Core thermal hydraulics analysis

Chu, K. J., and A. E. Dukler, 1974, Statistical Characteristics of Thin Wavy Films, AIChE J. 20 695. (3) Chu, P. T., H. Chelemer, and L. E. Hochreiter, 1973, THINC IV An Improved Program for Thermal Hydraulic Analysis of Rod Bundle Cores, WCAP-7965, Westinghouse Electric Corporation, Pittsburgh, PA. (App.)... 

Structural mechanics analysis indicates that the acceptable temperature difference between the primary temperatures from IHX outlets is about 40 K. Thermal hydraulics analysis indicates that with one module isolated in one loop, the temperature difference is 10 K at 90 % power. With 2 modules isolated, this temperature difference is 24 K at 67 % power. To achieve this, it is necessary to reduce the core flow (to 67 %) in proportion to the power as well as the sodium flow in the unaffected loop (50 %) as in the affected loop (50 %) The... 

The reactor coolant pumps are sized to deliver flow that equals or exceeds the design flow rate utilized in the thermal hydraulic analysis of the Reactor Coolant System. Analysis of steady-state and anticipated transients is performed assuming the minimum design flow rate. Tests are performed to evaluate reactor coolant pump performance during the post-core load hot functional testing to verify adequate flow. 

The computer codes which are used to carry out the anticipated operational occurrences and DBA analysis should be properly verified and validated. This includes the codes used to predict the behaviour of the reactor core, thermal-hydraulic codes and the radiological release and consequence codes. In addition, the analysts and users of the codes should be suitably qualified, experienced and trained. 

The thermal-hydraulic analysis is performed using special codes that receive the fission power spatial deposition from the reactor analysis pin power deconstruction and perform a heat transfer and fluid flow calculation to calculate the expected temperature profile throughout the core. Because neutron cross sections depend on temperature, this calculation must be iteratively linked into the previous six steps. In many cases, this linkage results in an expansion of the parametric expansions of the neutron cross sections (described as part of Step 6) to be a function of temperature and total assembly fission power. 

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. 

Over the next several years, the PRA was detailed to the point it included detailed fault trees of the mechanical, electrical, and instrumentation and control systems and the scope was expanded to include shutdown, fire, flood events, and large release frequency and off-site dose quantifications. Core damage frequency PRA was supported by extensive plant thermal-hydraulic analysis to justify success criteria. Extensive testing and thermal-hydrauUc analysis, to support containment integrity during core melt sequences, underpirmed the large release PRA. 

COBRAIIIc is a widely used code for the thermal-hydraulic analysis of nuclear reactor cores. Its approach is based on subchannel analyses where conservation equations are axially solved and coupled through mixing coefficients (Todreas Kazimi 2001). 

In accordance with the requirement of Slovak Nuclear Regulatory Body in BNS 1.4.2 and general international requirements (SSG-3 and SSG-4), the PSA model shall be a comprehensive description of the NPP. In order to have useful information from PSA, a realistic model of the plant is necessary. To account for all relevant dependencies, all necessary support systems are included in the model such as HVAC, electrical supply and heat sink are included in the PSA model. The success criteria for front-end safety systems are derived based on detailed thermal hydraulic analysis specific for the plant and the fuel under analysis, both for core damage and for radiological releases. 

From Table 2, mass flow distributions to the three core Mets exhibit (see Figure 2) a nearly inverted profile to those specified for the in-core thermal/hydraulic design. That is, more flow is going to the innermost fuel region of the core than was assumed by the in-core analysis), leaving the outer region with less flow. 

Success criteria were extracted from information contained in plant procedures, supplemented by the PRA team s experience with the plant reviews of existing SRS engineering/physics analysis and analyses done in support of the PRA. Additional thermal-hydraulic analysis to refine some of the success criteria is planned for 1991. It is expected that the impact of improved success criteria will be a reduction in the estimated core damage frequency. Therefore, this issue does not require resolution before restart. 

