The sixth Euratom framework programme 2003–2006: a driving force for the construction of the Nuclear European Research Area

At the Lisbon 2000 summit, a strategic goal was proposed for the European Union: “to become the most competitive knowledge-based economy with more and better employment and social cohesion by 2010”. Overall, in particular in the community of nuclear fission research, this EC initiative was well accepted by the main stakeholders. In Europe, the main stakeholders (i.e. suppliers and/or demanders) of nuclear research are actually: the research organisations (with mixed public/private funding), the manufacturing industry (or vendors), the utilities (or engineering companies), the regulatory bodies (or technical safety organisations) and the academia (or universities). In response to the Lisbon 2000 objectives, Commissioner P. BUSQUIN launched the concept of European Research Area (ERA) which sets also the frame for the FP-6.In this invited lecture, research and training in nuclear fission are looked at from a nuclear ERA perspective with emphasis on the three success factors of any European integration policy, namely: needs, vision and instruments, that ought to be strongly shared amongst the stakeholders in the EU-25.The following questions are addressed. What is driving the current EU trend towards more research, more education and more training in general? Regarding nuclear fission, in particular, who are the end-users of Euratom “research and training” and what are their expectations (needs) from EU programmes? Do all stakeholders share the same vision about European research and training in nuclear fission? What are the instruments proposed by the Commission to the end-users to conduct joint research programmes?The above questions are all linked to nuclear knowledge management. It is shown that, in Community research, production of knowledge remains the most important objective, be it to generate products or services. However, dissemination/transfer (e.g. education and training) and exploitation (e.g. innovation) of knowledge become equally important as production, if one wants all energy options to remain open for the future generations.The answers to these questions and the examples given are taken primarily from reactor safety research (i.e. the area covered by FISA-2003) but are thought to be easily extrapolable to the other areas of nuclear fission research (i.e. radwaste management, innovative reactor concepts and safeguards, radiation protection, medical applications).In conclusion, amongst the stakeholders there seems to be a wide consensus about needs and instruments but a real common vision about a EU nuclear research strategy is still lacking. As a result, it is felt that European research in nuclear fission is still too much fragmented across Europe. It has not yet reached the “critical mass” necessary to levy the governmental and industrial resources needed to face all challenges posed by an ageing nuclear power park and its renewal by the year 2015 (estimated end of lifetime).     

Effect of polarization force on the Jeans instability in collisional dusty plasmas

The Jeans instability in collisional dusty plasmas has been analytically investigated by considering the polarization force effect.Instabilities due to dust-neutral and ion-neutral drags can occur in electrostatic waves of collisional dusty plasmas with self-gravitating particles.In this study,the effect of gravitational force on heavy dust particles is considered in tandem with both the polarization and electrostatic forces.The theoretical framework has been developed and the dispersion relation and instability growth rate have been derived,assuming the plane wave approximation.The derived instability growth rate shows that,in collisional dusty plasmas,the Jeans instability strongly depends on the magnitude of the polarization force.     

International standard problem on containment thermal–hydraulics ISP47: Step 1—Results from the MISTRA exercise

The understanding of hydrogen distribution during severe accidents in a nuclear reactor containment is still an open issue. Several containment thermal–hydraulics international standard problems (ISP) have been conducted to address this topic. However, the predictions made by the available lumped parameter or CFD computer codes were generally not satisfactory. Therefore, a new exercise was launched in 1999 using new state-of-the-art experimental facilities TOSQAN, MISTRA and ThAI that included sophisticated 3D instrumentation and well-controlled boundary conditions. Predictive capabilities of important and still uncertain phenomena such as wall condensation, natural circulation and gas stratification are assessed. In addition, comparison between lumped parameter (LP) and CFD codes and assessment of the capability of CFD codes to deal with scaling effects are performed. This article reports on the part of the exercise which concerns the MISTRA facility including experimental results and blind benchmark exercises.     

