Type Sodium-cooled Fast Reactor Research Articles

Resonance self-shielded multigroup atomic displacement cross section for assessing neutron irradiation damage

Accurate calculation of neutron-induced displacement damage is a key step for the subsequent investigations on neutron irradiation damage. Due to the use of multigroup displacement damage cross sections, the resonance self-shielding effect of neutron-induced nuclear reaction must be taken into account. In addition to the method proposed by the authors in 2019, two additional methods are proposed in the present work to consider the self-shielding effect on the calculation of neutron irradiation-induced displacement damage. The two newly proposed methods allow directly using self-shielded multigroup nuclear reaction cross sections without infinite-dilution cross sections, even though they are not as consistent as the first one in the case of weak resonance self-shielding effect. The resulting resonance self-shielding corrections on the displacement damage of 56Fe in a typical sodium-cooled fast reactor are quite close for the three different methods. The slight differences (within ∼ 2 %) are much smaller than the self-shielding corrections, which are about −11.3 % for the dominant neutron elastic scattering-induced damage and −10.4 % for the total displacement damage.

Phenomena identification and ranking table of station blackout accidents for China sodium cooled fast reactor

In order to satisfy analysis of safety requirements for China pool type sodium-cooled fast reactor, a phenomena identification and ranking table (PIRT) is developed for station blackout (SBO) accidents by 15 experts. The SBO accidents could be divided into two phases, i.e. phase 1: force circulation and phase 2: natural circulation, and the most important phenomena but maximum disagreements among experts are the gap conductance model between fuel and clad, the inter wrapper flow, the heat transfer between coolant and inter wrapper. Those phenomena should be carefully assessed and done more theorical and experiment analysis to develop more accurate and realistic model, to reduce uncertainty of safety analysis. The PIRT could be also used for developing verification and validation matrix for thermal hydraulic safety code, guiding researchers and designers to conduct experiments to close technological gap and giving related researcher and designer a comprehensive understanding for pool-type sodium-cooled fast reactor.

Analysis of the natural circulation test of PHENIX reactor by the THACS code

For the pool type sodium-cooled fast reactor (SFR), whether the primary circuit can establish a stable natural cycle after the reactor accident, and whether the primary natural cycle can take away the decay heat of the reactor core in the long-term, are the key issues in the design and safety analysis of SFR. In order to verify the natural circulation capacity of the primary circuit, a natural circulation test was carried out on pool type SFR Phenix by CEA before the reactor is decommissioned. One objective of this test is to verify the system codes used to simulate sodium natural convection phenomena.In this paper, the system code THACS is used to analyze Phenix natural circulation test, and the capacity of THACS to analyze the natural circulation of SFR is evaluated. Firstly, this paper introduces the components of Phenix reactor and the process of natural circulation test. Then, the models of program and the computational modeling of Phenix reactor are described. The calculation involves the whole process of the test: dry out of steam generators, scram, stop of pumps, two phases of natural circulation. Then, the results are compared with the test data, and the influence of inter-wrapper flow (IWF) on core outlet temperature is emphasized. The results show good applicability of THACS to analyze natural circulation in SFR, and the IWF can reduce the peak value of core outlet temperature at the beginning of natural cycle. Finally, the deviation between calculation and test results are analyzed, and the limitation of the models are evaluated.

Uncertainty quantification of sodium-cooled fast reactor based on the UAM-SFR benchmarks: From pin-cell to full core

To quantify the uncertainties in the neutronics calculations of sodium-cooled fast reactor (SFR), a series of sensitivity and uncertainty (S&U) analysis were performed on two typical SFRs, which were a large-size MOX-fueled core and a medium-size metallic fueled core. The analysis was done from the pin-cell level to full core level to investigate the uncertainty propagation in the two-step neutronics calculation scheme. The responses analyzed contained the kinf, homogenized cross-sections, sodium void reactivity (SVR), Doppler constant (KD) and control rod worth (CRW) at the sub-assembly level, and the keff and power distribution at the full core level. The direct numerical perturbation method was used to get the relative sensitivity coefficients and the ‘Sandwich Rule’ was used to get the relative uncertainties. The results showed that the relative uncertainties propagated linearly from pin-cell to full core for the kinf and keff. Relative uncertainties of reactivity like SVR were much bigger than the ones of kinf and keff. However, the relative uncertainty of power distribution was smaller than other responses. Differences were observed in the S&U analysis between the MOX fueled and metallic fueled SFRs. Softer spectrum of the large-size MOX fueled core caused bigger uncertainty compared to the metallic fueled core.

