Safety analysis of the reference research reactor MTR during reactivity insertion accident using the code MERSAT

The IAEA’s reference research reactor MTR-10 MW has been modeled using the code MERSAT. The developed MERSAT model consists of detailed representation of primary and secondary loops including reactor pool, bypass, main pump, heat exchanger and reactor core with the corresponding neutronics and thermalhydraulic characteristics. Following the successful accomplishment of the steady state operation at nominal power of 10 MW, reactivity insertion accident (RIA) for three different initial reactivity values of $1.5/0.5 s, $1.35/0.5 s and $0.1/1.0 s have been simulated. The predicted peaks of reactor power, hot channel fuel, clad and coolant temperatures demonstrate inherent safety features of the reference MTR reactor. Only in case of the fast RIA of $1.5/0.5 s, where the peak power of 133.66 MW arrived 0.625 s after the start of the transient, the maximum hot channel clad temperature arrives at the condition of subcooled boiling with the subsequent void formation. However, due to the strong negative reactivity feedback effects of coolant and fuel temperatures the void formation persists for a very short time so that thermalhydraulic conditions remained far from exceeding the safety design limits of thermalhydraulic instability and DNB. Finally, the simulation results show good agreement with previous international benchmark analyses accomplished with other qualified channel and thermalhydraulic system codes.     

CFD analysis of fast loss of flow accident in typical MTR reactor undergoing partial and full blockage: The average channel scenario

CFD investigation of loss of flow accident (LOFA) in typical MTR reactor undergoing partial and full blockage under the average channel condition is considered. The blockage scenarios considered in this work describe changes in the geometrical configuration of the flow channels as a result of thermal stresses or any other reason. That is the fuel plates of the average channel are assumed to buckle inwards along the plate height. As a result, the flow area decreases along the height of the channel until it achieves minimum in the middle. Three adjacent channels are simulated. With the area of the blocked channel decreases, that of the adjacent channel increases while the third channel remains unaltered. Blockage ratios considered in this work includes 0%, 20%, 40%, 50%, 60%, 80%, and full blockage. As a result of the change in the geometrical configuration of the flow channels, the hydraulic resistance also changes resulting in flow and heat transfer load to redistribute among the three channels. During the course of LOFA, the decay heat load is taken up by natural convection. While under the hot channel conditions, previous work showed that boiling is inevitable for even small blockage ratios. In this work maximum clad temperature is found to be under the boiling temperature at the operating pressure up to approximately 80% blockage ratio. For blockage ratio larger than 80%, the maximum clad temperature exceeds the boiling temperature indicating that boiling may occur.     

Effects of cooling channel blockage on fuel plate temperature in Tehran Research Reactor

In this study, the variation of the temperature distribution of the fuel plate in Tehran Research Reactor core was studied in case of coolant channels blockage. While the experimental method is not possible, both the analytical and simulation methods were used to obtain the more reliable data. The results show that one channel blockage will increase the fuel temperature to about 100%, but it does not lead to clad melt down still. With further calculation and simulation it is understood that if the coolant velocity drops to 90% of its nominal value, it may causes the clad melt-ing down. At least two channels with complete blockage even at the positions far from the core center can also melt down the clad.     

Analysis of partial and total flow blockage of a single fuel assembly of an MTR research reactor core

The main aim of the following study is to perform a safety analysis of the IAEA 10 MW MTR Pool type Research Reactor [IAEA-TECDOC-233, 1980. IAEA Research Reactor Core Conversion from the use of high-enriched uranium to the use of low enriched uranium fuels Guidebook] under flow blockage of a single Fuel Assembly (FA) conditions. Such event was rarely investigated in the open literature notwithstanding the fat that it constitutes a severe accident that may lead to local dryout and eventually to loss of the FA integrity. The transients herein considered are related to partial and total obstruction of the cooling channel of a single Fuel Assembly of the reactor core. This study constitutes the first step of a larger work, which consists in performing a 3D simulation using the Best Estimate coupled code technique. However, as a first approach the instantaneous reactor power is derived through the point kinetic approach of the used thermal-hydraulic system code.     

