Band 75, Heft 4

Three dimensional computational-fluid-dynamics simulations are performed for the fluid flow within a 40 rod fuel bundle in a square arrangement with a central moderator channel. To ensure spacing between the rods, the design of the bundle uses thin wires wrapped counter-clockwise around each rod. This geometry is presently investigated in the framework of the European High-Performance Light-Water Reactor (HPLWR), which operates at supercritical pressure of 25 MPa. A section with one revolution located in the evaporator region of the HPLWR core is investigated using hydraulic (to ensure fully developed flow inlet boundary conditions and reference for heated cases) and thermal-hydraulic boundary conditions. The geometry of wrapped wires gives rise to additional mixing and a circulating or ‘sweeping’ flow near the outer and inner regions of the fuel element next to the wall of the so called fuel assembly and moderator box. Some interesting flow features associated with the complex three-dimensional flow with significant transverse velocity components are visualized as the first evaluated result of this diversified investigation.

Higher reactor-core-resonance frequency than expected was recognised at the 1800 MWt BWR Barsebäck 2 in Sweden during stability measurements in 2004, and has subsequently been analysed in some detail. Estimation of the void velocity was made by analysing the phase shift between two levels in the LPRM-detector strings, and compared to calculations. Discrepancies in calculated frequency response between simulated perturbation and results from noise analysis were observed, whereas good agreement between calculated and measured void transport was found. The observed deviations in frequency calculations are most likely affected by the pronounced downward shift in power during the stability measurements, and in such circumstances, the shift in reactor-core-resonance frequency can be estimated from void transport calculations.

In advanced water cooled reactors, an annular gap exists between pressure tube and calandria tube. The gap is closed from top but is open from bottom. Due to differential temperature between pressure tube and calandria tube, air flow is induced by natural convection. This leads to heat transfer from pressure tube to calandria tube. The quantification of the heat transfer between pressure tube and calandria tube is numerically carried out with the help of the CFD code PHOENICS. Validation of the CFD code with experimental results and some established computational work from the literature has been done in order to verify the accuracy of the code. The natural convection phenomenon in the annular gap is then simulated. The velocity and temperature fields obtained from the CFD simulation are used to compute local and average heat transfer coefficients. Heat transfer coefficients for various pressure tube temperatures are computed. The effect of water on the heat transfer in the annular gas is also studied.

The work realized under the Severe Accident Research Network of excellence (SARNET) project has shown that the SOPHAEROS module, part of Accident Source Term Evaluation Code (ASTEC) can be fully used to simulate the fission products transport and deposition phenomena in the CANDU Primary Heat Transport (PHT) system. This paper presents an uncertainty analysis for the fission products transport in the CANDU PHT system during a severe accident to obtain the domains of the output parameters, for this study masses of Caesium, Strontium and Iodine deposited in the PHT system and its nodes, taking into account the associated input parameters uncertainties. Five uncertain parameters, the starting time for the releasing process, the duration of the releasing process, the releasing fractions for Cs, Sr and I have been chosen. To generate aleatory values for the uncertain parameters, a method and software have been developed and Monte Carlo simulations to determine uncertainties propagation through the SOPHAEROS module has been carried out.

During the past decade a number of modifications were implemented in the Ignalina NPP reactors cores. A significant reactor core loading change allowed increasing the safety level of the plant. During previously performed safety analyses a question has arisen concerning conservatism of used initial boundary conditions for investigation. The impact of different axial fuel burn-up and power-density profiles in the reactor core on modelling results as well as estimation of conservative conditions for the analysis of reactivity initiated accidents in the modernized RBMK-1500 reactor core were analyzed. The results of the investigations are presented in this paper. Simulation results show that previously used initial boundary conditions for modelling of accidents are changing due to a change of composition and loading in the modernized RBMK reactor core.

