Handbook of Nuclear Engineering

This chapter describes the current status of evaluated nuclear data for nuclear technology applications. We start with evaluation procedures for neutron-induced reactions focusing on incident energies from the thermal energy up to 20 MeV, though higher energies are also mentioned. This is followed by examining the status of evaluated neutron data for actinides that play a dominant role in most of the applications, followed by coolants/moderators, structural materials, and fission products. We then discuss neutron covariance data that characterize uncertainties and correlations. We explain how modern nuclear evaluated data libraries are validated against an extensive set of integral benchmark experiments. Afterward, we briefly examine other data of importance for nuclear technology, including fission yields, thermal neutron scattering, and decay data. A description of three major evaluated nuclear data libraries is provided, including the latest version of the US library ENDF/B-VII.0, European JEFF-3.1, and Japanese JENDL-3.3. A brief introduction is made to current web retrieval systems that allow easy access to a vast amount of up-to-date evaluated nuclear data for nuclear technology applications.

The theory behind the generation of thermal cross sections is presented, concentrating on the phonon expansion method. Examples are given for graphite, water, heavy water, and zirconium hydride. The graphite example demonstrates incoherent inelastic scattering and coherent elastic scattering for crystalline solids. The water example demonstrates incoherent inelastic scattering for liquids with diffusive translations. Heavy water adds a treatment for intermolecular coherence. Zirconium hydride shows the effects of the “Einstein oscillations” of the hydrogen atoms in a cage of zirconium atoms, and it also demonstrates incoherent elastic scattering. Neutron thermalization is introduced using Monte Carlo simulations of several systems, followed by multigroup discrete-ordinates and collision-probability methods. Size effects in thermalization are demonstrated. Steady-state slowing down is discussed by illustrating typical cross-section data, and showing slowing down by elastic scattering, inelastic scattering, and resonance cross sections in the narrow resonance approximation. Intermediate resonance self-shielding effects are introduced using the NJOY flux calculator and the WIMS implementation. The effects of time and space on slowing down are demonstrated using Monte Carlo simulations, and the theoretical basis is summarized.

Today, new evaluated data are almost always prepared in the now universally accepted ENDF/B format. Between the originally evaluated data as coded in the ENDF/B format and our particle transport codes, which actually use the evaluated data, are the often overlooked data-processing codes. These data-processing codes translate and manipulate the data from the single universal ENDF/B format to a variety of formats used by our individual particle transport codes, that is, in contrast to our universally accepted evaluated data format, ENDF/B, as yet there is no universally accepted format used by all of our application codes.

This chapter covers in detail the work done by our data-processing codes to prepare the evaluated data for use in our applications: this includes reconstructing energy-dependent cross sections from resonance parameters, Doppler broadening to a variety of temperatures encountered in real systems, defining data for use in both continuous energy Monte Carlo codes, as well as multigroup Monte Carlo and deterministic methods codes.

In this chapter, both WHAT needs to be done by our data-processing codes and WHY have been defined; also, the overall perspective of a general plan, “The Big Picture,” for the historical and current development of the methods used over the last half century as well as today, has been given.

The importance that processing code verification projects have played over the past decades as well as today has been stressed here. It should be remembered that computer codes have always been very complicated, and it is almost impossible to verify the results calculated by any one code without any comparison with one or more other independently developed codes. A classic mistake is to assume that checking the results will impede progress, whereas in fact experience has shown that taking the time to verify results can actually in the long run lead to savings in time and major improvements in the reliability of our codes.

This chapter describes the basic theory underlying the neutron transport equation and the principal approximations used in this equation’s applications to reactor physics. In addition to presenting detailed classical derivations of various forms of the transport equation, we discuss several important topics in a more rigorous manner than is found in typical derivations. For instance, we include (i) a discussion of the lack of smoothness of the angular flux in multidimensional geometries (this has a negative impact on numerical simulations); (ii) derivations of the transport equation in specialized 1-D, 2-D, and 3-D geometries; (iii) a derivation of the time-dependent integral transport equation; (iv) an asymptotic derivation of the point kinetics equation; and (v) an asymptotic derivation of the multigroup P₁ and diffusion equations. The basic approach taken by the authors in this chapter is theoretical, in the hope that this will complement more intuitive presentations of related topics found in other chapters of this handbook.

Two types of materials are selected by the nuclear industry to be used in nuclear reactors: either materials having specific nuclear properties, or standard engineering alloys corresponding to the thermomechanical loadings and environment.

The first class corresponds to the fuels neutron absorbing isotopes or alloys of low neutron capture cross sections. These atomic properties do not preclude the chemical state of the species, and the best chemical state could be selected (e.g., B, a neutron absorber, can be used in water solution as boric acid or as refractory carbide: B₄C). The second class corresponds to alloys, such as structural steels, stainless steels (SS), aluminum alloys, etc. Few specific alloys have been developed for particular applications, such as Zr alloys in water reactors or vanadium alloys in fusion devices.

All these materials have to support the environment of a nuclear reactor. In addition to standard engineering constrains (mechanical loadings, corrosion, high temperatures, etc.), the irradiation itself induces major changes in structure, properties, and behavior of all the materials.

Irradiation hardening, reduction of ductility, irradiation creep and growth, and swelling are described in detail in their physical mechanisms and their specific characteristics for the alloys and ceramics for current use and of future interest. Other irradiation effects such as radiolysis on water, or changes in electrical properties of insulating ceramics are also described.

After a generic description of the physics of the transformations induced in the microstructures by irradiation, the phenomena of major concern are presented for the different components of various reactors. The corresponding conditions are analyzed for various types of experimental reactors, power reactors (thermal and breeder), and fusion devices.

