Thermal-Hydraulic Calculations

The thermal-hydraulic calculations are important for designing the Super LWR core, since the fission reactions by thermal neutrons are greatly affected by the 

Due to the constraints in the neutronic calculation models used for the core design, the fuel assemblies are axially divided into a number of fuel elements, and within each fuel element, the fuel assembly is homogenized. For the purposes of evaluating the thermal-hydraulic feedback to the neutronic calculations and evaluating the basic thermal-hydraulic characteristics of the core, detailed calculations involving the modeling of each fuel rod are not necessary. There are three fundamental thermal-hydraulic parameters required for the core design calculations  

Average coolant density (and temperature) of the fuel element for the neutronic calculations 

Estimated peak cladding temperature for roughly considering the effective cooling of fuel rods 

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Section 8.1 provided a description of a core melt. This section backs up to describe thermal-hydraulic calculations of the phenomena before, during, and after the accident, and other calculations to estimate the radioactive release from containment. In this accident physics cannot be analyzed separately from in-plant transport. 

This referential thermal-hydraulic calculation have been studied in details by Bamel and al (2(X)2 ). The main results were the following. Gas pressure does not seem to influence the saturation process. In fact, the saturation kinetics is governed by the darcean liquid flow. This flow is coupled with temperature, as can be seen in equation 3. It is accelerated when the heat source is taken into account. More precisely, this acceleration is entirely due to dynamic viscosity decrease while heating. 

Furthermore, according to water mass conservation equation, porosity does not affect liquid fluxes ratio. Therefore, darcean liquid flow still governs saturation kinetics. As can be seen on table 2, Qm" and Cim" calculations give the same saturation time, whereas in Cim" the saturation phenomenon is accelerated. As for thermal-hydraulics calculations, full saturation of the EB is reached earlier when both the heating source is being activated and liquid dynamics viscosity depends on temperature. Once again, this acceleration is only due to water dynamic viscosity decrease while heating. 

Six thermal-hydraulic calculations have been done. They are denoted by Ci " ,, ... 

Kirillov P., Yuriev Yu., Bobkov V. 1984. Reference book on thermal - hydraulics calculations (nuclear reactors, heat exchangers, steam generators). M., Energoatomizdat Publishers, in Russian, p.296. 

Determination of the loads through analysis of the reactor operating parameters combined with thermal hydraulics calculations for the specific thermal stratification and fluctuation areas. For this purpose, the recent R D progress made in the EFR (European Fast Reactor) project was used and transposed to a real installation. 

In the period of 1998-99, two sets of experiments focused on problems of rapid decrease of concentration of boric acid in reactor coolant at nuclear reactor core inlet were performed at the University of Maryland, US, under the auspices of OECD. The situation, when there is an inadvertent supply of boron-deficient water into the reactor vessel, could lead to a rapid (very probably local) increase of reactor core power in reactor, operated at nominal power, or to a start of fission reaction in shut-down reactor (secondary criticality). In the above mentioned experiments the transport of boron-deficient coolant through reactor downcomer and lower plenum was simulated by flow of cold water into a model of reactor vessel. These experiments were selected as the International Standard Problem ISP-43 and organisations, involved in thermal — hydraulic calculations of nuclear reactors, were invited to participate in their computer simulation. Altogether 10 groups took part in this problem with various CFD codes. The participants obtained only data on geometry of the experimental facility, and initial and boundary conditions. 

The analysis should use a logical approach which models how the event sequences progress from core damage to a radiological release. This is usually done by event tree analysis which models the accident sequence in a number of time frames and uses a set of nodal questions to model the sequence of events which occur. The construction of the event trees needs to be supported by thermal-hydraulic calculations and modelling of fission product release and transport inside the containment. 

Laslau, P. and D. Serghiuta. 1991. Reactor Physics and Thermal Hydraulic Calculations, Internal Report INR 3360, Pitesti, Romanian. 

Thermal- hydraulic calculations. Thermal-hydraulic modelling of a suspended core. Reliable codes for PWRs exist their applicability to calculation of suspended cores needs to be examined. 

Some system design conditions for thermal-hydraulic calculations of the STAR-LM, including the heat exchanger (HX) data, are provided in Table XXni-3 general data on core and thermohydraulics of the STAR-LM are presented in Table XXIII-1 in the beginning of this description. 

Owing to the iterative process of design, the volume fractions used for neutronic and for thermal-hydraulic calculations differ by a few percent at the conceptual design stage... 

Fuel channel box bow is a well known problem for BWRs. In principle, channel box bow is considered in the core design by neutron-kinetic and thermal-hydraulic calculations. Therefore, the degree of box bow must be supervised. Additionally, minimizing the box bow must be achieved by specific measures during design and construction of the channel boxes. 

As it was said previously the thermal hydraulic calculations are not very complex. Nonetheless a predse determination of the thermal behaviour just above the bundle required a radiative heat transfer model includii gas emissivity. The CATHARE2 model was change to better accommodate our physical conditions but, up to now it does not include the effect of the aerosol particles on the emission properties of the steam-hydrogen gas mixture. 

The initial information about the containment were obtained from the shared cost action phase B. There were sue partii ants from four countries. Thermal hydraulic calculations used CONTAIN, COCMEL, RALOC, WAVCO, CONTEMPT4 and JERICHO. Aerosol calculations used CONTAIN, AEROSIM, NAUA, HAARM-DTM and AEROSOLS-B2. Some Iodine chemistry studies used IMPAIR-2. The prindpal results were summarized in [8]. The main findii were ... 

