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The primary focus of the research activities within Thermal Hydraulics is to perform experimental and theoretical investigations of single- and two-phase flows and heat transfer in reactor pressure vessels. Due to the specificity of the nuclear reactor environment, such areas as the thermal-hydraulics/neutronics coupling and the fluid/structure interaction dynamics are included as well. The goal is to gain a detailed insight into the governing phenomena: Such knowledge is required to develop and promote computational tools and experimental methods to predict and determine the temperature and flow velocity fields throughout the reactor pressure vessel and other key parameters of interest such as the margins to the Critical Heat Flux (CHF), void fraction distributions, pressure drops as well as flow and structure stability margins.

To this end, the following focus areas of activity are envisaged:

  • Analysis of Critical Heat Flux in LWR fuel assemblies with the objective to improve the accuracy of predictions of thermal margins.
  • Multiphase flow and heat transfer research aiming at improved predictability of void fraction, pressure drop and temperature distributions in LWR fuel assemblies.
  • Analysis of two-phase flow dynamics and fluid/structure interactions in nuclear reactors with a goal to predict stability margins for fluid flows in elastic channels with neutronic power feedback.
  • Development of novel methods and instrumentation for the measurement of thermal-hydraulic parameters of interest for reactor operation and fuel design.
  • Development of macroscopic models to determine the fluctuations of the temperature and flow velocity fields throughout the reactor core for noise diagnostics applications.
  • Development of non ill-posed mathematical models and numerical schemes for the macroscopic simulation of flow dynamics.

Core Physics

The research is directed towards enhancing accuracy and efficiency in predictions of nuclear reactor core properties and phenomena related to nuclear reactions and particle transport. This includes development of new calculation methods, new modelling of core physics phenomena, benchmarking and experimental validation of code systems and coupling of codes.


  • to improve the accuracy, speed, reliability and functionality of core physics computational tools
  • to expand the range of core physics phenomena that can be accurately modelled
  • to expand the range of applicability of existing core physics codes and methods
  • to make full use of experimental data on core properties and phenomena in order to demonstrate enhanced computational capabilities
  • to take part in the exploration of core physical properties of various reactor designs that can be reoriented towards new reactor concepts and in the development of tools for their application.


  • A higher degree of accuracy in core simulation tools improves reactor safety
  • Demonstrating better predictability can relax the need for conservative safety margins, opening the possibility for improved core performance, more efficient use of fissile material and higher flexibility in reactor operation
  • An improved ability to model new, or less well-understood, core phenomena improves the basic understanding of the physics of core systems and may be helpful in improving the overall computational accuracy
  • Improving the applicability of core physics codes may support applications in neighbouring fields, such as core diagnostics, reactor dynamics, reactor experiments, reactor chemistry and fuel technology. By coupling different codes, the overall excellence of the complete computational system may benefit from the particular advantage of each component
  • Taking active part in the development of tools for various reactor concepts opens possibilities to apply new knowledge not only to future nuclear power systems but also to the current generation of nuclear reactors.

Core and Plant Dynamics

Although nuclear reactors are designed to be operated at nominal full power and steady-state conditions, they also have to withstand incident or accident situations leading to non steady-state conditions. Such situations arise typically from equipment malfunction, inappropriate operator action or other perturbations to the system. Thus, predicting the nuclear reactor behaviour in such situations is of prime importance. This requires models and methods that are detailed and accurate enough to represent the dynamic behaviour of the plant with an acceptable level of confidence and fidelity.

The modelling is challenging, since nuclear power reactors are very complex systems. The complexity comes first from the size of the system and the variety of physical and mathematical models that are needed including the strong heterogeneity of the nuclear core and the strong interaction between the neutron kinetics and the thermal-hydraulics, and second from the intervention of external factors (equipment failure, operator action) and the plant automatic control logic.

Concerning the size of the system, it is customary to combine modelling tools at different scales: the macro scales (representing large parts of the system in a coarse manner) and the meso scales (representing some parts of the system in a relatively detailed and sophisticated manner). Such a combined meso/macro scale approach has long been used for modelling the time- and space-dependence of the neutron flux in the nuclear core, where very detailed transport calculations at the fuel assembly level are first carried out to provide macroscopic cross-section data to be used in a core simulator. Such a combined meso/macro scale methodology does not exist yet for the thermal-hydraulic modelling of the time- and space-dependence of the flow and state fields within the nuclear core. Furthermore, most of the macroscopic thermal-hydraulic models are based on diffusive numerical schemes that cannot be applied to fast transients or reactor instabilities and on constitutive flow regime maps developed for steady-state conditions.

The Research Area Core and plant dynamics will thus aim at improving the existing simulation strategies and developing new ones for accurately representing the dynamics of nuclear reactors.

