DEVELOPMENT OF A TWO-DIMENSIONAL THERMOHYDRAULIC HOT POOL MODEL AND ITS EFFECTS ON REACTIVITY FEEDBACK DURING A UTOP IN LIQUID METAL REACTORS
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- Nuclear Engineering and Technology
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- v.41 no.8
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- pp.1053-1064
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- 2009
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.
Core-wide temperature distribution in sodium-cooled fast reactor plays a key role in its decay heat removal process, however the prediction for temperature distribution is quite complex due to the conjugate heat transfer between the assembly flow and the inter-wrapper flow. Hybrid medium model has been proposed for conjugate heat transfer modeling in the core. The core is modeled with a Realistic modeled inter-wrapper flow and hybrid medium modeled assembly flow. To validate present model, simulations for a three-assembly model were performed with Realistic modeling, traditional porous medium model and hybrid medium model, respectively. The influences of Uniform/Non-Uniform power distribution among assemblies and the Peclet number within the assembly flow have been considered. Compared to traditional porous medium model, present model shows a better agreement with in Realistic modeling prediction of the temperature distribution and the radial heat transfer between the inter-wrapper flow and the assembly flow.
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
Numerous losses of decay heat removal capability have occurred at U during stutodwn while its significance to safety is needless to say. A study is carried out as an attempt to assess what could be done to lower the frequency of these events and to mitigate their consequences in the unlikely event that one occurs. The shutdown risk model is developed and analyzed using Event/Fault Tree for the typical pressurized water reactor. The human cognitive reliability (HCR) model, two-stage bayesian approach and staircase function model are used to estimate human reliability, initiating event frequency and offsite power non-recovery probability given loss of offsite power, respectively. The results of this study indicate that the risk of a Pm at shutdown is not much lower than the risk when the plant is operating. By examining the dominant accident sequences obtained, several design deficiencies are identified and it is found that some proposed changes lead to significant reduction in core damage frequency due to loss of cooling events.
An influence analysis on multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout is performed to compare the plant responses according to the number of ruptured u-tubes under the assumption of a total of 10 ruptured u-tubes. In all calculation cases, the transient behaviour of major thermal-hydraulic parameters, such as the discharge flow rate through the ruptured u-tubes, reactor header pressure, and void fraction in the fuel channels is found to be overall similar to that of the base case having a single SG with 10 u-tubes ruptured. Additionally, as the conditions of low-flow coolant with high void fraction in the broken loop continued, causing the degradation of decay heat removal, the peak cladding temperature (PCT) would be expected to exceed the limit criteria for ensuring nuclear fuel integrity. However, despite the same total number of ruptured u-tubes, because of the different connection configuration between the SG and pressurizer, a difference is foud in time between the pressurizer low-level signal and reactor header low-pressure signal, affecting the time to trip the reactor and to reach the PCT limit. The present study is expected to provide the technical basis for the accident management strategy for mSGTR transient conditions of CANDU-6 plants.
In this study, the removal of Cs and Tc from a simulated fission products (FP) solution which were co-dissolved with U during the oxidative-dissolution of spent fuel in a mixed carbonate solution of
In order to estimate the effect of flow distributors connected to an upper nozzle of CMT(Core Makeup Tank) on the thermal-hydraulic characteristics in the tank, a simplified 2 inch Small Break Loss of Coolant Accident(SBLOCA) was simulated by skipping the decay power and Passive Residual Heat Removal System(PRHRS) actuation. The CMT is a part of safety injection systems in the SMART (System Integrated Modular Advanced Reactor). Each test was performed with reliable boundary conditions. It means that the pressure distribution is provided with repeatable and reproducible behavior during SBLOCA simulations. The maximum flow rates were achieved at around 350 seconds after the initial opening of the isolation valve installed in CMT. After a short period of decreased flow rate, it attained a steady injection flow rate after about 1,250 seconds. This unstable injection period of the CMT coolant is due to the condensation of steam injected into the upper part of CMT. The steady injection flow rate was about 8.4% higher with B-type distributor than that with A-type distributor. The gravity injection during hot condition tests were in good agreement with that during cold condition tests except for the early stages.