The OASIS code has been used for the thermal hydraulics analyses. A large number of events were analysed. The objective of these studies was to identify the worst-case scenario. Therefore, rupture position, rupture time, and break size spectra must be analysed. Currently, the detailed thermal hydraulics analysis for the primary coolant pipe break accident as a function of the break position is being performed. This study is based on conservative assumptions, as required for Category 4 DBE analyses. The main results of this analysis are the core inlet and outlet temperatures, the break flow rate, the IHX inlet and outlet temperatures, as well as the and clad temperature distribution. The results of the calculations show that the hotspot temperatures for the cladding and for the coolant are below the design safety limits (DSL). These temperatures are lower than the sodium boiling temperature. The regulator accepted the methods applied, as well as the results of the analysis. 

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... 

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. 

The MCST is defined as the maximum surface temperature of the cladding along the axial direction at a particular bumup. The MCST is shown for each fuel channel group at BOC, MOC, and EOC in Fig. 2.59 [9] for 1/4 core symmetry (for an explanation of fuel channel group see Sect. 2.3.2). The evaluations are based on the same methods as already explained so far (homogenized fuel assembly model with a single channel thermal-hydraulic analysis model). 

The MCST is evaluated with three-dimensional core calculations using the homogenized fuel assembly model and the single channel thermal-hydraulic analysis model as before. The peak value of the MCST is about 650°C, which is the same as that of the first trial core. As noted above, the removal of the flow separation plates for the first trial core decreases the core outlet temperature by about 40-50 C. Taking this reduction into account, the outer core downward flow cooling can effectively raise the average outlet temperature by about 70-80°C, which may have a great impact on the plant economy. 

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 core design procedure consists of two parts, nuclear design and thermal-hydraulic analysis. The former is based on the fine-mesh multi-group neutron diffusion solution. The latter is based on single channel analyses for the average and hot channels of all the fuel assemblies. This approach is the same as that in the Super LWR design. 

Preliminary reactor core concepts have been proposed at various institutions. Among them, the concept with mixed neutron spectrum proposed by SJTU [95,96] has achieved the special attention of Chinese researchers. The mixed spectrum SCWR core combines the merits of both thermal and fast spectra as far as possible. The basic idea is to divide the reactor core into two zones with different neutron spectra. In the outer zone, the neutron energy spectrum is similar to that of PWRs. To assess the performance of the reactor core, a coupled neutron-physics and thermal-hydraulics analysis was conducted [96]. 

Given the damage states, the analysis flows much as shown in Figure 6.3-1, depending on the process. For a nuclear power plant, thermal-hydraulic analyses determine the spatial temperature of the damaged core, and consequently the ability of the core to retain radioactive materials. Analysis of the physical processes reveals the amounts of hazardous materials that may be released. 

Kelly JM, Stewart CW, Cuta JM (1992) VIPRE-02 - A two-fluid thermal-hydraulics code for reactor core and vessel analysis Mathematical modeling and solution methods. Nucear Technology 100 246-259... 

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... 

RPV, and with helps of EDRS and CWCS. Details will be described in later (in LOCA analysis). Thus, the core flooding in a LOCA can be attained passively without ECC pumps or an accumulator if the containment initial water level is appropriate. Experiment [6] on the thermal hydraulic behaviour of water-filled containment was conducted and the relationships between the initial water level of containment and the balance pressure, etc., were obtained, and the principal function of core flooding was confirmed experimentally. 

Computer code CALPER — a thermal hydraulic subchannel analysis code for the assessment of coolant local conditions in the fuel assemblies and in the core of PWRAVWER-type nuclear reactors... 

The database information consists of the thermal-hydraulic/radiological variables of the core, reactor coolant system, containment system, and safety systems which are generated from integrated severe accident analysis code, such as MAAP or MELCOR program. 

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. 

During normal operation of the reactor, core heat is removed by the natural circulation of Pb-Bi coolant. The coolant at 1173 K enters the fuel tube in the lower plenum, absorbs the reactor heat, and at 1273 K reaches the upper plenum. Twelve sodium heat pipes transfer heat from the upper plenum to the system of heat utilizing vessels. Thermal-hydraulic analyses were carried out to study natural circulation and the effect of orificing in the primary loop. A computer model based on the law of conservation of momentum was developed for this analysis a simplified model of the primary loop is shown in Fig. XXIX-7. 

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