Updating of local blockage frequency in the reactor core of SFR and PRA on consequent severe accident in Monju

Fuel subassemblies of sodium-cooled fast reactors (SFRs) are densely arranged and have high power densities. Therefore, the local fault has been considered as one of the possible initiating events of severe accidents. In the conventional analyses for the license of Japanese prototype SFR (Monju), according to the local fault evaluation under the condition of one sub-channel flow blockage in the analyses of design basis accident (DBA), it was confirmed that the pin failures were limited locally without severe core damage. In addition, local flow blockage of 66% central planar in the subassembly was historically investigated as one of the beyond-DBAs. However, it became clear that these deterministic analyses were not based on a realistic assumption by experimental studies. Therefore, probabilistic risk assessment on local fault which was initiated from local flow blockage was performed reflecting the state-of-the-art knowledge in this study. As a result, damage propagation from local fault caused by local flow blockage in Monju can be negligible compared with the core damage due to anticipated transient without scram or protected loss of heat sink in the viewpoint of both frequency and consequence.     

Analysis of loss of coolant accident in MTR pool type research reactor

In MTR research reactors, heat removal is, safely performed by forced convection during normal operation and by natural convection after reactor shutdown for residual decay heat removal. However, according to the duration time of operation at full power, it may be required to maintain the forced convection, for a certain period of time after the reactor shutdown. This is among the general requirements for the overall safety engineering features of MTR research reactors to ensure a safe residual heat removal. For instance, in safety analysis of research reactors, initiating events that may challenge the safe removal of residual heat must be identified and analyzed.In the present work, it was assumed a total loss of coolant accident in a typical MTR nuclear research reactor with the objective of examining the core behavior and the occurrence of any fuel damage.For this purpose, the IAEA 10 MW benchmark core, which is a representative of medium power pool type MTR research reactors, was chosen herein in order to investigate the evolution of cladding temperature through the use of a best estimate thermalhydraulic system code RELAP5/mod3.2.     

Safety analysis of the IAEA reference research reactor during loss of flow accident using the code MERSAT

Using the thermal hydraulic code MERSAT detailed model including primary and secondary loop was developed for the IAEA's reference research reactor MTR 10 MW. The developed model enables the simulation of expected neutronic and thermal hydraulic phenomena during normal operation, reactivity and loss of flow accidents.Two different loss of flow accident (LOFA) have been simulated using slow and fast decrease time of core mass flow. In both cases the expected flow reversal from downward forced to upward natural circulation has been successfully simulated. The results indicate that in both accidents the limit of onset of subcooled boiling was not arrived and consequently no exceed of design limits in term of thermal hydraulic instability or DNB is observed. Finally, the simulation results show good agreement with previous international benchmark analyses accomplished with other qualified channel and thermal hydraulic system codes.     

Partial flow blockage analysis of the hottest fuel assembly in SNCLFR-100 reactor core

In this paper, we perform an unprotected partial flow blockage analysis of the hottest fuel assembly in the core of the SNCLFR-100 reactor, a 100 MW_(th) modular natural circulation lead-cooled fast reactor, developed by University of Science and Technology of China. The flow blockage shall cause a degradation of the heat transfer between the fuel assembly and the coolant potentially,which can eventually result in the clad fusion. An analysis of core blockage accidents in a single assembly is of great significance for LFR. Such scenarios are investigated by using the best estimation code RELAP5. Reactivity feedback and axial power profile are considered. The crosssectional fraction of blockage, axial position of blockage,and blockage-developing time are discussed. The cladding material failure shall be the biggest challenge and shall be a considerable threat for integrity of the fuel assembly if the cross-sectional fraction of blockage is over 94%. The blockage-developing time only affects the accident progress. The consequence will be more serious if the axial position of a sudden blockage is closer to the core outlet.The method of analysis procedure can also be applied to analyze similar transient behaviors of other fuel-type reactors.     

板状燃料组件流量分配CFD研究与优化

板状燃料组件被广泛应用于研究堆中,组件内的流量分配是设计时需要考虑的一项重要内容。计算流体动力学(Computational Fluid Dynamic,CFD)方法是研究流量分配的重要手段,但有限的计算资源限制了其在板状燃料组件流量分配研究中的推广。针对板状燃料组件冷却剂流道狭长、封闭的特点,提出了部分建模迭代求解的计算方式,将无流量分配组件与有流量分配组件两种工况下各流道流量的计算值与直接完整建模的结果进行了对比,最大误差分别为0.56%与0.81%。鉴于前者对计算资源的需求远小于后者,部分建模迭代求解可以作为板状燃料组件流量分配CFD研究的合理可信的优化方案。     

