Korea Advanced Liquid Metal Reactor Research Articles

Multi dimensional analysis of Design Basis Events using MARS-LMR

Abstract KALIMER-600 (Korea Advanced Liquid Metal Reactor), which is a pool type SFR (Sodium-cooled Fast Reactor), was developed by KAERI (Korea Atomic Energy Research Institute). DBE (Design Basis Events) for KALIMER-600 has been analyzed in the one dimension. In this study, the one dimensional analyzed sodium hot pool is modified to a three dimensional node system, because the one dimensional analysis cannot represent the phenomena of the inside pool of a big size pool with many compositions, such as UIS (Upper Internal Structure), IHX (Intermediate Heat eXchanger), DHX (Decay Heat eXchanger), and pump. The results of the multi-dimensional analysis compared with the one dimensional analysis results in normal operation, TOP (Transient of Over Power), LOF (Loss of Flow), and LOHS (Loss of Heat Sink) conditions. First, the results in normal operation condition show the good agreement between the one and multi-dimensional analysis. However, according to the sodium temperatures of the core inlet, outlet, the fuel central line, cladding and PDRC (Passive Decay heat Removal Circuit), the temperatures of the one dimensional analysis are generally higher than the multi-dimensional analysis in conditions except the normal operation state, and the PDRC operation time in the one dimensional analysis is generally longer than the multi-dimensional analysis. In particular, we can see the reverse and mixing flow phenomena in LOHS better than TOP and LOF, which affect the peak temperature in the one dimensional analysis at 330 s but is not observed in the multi-dimensional analysis. In addition, the different temperatures at the central line of fuel and cladding are shown due to different mass flow in the core. The two PDRCs show the almost same results in one and multi-dimensional analysis, respectively. However, in the comparison between the one and multi-dimension, the heat removal capability of PDRC in the multi-dimensional analysis is better than the one dimensional analysis at the beginning of calculation. The aspects change at the ending of calculation due to the different mass flow in PDRC and over flow sodium temperature. In wrap up, the sodium temperatures of the core in the one dimensional analysis are generally higher than the multidimensional analysis. The temperature peak point in the one dimensional analysis under LOHS does not show in the multi-dimensional analysis, because the reverse and mixing flow pattern between the hot pool and the core outlet region. Even thought there are some different results between the two different dimensional analyses, we can say that the KALIMER-600 system has the safety margins.

Comparison of various design parameters’ effects on the early-stage cooling performance in a sodium-cooled fast reactor

A numerical simulation for an investigation into the effects of three design parameters, heat removal rate from an RV (Reactor Vessel), the thermal center difference and the pump inertia force on early-stage cooling of the reactor pool in the KALIMER (Korea Advanced Liquid Metal Reactor)-600 during a loss of normal heat sink accident has been done. Based on the design values of the KALIMER-600, the thermal–hydraulic calculations for a steady and a transient state have been done using the COMMIX-1AR/P code. The effects of heat removal rate from the RV were evaluated by modulating the CCS (Cavity Cooling System) inlet velocity from 0.5% to 400% in comparison to reference value. The effects of the thermal center difference were evaluated by changing the IHX (Intermediate Heat eXchanger) and related structures’ elevation. The pump inertia force effects were evaluated based on the analysis of reactors employing various flywheels, which had different CDTs (CoastDown flow Times) that ranged from 0 (without a flywheel) to 300 s. The transient analysis was limited to early-stage cooling during a reactor trip before the onset of an overflow into the DHX (Decay Heat eXchanger) support barrel. It was found that the heat removal rate via CCS has little effects on early-stage cooling performance. The lifting IHX elevation and employing moderate coastdown flow contribute to improve it. It is expected that the appearance of reversed flow or a rapid increase of core peak temperature could be avoided by adopting those design modifications.

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.

