Some Aspect of Protective Systems in Nuclear Power Plants

Protective systems for nuclear power reactors have assumed a high degree of development. The primary motive has been the emphasis given to making reactors safe and more recently the desire to offset the site distance requirement through engineered safeguards. Increased reactor operating experience and improved safety analyses have contributed to a better definition of the events and conditions requiring protective systems. The judicious use of such philosophies as redundancy and coincidence in system designs has led to both greater safety and reactor operating continuity. Reliability of protective systems has been enhanced by parallel efforts in system and individual component design. Techniques and procedures for checking and testing protective systems have been developed and adopted at many installations to offset the inherent difficulty of assessing reliability of systems which experience little or no use under actual stressed conditions. Design practices have been effected to provide greater assurance that systems are independent although this remains as one of the outstanding problems. Protective systems in the context of monitoring devices are being developed by the application of noise analyses, digital computer control and the use of transistorized or solid state circuitry. Finally, the actual performance of protective systems is being manifested through analysis of existing operating records.     

RAPID Operator-Free Fast Reactor Concept without Any Control Rods Reactor Concept and Plant Dynamics Analyses

The 1,000kWe metal fueled sodium-cooled fast reactor concept “RAPID” to achieve highly automated reactor operation has been demonstrated. RAPID (Refueling by All Pins Integrated Design) is designed for a terrestrial power system which enables quick and simplified refueling. It is one of the successors of the RAPID-L, the operator-free fast reactor concept designed for lunar base power system. The essential feature of the RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small-size reactor core, 14,000 fuel pins are integrated and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years.

Unique challenges in reactivity control systems design have been addressed in the RAPID concept. The reactor has no control rod but involves the following innovative reactivity control systems: lithium expansion modules (LEM) for inherent reactivity feedback, lithium injection modules (LIM) for inherent ultimate shutdown, and lithium release modules (LRM) for automated reactor startup. All these systems adopt ⁶Li as a liquid poison instead of B₄C rods. In combination with LEMs, LIMs and LRMs, RAPID can be operated without an operator. In this paper, the RAPID reactor concept and its transient characteristics are presented.     

Comparison of accelerator-based with reactor-based nuclear waste transmutation schemes

An overview of the most significant studies in the last 35 years of partitioning and transmutation of commercial light water reactor spent fuel is given. Recent Accelerator-based Transmutation of Waste (ATW) systems are compared with liquid-fuel thermal reactor systems that accomplish the same objectives. If no long-lived fission products (e.g., ⁹⁹Tc and ¹²⁹I) are to be burned, under ideal circumstances the neutron balance in an ATW system becomes identical to that for a thermal reactor system. However, such a reactor would need extraordinarily rapid removal of internally-generated fission products to remain critical at equilibrium without enriched feed. The accelerator beam thus has two main purposes (1) the burning of long-lived fission products that could not be burned in a comparable reactor's margin (2) a relaxing of on-line chemical processing requirements without which a reactor-based system cannot maintain criticality. Fast systems would require a parallel, thermal ATW system for long-lived fission product transmutation. The actinide-burning part of a thermal ATW system is compared with the Advanced Liquid Metal Reactor (ALMR) using the well-known Pigford-Choi model. It is shown that the ATW produces superior inventory reduction factors for any near-term time scale.     

Time Optimal Digital Computer Control of Nuclear Reactors, (I)

Time optimal control of nuclear reactors has been studied using Pontryagin's maximum principle. Since the results are intended for application in a digital computer control system, restrictions on the maximum reactivity and on the maximum over (under) -shoot of reactor power are imposed to derive a practical control pattern. A study of continuous time systems is attempted in this paper with the view to deriving the basic characteristics of optimal solution and providing useful guides to future work on discrete time systems, including digital computer, to be treated in a forthcoming paper.

To facilitate synthesis of optimal control, the reactor model is simplified with the use of a one group model for delayed neutron and prompt jump approximations. As a result, the time optimal control pattern is found to be uniquely determined in a delayed neutron density (or reactor power) vs. reactivity phase plane for any range of power variation. The optimal control law is simple, and can be adopted in a feedback control system.     