Two RBMK-1500 water-cooled graphite-moderated channel-type power reactors at the Ignalina Nuclear Power Plant (INPP) are under decommissioning now. The total mass of irradiated graphite in the cores of both units is more than 3600 tons. The main source of uncertainty in the numerical assessment of graphite activity is the uncertainty of the initial impurities content in graphite. Nitrogen is one of the most important impurities, having a large neutron capture cross-section. This impurity may become the dominant source of C-14 production. RBMK reactors graphite stacks operate in the cooling mixture of helium-nitrogen gases and this may additionally increase the quantity of the nitrogen impurity. In this paper the results of the numerical modelling of graphite activation for the INPP Unit 1 reactor are presented. In order to evaluate the C-14 activity dependence on the nitrogen impurity content, several cases with different nitrogen content were modelled taking into account initial nitrogen impurity quantities in the graphite matrix and possible nitrogen quantities entrapped in the graphite pores from cooling gases.

Two experimental fuel elements with different UO 2 grain size were irradiated in the C2 capsule in the TRIGA MT reactor of INR Pitesti to a burn-up of 178.9 MWh kg −1 U and linear powers between 45 to 60 kW m −1 . Both elements were instrumented to measure internal gas pressure during irradiation. Experimental results presented in this paper represent data relative to two fuel elements with different UO 2 grain size. The predictions of fuel performance code ROFEM in terms of internal gas pressure were compared with the experimental data. Experimental programs are in progress at INR Pitesti. New instrumented tests are planed to be performed in the C2 capsule in order to establish the pellet microstructure influence on fission gas release. This paper describes measuring techniques used at INR Pitesti and present experimental results of the in-pile tests comparatively with the ROFEM code predictions.

The derivation of the time-dependent P 1 model by the spherical harmonics expansion approach to solve the neutron transport equation is reviewed and its main features and physical characteristics are illustrated. The telegrapher's equation is presented and its formulation for multiplying systems is discussed. In particular, it is evidenced that the presence of a time derivative of the source terms in the telegrapher's equation implies the appearance of time derivatives of the flux and of the precursor concentrations. The role of these terms is evaluated and discussed.

The critical slab problem of the reflected isotropic one-speed neutron transport equation is solved with the Chebyshev polynomial approximation. The efficiency of the reflection coefficient, R, in the neutron transport equation is obtained for different c values. For the solution, the Mark boundary condition is used. The values obtained from this approximation are compared with results obtained by the spherical harmonics method.

A formulation for the calculation of reactivity using a recursive process is presented in this paper, as well as the treatment to reduce noise intensity that is found in the nuclear power signal. Using the history of nuclear power considered as the memory of such power and the filter exponentially adjusted with the least squares method, it is possible to reduce the nuclear power fluctuations without causing attenuation for the calculation of reactivity and with a smaller delay than that for low-pass filter of first order delay filter.

To develop a model with scattering medium, a consistent backward forward scattering kernel is described and applied to the transport equation in plane geometry. Asymptotic solutions of critical half thicknesses are computed by using the direct second kind Chebyshev polynomial method (U N ).

Interactions of radionuclides in extremely low concentrations were determined by measurement of their apparent self-diffusion coefficients. Formation of trace amounts of hydrolysis products, hydroxides of Ru IV , Mn IV , Fe III , Eu III and of ionic precipitate of AgI were determined from apparent self-diffusion coefficients. Transition of 85 Sr 2+ by ion exchange or by coprecipitation into trace amounts of Fe III -hydroxide and Mn IV -hydrated oxide was revealed by this method. 85 Sr 2+ sorbed on trace amounts of Fe III -hydroxide and Mn IV -hydrated oxides lost its individual chemical properties. This can negatively affect its separation from solution. Formation of unstable embryos of Ag 131 I trace precipitate in unsaturated solution at concentration of Ag + 1 · 10 −9 mol · l −1 and 131 I −1 · 10 −8 mol · l −1 was detected by self-diffusion. Formation of trace amounts of Ag 131 I led to the decrease of sorption of 131 I − on hydrated ferric oxide.