Based on the presented specifics of RBMK, two categories of Beyond Design Basis Accidents (BDBA), i. e. core damage types in RBMK reactors were proposed by RBMK designers from Russia (Kurchatov Institute and Research and Development Institute of Power Engineering) [8]:

• damage of the core or its components with the reactor maintaining its overall structural integrity;

• total damage of the core, resulting in the loss of general structural integrity of the reactor system.

Such grouping of accident is very important regarding accident management. If the core or its components remain structurally intact (first category of core damage), then the controlling actions (accident management) for limiting and delaying damage of the core, as well as prevention of confinement damage and mitigation of fission products release are possible. If the general structural integrity of the reactor system is lost, then depending on the degree to which the general structural integrity of the reactor is maintained, (second category of core damage), the emergency plan has to be activated in order to protect the public (sheltering, evacuation, etc.).

The first category of core damage can be further subdivided into the following accident groups (Fig. 5):

— no severe damage of the core (1.1);

— severe core damage accompanied by containment of the core fragments in the reactor core, accident localization system or other reactor buildings (1.2).

Accidents, leading to a complete reactor core damage, with loss of structural integrity of the reactor can also be divided into two groups [9]:

— accidents when heat-up of the reactor core occurs during reactor operation or within the first seconds after the reactor shutdown, when decay heat is high (2.1);

—

accidents when heat-up of the reactor core occurs after the reactor shutdown (2.2).

As it was mentioned, such grouping of accidents is the starting point for the development of the measures for accident management. However, the development of accident management guidelines requires performing deterministic analysis of all possible accidents in each group. Based on this analysis the accident consequences, available time for possible operator actions and possible modifications of emergency systems may be determined.

In the next chapter the deterministic analysis of reactor core and reactor cooling system is presented, whereas the modeling of the process in the RBMK confinement is presented in the monograph [10].

Subtracting the structural damping ratio from the total yields the two-phase fluid-damping ratio (Noghrehkar et al., 1995). Total damping includes structural damping, viscous damping and a two-phase component of damping as explained by (Pettigrew et al. 1994). The damping ratio increases as the void fraction increases and peaks at 60% (Carlucci, 1983), then the ratio decrease with a (Figure 20). Damping also decreases as the vibration frequency increases (Pettigrew et al., 1985).

Damping in two-phase is very complicated. It is highly dependent upon void fraction and flow regime. The results for the two-phase component of damping can be normalized to take into account the effect of confinement due to surrounding tubes by using the confinement factor C (Pettigrew et al., 2000). This factor is a reasonable formulation of the confinement due to P / D. As expected, greater confinement due to smaller P / D increase damping. The confinement factor is given by equation below:

[1 + (D /De)3]

[1 — (D /De)2] 2

A schematic diagram of the experimental apparatus is given in Fig.1, which is adopted to study the boiling two-phase flow in tube-bundle channels. In the experiments, all measured

data are recorded at the intervals of 10ms. The error in the pressure drop measurement is less than ±2%. The error in flow rate measurement is less than ±3%, and for temperature measurement is less than ±1%.

The test section of tube-bundle channel is composed of four straight electrical heaters and a thimble, which is shown in Fig. 2. The thimble is made of PMMA pipe for the purpose of flow visualization, of which the inner diameter is 35mm and the length is 2100mm. The dimension of the electrical heater is Ф8шш and 1500mm long; the gap between electrical heaters is 5mm; the width between electrical heater and thimble is 4.5mm. There are, along flow direction, three pressure sensors and five thermocouples. The length between pressure sensors is 200mm and that between thermocouples is 100mm.

Plexiglas pipe

Electric heat tube P3

P2

P1

Thermocouples

Fig. 2. Structure of test section

A schematic diagram of sub-channels partition is shown in Fig. 3. There are 9 sub-channels in the present test section.