The shut-down of the Barsebäck 1 BWR: A unique opportunity to measure the power distribution in nuclear fuel rods

Reactor poolside measurements of gamma radiation specific for the fission product ¹⁴⁰La (1596 keV) have been used for an experimental determination of axial power distributions in 55 nuclear fuel rods irradiated in the Barsebäck 1 BWR nuclear power plant. The measurements take advantage of the unique situation of a very short last reactor cycle of only three months due to the out-phasing of the reactor unit at November 30 1999. ¹⁴⁰La whose decay is controlled by the mother nuclide ¹⁴⁰Ba with the half-life 12.75 days reflects an average power distribution, representative for the latest weeks of core operation (in this case basically during November 1999). The measured intensities have been transformed into a 25 nodal representation to allow a precise and direct comparison with the corresponding calculated power distribution. The 55 rods were selected from two different fuel assemblies with average burn-ups of 1.9 and 9.7 MWd/kgU, respectively (that is one fresh bundle and one slightly more than one cycle bundle). The stability and the linearity of the measurement system were evaluated. The linearity was checked using the two-source method. The stability was checked by recurrent measurements on a reference fuel rod. The results have been used in the validation of the pin power reconstruction model of Westinghouse 3D core simulator POLCA-7. The deviation between measured and calculated ¹⁴⁰Ba concentration (expressed as radial error) is typically a few percent on rod level. Results indicate that also Gd-rods are properly modelled over a broad range of conditions. It is indicated that predictions for fuel rods in their first month of operation are less accurate than for the rest of the rods.     

Results of the verification period 1985–1990 of the German-Swiss high converting reactor PWR-C1

A high converting reactor (HCR), the PWR-C1, is described, which was developed on the basis of the commercially proven pressurized water reactor (PWR) technology by Siemens AG Power Generation Group in cooperation with the Karlsruhe Nuclear Research Center (KfK), the Paul Scherrer Institute at Würenlingen (PSI) and the Technical University of Braunschweig (TUBS).The essential features of the concept in comparison to the conventional Konvoi-reactor core are a triangular fuel pin lattice with a tighter pitch and a higher density of control assemblies. The main goal was a better fuel utilization, which was achieved mainly by hardening of the neutron energy spectrum. Because of the lower moderation ratio due to the tighter pitch, the neutron spectrum is shifted into the epithermal energy range, whereby the conversion rate is augmented, from 0.35 for a Konvoi-reactor core to 0.75 for a PWR-C1. The PWR-C1 concept is the result of extensive investigations of tight and very tight fuel pin lattices, concerning neutron physics, thermohydraulics, emergency cooling and mechanical design. For establishment and improvement of codes several experiments were performed in the mentioned areas; (1) the PROTEUS experiments for physics investigations at PSI Würenlingcon, (2) the CHF experiments at Siemens Karlstein and KfK, and (3) the NEPTUN and FLORESTAN flooding experiments at PSI and KfK. All these experiments led to the validation or improvement of codes and/or the verification of calculational methods. Also methods used in conventional design activities could be improved.The investigations in mechanical design too led to results, which partly are of great use for conventional design work.     

Fission product release under severe accidental conditions: general presentation of the program and synthesis of VERCORS 1–6 results

The French Nuclear Protection and Safety Institute (IPSN) launched the HEVA-VERCORS program in 1983, in collaboration with Electricité de France (EDF). This program is devoted to the source term of fission products (FP) released from PWR fuel samples during a sequence representative of a severe accident. The analytical experiments are conducted in a shielded hot cell of the LAMA facility of the Grenoble center of CEA (Commissariat à l’Energie Atomique); as simplified tests addressing a limited number of phenomena, they give results complementary to those of the more global in-pile PHEBUS experiments. Six VERCORS tests have been conducted from 1989–1994 with higher fuel temperatures (up to 2600 K) compared with the earlier HEVA tests in order, in particular, to quantify better the release of lower volatile FPs. This paper gives an overview of the experimental facility, a synthesis of FP release from these tests and exhibits, as an example, some specific results of the VERCORS 6 test, performed with high burn-up fuel (60 GWd tU⁻¹). The on-going VERCORS HT–RT program, designed to reach fuel liquefaction temperatures, is described before conclusions are drawn.     

