Some aspects of Monte Carlo method application to nuclear reactors analysis

The MCU project is described. Its aim is to guarantee the high precision of calculation of reactor parameters. The project includes creation of Monte-Carlo codes, their verification and validation, development of constant base and methods of application Monte-Carlo calculations for increasing the precision of engineer methods for predicting of the main reactor parameters.     

Fuel burn-up in nuclear reactors

The paper describes the method of calculating fuel burn-up in nuclear reactors, taking into account the capture and multiplication of neutrons while slowing down. In the calculations, account is taken of the burn-up of U²³⁵ and the build-up and burn-up of Np²³⁹, Pu²³⁹, Pu²⁴⁰, Pu²⁴¹ and of the fission fragments.     

Thermal shock in nuclear reactors

Most types of liquid cooled reactors undergo rapid changes in coolant temperature during abnormal operating conditions, which cause components adjacent to the coolant to undergo rapid surface temperature changes. The differential expansion between surface and bulk material caused by repeated thermal shocks can induce fatigue or cyclic creep damage. In thick walled components individual thermal shocks may cause rapid fracture from pre-existing defects. This paper reviews the reported experimental and analytical work defining the temperature, stress and resultant damage in reactor structural components subject to thermal shock.     

Selected safety aspects of containments for nuclear power plants with WWER-440 reactors

Considerable attention has been and continues to be focused on the design and operational features that prevent the release of radioactive materials to the environment for a spectrum of accidents for the two classes of WWER-440 reactors: the older 230 model and the more recently designed 213 models.This paper, based on published and unpublished information, aims to clarify the perceptions of the Russian WWER-440 models 230 and 213 Nuclear Power Plant containment system designs and their relevance to selected aspects of accident mitigation. It should be noted that these are unclearly and often negatively perceived, primarily because of a lack of reliable information and a poorly assembled experimental database. Conflicting statements have been made regarding the nature and the features of the plant's containment system. The paper presents a brief outline of the design of both WWER-440 models with respect to their confinement functions. Selected safety-related aspects of the accident localization systems are discussed, and the recognized shortcomings and safety merits are pointed out. The older 230 units experience high leak rates and are designed to withstand medium-size pipe breaks. The possible implications for safety are pointed out in the paper. The on going studies that concentrate on improving the system are highlighted. Some of the proposed modifications of the system, which would significantly decrease risks associated with accidents that are beyond the original design basis, are discussed. The design of the newer 213 model differs in many aspects. It incorporates the simple and original application of passive natural processes to limit the large-break loss-of-coolant accident post accident pressure. Other features of this containment design, such as complicated geometry, dependence on several mechanical devices and interlocks, and insufficient experimental evidence, lead to doubts concerning the operation of this containment under accident conditions. For the newer 213 model, current work is devoted mainly to safety assessment and verification of the containment design. Some information concerning the on-going work is provided in the paper.     

The neutronic potential of nuclear power for long-term radioactivity risk reduction

An analysis to evaluate the comparative neutronic transmutation potential of different nuclear power system (standard or advanced fission reactors and accelerator driven hybrids) is presented. The analysis is based on an evaluation of neutronic constraints for the reduction of both long-lived fission product toxicity and fuel waste toxicity integrated over the life of the nuclide families, taking into account the overall neutron balance of the systems being considered.     

Introducing and validating a new method for coupling neutronic and thermal-hydraulic calculations

In this study, a new and innovative method is introduced for analyzing neutronic and thermal-hydraulic calculation. For this aim, VVR-S research reactor was selected, and the calculation procedure was performed for it. WIMS, CITATION and COBRA-EN codes were used for investigation. Calculation model consists of two sub-models: neutronic and thermo-hydraulic. The neutronic model uses WIMS and CITATION codes for neutronic simulation of the reactor core and calculating flux and power distribution over it. WIMS code simulates the fuel assemblies and CITATION models the core. The thermal-hydraulic model uses COBRA-EN code for performing the relative calculation. In this study, FORTRAN 90 program is used for linking two sub-models and performing the calculation. The proposed procedure is performed for VVR-S analysis and finally, the obtained results are compared with the experimental results that show good agreement with it.     

Automation of the modeling and some neutronic calculations of the Syrian miniature neutron source reactors

An automated software, BMAC, for modeling and performing the neutronics calculations of MNSRs and similar reactors (TRIGAs) has been developed. Calculation of initial excess reactivity, flux and power distributions, and all other neutronic parameters of the reactor, full core representation, can be made automatically using a 3-D model, by coupling WIMSD-4 and CITATION codes, in a very quick and simple way. No preliminary CITATION input file is needed. All required data are read from an external input file simply prepared. Accurate results for the parameters of the reactor, in the framework of Diffusion Theory, can be obtained.     

Trial decommissioning of nuclear reactors

Translated from Atomnaya Énergiya, Vol. 69, No. 3, pp. 157–160, September, 1990.     

Electrical simulation of nuclear reactors

The application of electrical simulation methods to nuclear reactor calculations considerably shortens the time required for the computational work. In this review the advantages and disadvantages of simulators are discussed and an example is given of the simulation of the reactor isotopic composition with the help of the simulator MN-7. The effectiveness of the electrical simulation method for the investigation of non-stationary reactor processes is shown by an example of design calculations carried out for an automatic power regulator for a reactor. It is shown that it is possible to simulate nonstationary processes occurring in a power reactor, taking into account the temperature coefficients of the reactivity; other applications of simulators in reactor design are indicated.     

Fuel elements of nuclear reactors

On the basis of foreign reports presented at the Second International Conference on the Peaceful Uses of Atomic Energy, Geneva, 1958, the characteristics of construction of fuel elements (FE) and basic data relating to them for a series of reactors are given. Problems on the selection of fuel and construction materials as well as the technology of preparing FE for various types of nuclear reactor are examined.     

Use of burnable poisons in nuclear reactors

This paper considers the nature and purpose of the use of burnable poisons in nuclear reactors. The points described are: possible ways of distributing the poisons in the active zone, constructional and engineering problems in adding poisons to nuclear reactors, and some peculiarities of the physics design of reactors with burnable neutron poisons.     

Spatial instability analysis in pressurized water reactors

A simple methodology has been developed to assess the spatial dynamic behavior of large PWRs against xenon spatial instability in different modes. Method of analysis aims to analyze xenon dynamic behavior against anticipated reactivity perturbations. Reactivity perturbations in different modes have been evaluated based on reactivity device movements as well as localized thermal variations in the core. Effect of individual core design and operating parameters on xenon spatial instability has been studied. Behavior of spatial stability index (SI) with core size is investigated. Based on SI-core size curve, a threshold core size has been determined beyond which a PWR core tends to become spatially unstable. Methodology has been used to assess the spatial xenon dynamic behavior of different modes of oscillations in VVER1000 and AP1000 reactor cores.     

Small nuclear power plants based on ship-propulsion nuclear reactors

NIKIéT. Translated from Atomnaya énergiya, Vol. 77, No. 6, pp. 407–414, December, 1994.