• Nuclear Engineering and Technology
• /
• v.54 no.11
• /
• pp.4338-4347
• /
• 2022

The CUPID code was developed and is continuously updated in KAERI. Verification and validation (V&V) is mainly done for light water reactors (LWRs). This paper describes a benchmarking of the detailed mesh level compared with sub-channel level for application to pressurized heavy water reactors (PHWRs), even though component scale comparison for the PHWR moderator system was done once before. We completed a sub-channel level comparison between the CUPID code and the ASSERT code and a CUPID code analysis. Because the ASSERT code has already been validated with numerous experiments, benchmarking with the ASSERT code will offer us more trust on the CUPID code. The target channel has high power and thus high pressure deformation. The high power channel tends to have a high possibility of critical heat flux (CHF), because a high void fraction and quality in channel exit region appear. In this research, after determining the reference grid and T/H model, we compared the sub-channel level results of the CUPID code with those of the ASSERT code.

• Nuclear Engineering and Technology
• /
• v.52 no.9
• /
• pp.1974-1982
• /
• 2020

Numerical prediction of fission product plateout and circulating coolant activities under normal operating conditions is crucial in the design of a high temperature gas-cooled reactor (HTGR). The results are used for the maintenance and repair of the components as well as the safety analysis regarding early source terms under loss of coolant accident scenarios. In this work, a new computer code named POSCA (Plate-Out Surface and Circulating Activities) was developed based on a one-dimensional model to evaluate fission product plateout and circulating coolant activities within the primary circuit of a HTGR. The verification and validation of study for the POSCA code was done using available analytical results and two in-pile experiments (i.e., OGL-1 and VAMPYR-1). The results of the POSCA calculations show that POSCA is able to simulate plateout and circulating coolant activities in a HTGR with fast computation and reasonable accuracy.

• Journal of KIISE:Computing Practices and Letters
• /
• v.6 no.2
• /
• pp.161-172
• /
• 2000

In this paper, we introduce the development of a LOTOS-based tool, supporting formal methods, called ForTIA (A Formalism for Telecommunication and Information Systems). By using LOTOS, an ISO standard formal specification language, the user requirements and system models can be abstracted and represented formally. Therefore, the system can be validated and verified on the specifications, before implementations. ForTIA supports light-weight formal methods based on validation to be used in real industry. Key functions of ForTIA are simulation and C++ code generation. In simulation, tree based visual validation mechanism is provided and in code generation, the full C++ source code is generated to be used for system implementations.

• Nuclear Engineering and Technology
• /
• v.54 no.11
• /
• pp.4049-4061
• /
• 2022

TRISO-coated particle fuel is widely used in high temperature gas cooled reactors and other advanced reactors. The performance of coated fuel particle is one of the fundamental bases of reactor safety. The failure probability of coated fuel particle should be evaluated and determined through suitable fuel performance models and methods during normal and accident condition. In order to better facilitate the design of coated particle fuel, a new TRISO fuel performance code named FRAT (Fission product Release Analysis Tool) was developed. FRAT is designed to calculate internal gas pressure, mechanical stress and failure probability of a coated fuel particle. In this paper, FRAT was introduced and benchmarked against IAEA CRP-6 benchmark cases for coated particle failure analysis. FRAT's results agree well with benchmark values, showing the correctness and satisfactory applicability. This work helps to provide a foundation for the credible application of FRAT.

• Nuclear Engineering and Technology
• /
• v.56 no.4
• /
• pp.1125-1134
• /
• 2024

This paper presents a validation study of the subchannel analysis code SLTHEN used for the core thermal-hydraulic design of the Prototype Gen-IV sodium-cooled fast reactor (PGSFR). To assess the performance of the ENERGY model of SLTHEN, four liquid metal heating experiments conducted by ORNL, WARD, and KIT with hexagonal assemblies of wire-wrapped rod bundles were analyzed. These experiments were performed with 19-and 61-pin bundles and varying power distributions of axial and radial peaking factors up to 1.4 and 3.0, respectively. The coolant subchannel temperatures measured at different axial locations were compared with the SLTHEN predictions with the Novendstern, Chiu-Rohsenow-Todreas (CRT), and Cheng-Todreas (CT) correlations for flow split and mixing in wire-wrapped pin bundles. The results showed that the SLTHEN predicts the measured subchannel temperatures reasonably well with root-mean-square errors of ~10 % and maximum errors of ~20 %. It was also observed that the CRT and CT correlations consistently outperform the Novendstern correlation.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1999.05a
• /
• pp.23-23
• /
• 1999

• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.05a
• /
• pp.7-8
• /
• 2005

• Proceedings of the Korean Nuclear Society Conference
• /
• 2002.10a
• /
• pp.188-188
• /
• 2002

• Proceedings of the Korean Nuclear Society Conference
• /
• 2002.10a
• /
• pp.146-146
• /
• 2002

• Proceedings of the Korean Nuclear Society Conference
• /
• 2003.10a
• /
• pp.181-181
• /
• 2003