Thermal-hydraulic considerations

This section describes some of the boiling phenomena that occur in water reactors with respect to safety analyses that require thermal hydraulic considerations. 

The reactor coolant pumps provide sufficient forced circulation flow through the Reactor Coolant System to assure adequate heat removal from the reactor core during power operation. A low limit on reactor coolant pump flow rate (i.e., design flow) is established to assure that Specified Acceptable Fuel Design Limits (SAFDLs) are not exceeded. Design flow is derived on the basis of the thermal-hydraulic considerations presented in Section 5.2. 

The rack employs two 0.174-in. Boral sheets between each row and each column of assemblies. The Boral sheets are composed of a 0,084-in. core of 35 wt v natural B4C in ALf, clad with 0.045 in. of aluminum. The two Boral sheets between assemblies are held in place. 2.714 in. apart, by a series of 0.125-in. aluminum channels. Which also serve to confine the 8.12-ft-sq fuel assembly when it is in the rack. The mechanical de.sign of the racks has been described by Allison and thermal-hydraulic considerations have been discussed by Melevia et al. ... 

One of the primary purposes of the water in the SEP is to ensure that the freshly discharged fuel assemblies, which continue to emit a significant amount of heat, are properly cooled. The thermal hydraulic considerations for the SEP racks and fuel assemblies may be broken down into the following categories ... 

This appendix summarizes thermal-hydraulic considerations for a plutonium-burning reactor and recommends for this purpose a low-power-density reactor that is cooled with low-temperature, low-pressure light water flowing at... 

Parameter and data uncertainty as well as uncertainly on the concept of the model should be considered for all models of a PSA, whereas uncertainly on the mathematical and numerical model should be taken into account mainly for the more complex thermal-hydraulics and integral codes. But so far, the uncertainty analysis in a PSA is generally restricted to the consideration of parameter and data uncertainties in the fault-tree and event-tree models. More specifically, the analysis is focused on uncertainties... 

These considerations apply to continuous process systems like thermal-hydraulic passive systems, fuUy relying on physical principles like natural circulation, for which the reliability figures are time specific, because of the involved thermal-lydraulic parameter evolution over time. 

With reference to our case of thermal-hydraulic passive system, let s consider the characteristic time-variant parameter W t) (its evolution during time will depend on the transient/accident scenario under consideration). The lower hoimd for natural circulation operation is denoted as Wi t), which, according to the failure criterion provided above (see chapter 4), is a fraction (0,8) of the flow-rate in nominal conditions. 

The preheater is essentially a plug flow reactor (L/D > 100) where the coal-solvent slurry and hydrogen gas are preheated to the liquefaction temperature. Extensive studies have been carried out to understand the exact nature of the processes, taking place in the preheater (2,11,13-16). Considerable work has also oeen performed to evaluate the dissolution process and its effect on thermal hydraulics of large-scale preheaters (13-16). 

The main purpose of the Swiss mixing-layer research programme is to investigate the thermal-hydraulic phenomena in horizontal shear layers with particular attention paid to the effects of Richardson (Ri) and Prandtl (Pr) number. Therefore, experiments are to be performed over a wide range of velocity and temperature differences between the two streams and with water and sodium. In connection with safety considerations for pool-type LMFBRs, a reliable experimental data base should allow one to validate codes used to calculate the flow fields in the pools. 

Though the fission in the reactor core is analyzed by the nuclear consideration, the heat generated by the reactor through fission and its use in the generation of power for a given reactor core is largely limited by thermal processes and material properties rather than by nuclear considerations. The safety analysis of the reactor during normal and abnormal operational conditions involves detailed thermal-hydraulics analysis. 

Vapor—aerosol interactions which are assumed to take place in the primary system were studied in detail in the British Falcon test facility. This facility, which is schematically shown in Fig. 7.19. (according to Beard et al., 1991), has been specifically designed to investigate the transport and deposition behavior of fission products under severe accident conditions. To produce representative aerosols, fuel pellets containing simulant fission products as well as trace-irradiated fuel pellets, both cladded in Zircaloy, were heated up to 2000 K in a steam—helium atmosphere in the presence of bulk-core materials. Fission product transport could be studied along a pathway which was designed to represent the upper plenum, hot-leg structures and containment. Considerable efforts were made to ensure that, as far as possible, the thermal-hydraulic conditions represented those of a selected accident. 

Appropriate neutronic, thermal-hydraulic, mechanical, material, chemical and irradiation related considerations associated with the reactor as a whole shall be taken into account in the design of fuel elements and assemblies, the reflectors and other core components. 

The European R D programme for EFR gave rise to more than one hundred "work packages" comprising more than a thousand individual tasks, some of which have involved considerable resources, e.g., reactor thermal-hydraulic experiments in large scale sodium and water mock-ups. 

Considerations of distributions across the flow channel, transverse to the primary flow direction, were first included in basically one-dimensional models by approximating the temperature distribution in the fluid parallel to the flow direction. Recently there is an increasing application of CFD to various single- and two-phase thermal-hydraulic analyses, including NCLs and supercritical fluid states, in nuclear power systems. These approaches also allow for resolution of the thermal stratification in horizontal and vertical sections of the loop as well as resolution of gradients normal to the primary flow direction and the consequent effects on calculated stability. Fully three-dimensional analyses are becoming the norm, but only for simple idealized single-phase cases. 

Material choices for structural components must balance neutronic considerations along with the manufacturing, mechanical, and thermal hydraulic needs... 

A significant consideration is the thermal-hydraulic criterion for limiting power for the FI phase of the accident. Initially, a "no bulk boiling" criterion was assumed to apply to the individual annular coolant channels within the fuel assemblies. Locally, subchannels could produce steam prior to the channel mixing cup temperature reaching the saturation temperature. 

The development of the power limits safety criteria is a two-step process (1) the extent of acceptable core damage, a macro consideration, must be defined and (2) the specific thermal hydraulic criteria that will conservatively bound the accident envelope within this acceptable core damage criteria must be established. 

A critical requirement for the acceptance of transient and accident analyses is the consideration of credibility of operator action under stress. The Thermal Hydraulic Limits Action Plan indicated that only during a LOPA is... 

The plant dynamics code for the analysis of plant control and startup thermal considerations are described in ref. [115]. The subchannel analysis code and the analysis are found in refs. [116, 117]. Thermal-hydraulic and coupled stability calculations at supercritical and at subcritical pressure as well as startup considerations are described in ref. [118]. 

These basic thermal-hydraulic parameters have been roughly determined from the considerations of reducing the BOP weight and improving the plant thermal efficiency (this is described in Chap. 3). The core design explained below is based on the core pressure of 25 MPa, inlet temperature of 280°C, and the average outlet temperature of 500°C. When these conditions are selected, the plant thermal efficiency becomes about 43.8%. These are the reference core characteristics. 

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