Investigation into the effects of a coastdown flow on the characteristics of early stage cooling of the reactor pool in KALIMER-600

A numerical investigation into the effect of a coastdown flow on the early stage cooling of the reactor pool in Korea Advanced Liquid Metal Reactor (KALIMER)-600 during a loss of normal heat sink accident has been carried out. Based on the design values of KALIMER-600, thermal-hydraulic calculations for steady and transient states have been done using the COMMIX-1AR/P code. Coastdown flow effect was evaluated based on a transient analysis of reactors employing various flywheels, which had coastdown flow time (CDT) values ranging from 0 (without a flywheel) to 300 s. The transient analysis has been done from a reactor trip to the onset of an overflow into the DHX support barrel. It was found that the coastdown flow range could be divided into three zones, based on its effect. Among them an excessive core coolant peak temperature and a reversed flow at the core region were observed for a medium coastdown flow range. The medium ranged coastdown flow induces the development of a high density layer near the core exit. This layer contributes to the development of an adverse effect in the core coolant flow, and finally results in increasing the core peak temperature. It was also found that the initiation of heat removal by DHX could be accelerated by the increase of the CDT, although it needs a large flywheel. From this analysis the best CDT is determined to be 25 s.

펌프 회전차의 관성모멘트 제공에 의한 KALIMER-600 원자로 풀 과도 성능 분석

The effect of an inertia moment of a pump flywheel on the thermal-hydraulic behaviors of the KALIMER-600(Korea Advanced LIquid MEtal Reactor) reactor pool during an early-phase of a loss of normal heat sink accident was investigated. The thermal-hydraulic analyses for a steady and a transient state were made by using the COMMIX-1AR/P code. In the present analysis a quarter of the reactor geometry was modeled in a cylindrical coordinate system, which includes a quarter of a reactor core and a UIS, a half of a DHX and a pump and a full IHX. In order to evaluate the effects of an inertia moment of the pump flywheel, a coastdown flow whose flow halving time amounts to 3.69 seconds was supplied to a natural circulation flow in the reactor vessel. Thermal-hydraulic behaviors in the reactor vessel were compared to those without the flywheel equipment. The numerical results showed a good agreement with the design values in a steady state. It was found that the inertia moment contributes to an increase in the circulation flow rate during the first 40 seconds, however to a decrease of it there after. It was also found that the flow stagnant region induced by a core exit overcooling decelerated the flow rate. The appearance of the first-peak temperature was delayed by the flow coastdown during the initial stages after a reactor trip.

An Enhanced Code for the Safety Analysis of Pool-Type Sodium-Cooled Fast Reactors

The Super System Code of the Korea Atomic Energy Research Institute (SSC-K) has been developed for the transient analysis of the Korea Advanced LIquid MEtal Reactor (KALIMER) system. Recently, a detailed three-dimensional (3-D) core thermal-hydraulic model was developed to describe nonuniformities of radial temperature and flow within a subassembly and to decrease the uncertainties in the reactor safety margins during accident situations. The Shutdown Heat Removal Test-17 (SHRT-17) performed in the Experimental Breeder Reactor-II (EBR-II) and the postulated unscrammed events for the KALIMER conceptual design have been analyzed using a code system that has coupled a detailed 3-D core thermal-hydraulic model with SSC-K. The coupled code predicted behaviors for the experimental trends for the protected loss-of-flow SHRT-17. The KALIMER-150 design was adopted for a plant application of the same code system. Three events, unprotected transient overpower (UTOP), unprotected loss of flow (ULOF), and unprotected loss of heat sink (ULOHS) were analyzed, and the simulation results were compared to those obtained using another code system that has coupled the Safety Analysis Section SYStem (SASSYS)-1 code with the same detailed 3-D core thermal-hydraulic model. The results, calculated with SSC-K coupled with the detailed 3-D core thermal-hydraulic model showed good agreement with the calculated results of the SASSYS-1 coupled code system for the UTOP and ULOF; however, some discrepancies were shown in the results for the ULOHS. These were found to have occurred because of a difference of the modeling for the decay heat removal system and primary coolant inventory. Through these analyses, the coupled code system was validated in order to be available for the safety analysis of a liquid-metal reactor (LMR) plant.