Jurgen Rudolph*, Steffen Bergholz, Benedikt Heinz and Benoit Jouan

AREVA NP GmbH Erlangen Germany

1. Introduction

1.1 General remarks

Within the continuously accompanying licensing process for NPPs until the end of their operational lifetime, the ageing and lifetime management plays a key role. Here, one of the main tasks is to assure structural integrity of the systems and components. With the help of the AREVA Fatigue Concept (AFC), a powerful method is available. The AFC provides different code-conforming fatigue analyses (e. g. according to the wide spread ASME code [1]) based on realistic loads. In light of the tightening fatigue codes and standards, the urge is clearly present that, in order to still be able to comply with these new boundaries, margins which are still embedded within most of the fatigue analyses in use, have to be reduced. Moreover, thermal conditions and chemical composition of the fluid inside the piping system influences the allowable fatigue levels, which have come under extensive review due to the consideration of environmentally assisted fatigue (EAF) as proposed in the report [2]. Therefore, for highly loaded components, some new and improved stress and fatigue evaluation methods, not overly conservative, are needed to meet the increasingly stringent allowable fatigue levels. In this context, the fatigue monitoring system FAMOS, central module of AFC, is able to monitor and record the real local operating loads. The different modules of the AFC are schematically represented in Figure 1.

Robertas Alzbutas12, Egidijus Norvaisa1 and Andrea Maioli3

Lithuanian Energy Institute 2Kaunas University of Technology 3Westinghouse Electric Company 12Lithuania 3USA

1. Introduction

Probabilistic Risk Assessment (PRA) tehniques applied to the definition of Emergency Planning Zone (EPZ) have not reached the same level of maturity when dealing with external events as PRA methodologies related only to internal events (Alzbutas et al., 2005). This is even of greater importance and relevance when PRA is used in the design phase of new reactors (IAEA-TECDOC-1511, 2006; IAEA-SSG-3, 2010; IAEA-SSG-4, 2010).

The design of the layout of a Nuclear Power Plant (NPP) within its identified site, with the arrangement of its structures, as well as the definition of the EPZ around the site can be used to maximise the plant safety related functions, thus further protecting nearby population and environment. In this regard, the design basis for NPP and site is deeply related to the effects of any postulated internal and external hazardous event and the possibilities of the reactor to cope with related accidents (i. e., to perform the plant safety related functions).

Among the objectives for advanced reactors there is the aim to establish such a higher safety level with improved design characteristics that would justify and enable revised emergency planning requirements. While providing at least the same level of protection to the public as the current regulations, ideally, but still not realistically, the total elimination of hazards’ consequences would result in the EPZ coincidinge with the site boundary, thus, there would be no need for off-site evacuation planning, and the NPP would be perceived as any other industrial enterprise.

In this chapter, the International Reactor Innovative and Secure (IRIS) is adopted as a prime example of an advanced reactor with enhanced safety. The IRIS plant (Carelli, 2003, 2004, 2005) used a Safety-by-Design™ philosophy and such that its design features significantly reduced the probability and consequences of major hazardous events. In the Safety-by­Design™ approach, the PRA played a key role; therefore a Preliminary IRIS PRA had been

developed along with the design, in an iterative fashion (Kling et al., 2005). This unprecedented application of the PRA techniques in the initial design phase of a reactor was also extended to the external event with the aim of reviewing the EPZ definition. To achieve this particular focus was dedicated to PRA and Balance Of Plant (BOP).

For the design and pre-licensing process of IRIS, the external events analysis included both qualitative evaluation and quantitative assessment. As a result of preliminary qualitative evaluation, the external events that had been chosen for more detailed quantitative assessment were as follows: high winds and tornadoes, aircraft crash and seismic activity (Alzbutas et al., 2005, Alzbutas & Maioli, 2008).

In general, the analysis of external events with related bounding site characteristics can also be used in order to optimize the potential future restrictions on plant siting and risk zoning. Due to this and Safety-by-Design™ approach, IRIS, apart from being a representative of innovative and advanced reactors, had the necessary prerequisite, (i. e., excellent safety), for attempting a redefinition of EPZ specification criteria, IRIS was therefore used as a test-bed.

The work presented in this chapter was performed within the scope of activities defined by the International Atomic Energy Agency (IAEA) Co-ordinated Research Project (CRP) on Small Reactors with no or infrequent on-site refuelling. Specifically, it was relevant to "Definition of the scope of requirements and broader specifications" with respect to its ultimate objective (revised evacuation requirements), and to "Identification of requirements and broader specifications for NPPs for selected representative regions" considering specific impact on countries with colder climate and increased interest for district heating co-generation.

The economic modelling and optimization presented in the second part of the chapter was concentrating on the evaluation of possibilities to construct a new energy source for Lithuania. The MESSAGE modelling tool was used for modelling and optimization of the future energy system development (IAEA MESSAGE, 2003). In this study, the introduced approach was applied focusing on Small and Medium nuclear Reactor (SMR), which was considered as one of the future options in Lithuania. As an example of SMR, the IRIS nuclear reactor was chosen in this analysis.

