Anticipated Operational Occurrences Research Articles

BISON analysis of FeCrAl and Zircaloy cladding deformation during simulated BWR cyclic dryout conditions

The BISON fuel performance analysis code has been used to study FeCrAl and Zircaloy cladding performance to better understand cladding deformation behavior in boiling water reactors during conditions caused by anticipated operational occurrences (AOOs) or by anticipated transients without scram (ATWS). BISON models of novel separate effects tests simulating ATWS events predicted mechanical responses to thermal cycling boundary conditions. We assessed FeCrAl and Zircaloy deformation and burst models to determine those best suited for cyclic dryout event analysis. In addition, we compared diametral strain profiles of the post-experiment claddings to simulation results. The capability of the BISON code for modeling BWR AOO or ATWS events is promising.

Study on dryout and rewetting during accidents including ATWS for the BWR at JAEA

Post-boiling transition (Post-BT) heat transfer is important to analyze the duration of surface dryout and the peak cladding temperature during abnormal transient and design basis accident (DBA) conditions in light water reactors. Furthermore, the Japanese new safety regulation revised after the Fukushima Daiichi accident requires the analysis of the beyond-DBA conditions including the anticipated transient without scram (ATWS), where experimental databases for the code validation are almost-non-existent especially for those with highly oscillatory power change caused by thermo-nuclear coupling. Motivated by the importance, Japan Atomic Energy Agency has conducted a series of experiments for BT and rewetting phenomena to develop mechanistic models based on physical understanding and to validate thermal hydraulic codes. So far, the experimental databases for the Post-BT heat transfer rate, deposition rates of liquid droplets, and the dryout and rewetting behavior were developed mainly for the conditions covering the anticipated operational occurrences (AOO) for BWRs. The significance of the precursory cooling just before the rewetting was identified from comparison with a theoretical model for the two-dimensional transient heat conduction in the cladding. Recently, experiments including the simulation of the ATWS were initiated using the single-tube test apparatus with simulated spacers, liquid film heat transfer visualization apparatus for visual observation, and the HIDRA facility with test sections of a full-length 4 × 4 bundle and a short-length 3 × 3 length bundle. Some promising results have already been obtained, which show effects of simulated ferrule and swirl spacers on the post-BT heat transfer and rewetting velocity, and mechanism of the rewetting by the propagation of the liquid film on the initially heated surface of the heater rod.

Fission products behaviour during a power transient: Their inventory in an intragranular bubble

The behaviour of fission products is a key issue during Anticipated Operational Occurrences (AOOs) or Condition II transients or accidental sequence for nuclear fuel. Here we characterized how fission products behaved inside chromium doped UO2 pellet during a power ramp. At the pellet centre fission products have left the UO2 lattice and can be found in bubbles. The composition of the bubbles was determined using an original experimental methodology. The existence of separated precipitates made of metallic fission products for the one, and volatile fission products for the other, was evidenced. This result is discussed with regards to the behaviour of fission products during a power ramp.

Rewetting front propagation under anticipated operational occurrences for boiling water reactors – development of two-dimensional analytical model

An analytical model to predict a rewetting velocity applicable to high pressure and high flow rate condition during anticipated operational occurrences (AOOs) is developed by applying Wiener–Hopf technique coupled with appropriate kernel substitutions. The model considers the effects of enhanced cooling in the vicinity to liquid film front termed “precursory cooling” and heat input from fuel pellets on back side of wall as boundary conditions of a heat conduction equation. A simplified two-dimensional model neglecting an effect of axial heat conduction is also proposed. It is found through the comparison among the models and experimental data that the contribution of the heat conduction in the wall-depth direction is essential in the prediction of the rewetting velocity at the thermal-hydraulic condition simulating AOOs and the axial heat conduction has little influence when an enhanced heat transfer coefficient in the dried-out region is appropriately given as a function of distance from the liquid film front.

Core Heat Transfer Coefficients Immediately Downstream of the Rewetting Front during Anticipated Operational Occurrences for BWRs

A heat transfer coefficient (HTC) model was developed for the prediction of post-boiling transition (post-BT) behavior that might occur during anticipated operational occurrences (AOOs) for boiling water reactors (BWRs). The model development was based on measurements of heat transfer coefficient, liquid droplet deposition rate, and droplet concentration in our experiments conducted at high pressure. The model focused on the heat transfer near the rewetting front where the cooling by droplet deposition significantly affects the propagation behavior of a liquid film. The correlation by Sugawara was validated for the prediction of the deposition by using the experimental data. The model was also expressed as a function of the distance from the rewetting front to use in analytical models for the rewetting propagation. Both expressions of the present model successfully predicted our experimental data simulating the BWR thermal-hydraulic conditions.

