Thermal Hydraulic Test of Advanced BWR Fuel Bundle With Spectral Shift Rod (SSR): Overview and Pre-Test Analysis by TRACG Code

2010 ◽  
Author(s):  
Takao Kondo ◽  
Masao Chaki ◽  
Yukiharu Ohga ◽  
Moriyasu Abe
Keyword(s):  
Water Level ◽  
Spectral Shift ◽  
Next Generation ◽  
Test Plan ◽  
Hydraulic Test ◽  
Test Analysis ◽  
Operating Condition ◽  
Water Reactor ◽  
Fuel Bundle ◽  
Shift Effect

Japanese national project of next generation light water reactor (LWR) development started in 2008. As one of its development items, the thermal-hydraulic test of spectral shift rod (SSR) is planned. A new component called SSR, which replaces conventional water rod (WR) of boiling water reactor (BWR) fuel bundle, was invented to enhance the BWR’s merit, spectral shift effect for uranium saving. In SSR, water boils by neutron and gamma-ray direct heating and water level is formed as a boundary of the upper steam region and the lower water region. This SSR water level can be controlled by core flow rate, which amplifies core void fraction change, resulting in the amplified spectral shift effect. In this paper, its test plan overview and pre-test analysis by TRACG code is presented. The test plan was developed with the purpose of evaluating SSR thermal-hydraulic characteristics at the actual BWR operating condition (7MPa), such as the controllability of SSR water level, and obtaining data for the validation of calculation method. In the test plan, several types of SSR simulation which covers SSR design in both next generation BWR and conventional BWR were designed. Also test operating conditions such as thermal-hydraulic parameters are determined. In order to evaluate these test specifications, pre-test analysis by TRACG code was conducted. Analysis results of each parameter’s effect on SSR characteristics are consistent with SSR mechanism, which shows that the actual operating condition for SSR fuel is simulated well.

Thermal Hydraulic Test of Advanced Fuel Bundle With Spectral Shift Rod (SSR) for BWR: Steady State and Transient Test Results and Analysis

2012 ◽  
Author(s):  
Takao Kondo ◽  
Kazuaki Kitou ◽  
Masao Chaki ◽  
Yukiharu Ohga ◽  
Takeshi Makigami
Keyword(s):  
Steady State ◽  
Water Level ◽  
Simulation Analysis ◽  
Spectral Shift ◽  
Test Results ◽  
Analysis Method ◽  
Fuel Bundle ◽  
State Test ◽  
Shift Effect ◽  
Steady State Test

Japanese national project of next generation light water reactor (LWR) development started in 2008. Under this project, spectral shift rod (SSR) is being developed. SSR, which replaces conventional water rod (WR) of boiling water reactor (BWR) fuel bundle, was invented to enhance the BWR’s merit, spectral shift effect for uranium saving. In SSR, water boils by neutron and gamma-ray direct heating and water level is formed as a boundary of the upper steam region and the lower water region. This SSR water level can be controlled by core flow rate, which amplifies the change of average core void fraction, resulting in the amplified spectral shift effect. This paper presents the steady state test with varied SSR geometry parameters, the transient test, and the simulation analysis of these steady state and transient tests. The steady state test results showed that the basic functioning principle such as the controllability of SSR water level by flow rate was maintained in the possible range of geometry parameters. The transient test results showed that the change rate of SSR water level was slower than the initiating parameters. The simulation analysis of steady state and transient test showed that the analysis method can simulate the height of SSR water level and its change with a good agreement. As a result, it is shown that the SSR design concept and its analysis method are feasible in both steady state and transient conditions.

Film Boiling Behavior of a Pressurized Water Reactor Type Fuel Bundle

1981 ◽  
Vol 52 (1) ◽  
pp. 86-99 ◽  
Author(s):  
F. S. Gunnerson ◽  
D. T. Sparks ◽  
D. K. Kerwin
Keyword(s):  
Film Boiling ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Fuel Bundle ◽  
Reactor Type ◽  
Type Fuel

Application of Genetic Algorithm methodologies in fuel bundle burnup optimization of Pressurized Heavy Water Reactor

2015 ◽  
Vol 281 ◽  
pp. 58-71 ◽  
Author(s):  
M.L. Jayalal ◽  
Suja Ramachandran ◽  
S. Rathakrishnan ◽  
S.A.V. Satya Murty ◽  
M. Sai Baba
Keyword(s):  
Genetic Algorithm ◽  
Heavy Water ◽  
Water Reactor ◽  
Fuel Bundle ◽  
Heavy Water Reactor

