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.     

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.     

Thermal hydraulic and safety analysis for core conversion (HEU–LEU) of Syrian Miniature Neutron Source Reactor

The paper presents the behavior and properties analysis of the low enriched uranium fuel compared with the original high enriched uranium fuel. The MNSR reactor core was modeled with both fuel materials and the reactor behavior was studied during the steady state and abnormal conditions. The MERSAT code was used in the analysis. The steady state thermal hydraulic analysis results were compared with that obtained from the experimental results hold during commissioning the Syrian MNSR. Comparison with experimental data shows that the steady-state behavior of the HEU core was accurately predicted by the MERSAT code calculations. The validated model was then used to analyze LEU cores with two proposed UO₂ fuel pin designs. With each LEU core, the steady state and 3.77 mk rod withdrawal transient were run and the results were compared with the available published data in the literatures for the low enriched uranium fuel core. The results reveal that the low enriched uranium fuel showed a good behavior and the peak clad temperatures remain well below the clad melting temperature during reactivity insertion accident.     

The individual channel monitoring (ICM) proposal to improve the safety performance of RBMK

A comprehensive study has been completed in relation to the accident analysis of RBMK with key results documented in three companion papers in the present journal volume. The study also aimed at the comparison between selected accident analysis topics in Light Water Reactors (LWR) and RBMK. The relevance of the Fuel Channel Blockage (FCB) event within the area of accident analysis was confirmed. Owing to the probability of occurrence, also connected with the large number of channels, estimated in the order of 10⁻²/reactor/year, the FCB is actually part of the Design Basis Accidents (DBA) for the RBMK.In case the FCB is part of the DBA, a noticeable difference occurs in results from safety evaluations between LWR and RBMK: in the latter case, a DBA causes a loss of integrity for the pressure barrier and, though to a limited extent, damaged fuel overpasses such a barrier. The FCB event also causes contamination in reactor cavity and in selected parts of the overall confinement, damage in various graphite blocks and, even excluding the MPTR risk owing to the findings from the second companion paper in this journal volume, mechanical loads on neighbouring fuel channels, graphite stacks and reactor tank that need extensive examination before reactor restart.Therefore, a proposal has been formulated for performing the feasibility analysis for the design of a system denominated ICM (Individual Channel Monitoring). The goal of the ICM is the early detection of the FCB event and the triggering of scram in such a way to prevent pressure tube damage and, definitely, over-passing of the pressure barrier by molten or damaged fuel during a DBA situation.The ICM is based upon the signals of pressure-drop (or flow-rate) and fluid temperature transducers installed in the bottom and the upper parts of the fuel channels, respectively. The performed study shows that steam superheating at fuel channel outlet occurs early after the blockage event and the related temperature signal can be used to cause scram. The availability of sophisticate computational tools including the detailed neutron kinetic model for each core channel, made possible the preliminary conclusion of the study.     

基于SAC-CFR系统分析软件的EBR-Ⅱ LOHSWS事故分析

利用自主开发的系统分析软件SAC-CFR对美国实验增殖堆2号(EBR-Ⅱ)的未能紧急停堆的丧失热阱(LOHSWS)事故全厂瞬态行为进行建模分析。SAC-CFR耦合了新开发的三维钠池计算模型,用于分析EBR-Ⅱ钠池内的流型。结果表明,SAC-CFR计算结果与实验数据相符合,SAC-CFR可用于快堆部分事故工况的瞬态计算,同时也证实了EBR-Ⅱ可在LOHSWS事故下依靠固有安全性停堆。     

Thermal-hydraulic analysis of accidents leading to local coolant flow decrease in the main circulation circuit of RBMK-1500

There are a few transient and loss-of-coolant accident conditions in RBMK-1500 reactors that lead to a local flow decrease in fuel channels. Because the coolant flow decreases in fuel channels (FC) leads to overheating of fuel claddings and pressure tube walls, mitigation measures are necessary. The accident analysis enabled the suggestion of the new early reactor scram actuation and emergency core cooling system (ECCS) initiation signal, which ensures the safe shutdown of the reactor and compensates the stagnation flow. Analysis of such conditions is presented in this paper. Thermal-hydraulic analysis was conducted using the state-of-the-art RELAP5 code. Results of the analysis demonstrated that, after implementation of the developed management strategy for destruction of local flow stagnation, the Ignalina nuclear power plant (NPP) would be adequately protected following accidents, leading to local coolant flow decrease in the primary circuit.     

