Identification of initiating events for pressurized water reactor accidents

In order to aid operators in identifying the different initiating events as defined in the Final Safety Analysis Report (FSAR), we develop a novel identification procedure. The procedure is based on the monitoring of three key system parameters in a pressurized water reactor (PWR), i.e., the pressure, the average temperature, and the temperature difference of the hot-leg and cold-leg of the reactor coolant system. By monitoring the system thermal state diagram in a pressure–temperature space, an operator can easily identify what initiating event is taking place while a static point in the diagram starts to move. The event data pool is first established by storing the transient analysis results for events of different types using the optimal estimated RELAP5 model. Since the variation ranges of system key parameters at a specific time represent the specific character for each initiating event, the identification procedure can easily determine which cases in which the event data pool can be fitted to on-line data using only variation range comparison without complex calculations. This identification method is believed to be able to help the plant operator to identify the different events and then execute the Emergency Operating Procedure more effectively.     

Dynamic analysis of a pressurized water reactor

An apparent malfunction in a pressurized water reactor system has been investigated using fluctuation analysis. Both frequency-domain and time-domain analyses have been used and the results obtained by the two methods have been compared. The recording was performed by a relatively simple, cheap system giving high recording precision and the analyses were performed on an IBM 370 digital computer. It is shown that, while considerable information can be derived from frequency-domain analyses, a misinterpretation can occur in some cases. Time-domain correlation, as normally performed, was not very informative. However, time-domain correlation on bandwidth-limited time-series proved to be very valuable and could remove the misinterpretation of the frequency-domain analyses. The bandwidth limitation was performed by digital filters.     

Concept for monitoring gaseous and aerosol-bound radioactive discharges from boiling and pressurized water reactor plants in the event of design basis and severe accidents

A new concept for monitoring radioactive discharge has been developed. The resulting system presented here is intended to meet the requirements set forth in the German Nuclear Safety Standard KTA 1503.2 (draft) for accident surveillance while at the same time being suitable for activity monitoring during containment venting. The system and typical modes of system operation are described for plants equipped with pressurized water reactors (PWRs) and plants equipped with boiling water reactors (BWRs). A combination of different methods of evaluation allows the space needed for instrumentation as well as the effort required for testing to be minimized.     

Control rod drop hydrodynamic analysis for a pressurized water reactor

During operation of nuclear power reactors, reactivity initiated accidents can take place such as a control rod drop. If this occurs, the reactivity increases significantly and leads to an enhancement in power, fuel temperature and damage of reactor eventually. Exact assessment of these accidents depends on the hydrodynamic information. In this research, it is tried to simulate the unsteady flow field around the control rod for a pressurized water reactor power plant. In order to simulate the flow field around the control rod inside the guide tube, averaged Navier–Stokes equations accompanied by the layering dynamic mesh strategy have been used. The information exchange between the two computational stationary and moving grids, the computational grid around the control rod and the grid next to the guide tube, has been taken place through the interface. It was concluded that the time duration of control rod to reach the bottom of the core depends on the leakage. It was also observed that the velocity and acceleration of the control rod would be reduced by decreasing leakage flow rate and in certain leakages, the acceleration of the control rod approaches zero due to equilibrium conditions. During this research, a correlation based on the achieved data was proposed which would provide useful information on the relation between the leakage and the time for control rod to reach the bottom of the core.     

Studies on LLFP transmutation in a pressurized water reactor

A systematic study on the long-lived fission product (LLFP) transmutation in a pressurized water reactor (PWR) is performed, aiming at an optimal transmutation strategy for present nuclear energy development. The LLFPs selected in the analysis include ⁹⁹Tc and ¹²⁹I discharged from light water reactors (LWRs). The isotope ¹²⁷I is also considered to avoid the difficulties in isotopes separation. To minimize the negative impacts of LLFPs on the core performance and safety parameters, metallic technetium or MgI₂ target pins mixed with ZrH₂ are designed and investigated. Through the numerical analysis on equilibrium cycles, the transmuted amounts of ⁹⁹Tc and ¹²⁹I equal to the yields from 1.94 and 4.22 PWRs with a power of 1000 MWe, respectively. Numerical results indicate that both ⁹⁹Tc and ¹²⁹I can be transmuted conveniently in present PWRs in the form of target pins.     