Development of advanced I&C in nuclear power plants: ADIOS and ASICS

In this paper Automatic Startup Intelligent Control System (ASICS) that automatically controls the PWR plant from cold shutdown to 5% of reactor power and Alarm and Diagnosis-Integrated Operator Support System (ADIOS) that is integrated with alarms, process values, and diagnostic information to an expert system focused on alarm processing are described. Nuclear Power Plant is manually controlled from cold shutdown to 5% according to the general operation procedures for startup operation of nuclear power plant. Alarm information is the primary sources to detect abnormalities in nuclear power plants or other process plants. The conventional hardwired alarm systems, characterized by one sensor-one indicator may lead the control room operators to be confused with avalanching alarms during plant transients. ASICS and ADIOS are designed to reduce the operator burden. The advances in computer software and hardware technology and also in information processing provide a good opportunity to improve the control systems and the annunciator systems of nuclear power plants or other similar process plants. It is very important to test and evaluate the performance and the function of the computer- or software-based systems like ASICS and ADIOS. The performance and the function of ASICS and ADIOS are evaluated with the real-time functional test facility and the results have shown that the developed systems are efficient and useful for operation and operator support.     

SSTAR: The US lead-cooled fast reactor (LFR)

It is widely recognized that the developing world is the next area for major energy demand growth, including demand for new and advanced nuclear energy systems. With limited existing industrial and grid infrastructures, there will be an important need for future nuclear energy systems that can provide small or moderate increments of electric power (10-700 MWe) on small or immature grids in developing nations. Most recently, the global nuclear energy partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium-sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the small secure transportable autonomous reactor (SSTAR) has been under ongoing development as part of the US advanced nuclear energy systems programs. It is a system designed to provide energy security to developing nations while incorporating features to achieve nonproliferation goals, anticipating GNEP objectives. This paper presents the motivation for development of internationally deployable nuclear energy systems as well as a summary of one such system, SSTAR, which is the US Generation IV lead-cooled fast reactor system.     

加速器驱动次临界堆的安全和控制特性初步研究

加速器驱动次临界反应堆的结构特点使其安全和控制特性有别于临界反应堆。本工作使用数值计算和仿真运行方法,驱动堆的安全和控制特性进行了初步研究。结果表明：驱动堆不易发生瞬发临界,其安全特性优于临界堆,次临界度越深,安全性越好;驱动堆控制回路具有小的时间常数和超调量,调整时间短,控制特性亦优于临界堆。     

Best Estimated and Conservative Analyses for a Feedwater Pipe Break Accident of an Integral Type Reactor

An integral type reactor, which is an innovative design to achieve a high degree of safety, is currently being developed at the Korea Atomic Energy Research Institute. A feedwater pipe break accident is one of the most important accidents regarding the safety of an integral type reactor. A best estimated calculation, a conservative calculation, and a parameter study for a feedwater pipe break have been carried out. The sensitivity analysis in this paper performed is to establish the parameters which greatly affect the feedwater pipe break accident. A power level, an initial system pressure, a moderator reactivity coefficient and a break size are the major parameters which maximize a system pressure. The important function that must operate following a feedwater pipe break accident is an opening of the pilot operated safety relief valves, and an initiation of the passive residual heat removal system. The integral reactor safety systems function properly and thus secure the reactor to a safe condition with respect to the safety parameters.     

Time Optimal Digital Computer Control of Nuclear Reactor, (II)

A sampled data time optimal control of nuclear reactors has been developed using discrete versions of the Maximum Principle. The reactor model used is a second order nonlinear system derived by one group delayed neutron and prompt jump approximations. The control variable of the system is the rate of change of reactivity, and for reasons of safety in reactor operation, this, as well as the reactivity, is restricted. In addition, the reactor power overshoot and undershoot are also restricted to devise a practical control.

The discrete version of the Maximum Principle has been extended to deal with restricted state space problem, and with the use of this extension, a time optimal control for nuclear reactors with pulse width modulated control input has been derived.

The optimal control law is modified in the region around the terminal state, to obtain smooth control, and the stability of this modified control is verified by the second method of Iiapunov.

The results are directly applicable to digital computer control systems for nuclear reactors.     

The case of an oscillating manometer with variable density and dissipation: experimental and numerical study

With the increasing use of computational fluid dynamics (CFD)-based simulations in the assessment of thermal hydraulic behaviour of nuclear reactors, there is a need to benchmark the predictions of CFD codes. The case of a frictionless U-tube manometer [Ransom, V., 1992. Oscillating manometer. In: Hewitt, G.F., Delhaye J.M., Zuber N. (Eds.), Multiphase Science and Technology, vol. 6. Hemisphere, New York, USA, p. 591] has been previously used as a test case to benchmark the dispersive and dissipative characteristics of a numerical prediction. In the present paper, this case is extended to cover two new aspects that are often found in nuclear reactor systems: variable density and dissipation. Through a theoretical model, it is shown that the dynamic behaviour of the new system is more complicated than that of a constant-density frictionless U-tube manometer. Experiments of the test case have been carried out to substantiate the essential features of the response. CFD simulations of the system show both dissipative and dispersive errors compared to the theoretical result. The results can therefore serve as a useful benchmark for computer codes used to study the thermal hydraulics of nuclear reactor systems, especially in evaluating the role of numerical damping in systems in which physical damping exists inherently.     

