An application of systems analysis techniques to the study of reactor seismic safety

Systems analysis is being used in conjunction with structural analysis to study the conservatisms and to provide insights into aspects of reactor seismic safety. An event-tree/fault-tree model of a commercial nuclear power plant is being constructed to determine the probability of release and probabilities of system and component failures caused by possible seismic events. The event-tree/fault-tree model is evaluated using failure data generated by applying the response a component sees to the component's fragility function. The responses are calculated by a structural analysis code using earthquake time histories as forcing functions. The quantification of the event-tree/fault-tree model is done conditional on a given seismic event and the conditional probabilities thus calculated unconditioned by integrating the results over the seismic hazard curve. In this way, most of the dependencies between event failures resulting from the seismic event itself are removed making known fault-tree analysis quentification techniques applicable. The outputs from the computations will be used in sensitivity studies to determine the key calculations and variables involved in seismic analyses of nuclear power plants.     

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.     

Nonlinear seismic response analysis of an embedded reactor building based on the substructure approach

A practical method to calculate the elasto-plastic seismic response of structures considering the dynamic soil-structure interaction is presented. The substructure technique in the time domain is utilized in the proposed method. A simple soil spring system with the coupling effects which are usually evaluated by the impedance matrix is introduced to consider the soil-structure interaction for embedded structures. As a numerical example, the response of a BWR-MARK II type reactor building embedded in the layered soil is calculated. The accuracy of the present method is verified by comparing its numerical results with exact solutions. The nonlinear behavior and the soil-structure interaction effects on the response of the reactor building are also discussed in detail. It is concluded that the present method is effective for the seismic design considering both the material nonlinearity of the nuclear reactor building and the dynamic soil-structure interaction.     

Conceptual evaluation of fluid–structure interaction effects coupled to a seismic event in an innovative liquid metal nuclear reactor

The seismic response analysis of such liquid storage systems, especially liquid metal reactors, as for example the eXperimental Accelerator Driven System (XADS), was examined taking into account mainly the coupling effects of the fluid–structure interaction and their influence on its relevant internal systems and components.Therefore this paper deals with the structural analyses of the seismically induced hydrodynamic responses, in the event of a safe shutdown earthquake (SSE), and the free oscillation (known as sloshing waves) of a metal liquid coolant as well as the dynamic buckling effects on involved structures.To the mentioned purpose the interaction and coupling effects among the main reactor vessel structures and the primary coolant response were investigated by means of a numerical evaluation (with a qualified finite element code) because of the lack of analytical linear theories that in any case are not adequate to describe all the complex phenomena related to the seismic loading.For the numerical modelling procedure, 3D finite element models were set up to analyse the propagation of seismic waves as well as its derived structural effects, such as the fluid steep waves motion, the local buckling bulges, etc., taking into account the geometrical and material nonlinearities of the RPV and the considered simplified internals.The obtained numerical results in terms of stress intensity and of the capability of the structures to resist relevant seismic loads are, thus, presented and discussed. Moreover the performed analyses allowed to highlight the structures mostly affected by the assumed loading conditions in order to achieve data useful for an upgrading of the design geometry, if any, for the considered reactor.     

Preliminary evaluation of aircraft impact on a near term nuclear power plant

For the assessment of the safety and durability of a nuclear power plant (NPP), the containment building behaviour shall be evaluated, under various service and extreme conditions, both natural or produced by natural accident or vicious man activities, like September 2001 jet aircraft crashes.The aim of this paper is to preliminary evaluate the effects and consequences of the energy transmitted to the outer containment walls (according to the international safety and design code guidelines, as NRC or IAEA ones) due to a military or civil aircraft impact into a nuclear plant, considered as a ‘beyond design basis’ event.To perform reliable analysis of such a large-scale structure and determine the structural effects of the propagation of this types of impulsive loads (response of containment structure), a realistic but still feasible numerical model with suitable materials characteristics were used by means of which relevant physical phenomena are reflected. Moreover a sensitivity analysis has also been carried out considering the effects of different containment wall thickness and reinforced/prestressed concrete features. The obtained results were analysed to check the NPP containment strength margins.     

Application of seismic analysis methodology to small modular integral reactor internals

The fluid–structure interaction (FSI) effect should be carefully considered in a seismic analysis of nuclear reactor internals to obtain the appropriate seismic responses because the dynamic characteristics of reactor internals change when they are submerged in the reactor coolant. This study suggests that a seismic analysis methodology considered the FSI effect in an integral reactor, and applies the methodology to the System-Integrated Modular Advanced Reactor (SMART) developed in Korea. In this methodology, we especially focus on constructing a numerical analysis model that can represent the dynamic behaviors considered in the FSI effect. The effect is included in the simplified seismic analysis model by adopting the fluid elements at the gap between the structures. The overall procedures of the seismic analysis model construction are verified by using dynamic characteristics extracted from a scaled-down model, and then the time history analysis is carried out using the constructed seismic analysis model, applying the El Centro earthquake input in order to obtain the major seismic responses. The results show that the seismic analysis model can clearly provide the seismic responses of the reactor internals. Moreover, the results emphasize the importance of the consideration of the FSI effect in the seismic analysis of the integral reactor.     

