Volume 74, Issue 1-2

The AVR pebble bed reactor (46 MWth) was operated 1967 – 1988 at coolant outlet temperatures up to 990°C, which are suitable for process heat applications. Also because of a lack of other experience the AVR operation is a basis for future HTRs. This paper deals with insufficiently published but unresolved safety problems of AVR and draws conclusions for future pebble bed HTRs: Although the AVR was operated only for about 4 years at coolant temperatures > 900 °C its primary circuit is heavily contaminated with dust bound metallic fission products ( 90 Sr, 137 Cs), which create major problems in current dismantling. The end of life contamination reached several percent of a single core inventory. The AVR contamination was mainly caused by inadmissible high core temperatures, and not – as presumed in the past – by inadequate fuel quality only. The high core temperatures, which cannot be easily measured in pebble beds, are probably caused by the insufficiently examined pebble bed mechanics. Safety of future reactors requires a gas tight containment and in addition elaborate R&D or reduction of requirements concerning temperatures and burn-up.

The “High Performance Light Water Reactor” (HPLWR) is a Light Water Reactor operating with supercritical water as coolant. At a pressure of 25 MPa in the core, water is heated up from 280 to 500 °C. For these conditions, the envisaged net plant efficiency is 43.5 %. The core design concept is based on a so-called “3-pass-core” in which the coolant is heated up in three subsequent steps. After each step, the coolant is mixed avoiding hot streaks possibly leading to unacceptable wall temperatures. The design of such a core comprises fuel assemblies containing 40 fuel rods and an inner and outer box for a better neutron moderation. Nine of these are assembled to a cluster with common head- and foot piece. The coolant is mixed inside an upper and inside a lower mixing chamber and leaves the reactor pressure vessel through a co-axial pipe, which protects the vessel wall against too high temperatures.

Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensors for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280 cal g –1 UO 2 . Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti.

One of the most versatile types of heat exchangers used is the plate heat exchanger. It has principal advantages over other heat exchangers in that plates can be added and/or removed easily in order to change the area available for heat transfer and therefore its overall performance. The cooling systems of Egypt's second research reactor (ETRR 2) use this type of heat exchanger for cooling purposes in its primary core cooling and pool cooling systems. In addition to the change in the number of heat exchanger cooling channels, the effect of changing the amount of mass flow rate on the heat exchanger performance is an important issues in this study. The inertia flywheel mounted on the primary core cooling system pump with the plate heat exchanger plays an important role in the case of loss of flow transients. The PARET code is used to simulate the effect of loss of flow transients on the reactor core. Hence, the core outlet temperature with the pump-flywheel flow coast down is fed into the plate heat exchanger model developed to estimate the total energy transferred to the cooling tower, the primary side heat exchanger temperature variation, the transmitted heat exchanger power, and the heat exchanger effectiveness. In addition, the pressure drop in both, the primary side and secondary side of the plate heat exchanger is calculated in all simulated transients because their values have limits beyond which the heat exchanger is useless.

The functions of the Instrumentation and Control (I&C) system in research reactors, the changes in its design according to the advances in the technology, and the internationally established safety requirements on the design and operational performance of this system are reviewed. The main features of the communication networks commonly used in the Supervision and Control systems (SCS) are presented. A methodology for the performance analysis of the communication networks of computer-based distributed SCS is developed and presented along with discussions. Application of this methodology to a modern SCS of a typical research reactor is illustrated.

A one dimensional steady state thermal analytical model has been developed to study the heat transfer and temperature distributions in a quartz ampoule filled with MoO 3 powder. The source heat generation inside the ampoule is released from the high neutron flux (1.4 × 10 14 neutrons/cm 2 s) interaction with MoO 3 . Natural and forced convections heat transfer boundary conditions are adopted during the irradiation process. The peak temperatures in MoO 3 powder and quartz are calculated and compared with their melting temperatures to ensure the irradiation safety criteria.