This chapter intends to provide a ready reference to mathematical concepts and tools customarily used in nuclear science and engineering, thus facilitating the reading of the other chapters in this handbook. The material presented in this chapter addresses the mathematical requisites at the graduate level in nuclear engineering, summarizing the following topics: vectors and vector spaces, matrices and matrix methods, linear operators and their adjoints in finite and infinite dimensional vector spaces, differential calculus in vector spaces, optimization, least squares estimation, special functions of mathematical physics, integral transforms, and probability theory. A list of suggested textbooks, covering many of the details omitted in this chapter, is provided in the Bibliography Section.

The transport equation is introduced to describe a population of neutral particles such as neutrons or photons, in a close domain, under steady-state (i.e., stationary) conditions. Its derivation is based on the principle of particle conservation. The transport equation describes the statistical behavior of a large population of particles. The exact number of particles per unit volume is continuously varying with time, even at steady-state conditions. Under steady-state conditions, the number density of particles oscillates about an average value related to the solution of the steady-state transport equation. A solution of the transport equation is required in many fields of nuclear engineering, notably in reactor physics, in safety and criticality, and in radiation shielding and protection. We review legacy approaches for solving the steady-state transport equation, namely, the method of spherical harmonics, the collision probability method, the discrete ordinates method, and the method of characteristics. The full-core calculation consists of solving a simplified transport equation, either the diffusion equation or the simplified P _n equation.

This chapter presents a detailed description of the elements that comprise a lattice physics code. Lattice physics codes are used to generate cross section data for nodal codes, where the nodal codes are used to model the coupled neutronics and thermal-hydraulics behavior of the entire reactor core during steady state and transient operation. Lattice physics codes analyze axial segments of fuel assemblies, referred to as lattices, to determine the detailed spatial and spectral distribution of neutrons and photons across the segment. Once the flux distribution is known, the cross sections can be condensed and homogenized into the structure needed by the nodal code. The nodal code then pieces the various lattices together to construct the various fuel assemblies in the reactor core. This chapter is split into individual sections representing the major pieces of a lattice physics code. Section 1 presents a general overview of the computational scheme used for a typical lattice physics code (Knott). The remaining sections of this chapter are used to describe the major pieces in detail. Section 2 describes the contents of the cross section library that accompanies a lattice physics code (Yamamoto). Section 3 discusses the various resonance treatments used in lattice physics calculations (Yamamoto). Section 4 describes a method for removing cross section energy detail without sacrificing too much accuracy (Knott). Section 5 describes the fine-mesh transport calculation on the heterogeneous lattice geometry (Knott). Section 6 discusses the burnup calculation (Yamamoto). Section 7 describes some of the details of a typical case matrix (Knott), and Sect. 8 discusses some of the edits that are provided by the lattice physics code (Knott). This chapter provides the interested reader with a broad understanding of a typical lattice physics code.

This chapter discusses how core isotopic depletion and fuel management are completed for reactor cores of nuclear power plants. First, core isotopic depletion is discussed, in particular, how the Bateman equation is numerically solved, and the behaviors of the fissile, fertile, burnable poison and transient fission products isotopes. The concepts of breeding, conversion, and transmutation are introduced. Nuclear fuel management is discussed next, with a strong emphasis on the fuel management for light water reactors (LWRs), given their predominance. The discussion utilizes the components of design optimization, those being objectives, decision variables, and constraints. The fuel management discussion first addresses out-of-core fuel management, which involves such decisions as cycle length; stretch out operations; and feed fuel number, fissile enrichment, and burnable poison loading and partially burnt fuel to reinsert, for each cycle in the planning horizon. In-core fuel management is introduced by focusing on LWRs, with the basis of making decisions associated with determining the loading pattern, control rod program, lattice design, and assembly design presented. This presentation is followed by a brief review of in-core fuel management decisions for heavy water reactors, very high temperature gas-cooled reactors, and advanced recycle reactors. Mathematical optimization techniques appropriate for making nuclear fuel management decisions are next discussed, followed by their applications in out-of-core and in-core nuclear fuel management problems. Next presented is a review of the computations that are required to support nuclear fuel management decision making and the tools that are available to accomplish this. The chapter concludes with a summary of the current state of depletion and nuclear fuel management capabilities, and where further enhancements are required to increase capabilities in these areas.

This chapter deals with shielding against nonionizing radiation, specifically gamma rays and neutrons with energies less than about 10 MeV, and addresses the assessment of health effects from exposure to such radiation. The chapter begins with a discussion of how to characterize mathematically the energy and directional dependence of the radiation intensity and, similarly, the nature and description of radiation sources. What follows is a discussion of how neutrons and gamma rays interact with matter and how radiation doses of various types are deduced from radiation intensity and target characteristics. This discussion leads to a detailed description of radiation attenuation calculations and dose evaluations, first making use of the point-kernel methodology and then treating the special cases of “skyshine” and “albedo” dose calculations. The chapter concludes with a discussion of shielding materials, radiological assessments, and risk calculations.

The aim of this chapter is to give some key points on the use of high performance computing (HPC) in the field of nuclear engineering. This chapter is divided into two main parts. This first one is an introduction to parallel computing. In this first part, we will describe not only the main computer and processor architectures which are used today but also some which are not so usual but which will allow the readers to better understand the key point of parallelism from the hardware point of view. Still in this first part, we will continue by describing the main parallelism models, in the same point of view as the description of the parallel architecture. In Sect. 4, we will give some basic ideas on how to design parallel programs. This section is the last of the first part. The second part is dedicated to the use of high performance computing in nuclear engineering. We will give first the main challenges which can be addressed using HPC. Some of them are illustrated in Sect. 6 on some of the main scientific domains in nuclear engineering (reactor physics, material sciences and thermal-hydraulic). For each of them we have tried to describe the main problems which can be addressed using HPC but some of them remain as scientific and industrial challenges.

The analysis of the behavior of light water reactor (LWR) fuel rods is described. The properties of relevant fuel and cladding materials are discussed and numerical data are given. The basic phenomena taking place in pellet-in-cladding nuclear reactor fuel are described systematically, including neutronic aspect of the fuel, the thermal and mechanical behavior, the fission gas behavior, and radiation effects. Finally typical phenomena and issues in the design and licensing of LWR fuels and their effects on fuel behavior are discussed: the high burnup structure, pellet-cladding interaction, pellet-coolant interaction, loss-of-coolant accidents (LOCA), and reactivity-initiated accidents (RIA).