水力驱动控制棒冷态极限落棒能力分析

为了验证控制棒水力驱动系统在卡棒和倒置等极限工况下的停堆可靠性，在200MW低温核供热堆控制棒水力驱动系统的1：1实验台架上进行了冷态极限落棒实验，通过对实验结果的分析。得到了在发生卡棒事故时控制棒的落棒能力，揭示了落棒机理，建立了正置时控制棒插入堆芯的模型，并用实验数据验证了该模型的正确性，建立了倒置时控制棒插入堆芯的模型，获得了倒置时控制棒的插棒能力。     

钠冷快堆反应性意外引入事故时停堆保护研究

在反应堆系统中,当反应堆处于异常工况时,如果运行参数超出保护限值,则由保护系统触发相关保护动作,以保证反应堆的状态符合事故验收准则的要求。本文将通过Simulink建立钠冷快堆主要系统模型,在发生反应性意外引入事故时,借鉴快堆事故分析中预期瞬态无停堆保护（ATWS）的分析方法,基于相应保护参数的测量误差和数据处理过程对反应堆一回路的保护参数及其整定值进行研究,并确保钠冷快堆的状态在整个反应性引入事故过程中符合钠冷快堆的事故验收准则。仿真结果表明,当发生补偿棒失控提升5 s和10 s时,目前的堆芯出口钠温、功率、功率流量比等保护参数的整定值、信号测量延迟及落棒时间可取其他值。当补偿棒失控提升15 s时,只要保证保护参数整定值、相应参数的信号测量延迟及落棒时间能使反应堆在36.45 s前进入深度次临界都是可以的。     

Scram and nonlinear reactor system seismic analysis for a liquid metal fast reactor

Earthquake vibrations cause large forces and stresses that can significantly increase the scram time required for safe shutdown of a nuclear reactor. The horizontal deflections of the reactor system components cause impact between the control rods and their guide tubes and ducts. The resulting frictional forces, in addition to other operational forces, delay the travel time of the control rods. To obtain seismic responses of the various reactor system components (for which a linear response spectrum analysis is considered inadequate) and to predict the control rod drop time, a non-linear seismic time history analysis is required. Nonlinearities occur due to the clearances or gaps between various components. When the relative motion of adjacent components is large enough to close the gaps, impact takes place with large impact accelerations and forces.This paper presents the analysis and results for a liquid metal fast rector system which was analyzed for both scram times and seismic responses such as bending moments, accelerations and forces. The reactor system was represented with a one-dimensional nonlinear mathematical model with two degrees of freedom per node (translational and rotational). The model was developed to incorporate as many reactor components as possible without exceeding computer limitations. It consists of 12 reactor components with a total of 71 nodes, 69 beam and pin-jointed elements and 27 gap elements. The gap elements were defined by their clearances, impact spring constants and impact damping constants based on a 50% co-efficient of restitution.The horizontal excitation input to the model was the response of the containment building at the location of the reactor vessel supports. It consists of a 10 sec safe shutdown eathquake (SSE) acceleration-time history at 0.005 sec intervals and with a maximum acceleration of 0.408 g. The analysis was performed with two Westinghouse special purpose computer programs. The first program calculated the reactor system seismic responses and stored the impact forces on tape. The impact forces on the control rod driveline were converted into vertical frictional forces by multiplying them by a coefficient of friction, and then these were used by the second program for the scram time determination.The results give time history plots of various seismic responses, and plots of scram times as a function of control rod travel distance for the most critical scram initiation times. The total scram time considering the effects of the earthquake was still acceptable but about four times longer than that calculated without the eathquake. The bending moment and shear force responses were used as input for the structural analysis (stresses, deflections, fatigue) of the various components, in combination with the other applicable loading conditions.     

Cobalt irradiation box ejection accident of ETRR-2

The new Egyptian Test and Research Reactor Number 2 ETRR-2, MTR type, is now under operational tests. It has a main central irradiation channel for the purpose of Co⁶⁰ isotope production with an intended rated capacity of 50 000 Ci per year. The reactivity introduced in the reactor due to accidental ejection of the Co⁶⁰ irradiation box (CIB) should be discussed. This reactivity insertion accident (RIA) may be fast or slow with maximum reactivity worth 2.9428 $. The CIB may move with constant speed or variable acceleration according to its initial speed and the applied forces. This results in a linear, parabolic or sinusoidal motion, which in turn affects the reactivity insertion rate (RIR). The present work analyzes this type of perturbation during normal operating conditions: 22 MW full power and 1900 kg s⁻¹ forced core cooling flow. The work serves as a part of the safety evaluation process applicable to similar MTR cores. The RIA code TRANSP20 is developed for this study. It simulates various types of RIR, fast or slow resulting from different CIB ejections. Scram signal due to power, period, inlet and outlet temperatures, or temperature difference is expected to activate the shutdown system. The work presents five case studies, two for fast ejection and three for slow. The transient behavior of the reactor during this is illustrated. The results show that the reactor can withstand slow ejection if the scram is available. However, for fast ejection the scram system does not prevent the clad temperature from exceeding safety limits. Recommendations to prevent or mitigate this accident are highlighted.     