Flood Risk Pattern Recognition Using Chemometric Technique: A Case Study In Kuantan River Basin

2014 ◽  
Vol 72 (1) ◽  
Author(s):  
Ahmad Shakir Mohd Saudi ◽  
Hafizan Juahir ◽  
Azman Azid ◽  
Mohd Khairul Amri Kamarudin ◽  
Mohd Fadhil Kasim ◽  
...  
Keyword(s):  
Neural Network ◽  
Artificial Neural Network ◽  
Water Level ◽  
River Basin ◽  
Stream Flow ◽  
Pearson Correlation ◽  
Suspended Solid ◽  
Test Analysis ◽  
Risk Class ◽  
Artificial Neural

Integrated Chemometric and Artificial Neural Network were being applied in this study to identify the main contributor for flood, predicting hydrological modelling and risk of flood occurrence at the Kuantan river basin. Based on the Correlation Test analysis, the relationship for Suspended Solid and Stream Flow with Water Level were very high with Pearson correlation of coefficient value more than 0.5. Factor Analysis had been carried out and based on the result, variables such as Stream Flow, Suspended Solid and Water Level turned out to be the major factors and had a strong factor pattern with the results of factor score with >0.7 respectively. Time series analysis was being employed and the limitation had been set up where the Upper Control Limit for Stream Flow, Suspended Solid and Water Level where at this level, it was predicted by using Artificial Neural Network (ANN) to be High Risk Class. The accuracy of prediction from this method stood at 97.8%.

The REBUS Experimental Programme for Burn-Up Credit

2004 ◽  
Author(s):  
Pierre D’hondt ◽  
Peter Baeten ◽  
Leo Sannen ◽  
Daniel Marloye ◽  
Benoit Lance ◽  
...  
Keyword(s):  
Water Level ◽  
Nuclear Research ◽  
Research Centre ◽  
Reactor Physics ◽  
Fission Rate ◽  
Mox Fuel ◽  
Experimental Programme ◽  
Fuel Bundle ◽  
International Programme ◽  
Good Agreement

An international programme called REBUS for the investigation of the burn-up credit has been initiated by the Belgian Nuclear Research Centre SCK-CEN and Belgonucle´aire with the support of EdF and IRSN from France and VGB, representing German nuclear utilities and NUPEC, representing the Japanese industry. Recently also ORNL from the U.S. joined the programme. The programme aims to establish a neutronic benchmark for reactor physics codes in order to qualify the codes for calculations of the burn-up credit. The benchmark exercise investigates the following fuel types with associated burn-up: reference fresh 3.3% enriched UO2 fuel, fresh commercial PWR UO2 fuel and irradiated commercial PWR UO2 fuel (54 GWd/tM), fresh PWR MOX fuel and irradiated PWR MOX fuel (20 GWd/tM). The experiments on the three configurations with fresh fuel have been completed. The experiments show a good agreement between calculation and experiments for the different measured parameters: critical water level, reactivity effect of the water level and fission-rate and flux distributions. In 2003 the irradiated BR3 MOX fuel bundle was loaded into the VENUS reactor and the associated experimental programme was carried out. The reactivity measurements in this configuration with irradiated fuel show a good agreement between experimental and preliminary calculated values.

A Large-Scale Numerical Simulation on Bubbly and Liquid Film Flows in Narrow Fuel Channels

2005 ◽  
Author(s):  
K. Takase ◽  
H. Yoshida ◽  
Y. Ose ◽  
H. Akimoto
Keyword(s):  
Large Scale ◽  
Two Phase Flow ◽  
Reactor Core ◽  
Light Water Reactor ◽  
Phase Flow ◽  
Analysis Method ◽  
Two Phase ◽  
Light Water ◽  
Water Reactor ◽  
Fuel Bundle

In order to predict the water-vapor two-phase flow structure in a fuel bundle of an advanced light-water reactor, large-scale numerical simulations were carried out using a newly developed two-phase flow analysis method and a highly parallel-vector supercomputer. Conventional analysis methods such as subchannel codes need composition equations based on many experimental data. Therefore, it is difficult to obtain highly prediction accuracy on the thermal design of the advanced light-water reactor core if the experimental data are insufficient. Then, a new analysis method using the large-scale direct numerical simulation of water-vapor two-phase flow was proposed. The coalescence and fragmentation of small bubbles were investigated numerically and the bubbly flow dynamics in narrow fuel channels were clarified. Moreover, the liquid film flow inside a tight-lattice fuel bundle which is used to the advanced light-water reactor core was analyzed and the water and vapor distributions around fuel rods and a spacer were estimated quantitatively.

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