This chapter deals with neutron fluctuations in nuclear systems. Such neutron fluctuations, or neutron noise, fall into two categories: neutron noise in zero power systems and neutron noise in power reactors. The concepts, the theory, and the methodology of these fluctuations as well as their various applications for extracting information in a nonintrusive way about the system in question are described. A number of specific applications are described, where detection and analysis of zero power and power reactor noise make it possible to extract diagnostic information about the system by determining some parameters of the system during normal operation, or by detecting, identifying, and quantifying developing anomalies at an early stage and determining their severity. This chapter ends with an outline of future developments and actual issues in the field.

The main theme of this chapter is the process and evolution of deterministic and probabilistic safety analyses that have played a backbone role in assuring public health and safety in the peaceful uses of nuclear power. The chapter begins with a discussion of the origin of nuclear power safety analysis together with the overall perspectives of both deterministic and probabilistic approaches that are still prevalent, although there is an increasing trend in application of probabilistic safety analysis in safety-related decision making. Deterministic approaches, such as the defense-in-depth or safety margin, are regarded as a means to cope with uncertainties associated with adequacy of safety features. As probabilistic methods and applications gain maturity and acceptance, the uncertainties associated with safety features are measured and described probabilistically. The chapter concludes with a detailed discussion of the probabilistic safety assessment and its uses in nuclear power safety analysis.

Effective field modeling of two-phase flow has provided a critical part of the foundation upon which light water (power) reactor technology was made to rest some 20-30 years ago. We can envision a similarly significant role in the future as simulation capabilities are poised to meet new kinds of practical demands at the interplay between economics, safety assurance, and regulatory needs. These new demands will require better predictive reliability for larger departures from past practices, and this in turn will require strengthening of the scientific component along with translating past empiricism into more and more fundamental terms. In this perspective, the mathematical formulation of the effective field model, as well as the numerical implementation of this formulation needs to be revisited and reassessed. Helping respond to this need is the purpose of this chapter.

We delineate a conceptual framework for addressing prediction of multiphase flows at the three-dimensional, phase distribution level. This is in terms of a local, disperse system description (bubbles/drops in a continuous liquid/vapor phase). The requirement that follows is a well-posed formulation and a high-fidelity numerical treatment that allows capturing of shocks and contact discontinuities over all (relative) flow speeds, consistently with what is physically allowable according to the density ratios involved — in particular, high relative Mach numbers for droplet/particle flows. The importance of inviscid interactions (between the phases) in this context is highlighted.

The scope is for disperse-phase volume fractions up to about 20% as pertinent to fluid-fluid systems. This provides the basis for addressing phase transitions through coalescence as this process becomes significant at still higher volume fractions. The theoretical framework provides also the basis for extensions (outlined only in general terms here) to the high-volume fractions pertinent to dense solid-particle systems. The computational approach is readily applicable to both these extensions.

A general disperse system formulation is derived by means of a new, “hybrid” method that incorporates features of a statistical approach and reveals more clearly the nature of phase interactions at the individual particle scale. Moreover in this manner the formulation lends itself to elaboration of the constitutive treatment by means of numerical simulations (based on the direct solution of the Navier-Stokes equations) resolved at the particle scale. The formulation is exemplified by successive applications to various increasingly complex situations, starting with non-dissipative systems, where one or the other phase may be incompressible. At each step we examine the hyperbolic character of the system of equations, and we include consideration of high (relative) Mach numbers. The basic constitutive treatment concerns pseudo-turbulent fluctuations of the continuous phase, and the resulting systems of equations are fully closed and hyperbolic even in their non-dissipative form (ready for computation), except for a non-hyperbolic corridor around the transonic region. Results obtained are discussed in relation to formulations that form the basis of current numerical tools (codes) employed in nuclear reactor design and safety analyses (mostly addressing bubbly flows), as well as formulations found in other contexts.

This mathematical formulation is pursued further to its numerical implementation. With an emphasis on flow compressibility, we focus on capturing shocks and contact discontinuities robustly for all flow speeds and at arbitrarily high spatial resolutions. As we learn from relatively recent progress in single-phase flows, the key role is that of “up-winding” applied on the basis of a scheme that emphasizes conservative discretization. This background is briefly reviewed, culminating with a rather detailed exposition of the most recent advance in this line of development: the Advection Upstream Splitting Method (AUSM). The essential and new step here is to extend the basic ideas of the AUSM to the compressible multi-hydrodynamics problems of interest here, and the above-mentioned EFM in particular. We also include calculations illustrative of numerical performance.

The presentation is arranged into two autonomous parts: Part I addresses the formulation of the EFM and Part II deals with the numerical implementation and testing. An overall summarization of where we stand at the completion of this work and what we see as needed future developments is provided in the following.

One basic objective was to address inviscid interactions by means of a coherent theoretical formulation and through to computational testing. In particular, we wanted to access high-fidelity simulations of the type needed to address flow regimes at all flow speeds; especially at the high relative Mach numbers pertinent to disperse particle/droplet flows. This requires that the system of equations be hyperbolic, and we wanted to achieve a solid foundation rather than adopting any of the several ad hoc constitutive models as post facto “remedying” the problem.

On theory, we begin with entropy rather than energy transport equations and we derive, consistently with thermodynamics and the momentum equations, a condition for satisfying conservation of total energy. This condition is of utmost importance showing the tight link between the conservation laws employed, and the transport equations of volume fraction and of pseudo-turbulent kinetic energies of the continuous (included) and disperse (not yet included in the derivation) phases. On this basis we demonstrate a systematic way to deduce closed systems of equations for non-dilute disperse flows, and thusly we arrive at an EFM that is hyperbolic except for a “corridor” around the transonic region. The key is a function of the disperse phase volume fraction E(α _d ). It enters as a coefficient of the disperse phase pseudo-turbulent kinetic energy. Awaiting further definition as a function of the Mach number, by means of the type of direct simulations noted above, it is employed here throughout in its zero-Mach form. A much needed extension would also involve the pseudo-turbulent kinetic energy of the disperse phase, along with physics of dense dispersions (collisions etc.).

