Band 70, Heft 4

Neutronic-thermal hydraulic oscillations are a well known phenomenon in the operation of boiling water reactors. Despite the progresses that have been made concerning their simulation, the computational determination of the stability threshold in the power-flow map still suffers from inaccuracies. In this work, the decay ratio of neutronic-thermal hydraulic oscillations and hence the stability threshold was calculated with the SIMULATE-3K code developed by Studsvik Scandpower. It is known from experience that there is a higher probability for a less stable reactor operation in the case of one or more pumps tripped while the remaining pumps run at minimum speed. For the analysis of this state, the characteristics of the pump system were modelled carefully. It is shown that the stability behaviour is determined by the axial and radial power profile due to control rod manoeuvring and fuel burn-up and by the number of pumps running. Furthermore the calculations reveal the impacts of the insertion of shallow rods and of the control-rod group insertion on the decay ratio. A comparison of the calculations with experimental results of stability measurements shows that by a thorough analysis and proper modelling of the plant characteristics the decay ratios can be calculated with the SIMULATE-3K code.

During the last years one focus of German PWR safety analyses was boron dilution events with the potential of reactivity transients. Coolant with a low boron concentration could be collected in localized areas of the reactor coolant system e. g. by separation of borated reactor coolant into highly concentrated and diluted fractions (inherent dilution) which can occur during reflux-condenser heat transfer after a small break loss of coolant accident with a limited availability of the emergency core cooling systems. The TüV NORD SysTec appraises in the framework of follower core assessments safety analyses of boron dilution events presented by the utilities. These analyses are based on the simulation of boron dilution and transport processes in conjunction with a number of dedicated experiments. The analyses shall demonstrate that boron dilution events cannot lead to recriticality of the core. Hence the boron concentration at the core inlet has to be determined. TüV NORD Sys Tec applies the CFD code FLUENT for the investigation of boron dilution events in pressurized water reactors. To affirm the FLUENT abilities for the simulation of boron dilution events a validation against the ROCOM experiment T6655_21 with density driven coolant mixing was performed. This validation proves that FLUENT is able to determine the boron concentration in the reactor core with sufficient accuracy and to solve the safety issue, whether the core gets critical or not.

The comparison of different methods of calculation and computer codes used to obtain helium isotope production cross-sections for nuclear reactions induced by nucleons at energies up to 1 GeV was performed. The cross-sections were calculated using the GNASH code, the modified ALICE code, the DISCA code and the CASCADE/INPE code. The evaluation of the helium production cross-section for tantalum and tungsten was performed using the results of model calculations, available experimental data and systematics predictions. The cross-sections were obtained for nuclear reactions induced by neutrons and protons at energies up to 1 GeV

The fire hazard analysis methodology and main results of Ignalina Nuclear Power Plant RBMK-1500 type reactor's systems and rooms with safety elements and also rooms adjacent to main rooms and other fire hazardous rooms are presented. In total this concerns about 2100 rooms in the control area of one Unit, where the fire could influence nuclear safety of the power plant. For this purpose the decision-making computerized system FIRE-INPP was developed. It includes modules for recording necessary information, data processing, analysis and assessment on compliance of rooms at Ignalina NPP with national fire safety requirements and international recommendations regarding fire hazard. Screening algorithms and decision-making module are the basis for the decision-making computerized system, that allows to perform unambiguous, reliable and fast fire hazard analysis and give recommendations to ensure fire safety adequacy.

One of main issues in planning future power system expansion is forecast of expected electricity generating costs. Due to the fact that the forecast should be extended to plants lifetime it is best to consider and compare lifetime levelized cost of generated electricity for each of the candidate plants. The analysis is a part of studies for the future expansion of the Croatian power system, but is valid generally. The main competitors to nuclear power plants are combined cycle natural gas fired plants whose economy strongly depends upon gas cost. The expected increase of gas cost in following decades can substantially influence the competitiveness of nuclear plants versus gas fired plants. In addition, the costs of electricity produced by coal fired plants and wind electricity generators were also considered. Due to uncertainties of input parameters it is advantageous to use probabilistic instead of deterministic method of analysis. The paper considers the problem of the probabilistic distribution of nuclear power plant investment costs for which such plants could be competitive to gas fired plants for an estimated probable range of gas cost increase rates.

