Volume 80, Issue 5

Steam condensation induced water hammer (CIWH) phenomena are investigated and new theoretical results are presented. We use the WAHA3 model based on two-phase flow six first-order partial differential equations that present one dimensional, surface averaged mass, momentum and energy balances. A second order accurate high-resolution shock-capturing numerical scheme was applied with different kind of limiters in the numerical calculations. The applied two-fluid model shows some similarities to RELAP5 which is widely used in the nuclear industry to simulate nuclear power plant accidents. This model was validated with different CIWH experiments which were performed in the PMK-2 facility, which is a full-pressure thermohydraulic model of the nuclear power plant of VVER-440/312 type in the Energy Research Center of the Hungarian Academy of Sciences and in the Rosa facility of the Japan Atomic Energy Agency. In our present study we show the first part of a planned large database which will give us the upper and lower flooding mass flow rates for various pipe geometries where CIWH can happen. Such a reliable database would be a great help for future reactor constructions and scheming.

During the last decades several new VVER-1000 units of different projects have been commissioned. At some of them similar initial fuel loadings were implemented. To determine experimentally neutron-physics characteristics of those units almost similar physical start-up test programs were performed and measurement equipment were also similar. The paper presents the analysis of some result of neutron-physics characteristics that were carried out during physics start-up of VVER-1000 units with similar initial fuel loading. The subject of analysis is an estimation of experimental uncertainties. An effort is made to consider individual test result as a single random realization of universal set. It is assumed the analysis of individual results spread helps to estimate general inherent uncertainty of experiment.

Station blackout anticipated transient without scram (SBO ATWS) is considered as loss of off-site and on-site power but no credit for automatic reactor trip. SBO ATWS causes reactor coolant pump (RCP) trip, loss of all main feedwater pumps and turbine trip, then the reactor coolant system (RCS) pressure rises rapidly due to loss of heat removal paths. The ASME Code Level C service limit criteria of 22.06 MPa (3200 psig) is assumed to be an unacceptable plant condition in SECY-83-293. The simulation is performed by TRACE which is a thermal-hydraulic code developed by U.S. NRC. Three different AFW flows are modeled to ensure the pressures will not be beyond the criteria. RCP seal-leakage is concerned as a SBLOCA due to loss of RCP seal-cooling. Four possible leakage flows are modeled to examine the reactor core water level and temperature variation.

The main objective of this study is to predict the thermal hydraulic behavior of nanofluids as the coolant in the fuel assembly of variable moderation high performance light water reactor (VMHWR). VMHWR is the new version of high performance light water reactor (HPLWR) conceptual design. Light water reactors at supercritical pressure (VMHWR, HPLWR), being currently under design, are the new generation of nuclear reactors. Water-based nanofluids containing various volume fractions of Al 2 O 3 nanoparticles are analyzed. The conservation equations and conduction heat transfer equation for fuel and clad have been derived and discretized by the finite volume method. The transfer of mass, momentum and energy between adjacent subchannels are split into diversion crossflow and turbulent mixing components. The governed non linear algebraic equations are solved by using analytical iteration methods. Finally the nanofluid analysis results are compared with the pure water results.

In the present investigation, a hybrid reactor system was designed by using 100 % Flibe, 90 % Flibe-10 % ThF 4 , 90 % Flibe-10 % UF 4 fluids, ENDF/B-VII, JEFF-3.1, JENDL-4.0, ROSFOND, BROND-2.2, CENDL-3.1 evaluated nuclear data libraries and Ferritic Steel, 9Cr2WVTa, V4Cr4Ti, SiC structural materials. The fluids were used in the liquid first wall, liquid second wall (blanket) and shield zones of a fusion–fission hybrid reactor system. The nuclear parameters of a fusion–fission hybrid reactor such as tritium breeding ratio ( TBR ), energy multiplication factor ( M ), heat deposition rate were computed in liquid first wall, blanket and shield zones. Three-dimensional nucleonic calculations were performed using the most recent version MCNPX-2.7.0 the Monte Carlo code.

This paper illustrates the neutronic and thermal hydraulic models that were implemented in the nuclear research reactor simulator based on LabVIEW. It also describes the system and transient analysis of the simulator that takes into consideration the temperature effects and poisoning. This simulator is designed to be a multi-purpose in which the operator could understand the effects of the input parameters on the reactor. A designer can study different solutions for virtual reactor accident scenarios. The main features of the simulator are the flexibility to design and maintain the interface and the ability to redesign and remodel the reactor core engine. The developed reactor simulator permits to acquire hands-on the experience of the physics and technology of nuclear reactors including reactivity control, thermodynamics, technology design and safety system design. This simulator can be easily customizable and upgradable and new opportunities for collaboration between academic groups could be conducted.

The self-assessments performed by thirty-eight countries on application of the Code of Conduct on the Safety of Research Reactors were analyzed and discussed. The results of this analysis were used to identify areas of satisfactory application of the Code and area needing improvements, and therefore require more attention worldwide. The results showed improvement in application of the Code provisions; ntformat otably in aging management, regulatory supervision, and consideration of human factors. However, there is a continuing need for further improvement in these areas, as well as in operational radiation protection, emergency preparedness and decommissioning planning. Additionally, increased attention needs to be given to periodic safety reviews, evaluation of site-specific hazards, and assessment of extreme external events. The results showed consistency with the feedback from other sources of information on generic safety issues for research reactors.