钠冷快堆停堆保护方案研究

在借鉴中国实验快堆(CEFR)热工模型建模经验的基础上,利用Relap5程序建立霞浦示范快堆(CFR)的主要系统模型,并参考快堆安全分析中的预期瞬态无停堆保护(ATWS)的分析方法,对发生反应性意外引入事故时的安全裕度和停堆保护进行仿真研究。仿真结果表明,额定功率下发生反应性引入时,不会触发短周期的报警和停堆;当发生补偿棒失控提升5 s和10 s时的反应性意外引入事故,目前一回路保护参数整定值、信号测量延迟及安全棒落棒时间可以取其他值;当补偿棒失控提升15 s时,在目前的设计下,核功率和功率流量比信号能确保事故下的反应堆状态符合事故验收准则。当其他保护信号失效,堆芯出口钠温所触发的停堆保护若要实现同样的功能,则需保证反应堆在14.85 s之前进入深度次临界。     

Safety analysis of small long life CANDLE fast reactor

Safety analysis for small long life fast CANDLE reactor was performed with ULOF (unprotected loss of flow), SDRW (unprotected shut down rods withdrawal), ULOHS (unprotected loss of heat sink) and LB (local blockage) accidents. The employed reactor system is based on the former steady state research. The core with 1.0 m radius and 2.0 m length produces 200 MW thermal power in steady state, using enriched N-15 natural uranium as fresh fuel and lead bismuth as coolant. The former 3 accidents were simulated without scram by neutronic-thermal hydraulic calculation coupled with stationary diffusion calculation. The LB accident was simulated by transient thermal hydraulic calculation only, because in this accident the neutronic factors basically do not change. The analysis results show that the proposed small CANDLE fast reactor can survive all the accidents without any active protection.     

ASTEC V2.0 computational evaluation of source term and its chemical forms under accidental conditions during mid-loop operation

The possibility of an accident or component failure during mid-loop operation has been identified in probabilistic safety studies as a major contributor to core melt frequency and source term risk. The fission products release and transport to the containment has been analyzed during mid-loop operation of a reference PWR 1000 MWe reactor using the severe accident integral code ASTEC V2.0. The analyses have been performed considering the loss of residual heat removal (RHR) system at various times after reactor shutdown for the reactor vessel configuration with the removed upper head (open reactor). In this configuration, the possible air ingress can have an impact on safety such as accelerated oxidation and increased volatility of certain FPs (particularly iodine and ruthenium). Sensitivity calculations have been performed in terms of air ingress simulation with a different intensity. Besides equilibrium chemistry model, most of the calculations have also used a limited kinetics model. The study has shown that without air ingress the only predicted gaseous form of iodine is HI (≤7.4% of the total mass of iodine released from core) and no gaseous RuO₄ is created. Sensitivity calculations have illustrated that the gross fraction of gaseous iodine (I₂ + HOI + HI) has an increased trend with growth of air ingress intensity and with the duration of sequence evolution. In most oxidative atmosphere the gross iodine gaseous fraction could increase by a factor form of two to several times as compared to the corresponding case without air ingress (particularly due to I₂ persistence). Creation of gaseous RuO₄ is sensitive to carrier gas temperature; therefore a considerable fraction (≤3%) is predicted only in the sensitivity cases with the shortest time of loss of RHR after reactor scram.     

Analysis of the Chernobyl accident from 1:19:00 to the first power excursion

Many researchers have reported that the root cause of the Chernobyl accident has not been clarified still now. Since many of them discussed the accident without a precise thermal-hydraulic investigation, thermal-hydraulic calculations coupled with neutronic calculations have been done on the basis of the recorded result at the Chernobyl Unit-4. Plant configurations and operational conditions were given to the code on the basis of reported result and published papers. Calculation could trace plant parameters from 1:19:00 to the first power excursion without any discrepancies measured at the Chernobyl Unit-4. Reactivity slightly smaller than 1β by the positive scram is concluded as a possible direct cause of the accident, which acts as a trigger to increase the reactor power. Other possibilities as a trigger of the accident such as cavitation in pumps and pump coast-down were investigated. The importance of the calculation from the stable condition is also described in this paper in order not to bring unnecessary assumptions into the calculation.     