Numerical simulation of Post Accident Heat Removal in a typical pool-type SFR under severe core relocation scenario

Under a hypothetical core meltdown and relocation scenario in Sodium cooled Fast Reactor (SFR), there may be a chance for reactor main vessel failure if the core debris are allowed to settle at the bottom of this vessel. To avoid this, an in-vessel core catcher assembly is installed in pool type SFRs just above the bottom wall of the main vessel. The function of this assembly is to collect, retain and passively cool the core debris by assisting natural convection decay heat removal from these debris. In this work, the passive decay heat removal from the core debris settled on single-tray-type design of core catcher has been numerically simulated under different amounts of core relocation using a validated 2-D axisymmetric CFD model. The maximum temperature experienced by the debris bed, core catcher plate and the welded joint between the core support structure and the main vessel (i.e., triple point) are evaluated in order to estimate the amount of core relocation that can be safely handled by the primary containment system under Post Accident Heat Removal (PAHR) scenario. The numerical study revealed that the single-tray-type design of core catcher can potentially assist the passive cooling even under the settling of debris material with heat generation equivalent to 60% of the full core on the core catcher. In addition to this, the effect of the size of opening created on grid plate during core relocation on Post Accident Heat Removal (PAHR) is studied for different amounts of core relocation. From this numerical study, the thermal capability of single-tray type in-vessel core catcher of SFR is assessed. The predictions are useful in determining the safe thermal margins available under core relocation scenarios and extending the functionality of single-tray-type core catcher for handling Whole Core Meltdown Accident (WCA) scenario in future SFRs.

Advanced Burner Reactors with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization

This paper assesses the feasibility of designing seed-and-blanket (S&B) sodium-cooled fast reactor (SFR) cores to generate a significant fraction of the core power from radial thorium-fueled blankets that operate in the breed-and-burn (B&B) mode. The radiation damage on the cladding material in both seed and blanket does not exceed the presently acceptable constraint of 200 displacements per atom (dpa). The S&B core is designed to have an elongated seed (or driver) to maximize the fraction of neutrons that radially leak into the subcritical B&B blanket and reduce the neutron loss via axial leakage. A specific objective of this study is to maximize the fraction of core power generated by the B&B blanket that is proportional to the neutron leakage rate from the seed to the blanket. Since the blanket feed fuel is very inexpensive and requires no reprocessing and remote fuel fabrication, a larger fraction of power from the blanket will result in a lower fuel cycle cost per unit of electricity generated by the SFR core. It is found possible to design the seed of the S&B core to have a lower transuranics (TRU) conversion ratio (CR) than a conventional advanced burner reactor (ABR) core without deteriorating core safety. This is due to the unique synergism between a low CR seed and the B&B thorium blanket. The benefits of the synergism are maximized when using an annular seed surrounded by inner and outer thorium blankets. Two high-performance S&B cores are designed to benefit from the annular seed concept: (1) an ultra-long-cycle core having a CR = 0.5 seed and a cycle length of ~7 effective full-power years (EFPYs) and (2) a high-transmutation core having a TRU CR of 0.0. The TRU transmutation rate of the latter core is comparable to that of the reference ABR with a CR of 0.5, and the thorium blanket can generate close to 60% of the core power. Because of the high blanket power fraction along with the high discharge burnup of the CR = 0 seed, the reprocessing capacity per unit of core power required by this S&B core is only approximately 1/6th that of the reference ABR core with a TRU CR of 0.5. Although the seed fuel CR is nearly zero, the burnup reactivity swing is low enough to enable a cycle length of more than 4 EFPYs. This is attributed to a combination of reactivity gain in the thorium blankets over the cycle and the relatively high heavy metal inventory. Moreover, despite the very low leakage, the S&B cores feature a less positive coolant reactivity coefficient and large enough negative Doppler coefficient even when using nonfertile fuel for the seed, because of the unique physics properties of the 233U and Th in the thorium blankets. With the long cycles, the S&B SFR is expected to have a higher capacity factor, and therefore a lower cost of electricity, than conventional ABRs. The discharge burnup of the thorium blanket fuel is typically 70 MWd/kg such that the thorium fuel utilization is approximately 12 times that of natural uranium in light water reactors. A sensitivity study is subsequently undertaken to quantify the trade-off between the core performances and several design variables: amount of zirconium in the inert matrix seed fuel, active core height, coolant pressure drop, and radiation damage constraint. The effect of the criterion used for quantifying acceptable radiation damage is evaluated as well. It is concluded that a viable S&B core can be designed without significant deviation from typical SFR core design practices.