If IRIS with reduced EPZ could be built near the cities with a big heat demand is, it could be used not only for electricity generation, but also for heat supply for residential and industrial consumers. This would allow not only to reduce energy prices but also to decrease fossil fuel consumption and greenhouse gas emissions.

Finally, the analysis of uncertainty and sensitivity enabled to investigate how uncertain were results of this modelling and how they were sensitive to the uncertainty of model parameters (Alzbutas et al, 2001).

In summary, the study presented in this chapter consists of two main parts: the analysis of EPZ in relation to PRA with focus on external events, and the economic optimization of future energy system development scenarios with focus on sensitivity and uncertainty analysis in relation to initial model parameters. The study explicitly uses features of IRIS technology and a potentially reduced EPZ.

(Kim et al., 2009) have carried flow induced vibrations (Experimental study of two circular cylinders in tandem arrangement) and examined three different experimental conditions both cylinders allowed to vibrate, the upstream cylinder is allowed to vibrate with the downstream cylinder fixed and downstream cylinder allowed to vibrate with upstream cylinder fixed. The results include five regimes depending upon ^/q, fluctuating lift forces and vibration characteristics of the cylinder as given in Table 3.

An RV mockup and a CSB mockup were designed and manufactured because the remote measurement system should be subjected to a test to verify its applicability in construction projects. This was done using the RV mockup and the CSB mockup developed here.

1.3.2 Design concept of the RV and the CSB mockup

To design the RV mockup and the CSB mockup, it was necessary to follow a number of principles. First, the measurement parts of the RV core-stabilizing lugs and the CSB snubber lugs should be designed to simulate the actual size of the construction site. Second, the 72 points of the RV core-stabilizing lugs and the CSB snubber lugs should be measured simultaneously. The factors not related to any measurement part should be designed to be as simple as possible.

The RV mockup and the CSB mockup were designed to evaluate the setting suitability of remote measurement system. In particular, the RV core-stabilizing lug of the RV mockup and the CSB snubber lug of the CSB snubber lug were designed to match the RV and the CSB of an actual nuclear power plant.

Other parts of the actual RV and CSB were simply designed for usability of the experiment and out of concerns for the manufacturing budget. Thus, the thickness of cylinders was designed to be thinner than that of an actual RV and an actual CSB.

The high ADS lines, figure 6, are horizontally oriented and connect the PRZ steam space with the HPC. A pneumatic motor operated globe valve is located in each line. Downstream from each isolation valve is a transition piece with an internal square-edge orifice. The two high ADS lines enter the HPC above the waterline, penetrate it and then terminate with a sparger.

The middle ADS lines are horizontally oriented and connect the RPV CL to the HPC. A pneumatic motor operated globe valve is located in each line. Downstream from each isolation valve is a transition piece with an internal square-edge orifice. These two lines enter the HPC, penetrate it and then turn downward before terminating below the HPC waterline. A sparger is considered at the end of these lines.

The ADS sump recirculation lines are horizontally oriented and connect the RPV lower CL to the HPC. A pneumatic motor operated globe valve is located in each line. Downstream from each isolation valve is a transition piece with an internal square-edge orifice. These two lines enter the HPC, penetrate it and then turn downward before terminating below the HPC waterline. No sparger is considered for these lines.

1.5 General remarks and context

The detailed fatigue calculation (DFC) is usually carried out after a certain time period of plant operation, every ten years for instance. These analyses are often performed in the framework of the periodic safety inspection (PSI). Loading data of the operational period as well as anticipated loads of future operation are used as essential input parameters. Hence, usage factors are calculated for the current state of the plant and some prognoses are taken into account to get results until the end of life.

The simplified elasto-plastic fatigue analysis based on elastic FE analyses and plasticity correction (fatigue penalty or strain concentration factors Ke) e. g. according to paragraph 7.8.4 of [4] or equally NB 3228.5 of [1] is known to yield often overly conservative results.

In the practical application this may yield high calculated usage factors. As a consequence, the less conservative elasto-plastic fatigue analysis method based on non-linear FE analyses will often be used for fatigue design. This is associated with an increased calculation effort. Computing times for complex geometries and numerous transients may be significant. Under these circumstances the specified transients have to be rearranged in a small set of covering transients, approximately ten, for calculation purposes.

The possible modification of design codes in respect of more severe fatigue curves and particularly the consideration of EAF will significantly influence the code based fatigue design. Of course, these developments are attentively followed and actively accompanied; see "supporting functions" in Figure 1.