The Ulchin 3&4, which are 2825 MWt PWRs, adopted Safety Depressurization System (SDS) to mitigate the beyond design basis event of Total Less of Feedwater(TLOFW). In this study the results and methodology of the analyses for the determination of SDS bleed capacity are discussed. The SDS design bleed capacity has been determined from the CEFLASH-4AS/REM simulation according to the following design criteria : 1) Each SDS flow path, in conjunction with one of two High Pressure Safety Injection (HPSI) pumps, is designed to have a sufficient capacity to prevent core uncovery if one SDS path is opened simultaneously with the opening of the Pressurizer Safety Valves (PSVs). 2) Both SDS bleed paths are designed to have sufficient total capacity with both HPSI pumps operating to prevent core uncovery if the Feed and Bleed (F&B) initiation is delayed up to thirty minutes from the time of the PSVs lift. To verify the results of CEFLASH-4AS/REM simulation a comparative analysis kas also been per-formed by more sophisticated computer code, RELAP5/MOD3. The TLOFW event without operator recovery and TLOFW event with F&B are analyzed. The predictions by the CEFLASH-4AS/REM of the transient too phase system behavior are in good qualitative and quantitative agreement with those by the RELAP5/MOD3 simulation. Both of the results of analyses by CEFLASH-4AS/REM and RELAP5/MOD3 have demonstrated that decay heat removal and core inventory make-up can be successfully accomplished by F&B operation during now event for the Ulchin 3&4.
The wall shear stress in the vicinity of end-to end anastomoses under steady flow conditions was measured using a flush-mounted hot-film anemometer(FMHFA) probe. The experimental measurements were in good agreement with numerical results except in flow with low Reynolds numbers. The wall shear stress increased proximal to the anastomosis in flow from the Penrose tubing (simulating an artery) to the PTFE: graft. In flow from the PTFE graft to the Penrose tubing, low wall shear stress was observed distal to the anastomosis. Abnormal distributions of wall shear stress in the vicinity of the anastomosis, resulting from the compliance mismatch between the graft and the host artery, might be an important factor of ANFH formation and the graft failure. The present study suggests a correlation between regions of the low wall shear stress and the development of anastomotic neointimal fibrous hyperplasia(ANPH) in end-to-end anastomoses. 30523 T00401030523 ^x Air pressure decay(APD) rate and ultrafiltration rate(UFR) tests were performed on new and saline rinsed dialyzers as well as those roused in patients several times. C-DAK 4000 (Cordis Dow) and CF IS-11 (Baxter Travenol) reused dialyzers obtained from the dialysis clinic were used in the present study. The new dialyzers exhibited a relatively flat APD, whereas saline rinsed and reused dialyzers showed considerable amount of decay. C-DAH dialyzers had a larger APD(11.70
The wall shear stress in the vicinity of end-to end anastomoses under steady flow conditions was measured using a flush-mounted hot-film anemometer(FMHFA) probe. The experimental measurements were in good agreement with numerical results except in flow with low Reynolds numbers. The wall shear stress increased proximal to the anastomosis in flow from the Penrose tubing (simulating an artery) to the PTFE: graft. In flow from the PTFE graft to the Penrose tubing, low wall shear stress was observed distal to the anastomosis. Abnormal distributions of wall shear stress in the vicinity of the anastomosis, resulting from the compliance mismatch between the graft and the host artery, might be an important factor of ANFH formation and the graft failure. The present study suggests a correlation between regions of the low wall shear stress and the development of anastomotic neointimal fibrous hyperplasia(ANPH) in end-to-end anastomoses. 30523 T00401030523 ^x Air pressure decay(APD) rate and ultrafiltration rate(UFR) tests were performed on new and saline rinsed dialyzers as well as those roused in patients several times. C-DAK 4000 (Cordis Dow) and CF IS-11 (Baxter Travenol) reused dialyzers obtained from the dialysis clinic were used in the present study. The new dialyzers exhibited a relatively flat APD, whereas saline rinsed and reused dialyzers showed considerable amount of decay. C-DAH dialyzers had a larger APD(11.70