GPU加速的球床式氟盐冷却高温堆堆芯热工水力程序研发

球床式氟盐冷却高温堆(Pebble Bed Fluoride-salt Cooled High Temperature Reactor,PB-FHR)是一种先进的第四代反应堆。三维堆芯热工水力程序能够模拟具有复杂空间效应的工况,但计算耗时较高。图形处理器(Graphics Processing Unit,GPU)具有大量计算单元,可有效提高程序的计算速度。本文研发了GPU加速的PB-FHR堆芯热工水力程序(GPU-accelerated Thermal Hydraulic Code,GATH),采用非热平衡多孔介质模型建立堆芯物理模型,研究并实现了GPU高速求解算法。对PB-FHR的堆芯模型进行了热工水力分析,与商用计算流体力学软件ANSYS CFX的计算结果进行了对比,验证了程序的正确性。GPU加速性能分析的结果表明,程序整体的加速比率可达8.39倍,证明所研发的GPU求解算法能有效提升堆芯热工水力分析的计算效率。     

Flow blockage analysis of a channel in a typical material test reactor core

The course of the partial and total blockage of a channel in the IAEA 10 MW MTR pool type research reactor core (IAEA-TECDOC-233, 1980. IAEA Research Reactor Core Conversion from the use of high-enriched uranium to the use of low enriched uranium fuels Guidebook.) without scram is investigated. The analysis is performed with the best estimate code RELAP5/MOD3.3. The interaction of the obstructed channel and its adjacent channels has been taken account of. Results indicated that even when the flow channel has been totally blocked, there is still no boiling occurrence, and the fuel temperature is low enough to maintain its integrity. This work indicates that the consideration of the conjugate heat transfer in the obstructed channel during this transient is very important.     

Mass flow rate sensitivity and uncertainty analysis in natural circulation boiling water reactor core from Monte Carlo simulations

Our aim was to evaluate the sensitivity and uncertainty of mass flow rate in the core on the performance of natural circulation boiling water reactor (NCBWR). This analysis was carried out through Monte Carlo simulations of sizes up to 40,000, and the size, i.e., repetition of 25,000 was considered as valid for routine applications. A simplified boiling water reactor (SBWR) was used as an application example of Monte Carlo method. The numerical code to simulate the SBWR performance considers a one-dimensional thermo-hydraulics model along with non-equilibrium thermodynamics and non-homogeneous flow approximation, one-dimensional fuel rod heat transfer. The neutron processes were simulated with a point reactor kinetics model with six groups of delayed neutrons. The sensitivity was evaluated in terms of 99% confidence intervals of the mean to understand the range of mean values that may represent the entire statistical population of performance variables. The regression analysis with mass flow rate as the predictor variable showed statistically valid linear correlations for both neutron flux and fuel temperature and quadratic relationship for the void fraction. No statistically valid correlation was observed for the total heat flux as a function of the mass flow rate although heat flux at individual nodes was positively correlated with this variable. These correlations are useful for the study, analysis and design of any NCBWR. The uncertainties were propagated as follows: for 10% change in the mass flow rate in the core, the responses for neutron power, total heat flux, average fuel temperature and average void fraction changed by 8.74%, 7.77%, 2.74% and 0.58%, respectively.     

一种液态燃料熔盐堆堆芯流量分配设计

在液态燃料熔盐堆(Molten salt reactor,MSR)热工水力设计中,为实现堆芯径向功率展平需对堆芯流量分配进行设计,使得堆芯进口流量分布正比于释热量分布,而下腔室结构和流场分布对堆芯流量分配起决定性作用。利用FLUENT软件对堆芯三维流场进行模拟,通过调节下腔室结构和流量分配装置,对下腔室流场分布进行优化,最终实现堆芯流量合理分配。数值模拟结果表明,喇叭状下腔室比椭球形下腔室熔盐通道流量标准差降低4.2%,设置流量分配板熔盐通道流量标准差降低29.2%;改变下腔室结构和设置流量分配装置能够较好调节流量分配和功率分布匹配性,该结果可为液态熔盐堆堆芯优化设计提供依据。     

Thermal-hydraulic analysis of a fluoride-salt-cooled pebble-bed reactor with CFD methodology

The Fluoride Salt Cooled High Temperature Reactor (FHR) is an innovative concept reactor that inherits the technical foundation and advantages of the six optional generation-IV reactors and pressurized water reactors, which is mainly in process in both China and the United States. In this paper, the porous and realistic modeling approaches are adopted to analyze the thermal hydraulic characteristics of a FHR core and a unit segment of pebbles in the core respectively. The distributions of temperature and pressure of the fluoride salt, as well as the reflector temperature profile, are obtained using the porous model. The detailed local flow and heat transfer are investigated by the realistic modeling method for the locations which may have the maximum coolant temperature based on the results of the porous model. The profiles of temperature, velocity, pressure and Nusselt number (Nu) of the coolant on the surface of the pebble are also obtained and analyzed. Numerical results showed that the flow field between the fuel pebbles is complex including secondary flow and back-flow phenomenon, which are hard to measure by experiments. This work can provide useful information for the experimental and mechanism research of FHRs.     