The emphasis will be on:

  • the coupling between simulation tools representing different physical phenomena (neutron kinetics/thermal-hydraulics)
  • the coupling between simulation tools modelling the same physical phenomena but at different scales (CFD/system code, transport/diffusion calculations)
  • increasing the fidelity of primary system simulation for transients and accidents. This includes improved reliability, better accuracy and improved understanding of biases and uncertainties
  • a better understanding of BWR instabilities, of their occurrence and thus a better capability of predicting them
  • a better understanding of the interactions between the fluids in the system and the physical boundaries of the system (Fluid-Structure Interaction, FSI)
  • expanding the knowledge on the range of applicability of existing tools and methods.


The research area Chemistry encompasses all chemical processes and problems of relevance in Nuclear Reactor Technology. This includes radiation chemistry, water chemistry, corrosion, radiation induced interfacial processes, fuel chemistry, actinide chemistry, crud formation, etc.


Increased understanding of chemical processes (in particular radiation induced processes) in nuclear reactors is a prerequisite for safety and performance optimization. Increased understanding could result in more effective management of coolant chemistry which in turn would reduce corrosion and crud formation and thereby reduce the activity in the coolant system. In addition, understanding the chemistry during the course of a possible accident will help in minimising the possible release of radionuclides to the environment and will also help in the subsequent decontamination procedure.

Material Physics and Engineering

In the area of Material Physics and Engineering, one focus area is the understanding of radiation damage in materials used in the reactor. This is an important area since the Swedish plants are aiming for plant life extension (PLEX). The Swedish reactor pressure vessel welds often have a slightly higher nickel content than similar vessels in other countries. The effect of the nickel content on the irradiation damage, as well as the combined effects of nickel and manganese needs to be clarified.

A project is underway at KTH to model the impact of radiation. At Chalmers, a project is planned to use positron annihilation measurements for material characterisation. This will give a broader understanding than just the impact of radiation – the impact of environmental factors and fatigue can also be understood better. Another area of focus is an understanding of the process for initiation of stress corrosion cracking. This is a combination of material condition, environment, stress and ageing of both the material and oxide.

Focus is also on ageing of materials (metals) with regard to the changing of microstructure and mechanical properties. Finally, Fuel Technology is also an investigation area as described separately.

Safety and Severe Accidents

This research area is a multidisciplinary discipline that combines scientific disciplines of reactor physics and nuclear chemistry, engineering subject of reactor technology and social sciences of risk perception and public safety. Its scope of research and education includes risk perception, risk assessment and risk management of rare, high-consequence hazards, which ranges from design-basis accidents to beyond design-basis accidents and severe accidents. Its research program is structured to address both components of Risk (Probability x Consequence) for design-basis, beyond design-basis and severe accidents. Such approach allows for comprehensive and thorough evaluation and resolution of important safety issues facing the nuclear power industry.

Scientific research and education is performed in two tracks aimed at maintaining knowledge and competence, and advancing state-of-the art in nuclear power safety.

Limitation of Consequence by Mitigation – addressing accident phenomena of risk importance to develop measures to limit the accident Consequence

The mission of this research track is to create new knowledge (e.g. data, insights, models, codes and methodology) on reactor severe accident phenomena of risk importance. The main motivation is to reduce the uncertainties in quantification of severe accident risks in a light water reactor. The goal is to help bring to the resolution the long-standing severe accident issues in nuclear power plants, such as steam explosion and corium coolability, that are of paramount importance to the safety of Swedish BWRs.

The activities are structured into following main domains:

  • methods on in-vessel retention (IVR) of the core melt (corium)
  • experimentation and analysis methods on ex-vessel corium coolability
  • experimentation and analysis methods on steam explosion energetics
  • tools and methodologies for integral severe accident analysis
  • deterministic/probabilistic methods in risk assessment.

Reduction of Probability by Prevention – analysis of reactor transients to reduce the Probability of an accident

The purpose of this research track is to perform high-fidelity numerical predictions of the reactor behavior in abnormal transient scenarios of safety significance. The main motivation is to evaluate effects of the Swedish reactor power uprate program by accurately determining the safety margin for plants undergoing power uprates. The goal is to increase the fidelity of primary system simulation, which has high potential for advancement in the state-of-the-art. This approach is at the heart of the excellent safety record of nuclear power – always striving to improve current tools.