Analytical investigation of a one-dimensional homogenized model for a pressurized water reactor core

A one-dimensional homogenized model for dynamic fluid-structure interaction in a pressurized water reactor core is used to study the influence of the virtual density and spacer's stiffness. The model consists of a linear system of partial differential equations for fluid velocity, rod velocity and pressure. For these equations analytical solutions are deduced for boundary conditions prescribing either periodic wall oscillations or linearly growing wall accelerations from rest. The theoretical model for the virtual density is verified by comparison to an experiment. For zero spacer stiffness, purely acoustic oscillations appear. For positive spacer stiffness, additional oscillations arise with relative rod motions. The wavelengths of the latter oscillations are small for weak spacers. Large numerical effort would be required in a more complete three-dimensional core-model to resolve such short wave lengths. In fact in a typical core the spacer's stiffness c_s is small in comparison to the fluid bulk modulus K. For it might be appropriate to neglect the influence of the spacers.     

Homogenized model for fluid-structure interaction of the pressurized water reactor core internals during blowdown

A method for the numerical simulation of the pressurized water reactor core internal's behaviour during a blowdown accident is described, by which the motion of the reactor core and the interaction of the fuel elements with the core barrel and the coolant medium is calculated. Furthermore, some simple models for the support columns, lower and upper core support and the grid plate are provided. In order to investigate the global core motion during the blowdown accident, the core model describes the coupled fluid-rod motion with Homogenization methods. The heterogeneous fluid-rod mixture thus is treated as a special continuum with anisotropic material properties. Furthermore, the core model considers elastical rod forces against bending and axial straining and the direct interaction of neighbouring fuel elements, which is a highly nonlinear process due to the finite gaps. Because this effect is very important, two simulation models have been developed and are compared. All these models have been implemented into the blowdown code FLUX-4. With the new code version FLUX-5 the PWR-blowdown is parametrically investigated.     

蒸汽发生器完全丧失给水引发的压水堆严重事故研究

采用严重事故最佳估算程序SCDAP/RELAP5/MOD3.4,建立了美国Surry核电站的详细计算模型,对完全丧失给水(TLFW)引发的堆芯熔化事故进行了研究分析.为准确预测压力容器内堆芯熔化的进程,给二级PSA提供可信的初始条件,计算中考虑了一回路压力边界的蠕变破裂失效,并评价了人为干预对堆芯熔化进程及事故后果的影响.     

Fission-product chemistry and aerosol behaviour in the primary circuit of a pressurized water reactor under severe accident conditions

This review considers fission-product chemistry and aerosol behaviour in the primary circuit of a pressurized water reactor (PWR) during severe accidents. Three key accident sequences (V, TMLB' and S₂D) are considered, and their principal thermal-hydraulic and physical characteristics affecting chemistry behaviour are identified. The inventories, chemical forms and timing of fission products released from the fuel are summarized together with the major sources of structural materials and their release characteristics. The chemistry of each main fission-product species within the primary circuit is reviewed from available experimental and thermodynamic data and/or theoretical predictions. Modelling studies of primary circuit fission-product behaviour are reviewed briefly and the principal requirements for further study assessed with respect to experimental and modelling programmes currently in progress.     

A 10 MWe pressurized water reactor for safe and reliable power

A design for an innovative, passively safe 10 MWe power plant based on the proven pressurized water reactor technology has been developed. The plant incorporates an innovative design approach to achieve “walk-away” safety and includes significant simplification and elimination of systems and components when compared to the current generation commercial nuclear power plants. The plant has been designed such that the majority of the equipment will be pre-assembled as modules at off-site facilities and shipped to the site on trucks for installation. This approach will provide shorter construction schedules and improved quality control.     