Development of probabilistic risk assessment methodology against extreme snow for sodium-cooled fast reactor

This paper describes snow probabilistic risk assessment (PRA) methodology development through external hazard and event sequence evaluations mainly in terms of decay heat removal (DHR) function of a sodium-cooled fast reactor (SFR). Using recent 50-year weather data at a typical Japanese SFR site, snow hazard categories were set for the combination of daily snowfall depth (snowfall speed) and snowfall duration which can be calculated by dividing the snow depth by the snowfall speed. For each snow hazard category, the event sequence was evaluated by event trees which consist of several headings representing the loss of DHR. Snow removal action and manual operation of the air cooler dampers were introduced into the event trees as accident managements. Access route failure probability model was also developed for the quantification of the event tree. In this paper, the snow PRA showed less than 10−6/reactor-year of core damage frequency. The dominant snow hazard category was the combination of 1–2m/day of snowfall speed and 0.5–0.75day of snowfall duration. Importance and sensitivity analyses indicated a high risk contribution of the securing of the access routes.

Validation of the finned sodium–air heat exchanger model in MARS-LMR

A Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR), for which the Korea Atomic Energy Research Institute (KAERI) has designed a pool type sodium-cooled fast reactor, has a decay heat removal system (DHRS). The DHRS consists of sodium-to-sodium and sodium-to-air heat exchangers and their connecting pipes. There are four loops, which have two active and two passive type sodium–air heat exchangers, that are called a finned-tube sodium-to-air heat exchanger (FHX) and a helical-tube sodium-to-air heat exchanger (AHX), respectively. Recently, Zukauskaus’s air–sodium heat transfer models have been added in MARS-LMR, which is a safety analysis code for the PGSFR. In this study, to validate the newly added heat transfer models for the FHX, two experiments are selected: one is a performance test for a sodium to air heat exchanger (AHX) in the steam generator test facility (SGTF) for the prototype fast breeder reactor (PFBR), and the other is a start-up test with a dump heat exchanger (DHX) in JOYO. All validation results indicate that Zukauskaus’s correlation slightly over-estimates the heat transfer. One possible reasons is a smaller number of rows in the test bundle, which was already mentioned by Zukauskaus and Karni. The RMS values for the prediction of sodium temperature for the PFBR and the JOYO are 16.25% and 27.5%, respectively. When a correction factor is applied, their RMS values improve to 3.19% and 20.71%, respectively. In addition, the MARS-LMR’s prediction for the JOYO shows a much better accuracy with RMS of 7.45% when corrected sodium flow rates are applied.

Event sequence assessment of deep snow in sodium-cooled fast reactor based on continuous Markov chain Monte Carlo method with plant dynamics analysis

ABSTRACTMargin assessment of a nuclear power plant against external hazards is one of the most important issues after Fukushima Dai-ichi Nuclear Power Plant Accident. In this paper, a new approach has been developed to assess the plant status during external hazards and countermeasures against them in operation quantitatively and stochastically. A continuous Markov chain Monte Carlo (CMMC) method is applied and coupled with a plant dynamics analysis. In the CMMC method, a subsequence plant status is determined by the latest state (Markov chain) and the status is evaluated from the plant dynamics analysis. A failure or success of safety function of plant component is also evaluated stochastically based on a latest state of plant or hazard. A numerical investigation of plant dynamics analysis against a snow hazard is also carried out in a loop type sodium-cooled fast reactor so as to assess the margin against the hazard.