Development and Experimental Verification of the Numerical Simulation Method for the Quasi-Steady SWR Phenomena in an LMR Steam Generator

A quasi-steady system analysis of the sodium-water reaction (SWR) phenomena in a liquid-metal reactor (LMR) was performed using the Sodium-water reaction Event Later Phase System Transient Analyzer (SELPSTA) computer simulation code. The code has been formulated by implementing various physical assumptions to simplify the complex SWR phenomena, and it adopts the long-term mass and energy transfer (LMET) model developed in the present study. The LMET model is based on the hypothesis that the system transient can be described by the pressure and temperature transient of the cover gas space, and it can be applied only to the reaction period characterized by bulk motion. To evaluate the feasibility of the physical model and its assumptions, a scale-down mock-up test was carried out, and it was demonstrated that the numerical simulation using the LMET model adequately replicates the overall phenomena of the experiment with reasonable understanding. Based on the findings, as a numerical example, the long-term system transient responses during the SWR event of the Korea Advanced LIquid MEtal Reactor (KALIMER) were investigated, and it was found that the long-term dynamic responses are strongly dependent on the design parameters and operational strategies. As a result, the numerical simulation method developed in the present study is practicable; furthermore, the SELPSTA code is useful to resolve the risk for the SWR event.

Computational Study of Thermal Striping in an Upper Plenum of Kalimer

A computational study of thermal striping in an upper plenum of the Korea Advanced Liquid-Metal Reactor (KALIMER) was performed. The primary objective of the present study was to find the distribution of the amplitude of temperature fluctuation in the hot pool of KALIMER. The computations were performed using the CFX-4 code with the differential stress and flux turbulence model and the Van Leer convection scheme. Two cases with different outlet velocity of the control rod fuel assemblies are considered. The distributions of the velocity vector, temperature, and temperature fluctuation were obtained from the calculation. In order to quantitatively understand the amplitude of temperature fluctuation at the bottom wall of the upper internal structure (UIS), the amplitude of the fluctuation of temperature in the radial and angular directions was investigated. The amplitude of temperature fluctuation at the UIS bottom plate was largely dependent on the magnitude of the outlet velocity of the control rod fuel assemblies. From the calculated results, it was found that the largest temperature fluctuation occurred at the radial edge of the UIS bottom in the KALIMER design. Since thermal striping is dependent on the amplitude of temperature fluctuations and frequency, the region of the UIS bottom edge needs to be analyzed with a detailed unsteady calculation.

Progressive buckling analysis for a cylindrical shell structure with the free edge subjected to moving thermal cycles

In the KALIMER (Korea Advanced Liquid Metal Reactor) design, the reactor baffle structure is adopted to prevent the hot pool sodium from directly contacting the reactor vessel and to guide the hot sodium overflow in severe transient operating conditions. The parts in the vicinity of the hot pool free surface region could be repeatedly subjected to a moving axial temperature gradient and this might result in thermal ratcheting deformation. In this paper, the progressive thermal buckling behaviour following thermal ratcheting due to the moving axial temperature gradients in a cylindrical shell structure with an open free edge is investigated using numerical inelastic analysis with Chaboche's model. To do this, the analyses of the moving temperature distribution are carried out with a simple model and the severe moving axial temperature gradients are assumed to be sufficient for the evolution of thermal ratcheting.