Neutron cross-sections of Pu in the energy range 5–20 MeV

The evaluation of neutron cross sections and angular distributions of ²⁴²Pu in the energy range 5–20 MeV was performed using the direct interaction treatment by coupled channel method and the compound nucleus mechanism (statistical model) taking account of pre-equilibrium effects. The calculated cross-sections are in very good agreement with the existing experimental data (total and fission cross-sections). More accurate theoretical data have been obtained by the extension of procedures and parameterizations (previously used only for the main compound nucleus) in the incident neutron energy range where more compound nuclei are involved.     

In situ XRD analysis of the oxide layers formed by oxidation at 743 K on Zircaloy 4 and Zr–1NbO

Two alloys, having different oxidation behaviour (Zy4 and Zr–1NbO), have been investigated during oxidation at high temperature (743 K) and low oxygen pressure (10 kPa) by in situ X-ray diffraction (XRD). Tetragonal phase content and ‘pseudo-stresses’ on the monoclinic phase have been measured as a function of the oxide layer thickness. The tetragonal phase contents are similar for both alloys and decreased with the oxide layer thickness. Pseudo-stresses were much more compressive on Zr–1NbO alloy, with limited changes at the corrosion kinetics transition. On cooling, the tetragonal fractions do not change, while ‘pseudo-stresses’ decreased in different ways for the two alloys. With respect to stress analysis, no correlation was found between ‘pseudo-stresses’ and tetragonal phase content. In addition, due to the thermoelastic properties of the highly anisotropic phases of the zirconia, large internal thermal stresses are expected to develop during any temperature changes. The orders of magnitude of them are similar to the stresses induced by swelling during oxidation from Zr to ZrO₂.     

On the round table discussion on reactor power margins published in Nuclear Engineering and Design 163 (1–2) 1996

Some comments concerning the critical heat flux (CHF) correlations, the usage of the direct substitution versus the heat balance methods for correlation evaluation and the definition of CHF margins considered in the round table discussion on reactor power margins are presented.     

Evolution of ionization waves in a multi-pulsed plasma jet: the role of memory charges

In this paper, we discuss the properties of ionization waves(IWs) in a multi-pulsed plasma jet while using the two-dimensional computational approach. The IWs are generated by application of three short negative pulses with a repetition frequency 12.5 MHz. The simulations are performed continuously during a single run while accounting for charges accumulated inside(surface charges)and outside(space charges) the tube. The plasma forming gas mixture(He/O2 = 99.8%/0.2%) is injected through the disc...     

Safety performance of the 4S reactor on the ATWS events – statistical estimation of uncertainty

4S reactor is a sodium-cooled fast reactor developed as a small-decentralized power supply. The name of “4S” in this reactor stands for Super-Safe, Small and Simple, and they show representative features of the reactor.

The purpose of the present work is to evaluate quantitatively the super-safety of 4S reactor, and the safety performance is analyzed with ARGO-3, which is a plant dynamics code of a sodium-cooled fast reactor.

In this evaluation, some events, such as Unprotected Loss of Flow (ULOF) and Unprotected Transient Overpower (UTOP), are selected as typical cases from various transients and accidents. After metrics concerned with safety design is defined for each event, it is evaluated with statistical methods whether each metric satisfies acceptance criteria in a given criteria level.

Result about ULOF is as follows. The coolant temperature in the nominal hottest assembly outlet, “T_c” is selected as metric, and the upper side value of 95% confidential section in T_c is below 900 °C that is acceptance criteria. Also in UTOP, it is shown that the fuel maximum temperature in the nominal hottest assembly, “T_f” satisfies acceptance criteria. This result shows that 4S reactor has margin for safety acceptance criteria.     