While terms such as those proposed previously for “interfacial pressure” and “added mass” phenomena can be identified, the complete formulation is not reducible to any of those ad hoc models. Notably, the disperse phase pressure appears nowhere in the momentum equations. Also we find that the claimed as hyperbolic, Baer-Nunziato model involves a volume fraction transport equation, which is not physically tenable for dispersions, or is it an appropriate means to dealing with ill-posedness. On the other hand, we find perfect agreement with the formulations obtained at the incompressible limit by Geurst (1985), employing a complex variational approach, and by Wallis (1989), employing a rather involved development based on potential flow theory.

On computations our objective is to capture shocks and contact discontinuities, for conditions that are within the hyperbolic regions in the Mach number space, and to explore (1) behaviors within the non-hyperbolic corridor, and (2) means of stabilization as necessary. Given the EFM development needs expressed above, it is understood that this testing in the Mach number space is strictly provisional. We begin with an adaptive mesh refinement infrastructure, and the Advection Upstream Splitting Method (AUSM), currently the method of choice for single-phase compressible flows. A key point of adaptation to our EFM is treating the pseudo-turbulent stresses within the pressure flux splitting, and ensuring that the discretization of the nonconservative terms is done in a way that satisfies propagation of contact discontinuities in uniform steady flow without disturbing the pressure field. Our approach is readily extendable to any equation of state and to adding any number of equations (volume fraction transport, multiple equations for the disperse phase for tracking multiple length scales as may be found when the disperse phase is subject to fragmentation). The testing performed for this work was done on 1D problems only. Extending this testing to 2D and 3D problems is underway.

Testing was carried out independently with two computer codes: ARMS (all-regime multiphase simulation) and MuSiC-ARMS (Multi-scale Simulation Code-ARMS). The ARMS was built on an open access platform, the structured adaptive mesh refinement infrastructure (SAMRAI) developed at Lawrence Livermore National Laboratory. The MuSiC-ARMS was built, more recently, on the MuSiC platform, our own specialized code, using irregular grids to “fit” areas of highest refinement (shocks, interfaces, etc.), which are embedded in a multilevel (adaptive) Cartesian mesh. This platform is also used for a DNS code, the MuSiC-SIM, and a pseudo-compressible (incompressible) code, the MuSiC-ISIM. We focus on dispersed being the heavy phase (droplet/particle flows) so as to access realistically high Mach numbers, and significant inviscid interactions.

The test cases were selected to include various kinds of Riemann problems with discontinuities in (a) Mach number only (Fitt’s problems) and (b) pressure, or pressure and disperse phase volume fraction (shock tube problems). In the Fitt’s problem case, we include parametric studies on the value of C that appears in function E(α _d ). In addition, we consider shock wave “impact” problems on particle clouds that are either with sharp or smooth (in particle volume fraction) outer boundaries, and as part of this class also the case of dilute clouds for which we have the analytic solution for comparison. Finally, we considered the capturing of contact discontinuities in “mild” situations such as the so-called Faucet problem and the simple convection of a coherent second phase by uniform flow. The Faucet problem is well known to be failed under grid refinement in all published tests to date. The convection problem is important check of the pressure non-disturbing condition, a requirement that is hard to meet due to the non-conservative terms found in all effective field models.

The emphasis being on stability and convergence under grid refinement, all problems were carried out in the inviscid limit (no interfacial drag), and all cases passed the test except for the high pressure ratio shock tube problems where instabilities developed within the expansion wave. However, these cases were stabilized with a minimal amount of dissipation effected by adding a small amount of interfacial drag (roughly one tenth of the normal amount). These numerical results render support to the idea that, notwithstanding the “mild” non-hyperbolic corridor found in the analysis of Part I, the present effective field model is hyperbolic, and along with the numerical treatment employed they provide access to rather extreme two-phase flow conditions in a robust and accurate manner.

In an overall perspective of computational fluid dynamics, the presently offered capability is complementary to that already available through the “standard”, non-hyperbolic two-fluid model as already found in the computational frameworks of the ICE (Harlow and Amsden 1968) and SIMPLE (Patankar and Spalding 1972) methods. The special purposes aimed here are to overcome limitations in grid refinement and to approach flows where the phasic-relative velocities are high enough to introduce significant compressibility effects. Rapid advancement in hardware makes computational analysis of complex multiphase flows, even direct numerical simulations, increasingly more practical and reliable. High-fidelity/resolution techniques such as those employed here can address problems of varying time and length scales and this paves the way for actual simulations of multiphase physics at the effective field level, and even allowing a seamless analysis transitioning across regimes of multiphase flows.

In practice, the results of experiments seldom coincide with the computational results obtained from the mathematical models of the respective experiments. Discrepancies between experimental and computational results stem from both experimental and computational uncertainties. Such discrepancies motivate the activities of model verification, validation, and predictive estimation. Following a brief review of the classification and origins of experimental uncertainties, this chapter presents widely used statistical and deterministic methods for computing response sensitivities to model parameters, highlighting, in particular, the novel adjoint sensitivity analysis procedure (ASAP) for augmented nonlinear large-scale systems with feedback. The practical use of ASAP is illustrated by a large-scale application for analyzing the dynamic reliability of an accelerator system design for the International Fusion Materials Irradiation Facility (IFMIF).

