. Therefore, this paper presents a scheme of PSA level-1 by implementing an integrated method of Repairable Event Tree (RET) and Repairable Fault Tree (RET) analysis. The Core Damage Frequency (CDF) is calculated from consequence probabilities of the RET. An initiating event of Decay Heat Removal (DHR) systems of ASTRID reactor is analyzed. The proportionate CDFs estimated with repair and without repair have been compared and found that the recoveries can reduce CDF. In sum, this paper attempts to deal with the possibility of repair of some safety systems in PSA and its impacts on CDF of the NPP. 

Pure lead is not completely exempt firom polonium formation because Pb (the most abundant natural isotope of lead) transmutes into Bi, and Po is eventually produced from neutron capture by ° Bi. The rate of polonium production in pure lead is, however, much lower than in the case of LBE, and it is negligible in terms of decay heat power. In fact, the polonium inventory at equilibrium in the primary system of a 1500 MWth, pure lead-cooled reactor (ie, ELSY) has been calculated to be less than 1 g after 40 years of irradiation (Cinotti et al., 2011). 

From -180 min to -200 min, the TMI-2 core liquid level decreased as decay heat from the degraded core boiled liquid from the reactor vessel. The liquid level at -200 min stood 79 inches (2 m) above the bottom of the active core. The low thermal diffusivity of the large consolidated region of primarily ceramic core debris above the bottom crust prevented the interior of this region from cooling even when the reactor vessel was subsequently filled with water. Calculations indicate that a pool of molten material formed in the center of the consolidated region and increased in size during this period. 

In order to parametrically test fiie effect of different pre-acddoit conditions, in particular the reactor operation time prior to the acddent, calculations were performed for two diffarent axial power distributions in tiie core Namely, in addition to the power distribution obtained from the project sponsor, in which a typical end-of-life (higWy nonuniform, bottom-peaked) power disfribution was used, the APRIL.MOD3 code was also nm for a situation simulating the otitia end of possible al power profiles, in which a uniform axial decay heat generation rate was assumed. 

By circuit we mean from the bimdle exit to the containment entrance and, so far, these calculations have been performed in two phases first the thermal hydraulics then the fission product transport. In a reactor the decay heat of deposited aerosols would be expected to heat the pipe walls and thus feed back to the thermal hydraiuics but this effect will be negligible in Phebus. It is therefore reasonable to perform the calculations rmcoupled. 

In the case of reactor calculations, the decay heat profile evolution linked with the... 

A new model for heat-up of structures due to radioactive decay heating caused by deposited fission products has been included in VICTORIA-92. Decay heating is calculated according to elemental mass distribution and time since reactor shutdown. Temperatures of the deposited film and structure are calculated. Radioactive decay heating can cause deposited materials to revaporize and migrate through the RCS. This process can affect the source term late in an accident sequence when the containment is likely to have failed. In some cases, decay heating may induce failure of the RCS. 

The calculation results are shown in Fig. 6.36. The power decreases to the decay heat level due to the reactivity feedback and reactor scram. Reverse flow occurs in the water rod channel because the buoyancy pressure drop dominates the pressure drop balance. Heat conduction to the water rods increases when the coolant temperature in the fuel channel increases. This implies that the water rods serve as a heat sink . As the coolant expands in the water rods due to heat-up, there is an increase in the flow rate downstream from the water rods, including the fuel channel inlet. Consequently, the fuel channel flow rate is maintained even though the coolant supply from the cold-leg has stopped. This is called the water source effect of the water rods. The heat sink and water source effects mitigate heat-up of the fuel rod cladding, and hence enable the AFS to have a realistic delay time. The hottest cladding temperature begins to decrease before the initiation of the AFS. The increase in the hottest cladding temperature is about 250°C while the criterion is 520°C. 

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