Design & construction of a digital real time reactivity meter for Tehran research reactor

Reactivity measurement is one of the challenges of monitoring, control and investigation of nuclear reactors. In this paper design and construction of a reactivity meter for continuous monitoring of reactivity in research reactors are described. The device receives amplified output of the fission chamber, which is in mA range, as the input. Using amplifier circuits, this current is converted to voltage and then digitalized with a microcontroller to be sent to serial port of computer. The device itself consists of software, which is a MATLAB real time programming for the computation of reactivity by the solution of neutron kinetic equations. After data processing the reactivity is calculated and presented using LCD. Tehran research reactor is selected to test the reactivity meter device. The results of applying this reactivity meter in Tehran research reactor (TRR) are compared with the experimental data of control rod worth, void coefficient of reactivity and reactivity changes during approach to full power. The maximum relative error in several experiments is calculated to be 13%.     

Design evaluation of decay tank for a pool-type research reactor from the required minimum flow residence time point of view

A decay tank shall be designed to provide enough flow residence time to ensure that the N-16 activity decreases before the coolant leaves the decay tank's shielding room. However, when a proper criterion for the flow residence time in a decay tank is not presented, the tank would be oversized/undersized. In this paper, design evaluation for a decay tank is performed by investigating the effect of the fluid distribution along the residence time on the total dose rate and the required minimum flow residence time. The evaluation is also carried out to resize the predesigned decay tank.     

Prevention and mitigation of severe accident developments and recriticalities in advanced fast reactor systems

GEN-IV nuclear systems, especially advanced sodium-cooled fast reactors (SFRs) are on the horizon and a key issue of their success is the promise of a higher and improved safety level. In Europe safety investigations are currently under way e.g. in the collaborative CP-ESFR project of the EU. Both on the prevention and mitigation side significant efforts are invested to fulfill the high safety goals. One route of assurance concentrates on the mitigation or even elimination of specific severe accident routes leading to core disruption and recriticalities. The accident phase with larger disrupted and molten fuel regions is coined the transition phase. A competition between fuel losses and in-pool material motion exists deciding over recriticalities and energetics potentials in this phase. To get a control of the transition phase recriticalities and energetics, ideas have been developed to install dedicated means in the core that enhance and guarantee a sufficient and timely fuel discharge - a controlled material relocation (CMR). Several proposals are under way to accomplish this CMR and especially in Japan significant theoretical and experimental work has been performed. In Europe the path to develop CMR measures was driven in the past by the development of the CAPRA reactors with a high Pu enrichment. In the current paper the status of analyses is described and some new concepts are discussed. The CMR measures are discussed along two accident scenarios, the unprotected loss of flow (ULOF) and the instantaneous blockage accident (TIB).     

Investigation on two-phase critical flow for loss-of-coolant accident of pressurized water reactor

The previous investigations were mainly conducted under the condition of low pressure,however,the steam-water specific volume and the interphase evaporation rate in high pressure are much different from those in low pressure,Therefore,the new experimental and theoretical investigation are performed in Xi‘an Jiaotong University.The investigation results could be directly applied to the analysis of loss-of -coolant accident for pressurized water reacor.The system transition characteristics of cold leg and hot leg break loss-of -coolant tests are described for convective circulation test loop.Two types of loss-of-coolant accident are identified for :hot leg” break,while three types for “cold leg”break and the effect parameters on the break geometries.Tests indicate that the mass flow rate with convergent-divergent nozzle reaches the maximum value among the different break sections at the same inlet fluid condition because the fluid separation does not occur.A wall surface cavity nucleation model is developed for prediction of the critical mass flow rate with water flowing in convergentdivergent nozzles.     

CFD analysis of flow and heat transfer of turbulent pulsating flow in a tube in rolling motion

The investigation of flow and heat transfer of turbulent pulsating flow is of vital importance to the nuclear reactor thermal hydraulic analysis in ocean environment. In this paper, the flow and heat transfer of turbulent pulsating flow is analyzed. The calculation results are firstly verified with experimental data. The agreement between them is satisfactory. The effect of spanwise and wall-normal additional forces is significant in small Reynolds number, and decreases with Reynolds number increasing. The rolling axis and rolling radius contribute slight to the flow and heat transfer. The effect of velocity oscillation period on the heat transfer is limited than that of Reynolds number and oscillating velocity Reynolds number. The traditional empirical correlations could not predict the flow and heat transfer of turbulent pulsating flow in rolling motion.