The usual workflow of the fatigue analysis of NPP components is shown in Figure 11. The structural analysis might be simplified elasto-plastic or fully elasto-plastic. The transient temperature fields are analyzed for all relevant N model transients according to Figure 11. These transient temperature fields are themselves the input data for the subsequent transient (linear or non-linear) structural mechanical analyses yielding the local stresses and strains required for code-conforming fatigue assessment. Cycle counting is done in accordance with the requirements of the ASME code as implemented in the ANSYS® Classic Post 1 Fatigue module. It is explained in more detail in the following section.

Daniela Hossu, Ioana Fagara§an, Andrei Hossu and Sergiu Stelian Iliescu

University Politehnica of Bucharest, Faculty of Control and Computers

Romania

1. Introduction

Poor control of steam generator water level is the main cause of unexpected shutdowns in nuclear power plants. Such shutdowns are caused by violation of safety limits on the water level and are common at low operating power where the plant exhibits strong non­minimum phase characteristics. In addition, the steam generator is a highly complex, nonlinear and time-varying system and its parameters vary with operating conditions. Therefore, there is a need to systematically investigate the problem of controlling the water level in the steam generator in order to prevent reactor shutdowns.

Difficulties on designing a steam generator (SG) level controller arise from the following factors:

— nonlinear plant characteristics. The plant dynamics are highly nonlinear. This is reflected by the fact that the linearized plant model shows significant variation with operating power.

— nonminimum-phase plant characteristics. The plant exhibits strong inverse response behavior, particularly at low operating power due to the so-called "swell and shrink’ effects.

— dynamics uncertainties,

— corrupted feed-water flow measurement signal with biased noises.

At low loads (less than 15% of full power) the non-minimum phase behavior is much more pronounced.

Various approaches have been reported in the literature: an adaptive PID level controller using a linear parameter varying model to describe the process dynamics over the entire operating power range (Irving et al. 1980); a model of the steam generator water level process in the form of a transfer function, determined based on first-principles analysis and expert experience has been presented in (Zhao et al., 2000); LQG controllers with "gain­scheduling" to cover the entire operating range (Menon & Parlos, 1992); a hybrid fuzzy-PI adaptive control of drum level, a model predictive controller to identify the operating point at each sampling time and use the plant model corresponding to this operating point as the prediction model (Kothare et al., 2000). Paper (Park & Seong, 1997) presents a self organizing fuzzy logic controller for the water level control of a steam generator. A

nonlinear physical model with a complexity that is suitable for model-based control has been presented by Astrom and Bell (Astrom & Bell, 2000). The model describes the behavior of the system over a wide operating range.

With the advent of the current generation of high-speed computers, more advanced control strategies not limited to PI/PID, can be applied (Hirota, 1993), (Pedrycz & Gomide, 2007), (Yen et al., 1995), (Ross, 2004).

Model predictive control (MPC) design technique has gained wide acceptance in process control applications. Model predictive control has three basic steps: output prediction, control calculation and closing the feedback loop (Camacho & Bordons, 2004), (Demircioglu & Karasu, 2000), (Morari & Lee, 1999).

In this chapter, we apply MPC techniques to develop a framework for systematically addressing the various issues in the SG level control problem.

Fuzzy models have become one of the most well established approaches to non-linear system modeling since they are universal approximations which can deal with both quantitative and qualitative (linguistic) forms of information (Dubois & Prade, 1997), (Zadeh, 2005), (Zadeh, 1989) This chapter deals with Takagi-Sugeno (T-S) fuzzy models because this type of model provides efficient and computationally attractive solutions to a wide range of modeling problems capable to approximate nonlinear dynamics, multiple operating modes and significant parameter and structure variations (Kiriakidis, 1999), (Yager & Zadeh, 1992), (Ying, 2000). Takagi-Sugeno (T-S) fuzzy models have a good capability for prediction and can be easily used to design model-based predictive controllers for nonlinear systems (Espinosa et al., 1999).

The objective of this chapter is to design, evaluate and implement a water level controller for steam generators based on a fuzzy model predictive control approach. The chapter includes simulations of typical operating transients in the SG. A new concept of modular advanced control system designed for a seamless and gradual integration into the target systems is presented. The system is designed in such a way to improve the quality of monitoring and control of the whole system. The project targets the large scale distributed advanced control systems with optimum granularity architecture.