Numerical Investigation on the Long-term System Transient Response of a SWR Event in a LMR

In order to investigate the later phase of a sodium-water reaction (SWR) event, the code SELPSTA (Sodium-water reaction Event Later Phase System Transient Analyzer) has been developed and the analysis for the long-term system dynamic responses of a SWR event in KALIMER (Korea Advanced Liquid MEtal Reactor) has been made. The SELPSTA code uses the very simple analysis model applied only to the reaction period characterized by a bulk motion, and makes the very quick and concise computation possible. The code reasonably predicts the quasi-steady system transients and has the superiority in the aspect that the various design parameters or operational characteristics are flexibly applicable. In the long-term period of a SWR event, the system dynamic responses analyzed by the code totally depend on the system design parameters such as the breaking pressure of the rupture disk, the variation of the steam injection rate and the sodium drain tank pressure,etc. Based on these analyses results, it is expected that the numerical quantification method of the SELPSTA code is practicable for the long-term system transient analysis and also makes the design of a pressure relief system against a SWR event in a liquid metal reactor (LMR) possible.

Numerical Investigation on the Long-term System Transient Response of a SWR Event in a LMR

In order to investigate the later phase of a sodium-water reaction (SWR) event, the code SELPSTA (Sodium-water reaction Event Later Phase System Transient Analyzer) has been developed and the analysis for the long-term system dynamic responses of a SWR event in KALIMER (Korea Advanced Liquid MEtal Reactor) has been made. The SELPSTA code uses the very simple analysis model applied only to the reaction period characterized by a bulk motion, and makes the very quick and concise computation possible. The code reasonably predicts the quasi-steady system transients and has the superiority in the aspect that the various design parameters or operational characteristics are flexibly applicable. In the long-term period of a SWR event, the system dynamic responses analyzed by the code totally depend on the system design parameters such as the breaking pressure of the rupture disk, the variation of the steam injection rate and the sodium drain tank pressure,etc. Based on these analyses results, it is expected that the numerical quantification method of the SELPSTA code is practicable for the long-term system transient analysis and also makes the design of a pressure relief system against a SWR event in a liquid metal reactor (LMR) possible.

Model development for analysis of the Korea advanced liquid metal reactor

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.

Design of LMR reactor structures in the vicinity of hot pool free surface regions subjected to moving elevated temperature cycles

In this paper, the structural damage on the reactor structures in the vicinity of the hot pool free surface regions are investigated to check design validity and to propose a modified reactor baffle (RB) design with a Y-piece type cylindrical thermal baffle (TB) structure. Detailed thermal and stress analyses for an envelope thermal transient load of the Passive Safety Decay Heat Removal System event are carried out for both the original design and the modified design. From the results of the analyses, the maximum stresses for the RB and the reactor vessel occur in the level of hot pool free surface regions after primary pump flow coastdown. The modified RB design with the TB structure leads to a significant reduction of the thermal stress, the inelastic creep ratcheting strain, and the creep-fatigue damage compared with those of the original design. Finally, it is verified that the proposed design modification is acceptable in the Korea Advanced Liquid Metal Reactor structural design for hot pool free surface regions.

Buckling limit evaluation for reactor vessel of KALIMER liquid metal reactor under lateral seismic loads

The purpose of this paper is to evaluate the buckling limit of a conceptually designed reactor vessel of the Korea advanced liquid metal reactor (KALIMER). For evaluation of the critical buckling load, evaluation formulae and numerical simulation methods using finite element analysis are used. In the numerical simulation methods, eigenvalue buckling analysis, and non-linear buckling analysis with and without plasticity effects are carried out. From the calculated results of the critical buckling load in the KALIMER reactor vessel, it is shown that plasticity behavior significantly affects the buckling strength but the initial geometrical imperfections have little effect on it. In checking the buckling limits of the KALIMER reactor vessel for safe shutdown earthquake (SSE) lateral seismic load using the ASME B and PV Section III, Subsection NH, a non-seismic isolation design does not satisfy the buckling limit requirements but a seismic isolation design sufficiently satisfies the requirements. From the experimental verification tests performed for the cylindrical shells of the reduced scale, it is confirmed that the non-linear numerical simulation method used in the buckling strength evaluation can assure accurate results for buckling strength and buckling mode shape.