‘Quantization’ of stochastic variables: Description and effects on the input noise sources in a BWR

A set of macrostochastic and discrete variables, with Markovian properties, is used to characterize the state of a BWR, whose input noise sources are of interest.The ratio between the auto-power spectral density (APSD) of the neutron noise fluctuations and the square modulus of the transfer function (SMTF) defines ‘the total input noise source’ (TINS), the components of which are the different noise sources corresponding to the relevant variables.A white contribution to TINS arises from the birth and death processes of neutrons in the reactor and corresponds to a ‘shot noise’ (SN). Non-white contributions arise from fluctuations of the neutron cross-sections caused by fuel temperature and steam content variations. These terms, called ‘flicker noises’ (FN) are characterized by cut-off frequencies related to time constants of reactivity feedback effects.The respective magnitudes of the shot and flicker noises depend not only on the frequency, the feedback reactivity coefficients or the power of the reactor, but also on the ‘quantization’ of the continuous variables introduced, such as fuel temperature and steam content.The effects of this last ‘quantization’ on the shapes of the noise sources and their sum are presented in this paper.     

The determination of the Nb(n,n′) Nb cross-section in the neutron energy range 1–6 MeV

Measurements of the 30 keV isomer production cross-section in ⁹³Nb, excited by neutron inelastic scattering have been made at 10 neutron energies in the range 1–6 MeV. Small foils of niobium and suitable monitor elements were exposed to high fluences of monoenergetic neutrons and the cross-section was determined from the K X-ray activities so induced. The high purity Ge detector used in these X-ray measurements was calibrated against a standard solution of ^93mNb which is used as an international reference material for reactor neutron dosimetry. The neutron fluence of the irradiation was measured with a low-efficiency ²³⁵U fission chamber in which the thin fissile deposit was located immediately behind the niobium and monitor foils. The latter acted as secondary measures of the neutron fluence but could be used as the primary standard in the event of failure of the fission chamber. Thus, the production cross-section was measured relative to the ²³⁵U fission cross-section which is a standard reference cross-section. Comparisons are made with other experimental data and with nuclear model calculations of the cross-section and recommended values based on these and our experimental data are presented.     

Experimental studies of the absolute total delayed neutron yields from neutron induced fission of U in the energy range 1–5 MEV

Measurements of the total delayed neutron yield from fast neutron induced fission of ²³⁸U have been made using the experimental method based on the periodic irradiation of the fissionable sample by neutrons from a suitable nuclear reaction. The preliminary results on the energy dependence of the total delayed neutron yield are presented. According to the comparison of the experimental data with our prediction based on correlation properties of delayed neutron characteristics, it is concluded that the value of the total delayed neutron yield near the threshold of the (n,f) reaction is not a constant.     

Role of nuclear energy in environment, economy, and energy issues of the 21st century – Growing energy demand in Asia and role of nuclear

The economic growth of recent Asia is rapid, and the GDP and the energy consumption growth rate are about 8–10% in China and India. The energy consumption forecast of Asia in this century was estimated based on the GDP growth rate by Goldman Sachs. As a result, about twice in India and Association of South East Asian Nations (ASEAN) and about 1.5 times in China of SRES B (Special Report on Emission Scenarios) are forecasted. The simulation was done by Grape Code to analyze the impact of energy increase in Asia. As for the nuclear plant in Asia, it is expected 1500 GWe in 2050 and 2000 GWe in 2100, in the case of the environmental constrain. To achieve this nuclear utilization, there are two important aspects, technically and institutionally.

A. Development of the CANDLE core and/or the Breed and Burn core.

B. The establishment of the stable nuclear fuel supply system like “Asian nuclear fuel supply organization”.