Thermal hydraulic analysis for grid supported pressurized water reactor cores

This paper presents the methodology and results for thermal hydraulic analysis of grid supported pressurized water reactor cores using U(45% wt)-ZrH_1.6 hydride fuel in square arrays. The same methodology is applied to the design of UO₂ oxide fueled cores to provide a fair comparison of the achievable power between the two fuel types. Steady-state and transient design limits are considered. Steady-state limits include: fuel bundle pressure drop, departure from nucleate boiling ratio, fuel temperature (average for UO₂ and centerline/peak for U-ZrH_1.6), and fuel rod vibrations and wear. Transient limits are derived from consideration of the loss of flow and loss of coolant accidents, and an overpower transient.In general, the thermal hydraulic performance of U-ZrH_1.6 and UO₂ fuels is very similar. Slight power differences exist between the two fuel types for designs limited by rod vibrations and wear, because these limits are fuel dependent. Large power increases are achievable for both fuels when compared to the reference core power output of 3800 MW_th. In general, these higher power designs have smaller rod diameters and larger pitch-to-diameter ratios than the reference core geometry. If the pressure drop across new core designs is limited to the pressure drop across the reference core, power increases of ∼400 MW_th may be realized. If the primary coolant pumps and core internals could be designed to accommodate a core pressure drop equal to twice the reference core pressure drop, power increases of ∼1000 MW_th may be feasible.     

The design and evaluation of a passively cooled containment for a high-rating pressurized water reactor

An integrated pressurized water reactor (PWR) containment was conceptualized that allows heat to be rejected passively to the environment. The proposed containment is based on the demonstrated Ebasco Waterford 3 design. The secondary concrete shell was equipped with inlet and outlet vents that create an air-convection annulus. These vents also permit the submersion of the lower part of the primary containment into an external water pool. An internal water pool located at the bottom of the lower containment was added to increase in-containment heat storage. The performance of the proposed passively cooled containment was evaluated using a subdivided volume code, version 3.4e; the relative novelty of subdivided volume analyses for containment performance evaluation requires experimental verification of principal code predictions. Two experiments were carried out; one to test the performance of the external moat, and one to verify the code’s ability to predict thermal-stratification inside the containment. To improve the subdivided-volume simulation of convection-related parameters, a modeling technique (boundary layer flow approximation) was devised. Finally, the behavior of the proposed containment was evaluated for the worst-case large break loss of coolant accident and the worst-case main steam line break accident. Peak pressures remained below 0.45 MPa during both transients; internal wall pressure differences, equipment qualification temperatures, pressure restoration time also remained below design limits. The mitigation capability of hydrogen recombiners was also evaluated.     

Design of the advanced accumulator for the pressurized water reactor

With the increased requirement for nuclear power generation as an effective countermeasure against global warming, Mitsubishi has developed the advanced pressurized water reactor (APWR) by adopting a new component of the emergency core cooling system (ECCS), a new instrumentation and control system, and other newfound improvements. The ECCS introduces a new passive component called the advanced accumulator which integrates both functions of the conventional accumulator and the low-pressure pump without any moving parts. The advanced accumulator uses a new fluidics device that automatically regulates flow rates of injected water in case of a loss of coolant accident (LOCA). This fluidics device is referred to as a flow damper. This paper describes the design method of the flow damper and the standpipe.     

Development of a thermal-hydraulic analysis software for the Chinese advanced pressurized water reactor

A point reactor neutron kinetics model, a drift-flow U-tube steam generator model, a non-equilibrium three-region pressurizer model and other models were established and a transient analysis code with Visual Fortran 6.5 has been developed to analyze the thermal-hydraulic characteristics of the Chinese advanced pressurized water reactor (AC-600). Visual input, real-time processing and dynamic visualization output were achieved with Microsoft Visual Studio.NET 2003, which greatly facilitate applications in the engineering. The software were applied to analyze the transient thermal-hydraulic characteristics of the loss of feed-water accident, the double loops loss-of-flow accident, the reactivity insertion accident, the sudden increase of feed-water temperature accident and the loss of offsite power accident for the Qinshan nuclear power plant in China. The obtained analysis results are significant to the improvement of design and safety operation of the plant.     