A seismic analysis method for a block column gas-cooled reactor core

An analytical method for predicting the behavior of a prismatic high-temperature gas-cooled reactor (HTGR) core under seismic excitation has been developed. In this analytical method, blocks are treated as rigid bodies, are constrained by dowel pins which restrict relative horizontal movement but allow vertical and rocking motions. Coulomb friction between blocks and between dowel holes and pins is also considered. A spring dashpot model is used for the collision process between adjacent blocks and between blocks and boundary walls.Analytical results are compared with experimental results and are found to be in good agreement. The analytical method can be used to predict the behavior of the HTGR core under seismic excitation.     

水冷聚变堆主回路活化产物源项计算分析

水冷聚变堆中结构材料活化腐蚀产物和冷却剂活化产物是正常运行工况下的最主要放射性来源,也是反应堆运行及维护过程中工作人员辐照剂量的直接来源。本文使用CATE V2.1程序对国际热核聚变实验堆(International Thermonuclear Experimental Reactor,ITER)LIM-OBB(Limiter-Out-Board Baffle)冷却回路的活化腐蚀产物和水活化产物进行模拟计算,并根据CATE模拟得到的放射性活度通过点核积分程序分别计算正常运行1.2 a及停堆15 d的剂量率。计算结果表明,反应堆运行期间冷却剂活化产物比活度和剂量率远大于结构材料活化腐蚀产物,而停堆后冷却剂活化产物迅速衰变完,结构材料的活化腐蚀产物成为比活度和剂量率的主要来源。     

Modelling and parametric analysis of an advanced gas-cooled reactor

This paper describes a parameter identification scheme that operates on a mathematical model of one of the CEGB's Commercial Advanced Gas-cooled Reactors. Parameters in a 10th-order non-linear deterministic model are varied to fit experimental data from two rod movement transients. Two sets of results are analyzed, at low and high power operation of the Hinkley Point B CAGR plant. The values obtained for one of the parameters, the fuel-temperature coefficient of reactivity, are compared with those obtained from the standard CEGB analysis procedure and found to agree within the estimated error bounds.     

Assessment of TAPINS code for application to thermal-hydraulic analysis of an integral reactor,REX-10

The thermal-hydraulic analysis program for integral reactor system (TAPINS) is a thermal-hydraulic system code developed by Seoul National University for transient analysis of an integral reactor, REX-10. Specialized for a fully passive integral pressurized water reactor, TAPINS adopts a one-dimensional four-equation drift-flux model for two-phase flows. It also consists of component models for the core, the helical-coil steam generator, and the steam-gas pressurizer. This paper presents the developmental assessment of TAPINS to validate its applicability to the thermal-hydraulic analysis of REX-10. Assessment problems are determined by taking into account thermal-hydraulic phenomena expected during design basis accidents of REX-10, including the loss-of-feedwater accident and the small-break loss-of-coolant accident. To confirm the predictive capability of TAPINS for these phenomena, the TAPINS model is validated against four sets of separate effects problems, including the pressurizer insurge test, the subcooled boiling experiment, the critical flow test, and the Edwards pipe problem. In addition, the calculation results of TAPINS are compared with the experimental data obtained from a series of integral effects tests using a scaled apparatus of REX-10. From the validation results, it is demonstrated that TAPINS can provide the reasonable prediction on the thermal-hydraulic responses of REX-10 during the transient and accident conditions.     

Overpressure protection analysis of an advanced integral reactor

Overpressure protection analysis of KAERI's advanced integral reactor, which has been developed to verify the performance of the System integrated Modular Advanced ReacTor (SMART), has been performed using the Transients And Setpoint Simulation/Small and Medium Reactor (TASS/SMR) code. In the analysis, the loss of feed-water and the regulating bank withdrawal events on behalf of the decrease in the heat removal by the secondary system and the reactivity and power distribution anomalies are selected as the initiating events for the analysis because the highest peak pressures of the primary system occur during these events. Conservative assumptions and the various initial/boundary conditions have been applied to the overpressure protection analysis for the advanced integral reactor. Although the pressurization of the primary system occurs due to an unbalance between the power generation in the core and the heat removal through the steam generator, the peak pressures in the cases of using the loss of feed-water and the regulating bank withdrawal event as an initiating event are well below the acceptance criteria of 18.7 MPa, due to the reactor protection system and three pilot operated safety relief valves installed in the advanced integral reactor.     