间接氦气透平循环高温堆紧急停堆动态仿真

通过对10 MW高温气冷堆氦气透平发电装置(HTR-10GT)的堆芯、热交换器和透平压气机组等主要设备的数学建模和程序编制,初步建立起了一套模拟该装置瞬态特性的仿真程序.通过对该装置于5s时刻堆内引入0.1$阶跃正反应性引发的紧急停堆事故的瞬态模拟,初步验证了该装置紧急停堆预案设置的安全性和合理性,证明了旁路快开阀的设...     

Investigating the execution of EOPs in LOTDFP + ATWS + LOCA scenarios for an ABWR using MAAP code

The effectiveness of the execution of emergency operation procedures (EOPs) for an advanced boiling water reactor (ABWR) during postulated accident conditions using MAAP 4 code is discussed in this paper. The simulation scenarios included the loss of turbine driven feedwater pump (LOTDRFP), the anticipated transients without scram (ATWS), and the loss of coolant accident (LOCA). Based on the comparisons of responses on different parameters for cases with and without EOP actions, we concluded that the EOPs could effectively mitigate the consequences of the accidents. In addition, the emergency depressurization (ED) timing and the times spent between executing the EOP steps were also considered. The simulation results clearly reveal that both the earlier execution of ED and the decrease of times spent between each EOP step could delay the boron injection and leave the operator ample time to take some other remedy actions for reactor safe shutdown.     

基于SAC-3D系统软件的FFTF未能紧急停堆失流实验的数值模拟

本文利用系统分析软件SAC-3D对美国快通量试验堆（FFTF）堆芯及一回路进行了建模,并根据国际原子能机构（IAEA）提供的FFTF未能紧急停堆的失流实验的边界条件数据进行了事故瞬态仿真计算。计算得到堆芯热工水力及中子物理关键参数,仿真结果与实验测量数据符合较好。对比结果验证了SAC 3D在模拟液态金属冷却快堆事故工况中的有效性与准确性,也证明了FFTF堆型具有可靠的非能动安全性。     

Requirements on fuel management for a safe and optimum operation of the German research reactor FRJ-2

In case of a failure of a coarse control arm (CCAs) at FRJ-2, reactivity is added to the reactor. The amount of this reactivity depends on the efficiency of the individual CCAs which has been measured as 180% of the average reactivity of the six arms for the central arm. For this design basis accident, it is required that only four out of five residual arms are capable of shutting down the reactor. This minimum shutdown reactivity is provided by an optimum fuel management including an experimental reactivity determination. Calculation of fuel burnup and material densities is performed by the depletion code SUSAN, which has been verified by separate calculations using ORIGEN. The difference in the reactivity values (between calculation and measurement) is mainly a consequence of the limitation of the inverse kinetic method, which is incapable of covering the effects of the flux deformation and interaction of the CCAs and core in the process of reactor scram.     

Thermal-hydraulic modeling of flow inversion in a research reactor

The course of loss of flow accident and flow inversion in a pool type research reactor, with scram enabled under natural circulation condition is numerically investigated. The analyses were performed by a lumped parameters approach for the coupled kinetic–thermal-hydraulics, with continuous feedback due to coolant and fuel temperature effects. A modified Runge–Kutta method was adopted for a better solution to the set of stiff differential equations. Transient thermal-hydraulics during the process of flow inversion and establishment of natural circulation were considered for a 10-MW IAEA research reactor. Some important parameters such as the peak temperatures for the hot channel were obtained for both high-enriched and low enriched fuel. The model prediction is also verified through comparison with other computer code results reported in the literature for detailed simulations of loss of flow accidents (LOFA) and the agreement between the results for the peak clad temperatures and key parameters has been satisfactory. It was found that the flow inversion and subsequent establishment of natural circulation keep the peak cladding surface temperature below the saturation temperature to avoid the escalation of clad temperature to the level of onset of nucleate boiling and sub-cooled void formation to ensure the safe operation of the reactor.     

机组临时停机停堆期间的化学监督与控制

目前的化学与放射化学程序和措施,都是针对正常功率运行和按部就班有计划的大修状态而设置,如遇到机组跳堆、跳机或冷停堆等紧急情况,则没有相应的应急预案或相关程序进行提前或有目的地干预。基于这种情况,电厂化学人员经过多年的实践和不断经验反馈,总结并编写了专门针对紧急停机停堆的化学监督与控制应急预案。通过停堆过程和停堆后的不同状态,启机过程的化学与放射化学监测,监督燃料包壳状态,控制一回路的剂量水平,以防止设备腐蚀。