Kwi-Seok Ha

Development of a multi-dimensional thermal-hydraulic system code, MARS 1.3.1

Abstract A multi-dimensional thermal-hydraulic system code MARS has been developed by consolidating and restructuring the RELAP5/MOD3.2.1.2 and COBRA-TF codes. The two codes were adopted to take advantage of the very general, versatile features of RELAP5 and the realistic three-dimensional hydrodynamic module of COBRA-TF. In the course of code development, major features of each code were consolidated into a single code first. The resulting source programs were rewritten in standard fortran 90, and then were restructured using modular data structures based on “derived type variables” and a new “dynamic memory allocation” scheme. In addition, the Windows graphics features were implemented for user friendliness. This paper presents the developmental activities up to mars version 1.3.1 including the code consolidation, the code restructuring and modernization, and the results of the developmental assessment.

ADVANCED SFR DESIGN CONCEPTS AND R&D ACTIVITIES

In order to meet the increasing demand for electricity, Korea has to rely on nuclear energy due to its poor natural resources. In order for nuclear energy to be expanded in its utilization, issues with uranium supply and waste management issues have to be addressed. Fast reactor system is one of the most promising options for electricity generation with its efficient utilization of uranium resources and reduction of radioactive waste, thus contributing to sustainable development. The Korea Atomic Energy Research Institute (KAERI) has been performing R&Ds on Sodium-cooled Fast Reactors (SFRs) under the national nuclear R&D program. Based on the experiences gained from the development of KALIMER conceptual designs of a pool-type U-TRU-10%Zr metal fuel loaded reactor, KAERI is currently developing Advanced SFR design concepts that can better meet the Generation IV technology goals. This also includes developing, Advanced SFR technologies necessary for its commercialization and basic key technologies, aiming at the conceptual design of an Advanced SFR by 2011. KAERI is making R&D efforts to develop advanced design concepts including a passive decay heat removal system and a supercritical CO 2 Brayton cycle energy conversion system, as well as developing design methodologies, computational tools, and sodium technology. The long-term Advanced SFR development plan will be carried out toward the construction of an Advanced SFR demonstration plant by 2028.

SIMULATION OF THE EBR-II LOSS-OF-FLOW TESTS USING THE MARS CODE

Abstract As part of the development of a safety analysis methodology for a liquid-metal reactor (LMR) in Korea, the Multidimensional Analysis of Reactor Safety (MARS) code was selected as a system transient safety analysis code. The Korea Atomic Energy Research Institute developed the MARS code to analyze safety and thermal-hydraulic phenomena related to a two-phase flow in the transients of water reactors a decade ago. The addition of thermal-hydraulic models related to liquid metal as a coolant and reactivity feedback models associated with the kinetics calculation of an LMR core is required for the application of the MARS to the transients of an LMR design. A table for various properties of liquid sodium, several heat transfer coefficients according to flow regimes and geometries, and the models for a pressure drop due to the wire spacers of the LMR core were newly implemented. The improved MARS code was verified through the analysis of three shutdown heat removal tests (SHRT)-17, -39, and -45 conducted in the Experimental Breeder Reactor (EBR)-II reactor. The SHRT-17 test involved a simultaneous loss of electrical power to all pumps and a reactor scram from 100% power and flow. Thus, the test simulated a thermal-hydraulic transition from a forced convection to the totally passive decay heat removal due to a natural circulation. SHRT-39 and SHRT-45 are loss-flow tests without a reactor scram. However, the pump coastdown periods and initial states of the plant are different from each other. Simulated results for the flow rate and temperature for an instrumented subassembly agree well with the experimental data.

Modeling of flow blockage in a liquid metal-cooled reactor subassembly with a subchannel analysis code