Dynamic characteristics of KALIMER IHTS hot leg piping system conveying hot liquid sodium

In this paper, the dynamic characteristics of the KALIMER (Korea Advanced LIquid MEtal Reactor) IHTS hot leg piping system conveying hot liquid sodium are investigated to eliminate the pipe supports for economic reasons. To do this, a 3-dimensional straight pipe element and a curved pipe element conveying fluid are formulated using the dynamic stiffness method of the wave approach and coded to be applied to any complex piping system. Using this method, the dynamic characteristics including the natural frequency, the frequency response functions, and the dynamic instability due to the pipe internal flow velocity are analyzed. As one of the design parameters, the vibration energy flow is also analyzed to investigate the disturbance transmission paths for the resonant excitation and the non-resonant excitations.

Seismic base isolation technologies for Korea advanced liquid metal reactor

This paper describes the status and prospects of the seismic base isolation technologies for Korea Advanced Liquid Metal Reactor (KALIMER). The research and development program on the seismic base isolation for KALIMER began in 1993 by KAERI under the national long-term R&D program. The objective of this program is to enhance the seismic safety, to accomplish the economic design, and to standardize the plant design through the establishment of technologies on seismic base isolation for liquid metal reactors. In this paper, tests and analyses performed in the program are presented.

A Preliminary Shielding Design of 150MWe Liquid Metal Reactor

A preliminary shielding design of the KALIMER (Korea Advanced Liquid Metal Reactor) was established by two-dimensional discrete ordinate radiation transport analyses. The shielding design has tentatively adopted to use two limits on the fast neutron fluence and the DPA (Displacement Per Atom) simultaneously as a base for the exposure limit for neutron damages to reactor structures. In addition, activities in the PSDRS air effluent and the IHX secondary sodium were also examined.The DORT two-dimensional transport code was used to evaluate the KALIMER shielding design. The reactor system was represented by four axial zones, each of which was modeled in the R-Z geometry. The KAFAX-F22 library was used in the analyses, which was generated from the JEF-2.2 of OECD/NEA for LMR applications by KAERI (Korea Atomic Energy Research Institute), and consists of 80 neutron and 24 gamma energy groups.The performance of the shielding design is compared against the shielding design criteria. The results indicate that the support barrel, upper grid plate, and other reactor structures meet both the maximum neutron fluence and DPA limits established in the shielding design criteria. Activities of the air effluent in the PSDRS were also evaluated and are shown to satisfy the effluent concentration limits in 10 CFR Part 20.It is found that maximum DPAs show larger margins than the case when the neutron fluence limit is used as a design criterion. Therefore, the use of DPA as a shielding design criterion allows for a more flexibility in the LMR shielding design.

Evaluation of core nuclear analysis code system for LMR using measured physics parameters of BFS-73-1 critical assembly

As the first stage of the critical experiment plan for developing the Korea Advanced LIquid MEtal Reactor (KALIMER) core design, the basic neutronics characteristics of the BFS-73-1 critical assembly representing a uranium metal fueled benchmark core were investigated. The prediction capability of the core nuclear analysis code system, being developed as a standard analysis system in the KALIMER core design and analysis, has been evaluated against the BFS-73-1 experiment by comparing the calculated results to the measurements. The comparison shows that the effective multiplication factor was overpredicted by 0.3% in the C/E value. The 235U and 238U fission reaction rate distributions were underpredicted within a 2% discrepancy in the core region but appeared to be in poor agreement in the blanket region. The calculated values of spectral indices at the core center and the effective delayed neutron fraction, β eff, agreed with the measured ones within 2% and 1% discrepancies, respectively. The first order perturbation method predicted the sample reactivity worths for 235U, 238U, 239Pu and 10B within 7% but much larger discrepancies existed for the other materials. The calculated Doppler effect showed about an 8% overprediction.