Estimation of evaporation rate of cesium from hot sodium pool to inert cover gas of typical SFR

Estimation of evaporative release of Cesium-137 (Cs-137) from the hot pool of Sodium cooled Fast Reactor (SFR) under normal operating condition and hypothetical core meltdown accident scenario is important in safety analysis due to the radiological consequences associated with Cs release. A methodology has been developed to estimate the cesium evaporation rate from hot pool to inert cover gas region of typical SFR under fuel pin clad failure scenario based on the temperature and concentration dependent activity coefficient of cesium estimated using the Flory–Huggins solution theory. The predicted equilibrium partition coefficient (KD) values of cesium are validated with theoretical and experimental data available in literature. It has been observed that the KD of cesium is decreasing with increase in Cs concentration. A correlation has been developed for KD as a function of temperature in the infinite dilute solution regime. The present methodology predicted Retention Factor (RF) values of Cs based on the estimated KD values are closer to the experimental results compared to the published results of empirical models in the SFR hot pool temperature range. The methodology has been used for evaluating the evaporative release of cesium along with sodium from hot pool to cover gas region of typical SFR under fuel pin clad failure scenario. These estimations are useful for evaluating the radioactive dose due to Cs transport from SFR sodium pool to its inert argon cover gas region.

Measurement of the Residual Stresses and Investigation of Their Effects on a Hardfaced Grid Plate due to Thermal Cycling in a Pool Type Sodium-Cooled Fast Reactor

In sodium-cooled fast reactors (SFR), grid plate is a critical component which is made of 316 L(N) SS. It is supported on core support structure. The grid plate supports the core subassemblies and maintains their verticality. Most of the components of SFR are made of 316 L(N)/304 L(N) SS and they are in contact with the liquid-metal sodium which acts as a coolant. The peak operating temperature in SFR is 550°C. However, the self-welding starts at 500°C. To avoid self-welding and galling, hardfacing of the grid plate has become necessary. Nickel based cobalt-free colmonoy 5 has been identified as the hardfacing material due to its lower dose rate by Plasma Transferred Arc Welding (PTAW). This paper is concerned with the measurement and investigations of the effects of the residual stress generated due to thermal cycling on a scale-down physical model of the grid plate. Finite element analysis of the hardfaced grid plate model is performed for obtaining residual stresses using elastoplastic analysis and hence the results are validated. The effects of the residual stresses due to thermal cycling on the hardfaced grid plate model are studied.

Numerical simulation of passive heat removal under severe core meltdown scenario in a sodium cooled fast reactor

A sequence of highly unlikely events leading to significant meltdown of the Sodium cooled Fast Reactor (SFR) core can cause the failure of reactor vessel if the molten fuel debris settles at the bottom of the reactor main vessel. To prevent this, pool type SFRs are usually provided with an in-vessel core catcher above the bottom wall of the main vessel. The core catcher should collect, retain and passively cool these debris by facilitating decay heat removal by natural convection. In the present work, the heat removal capability of the existing single tray core catcher design has been evaluated numerically by analyzing the transient development of natural convection loops inside SFR pool. A 1-D heat diffusion model and a simplified 2-D axi-symmetric CFD model are developed for the same. Maximum temperature of the core catcher plate evaluated for different core meltdown scenarios using these models showed that there is much higher heat removal potential for single tray in-vessel SFR core catcher compared to the design basis case of melting of 7 subassemblies under total instantaneous blockage of a subassembly. The study also revealed that the side pool of cold sodium plays a significant role in decay heat removal. The maximum debris bed temperature attained during the initial hours of PAHR does not depend much on when the Decay Heat Exchanger (DHX) gets operational, and it substantiates the inherent safety of the system. The present study paves the way for better understanding of the thermal hydraulics of Post Accident Heat Removal (PAHR) towards extending the functionality of core catcher for handling Whole Core Meltdown Accident (WCA) scenarios in future SFRs.