Availability of a pressurized water reactor pressure vessel after a small LOCA

The integrity of a pressure vessel after a loss-of-coolant accident is investigated. The pressure and temperature in the pressure vessel is found by RELAP-4 for different break sizes. The effect of the pressure and thermal shock on the pressure vessel is evaluated by a two-dimensional finite element method (code SAP). By assuming axisymmetry, stress resulting from the shocks is calculated in the nozzle shell and in the beltline region of the pressure vessel. Similar calculations are carried out for the closure head and the lower head regions of the pressure vessel. Fracture mechanics is used to evaluate the stress intensity factor as a function of crack depth, and the critical crack depths are then obtained at various locations in the pressure vessel. The post-LOCA usability of the pressure vessel is predicted by comparing the stress intensity factor with the fracture toughness of the steel.     

Concept of a pressurized water reactor cooled with supercritical water in the primary loop

A novel concept of a pressurized water reactor with a primary loop cooled with supercritical water is introduced and analyzed in this work. A steam cycle analysis has been performed to illustrate the advantages of such a nuclear power plant with respect to specific power and thermal efficiency. Moreover, a reactor pressure vessel concept including its internals and a suitable core and fuel assembly design are presented overcoming the problems, which arise due to the high heat up of the coolant and the density change involved with it. The core power and coolant density distributions are predicted with coupled neutronic and thermal-hydraulic analyses. The method features the definition of inlet orifices for coolant mass flow adjustment within the core as well as an additional tool for the interpolation of local pin power data. The latter one has been used for a successive sub-channel analysis of the hottest fuel assembly of the core, which provides a more detailed spatial resolution and thus predicts peak cladding temperatures, the maximum linear pin power of fuel pins, and maximum fuel temperatures. It can be shown that maximum temperatures of claddings and fuel are well below the material limits. Thanks to an average core exit temperature below the pseudo-critical temperature, the core concept leaves enough margin for additional uncertainties and allowances for operation.     

A class-based search for the in-core fuel management optimization of a pressurized water reactor

The In-Core Fuel Management Optimization (ICFMO) is a prominent problem in nuclear engineering, with high complexity and studied for more than 40 years. Besides manual optimization and knowledge-based methods, optimization metaheuristics such as Genetic Algorithms, Ant Colony Optimization and Particle Swarm Optimization have yielded outstanding results for the ICFMO. In the present article, the Class-Based Search (CBS) is presented for application to the ICFMO. It is a novel metaheuristic approach that performs the search based on the main nuclear characteristics of the fuel assemblies, such as reactivity. The CBS is then compared to the one of the state-of-art algorithms applied to the ICFMO, the Particle Swarm Optimization. Experiments were performed for the optimization of Angra 1 Nuclear Power Plant, located at the Southeast of Brazil. The CBS presented noticeable performance, providing Loading Patterns that yield a higher average of Effective Full Power Days in the simulation of Angra 1 NPP operation, according to our methodology.     

Impacts and fluid-structure interaction in pressurized water reactor safety analysis

A numerical method is described for the analysis of coupled three-dimensional fluid-structure motion with impacts between structural parts at rigid or flexible supports with small clearances. The method is used for the analysis of the blowdown loadings and the response of internal structures in the vessel of a pressurized water reactor (PWR) in the hypothetical event of a sudden break of a coolant inlet pipe. The method is a generalization of the existing code FLUX which simulates the three-dimensional fluid-structure motion by means of an implicit time integration scheme. The additional supports with clearances are taken into account by applying support forces to the freely moving fluid-structure system. The forces are determined such that the kinematic constraints are enforced at each time step. Numerically, these forces are determined efficiently using a precomputed influence matrix which defines the dynamic displacement per time step at each support due to a unit force at each other support. According to the actually “active” supports the relevant influence matrix in constructed. Energy is conserved for rigid supports and for supports which are so flexible that the impact time is large in comparison to the time steps. Treatment of plastic supports is possible.An application of the new method is demonstrated by analysis of the core barrel motion in a PWR with and without impacts at the lower core barrel edge and at the upper flange. The results show the large effects of such impacts in changing the global motions. Large local impact forces and accelerations appear. The interaction with the fluid reduces these loads. By proper design of the supports the resultant stresses can be minimized. Thus the method can be used to demonstrate and enlarge nuclear reactor safety.