Abstract The local blockage in a subassembly of a liquid metal-cooled reactor (LMR) is of importance to the plant safety because of the compact design and the high power density of the core. To analyze the thermal-hydraulic parameters in a subassembly of a liquid metal–cooled reactor with a flow blockage, the Korea Atomic Energy Research Institute has developed the MATRA-LMR-FB code. This code uses the distributed resistance model to describe the sweeping flow formed by the wire wrap around the fuel rods and to model the recirculation flow after a blockage. The hybrid difference scheme is also adopted for the description of the convective terms in the recirculating wake region of low velocity. Some state-of-the-art turbulent mixing models were implemented in the code, and the models suggested by Rehme and by Zhukov are analyzed and found to be appropriate for the description of the flow blockage in an LMR subassembly. The MATRA-LMR-FB code predicts accurately the experimental data of the Oak Ridge National Laboratory 19-pin bundle with a blockage for both the high-flow and low-flow conditions. The influences of the distributed resistance model, the hybrid difference method, and the turbulent mixing models are evaluated step by step with the experimental data. The appropriateness of the models also has been evaluated through a comparison with the results from the COMMIX code calculation. The flow blockage for the KALIMER design has been analyzed with the MATRA-LMR-FB code and is compared with the SABRE code to guarantee the design safety for the flow blockage.

DEVELOPMENT OF THE MATRA-LMR-FB FOR FLOW BLOCKAGE ANALYSIS IN A LMR

The Multichannel Analyzer for Transient and steady-state in Rod Array - Liquid Metal Reactor for Flow Blockage analysis (MATRA-LMR-FB) code for the analysis of a subchannel blockage has been developed and evaluated through several experiments. The current version of the code is improved here by the implementation of a distributed resistance model which accurately considers the effect of flow resistance on wire spacers, by the addition of a turbulent mixing model, and by the application of a hybrid scheme for low flow regions. Validation calculations for the MATRA-LMR-FB code were performed for Oak Ridge National Laboratory (ORNL) 19-pin tests with wire spacers and Karlsruhe 169-pin tests with grid spacers. The analysis of the ORNL 19-pin tests conducted using the code reveals that the code has sufficient predictive accuracy, within a range of 5℃, for the experimental data with a blockage. As for the results of the analyses, the standard deviation for the Karlsruhe 169-pin tests, 0.316, was larger than the standard deviation for the ORNL 19-pin tests, 0.047.

COMPARISON OF THE DECAY HEAT REMOVAL SYSTEMS IN THE KALIMER-600 AND DSFR

A sodium-cooled demonstration fast reactor with the KALIMER-600 as a reference plant is under design by KAERI. The safety grade decay heat removal system (DHRS), which is important to mitigate design basis accidents, was changed in the reactor design. A loss of heat sink and a vessel leak in design basis accidents were simulated using the MARS-LMR system transient analysis code on two plant systems. In the analyses, the DHRS of KALIMER-600 had a weakness due to elevation of the overflow path for the DHRS operation, while it was proved that the DHRS of the demonstration reactor had superior heat transfer characteristics due to the simplified heat transfer mechanism.

Assessment study about the risk of Wilks’ formula for uncertainty quantification of design extension condition scenarios in prototype Gen-IV sodium fast reactor

ABSTRACT Sensitivity analysis and uncertainty quantification using Wilks’ formula and Monte Carlo for Unprotected Loss of Flow (ULOF) and Unprotected Transient OverPower (UTOP) accidents of prototype Gen-IV sodium-cooled fast reactor were performed. Multi-dimensional analysis for reactor safety for liquid metal reactors code calculations were conducted while simultaneously varying the values of all uncertain parameters according to their distribution using parallel computing platform integrated for uncertainty and sensitivity analysis to obtain uncertainty bands for Figures of Merit (FOM) – coolant, fuel centerline, and cladding temperature at the hottest fuel rod. To specify the uncertainty range of the parameters for each accident scenario, literature survey and expert judgments were consulted. By the sensitivity analysis, the importance ranking of 25 parameters in model identification and ranking table based on phenomena identification and ranking table was identified. Through Monte Carlo calculation, 95% upper limit and 95% confidence level were obtained, and about 2% and 5% under-prediction (risk) of FOM of ULOF and UTOP accidents using Wilks’ formula were confirmed, respectively.