Basic verification of THACS for sodium-cooled fast reactor system analysis

For the transient thermal–hydraulic analysis of sodium cooled fast reactor (SFR), a universal code THACS is being developed in China. The one-dimensional modules for the main components of the primary loop, intermediate loop and heat removal loop have been compiled and the models are introduced briefly in this paper. For two-phase calculation, the multi-bubble model is used in the core module for sodium, while a simple incompressible water model is applied in the water side of steam generator. The point kinetic equations are employed to calculate the fission power of the multi-channel core, and various kinds of feedbacks like Doppler Effect, radial and axial expansions, density effect are taken into account. Preliminary verifications of the code are done based on the benchmark analyses of a BN-800 type SFR under loss of flow (LOF) accident, the EBR-II shutdown heat removal tests, and the 1-D simulation of the out-of-pile LOF experiment in KNS. Code-to-code and code-to-experiment comparisons show that THACS is proper for the transient analysis of SFR.

Mathematical modelling of performance of safety rod and its drive mechanism in sodium cooled fast reactor during scram action

Performance of safety rod and its drive mechanism which are parts of shutdown systems in sodium cooled fast reactor (SFR) plays a major role in ensuring safe operation of the plant during all the design basis events. The safety rods are to be inserted into the core within a stipulated time during off-normal conditions of the reactor. Mathematical modelling of dynamic behaviour of a safety rod and its drive mechanism in a typical 500MWe SFR during scram action is considered in the present study. A full-scale prototype system has undergone qualification tests in air, water and in sodium simulating the operating conditions in the reactor. In this paper, the salient features of the safety rod and its mechanism, details related to mathematical modelling and sensitivity of the parameters having influence on drop time are presented. The outcomes of the numerical analysis are compared with the experimental results. In this process, the mathematical model and the computer software developed are validated.

Estimation of radionuclide transport to reactor containment building under unprotected severe accident scenario in SFR

In the design of Sodium cooled Fast Reactor (SFR), Defense in Depth (DiD) approach is followed with multiple barriers to limit the release of radionuclides to the environment under accidental conditions. Under severe accident scenario, the multiple barriers provided as part of DiD, can fail one by one and finally the radionuclides may be released to Reactor Containment Building (RCB). Under severe accident scenario, the extent of radionuclides release into the environment has to be limited based on the permissible dose to the public at the site exclusion zone boundary. The radionuclides of concern for SFR under accidental release are noble gases and volatile fission products (Cs, I, Te). A lumped analytical methodology has been developed to estimate the source term in the core and transport of radionuclides to cover gas region/RCB under severe accident scenario in SFR. The methodology estimates the fission product inventory in the core based on saturation analysis method. In the event of severe accident, the core melts and fission products are released into the sodium pool. The transport of fission products (volatile and non-volatile) from pool to cover gas region/RCB takes place mainly by the short term phenomenon (noble gas bubble transport) and sodium spray fire release. The transport of Volatile Fission Products (VFPs) to cover gas region/RCB also takes place via the long term phenomenon (by evaporation). The evaporative transport of VFPs from hot pool to cover gas region is evaluated based on equilibrium partition coefficient (KD). The methodology is verified with the results for radionuclide inventory in SFR core, available in literature and their fractional release to the RCB. This methodology is useful for the evaluation of public dose at the site exclusion zone boundary of a typical SFR under severe accident scenario.

Transmutation of americium in a large sized sodium-cooled fast reactor loaded with nitride fuel

We have performed the transient analysis of a large sized sodium-cooled reactor loaded with nitride fuel modified by different fractions of americium. Unprotected loss-of-offsite power, unprotected loss-of-flow and unprotected transient-over-power accidents were considered as the postulated transient and simulated with the SAS4A/SASSYS code based on the geometrical model of a BN1200 type sodium-cooled fast reactor (SFR) with the rated power of 2900MWth. Safety parameters used in the transient simulation model were obtained from the SERPENT Monte Carlo calculations.If 100K margin to fuel melting was maintained, 2wt.% and 6wt.% Am content could be permitted for the nitride fuel loaded BN1200 operating with the maximum linear power rating of 37kW/m and 33kW/m respectively, providing Am transmutation rates of 3.9kg/TWhth and 13.7kg/TWhth. These transmutation rates are more than 6 times higher than the ones previously reported for oxide fuel loaded BN600 operating with the maximum linear power ratings of 37kW/m and metallic fuel loaded Integral Fast Reactor (IFR) operating with the maximum linear power rating of 33kW/m respectively, being their reference cases with permitting 1wt.% Am content.

