Won-Pyo Chang

Model development for analysis of the Korea advanced liquid metal reactor

Abstract The SSC-K code is under development for analysis of the Korea Advanced LIquid MEtal Reactor (KALIMER) design adopting a pool-type reactor in Korea. The SSC-L code which was originally developed at Brookhaven National Laboratory for analysis of a loop-type liquid metal reactor, is its precursory code. The main reason for the development is that SSC-L cannot be applied directly to the KALIMER design because its application is limited to only a loop-type reactor. The SSC-K code represents the core with multiple coolant channels incorporated with a point kinetics model for calculation of the reactivity feedback. It can provide detailed one-dimensional thermal-hydraulic simulations not only for the primary and secondary sodium coolant circuits, but also the steam/water circuit of the balance-of-plant. This paper presents an overview of the recent developments on the physical models for SSC-K. Those developments are concerned with the two-dimensional hot pool model for analysis of the thermal stratification phenomena in the hot pool, the model for the passive decay heat removal system, the sodium boiling model in the core, and other physical models necessary for the KALIMER analysis. It also demonstrates the analysis results for the unprotected accidents like unprotected transient over power, unprotected loss of flow, and unprotected loss of heat sink postulated in the preliminary KALIMER design. The major focus of these analyses is made on confirmation of the inherent safety characteristics for the reactivity feedback in the core.

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.

INHERENT SAFETY ANALYSIS OF THE KALIMER UNDER A LOFA WITH A REDUCED PRIMARY PUMP HALVING TIME

The 600 MWe, pool-type, sodium-cooled, metallic fuel loaded KALIMER-600 (Korea Advanced LiquId MEtal Reactor, 600 MWe) has been conceptually designed with an emphasis on safety by self-regulating (inherent/intrinsic) negative reactivity feedback in the core. Its inherent safety under the ATWS (Anticipated Transient Without Scram) events was demonstrated in an earlier study. Initiating events of an HCDA (Hypothetical Core Disruptive Accident), however, also need to be analyzed for assessment of the margins in the current design. In this study, a hypothetical triple-fault accident, ULOF (Unprotected Loss Of Flow) with a reduced pump halving time, is investigated as an initiator of a core disruptive accident. A ULOF with insufficient primary pump inertia may cause core sodium boiling due to a power-to-flow mismatch. If the positive sodium reactivity resulting from this boiling is not compensated for by other intrinsic negative reactivity feedbacks, the resulting core power burst would challenge the fuel integrity. The present study focuses on determination of the limit of the pump inertia for assuring inherent reactivity feedback and behavior of the core after sodium boiling as well. Transient analyses are performed with the safety analysis code SSC-K, which now incorporates a new sodium boiling model. The results show that a halving time of more than 6.0 s does not allow sodium boiling even with very conservative assumptions. Boiling takes place for a halving time of 1.8 s, and its behavior can be predicted reasonably by the SSC-K.

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

DEVELOPMENT OF A TWO-DIMENSIONAL THERMOHYDRAULIC HOT POOL MODEL AND ITS EFFECTS ON REACTIVITY FEEDBACK DURING A UTOP IN LIQUID METAL REACTORS

The existence of a large sodium pool in the KALIMER, a pool-type LMR developed by the Korea Atomic Energy Research Institute, plays an important role in reactor safety and operability because it determines the grace time for operators to cope with an abnormal event and to terminate a transient before reactor enters into an accident condition. A two-dimensional hot pool model has been developed and implemented in the SSC-K code, and has been successfully applied for the assessment of safety issues in the conceptual design of KALIMER and for the analysis of anticipated system transients. The other important models of the SSC-K code include a three-dimensional core thermal-hydraulic model, a reactivity model, a passive decay heat removal system model, and an intermediate heat transport system and steam generation system model. The capability of the developed two-dimensional hot pool model was evaluated with a comparison of the temperature distribution calculated with the CFX code. The predicted hot pool coolant temperature distributions obtained with the two-dimensional hot pool model agreed well with those predicted with the CFX code. Variations in the temperature distribution of the hot pool affect the reactivity feedback due to an expansion of the control rod drive line (CRDL) immersed in the pool. The existing CRDL reactivity model of the SSC-K code has been modified based on the detailed hot pool temperature distribution obtained with the two-dimensional pool model. An analysis of an unprotected transient over power with the modified reactivity model showed an improved negative reactivity feedback effect.

Scoping System Analysis of KALIMER-600 Design Concept

The Korea Atomic Energy Research Institute (KAERI) is developing KALIMER (Korea Advanced LIquid Metal Reactor), which is a sodium-cooled, metallic-fueled, pool-type reactor. The KALIMER-600 design concept (600 MWt) was selected as one of the reference GEN-IV sodium-cooled fast reactors (SFRs). The safety design philosophy of KALIMER-600 places maximum reliance on passive responses to abnormal and emergency conditions, and minimizes the need for active and engineered safety systems. KALIMER-600 utilizes the intrinsic negative reactivity feedback effect under design extended conditions where reactor scram failures are postulated. In order to assess the effectiveness of the inherent safety features, a scoping system analysis during unprotected overpower, loss of flow and under cooling events has been performed using the system-wide transient analysis code SSC-K. The present scoping analysis focuses on identification of enhanced safety design features that provide passive and self-regulating response to transient conditions and evaluation of safety margins. The results of the scoping analysis indicate an understanding of various inherent reactivity feedback mechanisms is very important in establishing design features. The analysis results show that the KALIMER-600 design concepts provide larger safety margins with respect to sodium boiling, fuel rod integrity, and structural integrity. The inherent safety can be enhanced by accounting for reactivity feedback mechanisms in the design process.Copyright

Assessment of RELAP/MOD3 using BETHSY 6.2TC 6-inch cold leg side break comparative test

This report presents the results of the RELAP5/MOD3 Version 7j assessment on BETHSY 6.2TC. BETHSY 6.2TC test corresponding to a six inch cold leg break LOCA of the Pressurizer Water Reactor(PWR). The primary objective of the test was to provide reference data of two facilities of different scales (BETHSY and LSTF facility). On the other hand, the present calculation aims at analysis of RELAP5/N4OD3 capability on the small break LOCA simulation, The results of calculation have shown that the RELAP5/MOD3 reasonably predicts occurrences as well as trends of the major phenomena such as primary pressure, timing of loop seal clearing, liquid hold up, etc. However, some disagreements also have been found in the predictions of loop seal clearing, collapsed core water level after loop seal clearing, and accumulator injection behaviors. For better understanding of discrepancies in same predictions, several sensitivity calculations have been performed as well. These include the changes of two-phase discharge coefficient at the break junction and some corrections of the interphase drag term. As result, change of a single parameter has not improved the overall predictions and it has been found that the interphase drag model has still large uncertainties.