The activities are structured into following main domains:

  • methods for high-fidelity simulation of reactor, containment and plant systems
  • methods for analysis of broad range of reactor transients and accidents
  • methods for analysis of advanced systems

Reactor Diagnostics

Power reactor surveillance and noise diagnostics is an area which includes core physics, kinetics and dynamics; theory and application of stochastic processes; finally advanced signal analysis methods and inversion methods. The goal is to elaborate and use methods, by which one can make a diagnostics of the core by measuring fluctuations in certain process signals, such as the neutron flux, temperature, pressure etc. The purpose is either to detect the appearance of some anomaly (vibrations, boiling) etc., and try to quantify it (locate the position, determine the margin to stability etc.) in order to support the operator in determining the severity of the anomaly; or, it can be used to determine reactivity coefficients, velocity of coolant etc. in the normal state, with non-intrusive methods.

Neutron noise diagnostics usually consists of several steps:

  • modelling of the noise source in form of cross section fluctuations
  • calculating the dynamic transfer function between the noise source and the induced neutron noise
  • elaborating an inversion method to unfold the perturbation (noise source) from the induced noise, in possession of the transfer function of the system.

There are also other, more direct types of diagnostics, which do not require the knowledge of a transfer function, e.g. BWR instability, PWR core-barrel vibrations etc. Both method development and application in concrete problems/cases at Swedish nuclear power plants is pursued. Measurement data are available from power plants for test and application of the methods. A large number of practical cases were treated, some of which are used routinely at Ringhals.

One challenge for this area is to consider non-linear and non-stationary systems, and to develop and use methods capable of treating such systems and corresponding signals. Another challenge is to apply noise analysis in all the power plants which plan to perform power uprates, so that the “fingerprints” of the system both before and after the uprates can be compared, and to see what effect, if any, the power uprate incurred for the margins to stability, vibration levels etc.

Work planned in the future comprises:

  • development of methods of noise source modeling
  • development of the methods of calculating the dynamic transfer function of the core (the so-called noise simulator)
  • test and use of intelligent computing methods (earlier called soft computing) for advanced signal analysis of the signals, and for developing powerful unfolding/inversion methods
  • tackling new practical problems as and when they emerge
  • investigation of the dynamics, stability, and noise properties of advanced reactors
  • getting involved into the planning of the instrumentation of, and developing new diagnostic methods for advanced reactors
  • on the long run, to make noise analysis available for all Swedish reactor units, in form of a Diagnostic Centre, possibly with on-line access to the signals, for continuous surveillance and monitoring and operator support at the plants (i.e. to recognise anomalous processes such as the Barsebäck mixer incident at a much earlier stage).

Detectors and Measurement

Detectors and measurements are of key importance for the safe operation of nuclear reactors. Activities within this research area will contribute to and benefit other research areas as they provide the means for validating codes and improving models.

Within the research area Detectors and Measurements the following issues are dealt with:

Development of new detectors for measurements of relevance to nuclear power

This concerns the design and commissioning of new detectors. One example is development of fibre mounted scintillators for neutron detection that can be doped with either Li for thermal neutron detection of Th for fast neutron detection and that are small enough to be inserted into fuel bundles. Another example is development of novel detector materials, for instance crystals for gamma detection that do not need cryogenic cooling.

Development of new detection methods

This refers to use of existing detector types in novel fashions, as well as improved analysis methods. An example of the former could be the use of fast-neutron ionization chambers in BWR diagnostics, and Fourier analysis of gamma-ray spectra to find weak signatures in the tomography of spent nuclear fuel could exemplify the latter. Another example is the development of a neutron source that emits a signal at each neutron emission by the combination of a Cf-252 neutron source and a detector, sensitive to the spontaneous fission of the californium. This type of source can be of great benefit for time correlation measurements both in reactor measurements and for detection of fissile materials.

Introduction of measurement techniques novel to nuclear power

There is a potential for application of measurement techniques that have already been developed in other areas of science, but hitherto not been used in the realm of nuclear power technology. Examples could be positron annihilation measurements for materials investigation or accelerator mass spectroscopy for investigation of the actinide content in spent nuclear fuel. Other examples are the application of emission tomography (on-going work) and neutron transmission tomography.

Combination of different measurement techniques into a system with added values

In nuclear power applications where several measurement techniques are, or may be, applied, the results of the measurement efforts can be maximized by taking systems aspects into account. An example of this type of work is the on-going analysis of the safeguards system.


Nuclear safeguards is a vital part of today’s utilisation of nuclear power. In the modern sense, nuclear safeguards may be defined as comprising all measures that can provide national, regional or international bodies sufficient knowledge, control and means to prevent proliferation of nuclear weapons. The area is therefore rather wide, and it contains both technical and a non-technical parts. In the framework of SKC only measures that belong to the technical part are considered.

The technical part consists of the following areas:

Materials control and accountability

The goal of the research in this area is to elaborate and test methods for surveillance of existing and known nuclear materials (such as in a fuel pool, repository etc.) and detection, identification and quantification of unknown fissile materials (illicit trafficking). From the technical point of view, there are two method areas: non-destructive assay (NDA) and destructive assay (DA).

