Shigeo Ebata

Caorso limit cycle oscillation analysis with three-dimensional transient code TOSDYN-2

The Caorso limit cycle data observed during the stability tests were analyzed by the three-dimensional transient code TOSDYN-2. The Caorso limit cycle oscillation was spatially out of phase, and both the amplitude and the extent of the large amplitude region were large. For this reason, it is very valuable for the qualification of the TOSDYN-2 code. TOSDYN-2 employs a three-dimensional neutronics model and a multichannel-type thermal-hydraulic model. The channel-type grouping is important for qualification analysis. It was determined by considering the test results and the more detailed three-dimensional steady-state code results. The analytical results imply that many unstable channels or unstable regions might exist separately in the core. To account for this, TOSDYN-2 could accurately simulate both the amplitude of the limit cycle oscillation and the spatial power change profile. Thus, TOSDYN-2 applicability to the spatial power change phenomenon has been well verified.

TRACG transient analysis code -- Three-dimensional kinetics model implementation and applicability for space-dependent analysis

Space- and time-dependent phenomena, mostly related to neutron flux oscillations, have been observed in several boiling water reactor plants. A time-dependent three-dimensional transient analysis code is indispensable for simulating such phenomena. In a joint effort between the General Electric Company and the Toshiba Corporation, a three-dimensional neutron kinetics model has been implemented into the best-estimate thermal-hydraulics code, TRACG. A neutronics model implementation and the applicability of the modified TRACG code for analyzing space-dependent phenomena are discussed. To verify the code, startup tests with selected rod insertions, where control rods are locally inserted, are simulated. Both corewide, spatially in-phase neutron flux oscillations and regional, spatially out-of-phase oscillations are modeled. The results show that the modified TRACG code has sufficient capability to simulate space-dependent transients and is also a useful tool for investigating the fundamental mechanisms behind such transients.

Space-dependent analysis of BWR core nuclear thermal hydraulic instability and thermal margin

Abstract Nuclear thermal hydraulic oscillations in BWR cores were analyzed by the space-dependent BWR core transient program STANDY. In a simulation of instability in the Lasalle-2 unit, the oscillations that caused a scram were successfully reproduced. The maximum thermal margin decrement was far smaller than the initial margin, and significant margin to thermal limits existed at the time of scram. An analysis of hypothetical control rod insertion suggested that the oscillations could have been suppressed by only a few control rods. Analyses of a core destabilized by various parameters were also carried out to examine thermal margin sensitivity during the oscillations. The results showed that, regardless of which parameters were assumed to make the core unstable, thermal margin changes were substantially smaller than the initial margin expected under operation conditions to cause an instability.

An experimental study on rewetting phenomena in transient conditions of bwrs

Abstract It is known that rod temperature rise after boiling transition (BT) is not excursive and that the peak cladding temperature (PCT) is suppressed by rewetting to return to nucleate boiling, even if BT occurs under severe conditions exceeding abnormal operational transients for a BWR. The purpose of this study is to develop and verify the rewetting correlation. The rewetting correlation was developed based on single rod data, as a function of quality, mass flux, pressure and heat flux. The transient thermal-hydraulic code used in the BWR design analysis (SCAT) with this rewetting correlation was compared with transient rod temperature result after the occurence of BT obtianed by the 8×8 and 4×4 rod bundle. It is concluded that the transient code with the developed rewetting correlation predicts the PCT conservatively, and the rewetting time well.

Development of advanced core noise monitoring system for BWRS

A BWR core noise monitoring system is developed for addressing core anomaly problems in future advanced core operation. In order to monitor in-core status from a limited number of signals, various up-to-date signal processing algorithms are introduced to compensate for a lack of information. These algorithms, such as independent component analysis, factor analysis and model based parameter estimation, are demonstrated to be effective through real plant data analysis to evaluate core and regional stability index, reactivity coefficients and core flow rate. Through these practices, we demonstrate that the core noise monitoring system is an effective general platform for providing a variety of monitoring tools to meet the requirements in future advanced core operation.

