Advanced Water Reactor Research Articles

Predictability of safety analysis codes for departure from nucleate boiling in bundle for safety evaluation of massive hydrogen production systems

The advanced water reactor was indicated as a candidate for massive hydrogen production system using the water electrolysis method. In order to utilize the advanced water reactor system for hydrogen production, it is crucial to demonstrate the safety of the nuclear system during normal operations and accidents. Departure from nucleate boiling (DNB) is a critical phenomenon in the reactor core which should be addressed to demonstrate the integrity of the nuclear core during normal operations and accidents. Therefore, DNB has a particular importance to the reactor safety and precise prediction has been required for thermal-hydraulic analysis codes including subchannel and safety analysis codes. In this study it has been assessed the DNB prediction capability of thermal-hydraulic safety analysis codes used for the safety evaluation of nuclear reactor system against experimental data. For the assessment, thermal-hydraulic safety analysis codes, MARS-KS and TRACE, have been utilized. The DNB experiments conducted at the NUPEC experimental facility have been employed as a reference experiment for assessment. All experiments with bundle geometries under various steady-state conditions have been analyzed. The results show that both safety analysis codes generally predict the DNB power lower than the experimental database by 20% and the under-prediction occurs systematically with a linear characteristic. It is found that no significant difference in predictability of the DNB occurrence is observed between MARS-KS and TRACE. Therefore, it is concluded that both codes predict DNB conservatively, and MARS-KS and TRACE have almost identical predictability for the DNB occurrence.

Flow-induced vibration test of an advanced water reactor model Part 1: turbulence-induced forcing function

A 1:9 scale model of a proposed advanced water reactor was tested for flow-induced vibration. The main objectives of this test were: (1) to derived an empirical equation for the turbulence forcing function which can be applied to the full-sized prototype; (2) to study the effect of viscosity on the turbulence; (3) to verify the “superposition” assumption widely used in dynamic analysis of weakly copled fluid-shell systems; and (4) to measure the shell responses to verify methods and computer programs used in the flow-induced vibration analysis of the prototype. This paper describes objectives (1), (2), and (3); objective (4) will be discussed in a comparison paper. The turbulence-induced fluctuating pressure was measured at 49 locations over the surface of a thick-walled, non-responsive scale model of the reactor vessel/core support cylinders. An empirical equation relating the fluctuating pressure, the frequency, and the distance from the inlet nozzle center line was derived to fit the test data. This equation involves only non-dimensional, fluid mechanical parameters that are postulated to represent thefull-sized, geometrically similar prototype. While this postulate cannot be verified until similar measurements are taken on the full-sized unit, a similar approach using a 1:6 scale model of a commerical pressurized water reactor was verified in the mid-1970s by field measurements on the full-sized reactor.

Evaluation ofpfor239Pu: Impact for Applications of the Fluctuations at Low Energy

The average number of prompt neutrons p emitted per fission event has been evaluated for 239Pu with a special emphasis on the fluctuations experimentally observed in the low-energy range. These fluctuations have a significant impact on applications, especially the reactivity coefficient of advanced water reactors. Consequently, the p curve has to be defined in the same fine energy mesh as the fission cross section for accurate neutron source calculations. In this range, formalisms are proposed to calculate p from the resonance parameters, resolved or averaged. Using the JEF-1 library as a data base, an analysis of several thermal, low-moderated, or fast systems shows a good convergence of the selected microscopic and integral information.

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