Core Heat Transfer Coefficients Immediately Downstream of the Rewetting Front during Anticipated Operational Occurrences for BWRs

A heat transfer coefficient (HTC) model was developed for the prediction of post-boiling transition (post-BT) behavior that might occur during anticipated operational occurrences (AOOs) for boiling water reactors (BWRs). The model development was based on measurements of heat transfer coefficient, liquid droplet deposition rate, and droplet concentration in our experiments conducted at high pressure. The model focused on the heat transfer near the rewetting front where the cooling by droplet deposition significantly affects the propagation behavior of a liquid film. The correlation by Sugawara was validated for the prediction of the deposition by using the experimental data. The model was also expressed as a function of the distance from the rewetting front to use in analytical models for the rewetting propagation. Both expressions of the present model successfully predicted our experimental data simulating the BWR thermal-hydraulic conditions.

Safety Evaluation of the HTTR-IS Nuclear Hydrogen Production System

The establishment of a safety evaluation method is one of the key issues for the nuclear hydrogen production demonstration since fundamental differences in the safety philosophy between nuclear plants and chemical plants exist. In the present study, a practical safety evaluation method, which enables to design, construct, and operate hydrogen production plants under conventional chemical plant standards, is proposed. An event identification is conducted for the HTTR-IS system, a nuclear hydrogen production system by thermochemical water splitting iodine-sulfur process (IS process) utilizing the heat from the high temperature engineering test reactor (HTTR) in order to select abnormal events, which would change the scenario and quantitative results of the evaluation items from the existing HTTR safety evaluation. In addition, a safety analysis is performed for the identified events. The results of safety analysis for the identified five anticipated operational occurrences (AOOs) and three accidents (ACDs) show that evaluating items such as a primary cooling system pressure, temperatures of heat transfer tubes at pressure boundary, etc., do not exceed the acceptance criteria during the scenario. In addition, the increase of peak fuel temperature is small in the most severe case and therefore, the reactor core was not damaged and cooled sufficiently. These results will contribute to the safety review from the government and demonstration of the nuclear production of hydrogen.

Application of the Post-Boiling-Transition Standard to Licensing Analysis

To ensure fuel integrity, light water reactor cores are designed to avoid the onset of boiling transition (BT) inside the fuel assembly that leads to a deterioration of the heat transfer characteristics and subsequent excessive rise of the fuel-cladding temperature in the anticipated operational occurrences (AOOs). However, some boiling water reactors’ AOO events result in immediate scram or suppression of the reactor power due to an increase in the reactor coolant void fraction. Recent studies show that a short duration of dryout inside the fuel assembly only leads to a small rise in the fuel-cladding temperature and thus does not pose a threat to fuel integrity. Many tests on BT and an improved comprehension of its mechanism have led to the development of a methodology to appropriately assess the fuel-cladding temperature after BT has been reached. The Standards Committee of the Atomic Energy Society of Japan has therefore proposed a cladding temperature criterion after BT. Applying the post-BT standard enables the value of the operating limit minimum critical power ratio (OLMCPR) to be decreased by allowing for a short duration of dryout. We calculated the fuel-cladding temperature and dryout duration in the load rejection condition without a bypass event. The calculated results show that both the fuel-cladding temperature and dryout duration meet the post-BT standard in the case of a small OLMCPR, which is determined by the loss of feedwater heating. This enables a more efficient reactor core to be designed by applying the post-BT standard to licensing analysis. The possibility of applying a post-BT standard is demonstrated from the results of this work.

Current Status of the Post Boiling Transition Research in Japan

Development of rewetting correlation formula was the key to predict fuel-cladding temperature after Boiling Transition (BT). Japanese BWR utilities and vendors performed some tests of rewetting and made two rewetting correlation formulas. The effect on fuel integrity after BT depends on temperature of fuel rod and time of dryout. Main cause of losing fuel integrity during BWR's Anticipated Operational Occurrences (AOO) after BT is embrittlement of the claddings due to oxidation. Ballooning of fuel rod is excepted because its pressure boundary isn't broken. In Japan, the Standards Committee of Atomic Energy Society of Japan (AESJ) is making post BT standard. This standard provides guidelines based on the latest knowledge to judge fuel integrity in case of BT and the validity of reusing the fuel assembly that experienced BT in BWRs.