Trip setpoint analysis for the reactor protection system of an advanced integral reactor

The trip setpoints for the reactor protection system of a 65-MWt advanced integral reactor have been analyzed through sensitivity evaluations by using the Transients and Setpoint Simulation/System-integrated Modular Reactor code. In the analysis, an inadvertent control rod withdrawal event has been considered as an initiating event because this event results in the worst consequences from the viewpoint of the minimum critical heat flux ratio and its consequences are considerably affected by the trip setpoints. Sensitivity evaluations have been performed by changing the trip setpoints for the ceiling of a variable overpower trip (VOPT) function and the pressure of a high pressurizer pressure trip function. Analysis results show that a VOPT function is an effective means to satisfy the acceptance criteria as the control rod rapidly withdraws: on the other hand, a high pressurizer pressure trip function is an essential measure to preserve the safety margin in the case of a slow withdrawal of the control rod because a reactor trip by a VOPT function does not occur in this case. It is also shown that the adoptions of 122.2% of the rated core power and 16.25 MPa as the trip setpoint for the ceiling of a VOPT function and the pressure of a high pressurizer pressure trip function are good selections to satisfy the acceptance criteria.     

Current status and innovative improvement of small commercial reactor in China

The current status and developing plan of China’s nuclear energy are introduced, and features of the small commercial reactor, CNP-300, are described. The ongoing improvements including power uprate, life extension, core management, and integration of passive safety systems are generally presented with the efforts to enhance the safety and economy.     

氢化锆慢化熔盐堆钍铀转换性能初步分析

中子能谱对钍基燃料在熔盐堆中的利用效率及温度反馈系数等安全问题有较大影响,所以对熔盐堆新型慢化剂的研究具有重要意义。本工作基于SCALE6计算程序,对不同几何栅元结构的氢化锆栅元组件在熔盐堆的物理性能进行了研究,分别计算了中子能谱、钍铀转换比、~(233)U浓度、总温度反馈系数以及燃耗等中子物理参量。结果表明,减小六边形栅元对边距或者增加熔盐占栅元体积比可以增加钍铀转换比和改善温度反应性系数;当加入的氢化锆慢化剂体积份额为0.1时就可以将熔盐堆~(233)U初始浓度降低到2.5×10~(-2)以内;氢化锆慢化熔盐堆在超热谱条件下,其~(233)U初装载量和超铀核素产量较小,同时堆芯较为紧凑。     

Incorporating reliability analysis into the design of passive cooling systems with an application to a gas-cooled reactor

A time-dependent reliability evaluation of a two-loop passive decay heat removal (DHR) system was performed as part of the iterative design process for a helium-cooled fast reactor. The system was modeled using RELAP5-3D. The uncertainties in input parameters were assessed and were propagated through the model using Latin hypercube sampling. An important finding was the discovery that the smaller pressure loss through the DHR heat exchanger than through the core would make the flow to bypass the core through one DHR loop, if two loops operated in parallel. This finding is a warning against modeling only one lumped DHR loop and assuming that n of them will remove n times the decay power. Sensitivity analyses revealed that there are values of some input parameters for which failures are very unlikely. The calculated conditional (i.e., given the LOCA) failure probability was deemed to be too high leading to the identification of several design changes to improve system reliability. This study is an example of the kinds of insights that can be obtained by including a reliability assessment in the design process. It is different from the usual use of PSA in design, which compares different system configurations, because it focuses on the thermal–hydraulic performance of a safety function.     

Estimates on seismic amplification factors of reactor components

Employing an averaging technique we obtain estimates on seismic amplification factors for different components in nuclear reactors.     

Development of a three-dimensional CDA analysis code: SIMMER-IV and its first application to reactor case

For the transition phase analysis of core disruptive accidents, the development of a three-dimensional reactor safety analysis code, SIMMER-IV, has been carried out based on the technology of the two-dimensional SIMMER-III code. The world first application of SIMMER-IV to a small-sized sodium-cooled fast reactor has also been attempted to clarify event progression in the early stage of the transition phase. This SIMMER-IV calculation is compared to the two-dimensional case calculated by SIMMER-III, neglecting the presence of control rod guide tubes. The present analysis with the three-dimensional representation suggests that the conventional scenario leading to rather early high-mobility fuel pool formation is unrealistic and the degraded core tends to keep low mobility in the early stage of transition phase.     

New discrete models and their application to seismic response analysis of structures

New discrete models and their application to seismic response analysis of structures is proposed in this paper. These models consist of finite number of small rigid bodies connected with springs distributed over the contact area of two neighbouring bodies. In general size of stiffness matrices of these elements are at most (6 × 6) which are equal to or even smaller than of those of conventional finite elements so that considerable reduction of computing time can be expected. Effectiveness of these elements in nonlinear structural analysis, especially dynamic response analysis of structures will be demonstrated by several numerical examples.