Weitere Kapitel dieses Buchs durch Wischen aufrufen
To perform decommissioning of the Fukushima Daiichi NPP safely, it is very important to measure the components of the fuel debris. Therefore, a new nondestructive assay to identify and quantify the target nuclide in fuel debris using a pulsed-neutron source is under development in Kyoto University Research Reactor Institute.
We use the self-indication method for the nondestructive assay. This method is a neutron transmission method. The neutron transmission method is focused on resonance reactions (i.e., capture, fission) at the target nuclide. In the self-indication method, the transmitted neutrons from the sample are injected into an indicator. The indicator consists of a high-purity target nuclide. The transmitted neutrons are obtained by the time-of-flight (TOF) technique via resonance reactions in the indicator. The self-indication method has a high signal-to-noise (S/N) ratio compared to the conventional method.
In this study, numerical validation for the self-indication method to identify and quantify nuclides in a BWR-MOX pellet is described. The burn-up of the MOX pellet is 0 GWd/t, 10 GWd/t, 20 GWd/t, 30 GWd/t, 40 GWd/t, and 50 GWd/t. The 12-m measurement line in KUR-LINAC is simulated as a calculational geometry. Numerical calculations are carried out by continuous-energy Monte-Carlo code MVP2 with JENDL-4.0 as the nuclear data library. The burn-up calculations of the BWR-MOX pellet are performed by the deterministic neutronics code SARC 2006 with JENDL-4.0.
Numerical validation for application of the self-indication method is carried out. From the results, it is noted that the self-indication method has a good S/N ratio compared to the neutron transmission method for quantifying the amount of target nuclides in the fuel debris.
Working Party on Physics of plutonium fuels and innovative fuel cycles (2003) Physics of Plutonium fuels BWR MOX benchmark specification and results. OECD/NEA, ISBN: 92-64-19905-5
Okumura K et al (2007) SRAC 2006: a comprehensive neutronics calculation code system. JAEA-Data/Code, 2007–004
Shibata K et al (2011) JENDL-4.0: a new library for nuclear science and engineering. J Nucl Sci Technol 48(1):1–30 CrossRef
Nagaya Y et al (2005) MVP/GMVP II: general purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods. JAERI 1348, Japan Atomic Energy Research Institute
- Development of Nondestructive Assay of Fuel Debris of Fukushima Daiichi NPP (2): Numerical Validation for the Application of a Self-Indication Method
- Springer Japan
- Chapter 4
Systemische Notwendigkeit zur Weiterentwicklung von Hybridnetzen