J. N. Lillington, A. J. Lyons, I. M. Lovely, "Thermal-Hydraulic Calculation in TMI-2 Accident Analysis," Reactor Systems Analysis Division, AEE Winfrith, Report No. AEEW-R 2434, January 1989. 

As is briefly mentioned above, COREBN is based on the macro-cross section interpolations by bumups and the finite difference diffusion method for the neutron flux calculations. The macro-cross section sets for each fuel assembly type are prepared by ASMBURN as described above. COREBN linearly interpolates the macro-cross section sets tabulated for the three parameters, namely, the bumup, fuel temperature, and the moderator temperature. The bumup process of COREBN is similar to that of ASMBURN. Since COREBN is not equipped with a coupling function to the thermal-hydraulic calculations, the user has to give the input data of fuel temperature and moderator temperature for the calculations. 

The Nusselt number is calculated by the Oka-Koshizuka correlation [7] as described in (2.1). This correlation can be easily applied to the thermal-hydraulic calculations for the core design because it does not require the wall temperature. 

The single channel analysis model for the core thermal-hydraulic calculations does not consider the pressure drops and only takes into account the conservations of... 

S Applying the Single Channel Model to Core Thermal-Hydraulic Calculations... 

The core thermal-hydraulic calculations are based on the single channel analysis model. On the other hand, the three-dimensional core power distribution is obtained by COREBN for the calculation mesh described in Fig. 2.20 [9]. In the core thermal-hydraulic calculations, the neutron flux calculation mesh of the COREBN is assumed to compose a fuel channel group. The fuel channels in this fuel channel group are assumed to be identical. 

Figure 2.29 [9] shows the core thermal-hydraulic calculations by the single channel model. Each fuel assembly is assumed to be composed of 36 fuel channel... 

Fig. 2.29 Core thermal-hydraulic calculations by the single channel model. (Taken from doctoral thesis of A. Yamaji, the University of Tokyo (2005) [9])...

The coupling of neutronic and thermal-hydraulic calculations is especially important for designing the Super LWR core. The density change of the coolant (and moderator) is large and sensitive to the enthalpy rise of the coolant as it flows from the core inlet to the outlet. On the other hand, the core neutronic characteristics strongly depend on the coolant and moderator density distributions. 

The COREBN code does not have the coupling function. Hence, the bumup calculations for one cycle of the core operation is divided into a number of bumup steps. Within each bumup step, the neutronic and thermal-hydraulic calculations are coupled by the core power and density distributions (within each bumup step, the coolant density distribution is assumed to be constant). These calculations are repeated until the core power distribution and the density distributions are converged. Once the convergence is obtained, the bumup step proceeds to the next step. For the coupling calculations, the macro-cross section sets of the fuel assemblies are prepared for different coolant and moderator densities and these are interpolated by bumups. 

The coolant outlet temperature distributions at BOC, MOC, and EOC are shown in Fig. 2.58 [9] for 1/4 core symmetry. These thermal-hydraulic calculations are also based on the homogenized fuel assembly model and use the single channel analysis model as explained in Sect. 2.3.2. The detailed subchannel analysis results are explained in Sect. 2.5. 

The core water level is calculated from a detailed core thermal-hydraulic calculation. The quench front is calculated by a theoretical correlation proposed by Yamanouchi, et al. [24]. The heat transfer coefficient sharply changes by about 2 orders of magnitude in the vicinity of the quench front. In order to prevent the numerical instability caused by the abrupt change in the heat transfer coefficient, the neighboring nodes of the quench front are more finely divided into a size 1/100 of the thickness of the normal node as shown in Fig. 6.21. The flow regimes assumed in the reflood analysis are described in Fig. 6.22. Various heat transfer correlations are prepared according to the flow conditions. Table 6.9 [21] summarizes them. 

The SRAC system is used for core design analysis of the Super FR just as it was used for the Super LWR core design (see Chap. 2). It contains many kinds of calculation modules based on integral or differential neutron transport and FDM solutions. However, the SRAC system does not contain a thermal-hydraulic calculation module for coupling. The thermal-hydraulic calculation module is coupled with the original SRAC system. 

The thermal-hydraulic calculation is based on single charuiel analyses for multi coolant channels. A fuel assembly is expressed as two representative coolant channels. One is the average channel, and the other is the peak power channel. Both channels have the same thermal-hydraulic parameters, and only the linear heat rate distributions are different from each other. 

The neutron calculations and thermal-hydraulic calculations stated in Sects. 7.5.2.1 and 7.5.2.2 are all implemented into an automatic calculation scheme written in Perl and Awk script languages. The macroscopic cross section sets and HFFs are prepared for given coolant densities and bumup states by the staged homogenization of the unit cell and the assembly transport calculation. The auxiliary mesh generation module produces the Tri-Z mesh stmcture to be used in COREBN and CITATION. 

The main control module automatically prepares the inputs required for both neutronic and thermal-hydraulic calculations. It also produces the pin power distribution and bumup distribution, and then prepares the peak and average channels for each fuel assembly. The neutronic and thermal-hydraulic calculations are executed by the main control module internally and coupled to each other by pin power and coolant density distributions. 

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