Nowadays, superconducting magnetic coils are used in some tokamaks such as EAST, KSTAR, JT-60, and T-15 to generate strong magnetic fields and also in ITER magnetic fields of about 13 tesla will be produced with the help of superconductors. The tokamak superconductors are exposed to the variety of radiations (neutron, ions beam, and gamma) from plasma nuclear reactions which will affect some of the superconductor properties. Therefore, study of the irradiation effects on the superconductor structure and properties are very crucial from technological and scientific point of view. One of the superconductor irradiation effects to be investigated under different conditions of energy and dosage is the potential resistance of the material used in tokamak reactor magnetic coils against activation by radiation. In this work, pellets of high T c Bi-based superconductors have been prepared and after measurement of parameters, a sample of pellet has been irradiated with 15 MeV protons using Karaj cyclotron facility. The sample's parameters have been measured again after irradiation treatment. X-ray diffraction patterns and SEM images of the sample before and after irradiation treatment have been studied.

Radiation damage of Iron by a gamma spectrum of a high flux isotope reactor is estimated using our JAIPU Monte Carlo code developed for the estimation of radiation damage by neutron and gamma radiation up to 10 MeV energy. The rate of displacement per atom is estimated to be ∼1.389E-10/sec at 30 cm radial distance in the HB-2 tube of the reactor. Production of displaced atoms is found to be an order of magnitude higher than the number of defects. Ratio of the rates of displacement per atoms by gammas to neutrons of the reactor deduced from calculations is found to be consistent with the earlier reported result.

UO 2 F 2 is an intermediate in conversion of UF 6 to uranium oxide or metal form and is a direct product of the reaction of UF 6 with moisture in the air. UF 6 will be continually hydrolyzed in the presence of water vapor. The amount of UF 6 and H 2 O can be controlled during the hydrolysis reaction, if the concentration of U in the solution product of the hydrolysis be determined. In this paper, we have designed a γ-ray densitometer with 137 Cs as gamma source with a medium energy of 662 keV for measuring the U concentration. In this research we have also studied the effect of two different activities of 10 mCi and 30 mCi on the measurement errors, and according to the results we chose the activity of 30 mCi with a relative error of 1.7 %. Finally, in order to investigate the effect of the temperature of the solution on the calibration of the apparatus, we calibrated our γ-ray densitometer for two temperatures of 15 and 80 °C and it was found that the temperature modification can be attained by software correction.

In nuclear criticality safety analysis it is essential to ascertain how various components of the nuclear system will perform under certain conditions they may be subjected to, particularly if the components of the system are likely to be affected by environmental factors such as temperature, radiation or material composition. It is therefore prudent that a sensitivity analysis is performed to determine and quantify the response of the output to variation in any of the input parameters. In a fissile system, the output parameter of importance is the k eff . Therefore, in attempting to prevent reactivity-induced accidents, it is important for the criticality safety analyst to have a quantified degree of response for the neutron multiplication factor to perturbation in a given input parameter. This article will present the results of the perturbation of the parameters that are important to nuclear criticality safety analysis and their respective correlation equations for deriving the sensitivity coefficients.

The steady-state one speed neutron transport equation without sources is studied in slab geometry using Henyey-Greenstein scattering function. The critical thicknesses are calculated for various c (mean number of secondary neutrons per collision) and t scattering parameters using U N method and obtained values are compared with P N method. All the numerical results for different collision scattering functions and t parameters of U N and P N methods are presented in the tables.

An increasing number of developing countries are showing interest to become the emerging countries to nuclear energy. Most of these countries lack human resources and adequate infrastructures to enter such a venture. The principle objective of activities of FBNR Group is to train human resources for the countries that at the present lack the necessary conditions, but aim at the future clean and safe nuclear energy through the fourth generation and INPRO compatible nuclear reactors. The preparation for the future nuclear energy is done through development of innovative nuclear reactor that meets the INPRO philosophies and criteria. These countries may or may not have decided as yet to utilize nuclear energy, but are interested to gain a strong educational foundation for their future. The research and development of a small innovative nuclear reactor FBNR is used as the instrument for learning. The young scientists will learn how to be innovative with the vision of INPRO philosophy and criteria.

A network database consists of a collection of records connected to one another through links. Presence of database network increases transmission speed and accuracy of information, which are the most important factors for good regulatory functions. Therefore, the data bank for Egyptian Nuclear and Radiological Regulatory Authority (NRRA) is proposed to be founded. Interface network group (ING) is the workers in the data bank. It is divided into three subgroups; scientific management (SMG), system analyst (SAG), and network management (NMG) subgroups. Their primary task is collect of data and information from different sectors in NRRA, and to transform these data to the organized database forms, which are available for all workers on the computer network. Meanwhile to face the very broad range of technical matters on which the regulatory related research activities, NRRA will need to rely on the ING. Accordingly, NRRA can address to the ING subsidiary tasks, which are not less important than their main duties, such as; scientific communications and demands, follow up plans and track problem solutions.