Non-Destructive Assay (NDA) addresses methods that provide information on the physical properties of nuclear materials in a non-intrusive way. Typically, measuring techniques based on neutron and gamma radiation are utilised even though optical devices such as Cerenkov cameras are also used. The methods can be either passive (using spontaneous emission) or active (using external neutron or gamma sources). Research concerns both the development of advanced detection methods and detectors, as well as development of theoretical and numerical methods for effective data analysis and interpretation and for devising new detection methods.

Destructive Assay (DA) concerns methods that are based on measurements on detached samples of nuclear materials that get physically or chemically processed. Both radiation/particle detection and chemical methods are used. A general feature of DA is the necessary inclusion of chemistry in order to prepare the samples as well as access to a hot cell laboratory. Currently DA is not an active area within SKC, but future work is envisaged.

Containment and Surveillance (C/S)

This area comprises two parts: Surveillance and Seals. Surveillance is performed using a combination of grounded surveillance measures, such as cameras (optical, IR etc) and satellite surveillance (IR, optical, SAR). Sealing is performed either by using a system of passive seals or by using active seals in order to prevent and detect unauthorised access to nuclear materials or buildings etc. An important research area within C/S is to find analytical methods how to quantify performance and assurance of safeguards devices. This research aims at minimising the false-alarm rate and increasing the ability to detect irregular activities.

System analysis

Modern sensor technology potentially generates an exceedingly large amount of safeguards relevant information and the need for data fusion and automated decision support has therefore been recognised as more sophisticated technology is implemented. As a response, the techniques and methods of system analysis have been employed. The overall goals of system analysis in this context are: 1) to increase the efficiency, regarding costs and detection ability of implemented safeguards systems and 2) making the safeguards implementations as little intrusive as possible.

System analysis treats what kind of data each part of a safeguards implementation should output in order to maximise the amount of useful information and minimise the amount of irrelevant information. A related area of research is to find methods of communication between the various parts of a safeguards system that enables the system to create a confident picture of the safeguards status of a nuclear facility (or state). Another area of research is to find methods of quantifying detection limits of implemented safeguards systems and how such methods can be implemented in various algorithms to facilitate decision support.

Fuel Technology

One of the important objectives of nuclear energy engineering today is to improve the performance of reactor fuels. Using burnable poisons and increasing enrichment up to the present regulatory limit of 5%, average burnups of 50 GWd/ton today is common practice in light water reactors. New fuel designs may allow to increase this value up to 60-70 GWd/ton, beyond which however increasing difficulties with fuel restructuring, fission gas release, clad corrosion and pellet-clad mechanical interaction may restrict further improvement, even if higher enrichments would be permitted.

Safety limits and core loading restrictions, established to prevent fuel failures, constrain fuel and core design and limit fuel utilisation and plant operation flexibility. In-core fuel management research aims at finding ways to optimise the fuel utilisation within given safety limits and core operation constraints.

The standard once-through use of uranium based fuel predominantly adopted in Sweden and worldwide represents a non-optimal usage of limited natural resources. Uranium prices are expected to increase as easily available resources are continuously exhausted. Alternative fuel cycle options may provide substantial improvements in resource availability as well as other benefits in fuel performance.

Therefore, it is of interest to perform research on fuel science and fuel technology which would permit to increase both burnup and resource utilisation while maintaining a high degree of fuel reliability. A combination of basic fuel science and technology development is necessary to make progress as efficient as possible.


The long term objective of this research area is to develop fuels for light water reactors which could be licenced for a burnup exceeding 100 GWd/ton, and/or to increase available fuel resources by more than 50%.

The working goals of the research is:

  • to perform experimental and theoretical work in support of the development of new fuel materials
  • to develop advanced in-core fuel management and nuclear fuel design
  • to explore alternative fuel cycles and fissile/fertile fuel components.


The search for new fuel materials and compositions is driven by the potential benefit of increasing the burnup of discharged fuel, enabling to improve fuel utilisation and economy. Advanced fuel materials bear the possibility to improve the resistance towards various fuel failure mechanisms, thereby improving reactor safety and reliability of plant operation. They may also provide material properties more favourable for spent fuel management.

Expected benefits from research activities in the field of in-core fuel management are more streamlined nuclear fuel and core design that make better use of fissile material and more easily fulfil reactor design criteria, including the dynamic response to reactor transients.

Alternative fuel cycle studies may provide a more sustainable handling of natural resources, which may be highly desirable, and even necessary, in a long-term perspective. In the short term, innovative fuel cycle concepts may offer improvements in performance parameters such as core neutronics and dynamics, proliferation resistance and waste characteristics.

Tillhör: SKC
Senast ändrad: 2014-04-24