Investigation of Wire Effect in a Wire-Wrapped 37-Pin Fuel Assembly

The wire effect in a wire-wrapped 37-pin fuel assembly mock-up of a Japanese loop-type sodium-cooled fast reactor (SFR), Monju, has been investigated through a numerical analysis using a general-purpose commercial computational fluid dynamics (CFD) code, CFX. Complicated and vortical flow phenomena in the wire-wrapped 37-pin fuel assembly were captured by a Reynolds-Averaged Navier-Stokes (RANS) flow simulation with a shear stress transport (SST) turbulence model. The CFD results show good agreement with Rehme’s friction factor correlation model, which can consider the number of wire-wrapped pins in the fuel assembly. Three-dimensional multi-scale vortex structures start to be formed by an interaction between secondary flows around each wire-wrapped pin. Large-scale and small-scale vortex structures are generated in the corner and edge, and interior sub-channel, respectively. The behavior of the large-scale vortex structures in the corner and edge sub-channel are closely related to the relative position between the hexagonal duct wall and the wire spacer. Regardless of the relative position between the adjacent rod and wire spacer, a small-scale vortex is axially developed in the interior sub-channels. Furthermore, a driving force on each wire spacer surface is closely related to the relative position between the hexagonal duct wall and wire spacer. It is expected that the multi-scale vortex structures in the fuel assembly play a significant role in the convective heat transfer characteristics.Copyright

The Analysis of Partial Flow Blockage Accidents for a Sodium Cooled Fast Reactor

Analyses were performed for flow blockage accidents postulated in a conceptual design of a 600 MWe demonstration sodium cooled fast reactor with 3 types of core designs, i.e., Uranium, *L-TRU (TRansUrium) and **M-TRU cores, using the MATRA-LMR-FB code. The analysis was addressed for the 6 sub-channel blockage which is a design basis event (DBE). The accidents for 24 and 54 sub-channel blockages were also analyzed to estimate the extent of the blockage size which could lead to sodium boiling or fuel melting. Three radial blockage positions were also taken into account in the analysis.In result, a higher maximum coolant temperature in the subassembly was obtained as the number of blocked subchannels increased. A recirculation region was usually developed right above the blockage for large blockage cases. The analysis results showed that a favorable safety margin was assured for the design basis event, i.e., the 6 sub-channel blockage accident. For the 24 and 54 sub-channel blockage cases, the peak cladding temperature limit was breached, and there was a case in which fuel melting could be threatened.* TRU extracted from LWR fuel** Mixture of L-TRU and TRU extracted from self-recycled SFR fuelCopyright

Pre-Test Analysis of Natural Circulation Test of PHENIX End-of-Life With the MARS-LMR Code

PHENIX, a prototype sodium-cooled fast reactor (SFR), has demonstrated a fast breeder reactor technology and also achieved its important role as an irradiation facility for innovative fuels and materials. In 2009 PHENIX reached its final shutdown and the CEA launched a PHENIX end-of-life (EOL) test program, which provided a unique opportunity to validate an SFR system analysis code. The Korea Atomic Energy Research Institute (KAERI) joined this program to evaluate the capability and limitation of the MARS-LMR code, which will be used as a basic tool for the design and analysis of future SFRs in Korea. For this purpose, pre-test analyses of PHENIX EOL natural circulation tests have been performed and one-dimensional thermal-hydraulic behaviors for these tests have been analyzed. The natural circulation test was initiated by the decrease of heat removal through steam generators (SGs). This resulted in the increase of intermediate heat exchanger (IHX) secondary inlet temperature, followed by a manual reactor scram and the decrease of secondary pump speed. After that, the primary flow rate was also controlled by the manual trip of three primary pumps. For the pre-test analysis the Phenix primary system and IHXs were nodalized into several volumes. Total 981 subassemblies in the core were modeled and they were divided into 7 flow channels. The active 4 IHXs were modeled independently to investigate the change of flow into each IHX. The cold pool was modeled by two axial nodes having 5 and 6 sub-volumes, respectively. The reactor vessel cooling system was modeled to match the flow balance in the primary system. The flow path of vessel cooling system was quite complicated. However, it is simplified in the modeling. For a MARS-LMR simulation, the dryout of SGs have been described by the use of the boundary conditions for IHTS as a form of time-to-temperature table. This boundary condition reflects the increase in IHTS temperature by SG dryout during the initial stage of the transient and the increase in heat removal by the opening of the two SG containments at 3 hours after the initiation of the transient. Through the comparison of the pre-analysis results with the prediction by other computer codes, it is found that the MARS-LMR code predicts natural circulation phenomena in a sodium system in a reasonable manner. The final analysis for validation of the code against the test data will be followed with an improved modeling in near future.Copyright