Applying mechanistic models to reliability evaluation of mechanical components – An illustration

The traditional reliability analyses, considers components to be in binary state, either functional or faulty, and does not consider the concept of multi state or intermediate states between these two binary states. However, there are several components, which need to be operated in different states and their failure criterion also depend on these states. Hence, when dealing with these types of components one should use multi state concept. This can be achieved by modeling the components with mechanistic models, which can give a new dimension for reliability analysis for multiple states. The mechanistic model approach is based on the first principles of science and engineering which provides details about the various failure mechanisms and thereby improved understanding of the associated root causes of the failure and driving forces responsible for component failures. In this paper a general methodology for modeling the components with mechanistic models has been explained and is further illustrated with an example component. A case study on feed water system (consisting of control valves and other mechanical components) of a typical nuclear reactor has been presented.     

Reactivity control schemes for fast spectrum space nuclear reactors

Several reactivity control schemes are considered for future space nuclear reactor power systems. Each of these control schemes uses a combination of boron carbide absorbers and/or beryllium oxide reflectors to achieve sufficient reactivity swing to keep the reactor subcritical during launch and to provide sufficient excess reactivity to operate the reactor over its expected 7-15 year lifetime. The size and shape of the control system directly impacts the size and mass of the space reactor's reflector and shadow shield, leading to a tradeoff between reactivity swing and total system mass. This paper presents a trade study of drum, shutter, slat, and petal control schemes based on reactivity swing and mass effects for a representative fast-spectrum, gas-cooled reactor. For each control scheme, the dimensions and composition of the core are constant, and the reflector is sized to provide $5 of cold-clean excess reactivity with each configuration in its most reactive state. Reactivity insertion behavior is analyzed for each control scheme, along with the submersion behavior following a launch abort. The advantages and disadvantages of each configuration are discussed, along with optimization techniques and novel geometric approaches for each scheme.     

用于压水堆负荷跟踪运行的硼浓度模糊控制系统

对压水堆负荷跟踪运行进行了研究，提示将一种三维模糊控制系统应用于硼浓度自动调节的设计方案，在设计中，将核电站全范围模拟机的数学模型移植到微机上作为控制对象，对硼浓度模糊控制系统进行仿真实验的结果证实，该模糊控制方法不仅可行，而且效果良好。     

Buoyantly-driven two-phase countercurrent flow in liquid discharge from a vessel with an unvented gas space

This paper details experiments and analyses regarding the phenomenon of liquid discharge into a gaseous atmosphere from the bottom of a vessel with an unvented, upper gas space. The primary goal is the development of a simple model that predicts the rate of liquid discharge under the prevailing unvented condition. A literature survey of previous work on this phenomenon yielded only simple experiments and analyses that were limited in scope. Experiments were subsequently undertaken with an air-water system, using a larger volume and a wide range of drain line diameters. In addition to flowrate data, visual information was acquired regarding the physical mechanism possibly governing the prevalent flow regimes. The governing physical mechanism is identified as the stability of a gas-liquid interface, perturbed by buoyancy, at the drain line entrance. G.I. Taylor's fundamental analysis of interfacial stability lead to the determination of criteria for flow regime transition among the three prevalent flow regimes, corresponding to so-called small, medium, and large diameters. Also, analysis of the growth of interfacial instabilities lead to the application of flooding models for drainage rates within each regime. The models for moderate and large diameters were then compared against data, which confirmed their success in predicting discharge rates under the unvented condition.The motivation for this effort, besides the basic scientific significance of studying such a fundamental phenomenon, was its numerous applications, one of which is commercial nuclear reactor systems. Specifically, the phenomenon prevails in liquid coolant discharge from a PWR pressurizer, with an unvented steam volume, into a steam atmosphere existing in the adjoining hot coolant leg. Such a phenomenon could occur as part of a transient, or severe accident, scenario, entailing saturated conditions and steam production in the normally subcooled primary heat transport loop. The developed model was implemented in the Modular Accident Analysis Program (MAAP), a computer code designed to predict reactor system behavior in response to postulated off-normal conditions, including severe accident scenarios.     