Response sensitivities to parameters and the corresponding uncertainties are the fundamental ingredients for predictive estimation (PE), which aims at providing a probabilistic description of possible future outcomes based on all recognized errors and uncertainties. The key PE-activity is model calibration, which uses data adjustment and data assimilation procedures for addressing the integration of experimental data for updating (calibrating or adjusting) parameters in the simulation model. This chapter also presents a state-of-the-art mathematical framework for time-dependent data assimilation and model calibration, using sensitivities and covariance matrices. The basic premise underlying this mathematical framework is that only means and covariance matrices are a priori available, which is the usual situation when analyzing large-scale systems. Under this premise, the maximum entropy principle of statistical mechanics is employed in conjunction with information theory to construct a Gaussian prior distribution that takes all of the available information into account while minimizing (in the sense of quadratic loss) the introduction of spurious information. This prior distribution also comprises correlations among model parameters and responses, thus generalizing the state-of-the-art data assimilation algorithms used in geosciences.

Open issues (e.g., explicit treatment of modeling errors, reducing the computational burden, removing the current restriction to Gaussian distributions) are addressed in the concluding section of this chapter. Since predictive “best-estimate” numerical simulation models are essential for designing new technologies and facilities, particularly when the new systems cannot be readily tested experimentally, the only path to progress is to reduce drastically the uncertainties associated with such simulation tools while enlarging the respective validation domains.

The CEA (Commissariat á l’Energie Atomique) is strongly involved in R&D research programs concerning the use of nuclear energy as a clean and reliable source of energy and consequently is working on the present and future generations of reactors on various topics such as ageing plant management, optimisation of the plutonium stockpile, waste management and innovative systems exploration. Core physics studies are an essential part of this comprehensive R&D effort. In particular, the Zero Power Reactors (ZPR) of CEA: EOLE, MINERVE and MASURCA play an important role in the validation of neutron physics calculation tools (codes and nuclear data). Most recent programs notably contributed to: Obtain a very large and accurate experimental database for nuclides arising in plutonium and waste management (heavy nuclides and long lived fission products). Explore long-lived nuclides transmutation. Support the present French PWR fleet and the future reactors such as EPR. Explore innovative systems and new concepts in terms of new materials and fuels (ABWR, new MTR such as the Jules Horowitz Reactor under construction in Cadarache). Improve the physics of hybrid systems, involving a sub-critical reactor coupled with an external accelerator (ADS). A vast majority of theses programs are carried out within the frame of international collaboration. The Zero Power Reactors are also essential tools in the activities of teaching various topics related to experimental reactor physics. In particular, practical work on the operation and control of reactors and the associated methods of experimentation is done by the students enrolled in the “Génie Atomique” French nuclear engineering school. The experimental programs defined in the EOLE, MINERVE and MASURCA facilities aim at improving the calculation routes by reducing the uncertainties of the experimental databases. They also provide accurate data on innovative systems in terms of new materials (moderating and decoupling materials) and new concepts (ADS, ABWR, GEN IV) involving new fuels, absorbers and coolant materials. After a description of the previous experimental programs, an overview of the future experimentation is given in this chapter.

This chapter describes Korean experiences and accomplishments in construction, operation, and design development of nuclear power plants. Recognizing domestic energy-resource scarcity, Korea has steadily constructed, operated, and improved nuclear power plants based on self-reliant technologies and accumulated experiences. This chapter describes specifically i) two pressurized light water reactor models (OPR1000 and APR1400) currently being built and to be built in the future, and ii) experience with the pressurized heavy water reactors (CANDUs) built and in operation, in the Republic of Korea.

This chapter contains detailed description of the design and technical layout of Russian VVER-type reactors. Both the VVER-440 and the VVER-1000 reactor types are described. VVER reactors are a special design of Pressurized Water Reactors with some particular design features listed in the introduction. The most important of them are:

VVER reactors are the most frequently built reactor type in the world. Presently, there are 23 units of VVER-440 and 28 units of VVER-1000 worldwide in operation (see tables in Sects. 2.3 and 3.3). In the introduction, a historical overview on the design of VVER-440 and VVER-1000 reactors is given. First units with predecessors of VVER-440 type reactors were erected at the Novovoronesh NPP site in 1972 and 1973. The second step in the development of VVER-440 type reactors was the V-230 design, where the number of mechanical control rods was reduced from 73 to 37 due to introduction of boron as a moderator. In the period from 1973 to 1982, all 14 units were constructed with the V-230 design. The third step in VVER-440 development was V-213 reactor design referred to as the second generation of the standard VVER-440 reactors; their design basis included a double-ended instantaneous guillotine break of the maximum diameter primary pipeline. The development of VVER-1000 reactors was started by OKB Gidropress in 1966. The first reactor with an electrical power of 1,000 MW was commissioned at Novovoronezh NPP Unit 5 in 1980. In the design, the traditional engineering solutions of VVER were used with the appropriate modernization from the experience obtained in the design, manufacture, and operation of the VVER prototypes. The design concept was oriented to an increase in economic efficiency of the nuclear power plant (NPP) construction and operation, ensuring safety in accordance with the regulatory documents that were valid at the time. An instantaneous double-ended guillotine break of the main coolant pipeline was considered as the maximum design basis accident. The reactor plant was placed into a containment of prestressed concrete. Further on, the design modifications ended up in elaboration of V-302 Project implemented at the South-Ukraine NPP Unit 1 and V-338 Project implemented at the South-Ukraine NPP Unit 2 and the Kalinin NPP Units 1 and 2. Elaboration of these projects classified as small series designs and their realization was under way from 1976 to 1987. All the power units of VVER-1000 NPPs, beginning with 1985, were constructed to a standard design that contains the V-320 reactor plant, capable of being sited in seismic areas with earthquakes up to magnitude 9 (SSE), the reactor of small series; D nom 850 circulation loops without gate valves; wet reloading of internals; PGV-1000 horizontal steam generators; and GTsN-195M reactor coolant pumps (large series). The main feature of the enhanced standard VVER-1000 reactor is the application of jacket-free fuel assemblies (their quantity increased from 151 pcs. to 163 pcs. and the reduction of control rods from 109 to 61 [ up to 49 at the South-Ukraine NPP Unit 1]) and the application of ShEM drives in reactor trip system. The large series designs have been realized since 1978 up to now. Twenty-eight Generation II power units with VVER-1000 have been constructed and are in operation at NPPs. The accidents at TMI-2 and Chernobyl-4 NPPs have shown that it is necessary to take into consideration the beyond design basis accidents (BDBA) during their design and operation. In 1988 a new V-1000 design was launched (V-392 Project), which focused on safety improvements in response to new requirements of regulatory documents in order to prevent occurrence of BDBAs and to mitigate their consequences in the case that they happened. The idea was implemented in the RP design of V-392 Project incorporated into NPP-92 design. The main RP equipment of the V-392 design, including the reactor, was implemented in a set of RP V-428 at “Tianwan” NPP. The information on this NPP design was incorporated into the chapter on VVER-1000 reactors. The modifications to V-392 RP design can be found in the design of RP V-412, which are being implemented now at “Kudankulam” NPP. Designs of Units with RP V-392, V-428, and V–412 are referred to Generation III reactors.In both VVER-440 and VVER-1000 sections of the chapter, first, the main design parameters of these reactors are given. Second, the buildings and structures are described that house the reactor plant and auxiliary systems. As a special feature of VVER-440, two power units of this reactor type are incorporated into one main building of the NPP.. Further, a tower, being a part of the reactor building with V-213 reactor plant, houses the vacuum-bubbler passive system to reduce pressure in the containment. Quite comprehensive sections are devoted to the primary circuit systems and equipment. The reactor coolant system, reactor, main circulation pumps, pressurizer, steam generators, and chemical and volume control systems are described. While the VVER-440 reactors dispose of six primary circuit loops, the VVER-1000 has four loops. Special attention is paid to the core and fuel design. Special features of the VVER-440 core design are the fuel assemblies with housings and control assemblies, which consist of two parts: an absorber part and a fuel follower. When the absorber part is withdrawn from the reactor core, the fuel follower is inserted into the core. In VVER-440, profiled fuel assemblies with Gd₂O₃ burnable absorber are used to decrease the power peaking factor in the core.. The control elements for VVER-1000 reactors are of the cluster type, similar to western PWR designs. Typically, 18 absorber rods are placed in a control assembly. In subsequent sections, the secondary circuit components (Main Steam Line System, Main Feedwater System, Turbine, Generator, and Moisture Separator Reheater) are described. The power units with VVER-440 are equipped with two turbines and generators of 220 MW electrical power and the power units with VVER-1000 have one 1,000-MW-turbine driving one generator. Further I&C as well as electrical systems are briefly described. The instrumentation and control systems of some of the VVER-440 units have been recently updated. The refurbishments were mainly aimed at the introduction of advanced features for data processing, transmission, and archiving. In Sects. 2.2 and 3.2, the safety philosophy and safety systems ofVVER-440 and VVER-1000 are described. The applied concept of safety assurance for power units of Generations II and III is outlined. For the Generation III VVER designs, significant improvements were implemented in accordance with the up-to-date international requirements for NPP safety assurance. The data on the VVER reactors under construction, operation, and decommissioning are presented in Sects. 2.3 and 3.3.

This chapter describes the various families of reactors in which the primary fluid cooling the core is a gas, usually carbon dioxide or helium.

Early on, gas cooling was mostly used in graphite moderated reactors fueled with natural uranium, the British Magnox, and the French natural uranium graphite gas (NUGG). The availability of low enriched uranium fuel allowed the British to develop the advanced gas-cooled reactor as a successor to the Magnox.

In a world progressively dominated by the water cooled reactors, mostly PWR and BWR, gas cooling remained alive in the high temperature reactor families, prismatic, and pebble bed HTR, associated with graphite moderation. Both are based on the use of a very innovative fuel element, the coated particle.

The same fuel was used in a seldom known US program to develop nuclear propulsion for rockets: this NERVA story will be briefly recalled.

Still marginal, gas cooling is also present among the “Generation IV” concepts, through the very high temperature reactor system aimed at both electricity generation and hydrogen production and the GFR, gas cooled fast neutrons reactor.

The lead-cooled fast reactor (LFR) has both a long history and a currency of innovation. With early work related to the mission of submarine propulsion dating to the 1950s, Russian scientists pioneered the development of reactors cooled by heavy liquid metals (HLM). More recently, there has been substantial design interest in both critical and subcritical reactors cooled by lead (Pb) or lead–bismuth eutectic (LBE), not only in Russia, but also in Europe, Asia, and the USA. This chapter reviews the historical development of the LFR and provides detailed descriptions of the recent and current initiatives to design a variety of LFR concepts with several different missions in mind: accelerator-driven subcritical (ADS) systems for nuclear materials management, small modular systems for deployment in remote locations, and central station plants for integration into developed power grids. It describes design criteria and system specifications; features particular to the LFR in terms of neutronics, coolant properties, and material compatibility issues; approaches taken to core and reactor design; and considerations related to the balance of plant and plant layout.

The technology of nuclear power could be quite different from today’s if it had been practical in the beginning to supplement fission neutrons with accelerator-produced neutrons. The purpose of this chapter is to illustrate the possible benefits of implementing supplementary neutrons from accelerators in an optimized reactor. GEM^∗STAR (Green Energy Multiplier^∗Subcritical Technology for Alternative Reactors developed by Accelerator Driven Neutron Applications (ADNA Corp) is a subcritical thermal-spectrum reactor operating with molten salt fuel in a graphite matrix and in a continuous flow mode initially at k_eff = 0. 99. The model described is able to use natural uranium as fuel and generate twice as much electric power as a light water reactor (LWR) generates from the same mined uranium. GEM^∗STAR at k_eff = 0. 99 also can be fueled with unreprocessed LWR spent fuel, and it can generate as much electricity as the LWR had generated from the same fuel. Because GEM^∗STAR uses liquid fuel, it can recycle its own fuel at k_eff = 0. 95 without any operations on the fuel. This recycle can be repeated several more times, always without reprocessing, as accelerator or fusion neutron generation technology development reduces the cost of neutrons. GEM^∗STAR therefore increases the electricity from mined uranium many times while avoiding the serious problems of current nuclear-power technology arising from enrichment, reprocessing, fast reactor deployment, and near term high-level waste storage. GEM^∗STAR also offers technology for nuclear energy generation that promises reductions in nuclear electricity cost and eliminates major proliferation concerns. The technology can use a modest source of intermittent “green” electricity such as wind or solar as input power to drive an accelerator that, in effect, multiplies the green energy by a factor of about 30 with 24–7 continuity and without compromising any environmental objectives of green energy sources. This chapter is not a complete history of molten salt, graphite, and accelerator technologies, but a description of how these orphan elements of nuclear power development may be integrated for a GEM^∗STAR solution to the main barriers that constrain the full deployment of today’s nuclear power technology.