One of the main issues in a fusion reactor is the need for the frequent replacement of first wall structures during its lifetime. In order to eliminate this problem completely, use of a protective liquid wall between the fusion plasma and solid first wall has been proposed by various researchers. One of the major fusion energy reactor designs is called HYLIFE-II, an inertial fusion energy (IFE) type reactor that also utilizes such a flowing liquid wall to extend the lifetime of the first wall to its lifetime. This study presents the neutronic analysis of the HYLIFE-II reactor using Li 2 BeF 4 (Flibe) with various grades of plutonium as a flowing liquid wall. For this purpose, four different coolants containing Flibe with 0.5 % or 1 % of weapon grade (WG) PuF 4 and Flibe with 0.5 % or 1 % reactor grade (RG) PuF 4 were considered. Main neutronic parameters: tritium breeding ratio (TBR), energy multiplication factor (M) and radiation damage at the first wall were computed with respect to the liquid wall thickness and coolant type. The highest TBR and M values were reached when the Flibe with 1 % WG PuF 4 was used in the liquid wall.

In this article, a number of due changes in both a discrete ordinates neutron transport model and a companion void fraction evaluation scheme are introduced. These changes are due to the explicit consideration of a weakly divergent neutron beam, i. e. a neutron beam consisting of a monodirectional component with normal incidence, which is modeled by a Dirac delta distribution, and an angularly continuous component, which is modeled by a smooth function of the angular variable. Computational tests are performed in order to illustrate the numerical consistency of the evaluation scheme for weakly divergent beams with respect to the order of angular quadrature, as well as its low sensitivity to experimental inaccuracies in the detector responses.

In this paper the application of the Laplace transform method is described in order to determine the energy-dependent albedo matrix that is used in the boundary conditions multigroup neutron diffusion eigenvalue problems in slab geometry for nuclear reactor global calculations. In slab geometry, the diffusion albedo substitutes without approximation the baffle-reflector system around the active domain. Numerical results to typical test problems are shown to illustrate the accuracy and the efficiency of the Chebysheff acceleration scheme.

A general theoretical scheme for the U N method and the evolution of a general eigenvalue spectrum are described. The applicability of U N method to one dimensional slab geometry neutron transport problems is discussed. The eigenvalue spectrum is calculated for isotropic scattering with different values of the parameter c, the mean number of secondary neutrons per collision, known as the fundamental eigenvalue. Then the critical slab problem has been studied. The critical half thicknesses are computed for different values of c. For the solution, Mark and Marshak boundary conditions are used. Results are obtained for both, U N and P N approximations for comparison.

In this paper, the Particle Swarm Optimization (PSO) algorithm is modified to deal with Multiobjective Optimization Problems (MOPs). A mathematical model for predicting the dynamic response of the H. B. Robinson nuclear power plant, which represents an Initial Value Problem (IVP) of Ordinary Differential Equations (ODEs), is solved using Runge-Kutta formula. The resulted data values are represented as a system of nonlinear algebraic equations by interpolation schemes for data fitting. This system of fitted nonlinear algebraic equations represents a nonlinear multiobjective optimization problem. A Multiobjective Particle Swarm Optimizer (MOPSO) which is based on the Pareto optimality concept is developed and applied to maximize the above mentioned problem. Results show that MOPSO efficiently cope with the problem and finds multiple Pareto optimal solutions.

Estimation of indoor aerosol concentrations during outdoor release of radioactive materials is important since most people spend a lot of their time indoors. Whereas most of the mass transfer models are too complex to assess the indoor air quality by the first responders or radiation safety officers, this report uses the IAQX code to simulate the airborne concentration and inhalation exposure in a multi-zone building in the event of an outdoor radiological release. The way to establish a multi-zone well-mixed model is introduced first. The contamination in each zone is demonstrated by scenario and case study. Finally, the inhalation exposure will be derived from the air concentration versus time results. The analysis results show the major factors that influence the indoor air quality and health effects from an ambient release and the effectiveness of using IAQX to simulate complex multi-zone mass transfer model.

Within the framework of safety analysis the TÜV NORD SysTec GmbH & Co. KG has analyzed a steam generator tube rupture for a pressurized water reactor under realistic initial and boundary conditions. During the accident an undesirable backing of the coolant pressure by switching-on the safety injection pumps is avoided, which can endanger the coolant retention in the steam generators, because the shut off heads of the safety injection pumps are larger than the response pressure of the safety valves.