A method is proposed for the numerical and analytical calculations of the adjoint neutron spectrum in the resonance energy region, for narrow resonances, in order to reduce the numerical computations involved. The adjoint neutron spectrum of infinite cylindrical fuel cell is calculated according to a very simple analytical method with reasonable level of accuracy. The results obtained with this analytical method were compared to a reference solution obtained with a numerical solution of the neutron balance adjoint equations. The adjoint neutron spectrum is useful to determine the neutron resonance absorption, so that multigroup adjoint cross sections used by the adjoint diffusion equation can be obtained.

A positive isothermal temperature reactivity coefficient has been measured at zero power conditions at the TRIGA research reactor of the J. Stefan Institute in Ljubljana. The coefficient is measured in a critical reactor for different core configurations at room temperature and at low power. The reactor is filled with standard, 20 % enriched, fuel elements containing 12 wt % uranium with 6 % average burn-up. The experiments are analysed by reactor calculations using WIMSD code to provide the reactor physics explanations of the results measured. The positive temperature reactivity effect can be explained by reduced absorption in water prevailing at low fuel temperature over negative temperature effects on the reactivity in the fuel elements. The positive effect is due to the thermal spectrum shift in water and is not the consequence of a change in water density. The positive effect decreases with increasing temperature. The negative reactivity effect in fuel prevails at fuel temperatures above 50°C. The positive isothermal temperature reactivity coefficient has no effect on normal reactor operation in steady state and pulse mode.

The critical slab problem for linearly anisotropic scattering is investigated using the singular eigenfunction method. The third form of the transport equation is considered. The singular eigenfunctions for linearly anisotropic scattering are inserted into the Green's function. This Green's function with the full range orthogonality relations of the singular eigenfunctions together with the appropriate boundary conditions provide the criticality equation. This equation is exact and leads as shown in tables to fast converging numerical results.

The main objective is to study the effects of spectral shifting in an inertial confinement system for kT/shot energy regime on the breeding performance for tritium and for high quality fissile fuel. A protective liquid droplet jet zone of 2 m thickness is used as coolant, energy carrier, and breeder. Flibe as the main constituent is mixed with increased mole-fractions of heavy metal salt (ThF 4 or UF 4 ) starting by 2 moles% up to 12 moles%. Spectrum softening within the inertial confinement system reduces the tritium production ratio (TBR) in the protective coolant to a lower level than unity. However, additional tritium production in the 6 Li 2 DT zone of the system increases TBR to values above unity and allows a continuous operation of the power plant with a self-sustained fusion fuel supply. By modest fusion fuel burn efficiencies (40 to 60 %) and with a few mol.% of heavy metal salt in the coolant in form of ThF 4 or % UF 4 , a satisfactory TBR of > 1.05 can be realized. In addition to that, excess fissile fuel of extremely high isotopic purity with a rate of ∼ 1000 kg/year of 233 U or 239 Pu can be produced. Radiation damage through atomic displacements and helium gas production after a plant operation period of 30 years is very low, namely dpa < 1 and He < 2 ppm, respectively.

The non-proliferation characteristics of the Fixed Bed Nuclear Reactor (FBNR) is based on both, the extrinsic concept of sealing and the intrinsic concept of isotope denaturing. Its small spherical fuel elements are confined in a fuel chamber that can be sealed by the authorities for inspection at any time. Only the fuel chamber is needed to be transported from the fuel factory to the site and back. There is no possibility of neutron irradiation to any external fertile material. Isotopic denaturing of the fuel cycle either in the U-233/Th or Pu-239/U cycle increases the proliferation resistance substantially. Therefore, both concepts of “sealing” and “isotope denaturing” contribute to the fool proof non-proliferation characteristics of the proposed reactor. This contribution is focussing only on the won-proliferation aspect of the reactor.