This chapter describes the various industrial steps which constitute the front end of the nuclear fuel cycle, i.e., the complete set of operations needed to produce a functional fuel element ready to be loaded in a nuclear reactor. This chapter also provides data concerning the element uranium, its abundance and its most relevant properties. The exploration, mining, concentration and site rehabilitation processes are then described.

Light water reactors (LWR), which make the vast majority of the nuclear reactors operating today, and those under construction, cannot use “natural” uranium for their fuel: it must be “enriched” in isotope 235 and the enrichment process itself requires the uranium to be “converted” into a gaseous compound.

Once enriched to the required assay, uranium is then fabricated into solid ceramic “pellets,” piled into leaktight metallic “pins.” These pins are then “assembled” to constitute the fresh fuel element.

The chapter also provides additional information on MOX fuel assemblies used to recycle plutonium in LWR, as well as some data on plutonium and thorium.

One shall also find an explanation of the fascinating “Oklo Phenomenon,” which occurred almost 2 billion years ago in some uranium deposits of Gabon, in this chapter.

This chapter is expanded and updated from part of a previous Springer publication (Barré, 2005a).

Transuranium elements, neptunium, plutonium, americium, and curium, are formed via neutron capture processes of actinides, and are mainly by-products of fuel irradiation during the operation of a nuclear reactor. Their properties significantly impact the nuclear fuel cycle, affecting and often determining requirements and procedures related to handling, storage, reprocessing, and disposal of fuels and high-level waste. It is still debated if, in particular, plutonium is an unwanted waste or, possibly, a resource for the production of energy. A standard universally agreed route for the treatment of transuranium elements is not yet established. This chapter provides an overview of past and ongoing experience and perspectives related to studies on transuranic recovery and incorporation in fuels and targets for advanced nuclear fuel cycles and their disposal as the main component of high-level nuclear waste. In particular, the chapter describes the main properties of transuranium fuels, the specific requirements for their fabrication, their irradiation behavior, and their impact on the back-end of the fuel cycle. For the latter, a major issue is the development of options for reprocessing and separation of transuranium elements from spent fuel to make them available for further treatment. The effects caused by their presence in irradiated fuel and high-level nuclear waste on long-term storage and final disposal are also discussed.

The final destination of transuranium elements is still an open issue. The global context is characterized by a diversified set of options being pursued, which is reflected in this chapter. It is important to have a picture of the knowledge and experience gathered until now through relevant investigation campaigns worldwide. This is necessary to ensure that the renewed interest in nuclear energy as a key component of sustainable development of energy production brings the necessary focus to implement viable, safe, and technologically effective options for the treatment of transuranium elements.

Decommissioning a nuclear plant can be defined as the termination of operations and the withdrawal of the facility from service, followed by its transformation into an out-of-service state without radiological risks and, in some cases, its complete removal from the site. Decommissioning activities shall be carried out in a cost-effective manner assigning top priority to health and safety of the general public and the environment, as well as of the decommissioning workers. This chapter covers all aspects related to the closure of the operating life of nuclear plants and provides a description of all the activities and tools involved in both the decision-making and operative processes of decommissioning.

Nuclear plant decommissioning is a complex, long, and highly specialized activity. In some countries, therefore, it is even called “de-construction” because it is in many respects similar to the construction activity and, in addition, it deals with partly activated and contaminated structures. Activities to perform include technological tools, industrial safety, environmental impact minimization, licensing, safety analysis, structural analysis, etc. Other aspects are short- and long-term planning, calculation of cash flow and financing, waste disposal, and spent fuel strategy.

A lot of technical information is drawn from direct experience of nuclear operators. The widely used references are those from the OECD-NEA, UNO-IAEA, US-NRC, and the European Commission. They cover the results of working groups, special studies, comparisons of technologies, and recommendations.

Waste is produced at every stage of the nuclear fuel cycle. While large volumes of short-lived radioactive waste are already handled by the nuclear industry in surface storage facilities, the management mode of high-activity, long-lived waste has not been decided in detail and is still under study in all nuclear countries. Scientific knowledge is in progress, technical solutions are emerging, in a context where science and technology interact strongly with social and economical issues.

With a closed fuel cycle, waste management from its production to its final destination looks like a chain whose links are treatment recycling, conditioning, storage, and disposal of the final waste. With the open cycle option, the first link is absent.

This chapter provides the concepts and data that form the scientific basis of nuclear waste management.

Section 1 deals with the origin, nature, volume, and flux of nuclear waste, and describes the management options.

Section 2 deals with waste conditioning, with special emphasis on two important conditioning matrices: cement-like materials and glass. The elaboration and long-term behavior of these matrices are treated successively. In many countries, spent fuel is considered as waste, and must be conditioned as such. A special section is devoted to this issue.

Section 3 deals with waste storage and disposal. Interim storage of long-lived waste is already an industrial reality, and the design and properties of the corresponding installations are described. The final disposal of ultimate waste in deep geological repositories is more prospective, but the main concepts are described, with emphasis on the mechanisms, models, and orders of magnitude of the main physical and chemical phenomena that come into play in the long-term evolution of these installations. Finally, a short description of the methodology used to evaluate the safety of these installations is given. A simplified example of application of this methodology is given to evaluate the order of magnitude of the radiological impact of geological disposal of long-lived waste.

