Subchannel Analysis

B22. Bowring, R. W., Burn-out prediction in multi-rod clusters by subchannel analysis, paper presented at Sympo. Two-Phase Flow, Grenoble, May 1965. 

Subchannel analysis for PWR cores CHF predictions in BWR fuel channels... 

The worst operating condition in a common design practice consists of overly conservative assumptions on the hot-channel input. These assumptions must be realistically evaluated in a subchannel analysis by the help of in-core instrumentation measurements. In the early subchannel analysis codes, the core inlet flow conditions and the axial power distribution were preselected off-line, and the most conservative values were used as inputs to the code calculations. In more recent, improved codes, the operating margin is calculated on-line, and the hot-channel power distributions are calculated by using ex-core neutron detector signals for core control. Thus the state parameters (e.g., core power, core inlet temper-... 

The nonuniform heat flux prediction of W-3 DNB correlation (developed from single channel data as shown in Fig. 5.70) was verified in axially, nonuniformly heated bundles by comparison of the 284 DNB data points (Fig. 5.72) with the W-3 correlation through a subchannel analysis including spacer factor, showing excellent agreement. The standard deviation was 7.4%. 

Comparisons of rod bundle data with the Columbia correlation and other existing correlations were made using COBRA-IIIC code for predictions of all correlations. The DNBR (or CHFR) reported is not the critical power ratio as used by other authors. The DNBR errors reported by Reddy and Fighetti (1983) are based on the following analysis The measured local heat flux at the experimental location of the first or higher-rank CHF indications is compared with the predicted CHF calculated using local conditions from the subchannel analysis for the... 

A detailed subchannel analysis for BWRs is described by Lehey and Moody (1977). 

Reddy, D. G., and C. F. Fighetti, 1982, Subchannel Analysis of Multiple CHF Events, NUREG/CR 2855, Columbia University, New York. (5)... 

Rouhani, S. Z, 1973, Axial and Transverse Momentum Balance in Subchannel Analysis, Topical Meeting Requirements and Status of the Prediction of the Physics Parameters for Thermal and Fast Reactors, Julich, Germany. (3)... 

Rowe, D. S, 1970, COBRA II Digital Computer Program for Thermal Hydraulic Subchannel Analysis of Rod Bundle Nuclear Fuel Elements, BNWL 1229, Battelle Northwest Laboratory, Richland,... 

Weisman, J., 1973, Review of Two-Phase Mixing and Division Cross Section in Subchannel Analysis, Rep. AEEW-R-928, UK Atomic Energy Authority, Winfrith, England. (3)... 

Total calculated subassembly flow rate, as well as design data on core neutronics are input to subchannel analysis codes that predict coolant flow and temperature distribution in the subchannels of the core subassemblies. The peripheral subchannel temperatures and flow rates used for duct temperature prediction, as well as peak subchannel coolant temperature used for prediction of hot spot temperature of the hottest fuel elements are of particular interest. 

Subchannel analysis is commonly used for thermal hydraulic analysis of single fuel subassemblies. Bulk average values characterize the fluid dynamics and thermal coolant conditions in each subchannel. Thermal and hydraulic interactions between subchannels are taken into account. 

Subchannel analysis code for the calculation of coolant local thermo hydraulic conditions based on the rod bundles data obtained from the CHF data bank (subchannel CHF data bank)... 

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... 

Subchannel analysis codes, ASFRE for single-phase flow and SABENA for two-phase flow, have been developed for the purpose of predicting fuel element temperature and thermalhydraulic characteristics in the FBR fuel assemblies. ASFRE has the detailed wire-spacer model called distributed flow resistance model, which calculates the effect of wire-spacer on thermalhydraulics. Also planer and porous blockage models are implemented for fuel assembly accident analysis. In this reporting period, three dimensional thermal conduction model was used for the evaluation of local blockage in a fuel assembly. In addition, the comparison of pressure losses in the assembly with the water experimental data has been performed. Regarding SABENA, based on the two-fluid model, no activity is reported. 

Gaspaii, G.P., Hassid, A., and F. Lucchini. 1975. A rod-centered subchannel analysis with turbulent (enthalpy) mixing for critical heat flux prediction in rod clusters cooled by boiling water. Proceedings of the 5th International Heat Transfer Conference, September 3-7,1974. CONF-740925, Tokyo, Japan. 