Inertial impaction behavior of fuel particles under severe accident scenario in pool type sodium cooled fast reactor

During a highly unlikely event of energetic core disruptive accident in sodium cooled fast reactor (SFR), a high temperature and high pressure volume is formed due to neutronic excursion and expands rapidly (∼ms) into the sodium pool in the form of a fuel vapor bubble. This fuel vapor bubble further rises into the cover gas space, carrying radioactive materials along with it, which will contribute to the radiological source term in reactor containment building. In relation to this, a computer code PTRACK has been developed to describe the motion of an expanding and oscillating bubble and to evaluate trajectories of particles inside the bubble. The code considers adiabatic process for bubble expansion and solves particle momentum equation for evaluating the particle trajectories. PTRACK code has been validated against FAUST-1A benchmark experiments and the code predictions are in fairly good agreement with experimental results. Subsequently, a parametric analysis has been carried out to evaluate particle trajectories in a typical pool type SFR with initial fuel vapor bubble pressure of 1, 2 and 4MPa and for various particle diameters from 3 to 500μm. From this study, it has been observed that during fuel vapor bubble expansion, fuel particles are effectively retained by the sodium pool due to the inertial impaction behavior over a wide range of particle diameters and initial pressures. A parameter (inertial impaction parameter) has been defined to explain the above mentioned phenomenon. Results show that the fuel particle removal from the oscillating fuel vapor bubble due to inertial impaction phenomenon becomes significant when the inertial impaction parameter is close to unity.

Seismic modeling and analysis for sodium-cooled fast reactor

In this paper, the seismic analysis modeling technologies for sodium-cooled fast reactor (SFR) are presented with detailed descriptions for each structure, system and component (SSC) model. The complicated reactor system of pool type SFR, which is composed of the reactor vessel, internal structures, intermediate heat exchangers, primary pumps, core assemblies, and core support structures, is mathematically described with simple stick models which can represent fundamental frequencies of SSC. To do this, detailed finite element analyses were carried out to identify fundamental beam frequencies with consideration of fluid added mass effects caused by primary sodium coolant contained in the reactor vessel. The calculation of fluid added masses is performed by detailed finite element analyses using FAMD computer program and the results are discussed in terms of the ways to be considered in a seismic modeling. Based on the results of seismic time history analyses for both seismic isolation and non-isolation design, the functional requirements for relative deflections are discussed, and the design floor response spectra are proposed that can be used for subsystem seismic design.

Feasibility study on shipping cask for JSFR fresh fuel contained minor actinide

Japan Atomic Energy Agency (JAEA) initiated “Fast Reactor Cycle Technology Development (FaCT)” project since 2005 to accomplish necessary R&Ds goals for Japan sodium-cooled fast reactor (JSFR) cycle (the loop type sodium-cooled fast reactor with MOX fuel and its related fuel cycle system). Since the concept of multi-transuranic (TRU) fuel recycling has been adopted to reduce environmental burden and to enhance proliferation resistance, decay heat and radiation of MA, etc. became major consideration in designing fresh fuel shipping cask. The use of dry cask and helium gas filling for its high thermal conductivity has been selected, since water as a coolant is not preferable in order to avoid corrosion of fuel steel by sodium–water reaction products. Hence the helium gas cask with five fuel subassemblies storage was designed. To assure the thermal integrity, the cladding temperature of the fuel assembly has to be kept below the creep temperature to prevent any creep damage before fuel loaded into the core. In this study, a whole cask analysis using three-dimensional FEM model has been carried out to confirm absence of creep damage on fuel cladding and thermal integrity complying with the IAEA safety standards. Furthermore, detailed fuel subassembly analysis using two-dimensional FEM model estimated the maximum cladding tube temperature in detail to confirm the heat transport performance. In addition, the cask filled with water is assessed by Monte Carlo method for its integrity at criticality.

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.