Development of an on-line reactor stability monitoring system in a boiling water reactor

Abstract Reactor stability is an important topic in BWR plant design and operation. The objective of the present study is to establish a method for estimating the reactor stability in an operating BWR by measuring process noise signals, such as neutron flux or core flow rate. In the present study, several methods, based on autoregressive modeling, are presented and their reliability is quantitatively evaluated, using both simulated and real plant noise data. On the basis of these examinations, an on-line stability monitoring system was developed. Through field tests on the system over a year in an operating BWR, the validity of the system has been confirmed. Also, valuable data about reactor stability has been obtained.

Experience of on-line surveillance at ONAGAWA-1 BWR plant

Abstract An on-line system was installed on ONAGAWA-1 to monitor plant situations and to accumulate basic data for about 4 years, from start up test to 3rd cycle operation. Some kinds of anomalies were undergone during that period. Anomalies can be classified into several categories by growing modes of phenomena. It was verified that the system is useful to identify anomaly cause for the various kinds of anomaly modes. Some examples of experiences are introduced and discussed here.

Study on Solid-Liquid Two-Phase Flow on PWR Sump Clogging Issue

It has been a concern that sump screen clogging would occur in pressurized water reactors (PWRs) in the case of a loss-of-coolant accident (LOCA), because two-phase jet flow would strip off thermal insulation from the piping and wash down the broken and fragmented debris to sump screens. It is necessary for the evaluation of the effectiveness of sump screens to estimate the amount of transported debris from a break position to sumps. In general, conservative logic trees have been used to determine debris transport rates. Realistic debris transport evaluation is useful for considering measures and rational decision making in licensing. The purpose of this study is to develop a debris transport evaluation model and to apply the model to this issue. We developed a solid-liquid multiphase model that is capable of simulating debris transport, settling, and resuspension. The model is able to treat solid particles of different sizes, which are smaller than uniform-sized liquid particles. This approach contributes to reducing the calculation cost in a large-scale simulation. The model and a turbulence model were implemented into a code based on the moving particle semi-implicit (MPS) method. Several open-channel hydraulic experiments with fibrous debris were conducted. The code named SANSUI 2.0 was validated by the comparison of the analytical results with experiments. This method was applied to the debris transport analysis of a full-scale PWR containment vessel floor, and the debris transport behavior was evaluated.

Development of Advanced Core Noise Monitoring System for a Boiling Water Reactor

For an efficient nuclear power plant operation, advanced core design is very important. However, advances in design may result in approaching unexplored areas where unforeseen events are increasingly common. The purpose of this paper is to describe a BWR core noise monitoring system for addressing core anomaly problems in future advanced core operation. To monitor in-core status from a limited number of signals, we introduce various new algorithms to compensate for a lack of information. We demonstrate that up-to-date signal processing technologies, such as independent component analysis, nonlinear principal component regression, factor analysis and neural network, are effective tools for core monitoring. Also, the combination of the first principle or design model and the empirical fitting model is applied to estimate unobserved in-core state variables. Several concrete examples, such as stability monitoring, reactivity coefficient estimation, core flow estimation and in-core signal validation, are shown based on long-term monitoring experiences in real BWR plants. Through these examples, we demonstrate that the core noise monitoring system is a general platform for providing a variety of monitoring tools to meet the requirements of an advanced core operation strategy.

Proving Test on Thermal-hydraulic Performance of High Burnup 8*8 Fuel Assembly for BWR.

Nuclear Power Engineering Corporation (NUPEC) has completed the proving test for thermal-hydraulic performance of high burnup 8×8 fuel assembly. The high-burnup 8×8 fuel is a new type of BWR fuel assembly. New concepts such as ferrule type spacer and a large diameter central water rod are introduced in order to improve the thermal margin and reactor shut down margin for the higher uranium enrichment. The purpose of this test is to prove the thermal-hydraulic performance of this fuel and to verity the current BWR thermal-hydraulic design base.The tests were carried out in an out-of pile test facility that can simulate the high pressure and high temperature conditions of BWRs. An electrically heated test assembly simulated a BWR fuel bundle on full scale.The current BWR design method was compared with measured thermal-hydraulic data (data for critical power under steady state and unsteady state conditions, pressure drop and flow induced vibration of fuel rod). It was verified that the current BWR core design base has the enough reliability.