Once the individual channel properties have been derived from the HAMBO subchannel analysis, it is necessary to consider the flow distribution problem in the primary circuit. This question arises because the higher power channels tend to contain more steam and offer a greater resistance to flow than the low power channels. Moreover, there are differences in the pipe runs to and from the channels which can cause small but significant differences in flow. 

Subchannel analysis models have been investigated for CSRIOOOEA by using the experimental data available and the computational fluid dynamics (CFD) code (Du et al., 2013). The analysis results are used to improve a subchannel code. The steady-state subchannel analysis is conducted on the CSRIOOO FA to obtain the temperature distribution of coolant and cladding and pressure drop in the FA. The results show that smaller pitch will flatten the profile of the coolant temperature and reduce maximum cladding surface temperatures, but it increases the pressure drop in the assembly. 

The cladding temperature that was obtained by the three-dimensional coupled core calculation is the average temperature over the assembly. The peak cladding temperature of a fuel rod is necessary for the evaluation of the fuel cladding integrity. The subchannel analysis code of the Super LWR is coupled with the fuel assembly bum-up calculation code for this purpose [25]. Fuel pin-wise power distributions are produced for various bum-ups, coolant densities, and control rod positions. The pin-wise power distributions are combined with the homogenized fuel assembly power distribution to reconstmct the pin-wise power distribution of the core fuel assembly. The power distribution over the fuel assembly is taken into account as shown in Fig. 1.11. The reconstracted pin-wise power distribution is used in the evaluation of peak cladding temperature with the subchaimel analysis. 

Fig. 1.11 Coupling of subchannel analysis with three-dimensional core calculation (Reconstruction of pin-wise power distribution for the subchannel analysis)...

The maximum cladding temperature predicted by the subchannel analysis is higher than that predicted by single channel analysis which is used for the three-dimensional core calculation. 

The evaluation of peak cladding temperature is summarized in Fig. 1.12. The radial and local flux factors are evaluated separately, but further improvement was made. Incorporating subchannel analysis into the three-dimensional core coupled calculation, iterating the subchannel analysis with the core calculation rationalizes the evaluation of radial and local flux factors [28]. The nominal peak steady state temperature decreases 25°C from the value of the separate evaluation of Fig. 1.12. 

The change of cross flow within a subassembly may occur during transients. The MCST may change from the result of the single-channel calculation. A transient subchannel analysis code was developed and the safety analysis of a Super LWR was carried out [70]. The temperature rises from the steady state value are about 20°C at the abnormal transients and about 130°C at accidents. The maximum values still stay below the MCST criteria for transients and accidents. The development and application of the transient subchannel analysis code are sununarized in Sect. 6.8. 

Single channel thermal hydraulics (SPROD), 3D coupled core neutronic/thermal-hydraulic (SRAC-SPROD), Coupled subchannel analysis, Statistical thermal design method. Fuel rod behavior (FEMAXI-6), Data base of heat transfer coefficients of supercritical water (Oka-Koshizuka correlation)... 

Safety Transient and accident analysis at supercritical-and subcritrical pressure (SPRAT-F, SPRAT-DOWN), ATWS analysis (SPRAT-DOWN), LOCA analysis (SCRELA,SPRAT-DOWN-DP), Time-dependent subchannel analysis Start-up (sliding pressure and constant pressure)... 

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]. 

An improved core design procedure of the Super LWR that coupled the subchannel analysis with three-dimensional coupled core calculations is described in ref. [28]. The time-dependent subchannel analysis code for safety analysis of the Super LWR is described in ref. [123]. 

K. Yoshimura, Y. Ishiwatari, et al., Development of Transient Subchannel Analysis Code of Super LWR and Application to Flow Decreasing Events, Proc. NURETH-13, Kanazawa, Japan, Septemberl7-October 2, 2009, N13P1434 (2009)... 

J. Yoo, Y. Ishiwatari, Y. Oka and J. Liu, Subchannel Analysis of Supercritical Light Water-Cooled Fast Reactor Fuel Assembly, Nuclear Engineering and Design, Vol. 237, 1096-1105 (2007)... 

L. Cao, Y. Oka, Y. Ishiwatari and Z. Shang, Fuel, Core Design and Subchannel Analysis of a Super Fast Reactor, Journal of Nuclear Science and Technology, Vol. 45(2), 138-148 (2008)... 

T. Mnkohara, Design of Supercritical-Pressure Light Water Cooled Fast Reactor and the Subchannel Analysis, Doctoral thesis, the University of Tokyo, (2001) (in Japanese)... 

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