计算机模拟γ测量谱

一个建立在微机DOS环境下的计算机模拟γ射线测量谱系统以一个实测的^135Cs谱作为模板,对活化测量谱和同位素源的γ射线谱进行了成形模拟。谱中包含了对全能峰、康普顿沿、康普顿坪和反散射峰的模拟。模拟系统具有谱峰面积计算、谱成形、谱显示、谱处理、谱存盘等功能。利用该模拟程序对^137Cs源、^152Eu源、^60Co源的γ射线谱以及锆样品和岩石样品的中子活化测量谱进行了模拟,模拟谱与实际测量谱非常接近。     

A neuro-fuzzy controller for axial power distribution an nuclear reactors

A neuro-fuzzy control algorithm is applied for the core power distribution in a pressurized water reactor. The inputs of the neural fuzzy system are composed of data from each region of the reactor core. Rule outputs consist of linear combinations of their inputs (first-order Sugeno-Takagi type). The consequent and antecedent parameters of the fuzzy rules are updated by the backpropagation method. The reactor model used for computer simulations is a two-point xenon oscillation model based on the nonlinear xenon and iodine balance equations and the one group, one-dimensional neutron diffusion equation having nonlinear power reactivity feedback. The reactor core is axially divided into two regions, and each region has one input and one output and is coupled with the other region. The interaction between the regions of the reactor core is treated by a decoupling scheme. This proposed control method exhibits very fast response to a step or a ramp change of target axial offset without any residual flux oscillations between the upper and lower halves of the reactor core.     

BWR online monitoring system based on noise analysis

A monitoring system for during operation early detection of an anomaly and/or faulty behavior of equipment and systems related to the dynamics of a boiling water reactor (BWR) has been developed. The monitoring system is based on the analysis of the “noise” or fluctuations of a signal from a sensor or measurement device. An efficient prime factor algorithm to compute the fast Fourier transform allows the continuous, real-time comparison of the normalized power spectrum density function of the signal against previously stored reference patterns in a continuously evolving matrix.The monitoring system has been successfully tested offline. Four examples of the application of the monitoring system to the detection and diagnostic of faulty equipment behavior are presented in this work: the detection of two different events of partial blockage at the jet pump inlet nozzle, miss-calibration of a recirculation mass flow sensor, and detection of a faulty data acquisition card. The events occurred at the two BWR Units of the Laguna Verde Nuclear Power Plant.The monitoring system and its possible coupling to the data and processing information system of the Laguna Verde Nuclear Power Plant are described. The signal processing methodology is presented along with the introduction of the application of the evolutionary matrix concept for determining the base signature of reactor equipment or component and the detection of off normal operation conditions.     

Gas-cooled fusion-fission hybrid reactor systems

In this paper, the concept of the fusion-fission hybrid reactor is reviewed, and a system of classification for hybrid blanket designs is suggested. The advantages and disadvantages of gas cooling for hybrid reactor systems are discussed and the design implications of using gas cooling in a hybrid blanket are presented. Five of the more complete gas-cooled hybrid reactor conceptual design studies are discussed, and the fission-suppressed hybrid blanket concept is identified as offering potentially significant advantages in terms of inherent safety features and reduced technology development requirements compared to higher power fission blankets. It is concluded that helium is attractive as the coolant for hybrid reactor systems, and that technically viable reactor designs have been developed using helium cooling. The helium-cooled fission-suppressed hybrid blanket, based on thorium fuel for production of²³³U, is identified as being a particularly attractive candidate for further hybrid reactor development work.     

反应堆系统数字化实时仿真的高效实现

根据实验反应堆的物理特性，建立堆芯动态模型，探讨多种实时仿真算法的实现途径。研究提出了进行数字化实时仿真的一种高效实现方法。为配合功率调节系统半实物仿真试验而实现了一座实验反应堆在Windows平台下的实时仿真系统。      

Stability Analysis of a Nuclear Reactor Control System with Multiple Transport-Lags and an Asymmetrical Nonlinearity

In this paper, the stability-equation method is applied to the analysis of a nuclear reactor control system with multiple transport-lags and an asymmetrical nonlinearity. The characteristics of limit-cycles can be defined easily, and the effects of adjustable parameters can be realized directly. Based upon the presented analyses, the considered system has asymmetrical oscillation. These are checked by computer simulations.