The Nuclear Nonproliferation Treaty (NNPT or NPT) is the primary cornerstone of international efforts to prevent the proliferation of nuclear weapons. Currently, 189 countries are party to the treaty, with only four sovereign states abstaining: India, Israel, Pakistan, and North Korea. The treaty is broadly interpreted as having three pillars: (1) nonproliferation, (2) disarmament, and (3) the right to the peaceful use of nuclear technology.

Five countries are recognized by the NPT as nuclear weapon states (NWSs): the United States (US), the Soviet Union (obligations and rights now assumed by Russia), France, the United Kingdom, and the People’s Republic of China. These five nations are also the five permanent members of the United Nations (UN) Security Council. In accordance with the NPT, the NWSs agree to not transfer nuclear weapons to a nonnuclear weapons state (NNWS) or assist NNWSs in acquiring nuclear weapons. Additionally, the NNWSs party to the NPT agree not to receive or manufacture nuclear weapons. NNWSs also agree to accept safeguards monitoring by the International Atomic Energy Agency (IAEA) to verify that they are not diverting material derived from the peaceful use of nuclear technology to weapons.

The NPT’s preamble also contains language affirming the desire of all signatories to halt the production of nuclear weapons worldwide and to develop an additional treaty related to complete nuclear disarmament and liquidation, including their delivery vehicles. However, the NPT wording does not strictly require all signatories to actually conclude a disarmament treaty, but rather to negotiate in good faith. Some NNWSs belonging to the Non-Aligned Movement (an international organization of states considering themselves not formally aligned with or against any major power block) have interpreted the NPT as requiring the NWSs to disarm themselves and argue that these states have failed to meet their obligations.

The Strategic Arms Reduction Treaty (START) between the United States (U.S.) and Soviet Union is considered by many to be the largest and most complex arms control treaty in history. The treaty was signed on 31 July 1991, but entry-into-force was delayed until 5 December 1994 due to the collapse of the Soviet Union. By way of the Lisbon Protocol to the START treaty signed 23 May 1992, Russia, Belarus, Kazakhstan, and Ukraine became Parties to the treaty as legal successors to the Soviet Union. Upon initiation of the START II negotiations, the original START was renamed to START I. START II was signed by the U.S. and Russian presidents on 3 January 1993, banning the use of multiple independently targetable reentry vehicles (MIRVs) on intercontinental ballistic missiles (ICBMs). Russian ratification of START II was contingent on preservation of the Antiballistic Missile (ABM) treaty. START II was never entered-into-force because the U.S. withdrew from the ABM treaty 13 June 2002 in order to pursue a missile defense system, whereupon Russia withdrew from START II one day later. As a result of START I, there has been a significant reduction in the number of deployed warheads for both the U.S. and Soviet Union.

Under the May 2002 Strategic Offensive Reductions Treaty (SORT), both the U.S. and Russia pledged to reduce the number of deployed strategic nuclear weapons to between 1700 and 2200 by the year 2012. SORT is different from START in that it limits actual warheads, whereas START I limits warheads only through their means of delivery (ICBMs, SLBMs, and Heavy Bombers). Experts have estimated that by the year 2009, the U.S. and Russia arsenals for strategic nuclear weapons ranged from 2200 to 3000 each. The U.S. and Russian presidents signed a preliminary agreement on 6 July 2009 to further reduce the number of active nuclear weapons to between 1,500 and 1,675 from 2,200. In accordance with the agreement, the new caps on nuclear arsenals will need to be fully implemented by 2012.

Although the START and SORT treaties have been the backbone of joint US and Russia efforts toward nuclear disarmament, the treaties have not addressed the discontinuation of weapons-grade fissile material production and disposition of excess weapons-grade materials. The 1993 UN Assembly resolution 48/75L called for negotiations leading to a verifiable treaty banning the production of fissile materials for nuclear weapons. Additional UN activities have followed the initial resolution; however, a final treaty has not yet been completed. The current situation is that the US, France, and the United Kingdom have ceased production. In 1997, the US and Russia signed the Plutonium Production Reactor Agreement (PPRA) to cease production of plutonium for weapons production, which included provisions for monitoring. Although Russia still operates nuclear reactors used previously for production of weapons material, to generate heat and electricity, they do not process the spent fuel. Plans are in place to decommission the Russian production reactors. Unsubstantiated reports indicate that China also has instituted a moratorium on production. Both India and Pakistan apparently are still producing weapons-grade material, and Israel’s position is unclear.

Throughout the Cold War, the US and the Soviet Union produced ∼ 100 and 150 metric tons (MT) of weapons-grade plutonium, respectively. In September 2000, the US and Russia each formally agreed to transform 34 MT of excess military plutonium into a more proliferation-resistant form over the course of 20 years. Current plans for both countries are to irradiate all 34 MT of its plutonium in nuclear power reactors. Plutonium disposition programs in both countries are still in the early stages. The start-up costs of plutonium disposition are extremely high. Currently, Russia favors irradiation in a new generation of fast reactors yet to be developed, and the US favors irradiation in their existing commercial light-water-reactor (LWR) fleet. Additionally, a joint program was developed by the US and Russia to disposition excess highly enriched uranium (HEU). Excess HEU is currently being dispositioned by way of the joint HEU downblend program. The HEU downblend program includes 500 MT of HEU from Russia (of the 1,000–1,500 MT of HEU produced by the USSR during the Cold War), whereas ∼ 175 MT of HEU from the US (from 500 to 750 MT of HEU produced during the Cold War) has been declared excess. The HEU is downblended with natural uranium to produce low enriched uranium (LEU) for commercial power reactor fuel. Over 300 MT of HEU from Russia and tens of MT from the US have already been downblended.