M. Varvayanni

M. Varvayanni
National Center for Scientific Research Demokritos | ncsr · Research Reactor Laboratory (RRL)

PhD

About

100
Publications
7,926
Reads
How we measure 'reads'
A 'read' is counted each time someone views a publication summary (such as the title, abstract, and list of authors), clicks on a figure, or views or downloads the full-text. Learn more
404
Citations
Introduction
M. Varvayanni (Melpomeni Varvagianni) currently works at the Research Reactor Laboratory (RRL), National Center for Scientific Research Demokritos. The last years she does research in Nuclear Technology. Their current project is 'Hybrid Reactor for Civil Naval Propulsion Associating a Particle Accelerator and a Fission reactor - academic research'.

Publications

Publications (100)
Conference Paper
This work is a preliminary study on the nuclear propulsion of a large commercial ship and more specifically on the integration of a small modular Accelerator Driven System (ADS) into a Suezmax tanker. The goal is to propose a sub-critical pressurized water reactor utilizing whenever possible standard components used by operating conventional PWRs....
Article
Full-text available
The safe introduction of Generation IV (Gen IV) reactor concepts into operation will require extensive testing of their components. This must be performed under neutronic conditions representative of those expected to prevail inside the new reactor cores when in operation. In a thermal Material Testing Reactor (MTR) such neutronic conditions can be...
Article
In the field of nuclear reactor analysis, multi-physics calculations that account for the bonded nature of the neutronic and thermal-hydraulic phenomena are of major importance for both reactor safety and design. So far in the context of Monte-Carlo neutronic analysis a kind of “serial” algorithm has been mainly used for coupling with thermal-hydra...
Article
In the field of computational reactor physics, Monte-Carlo methodology is extensively used in the analysis of static problems while the transient behavior of the reactor core is mostly analyzed using deterministic algorithms. However, deterministic algorithms make use of various approximations mainly in the geometric and energetic domain that may i...
Article
ANET (Advanced Neutronics with Evolution and Thermal hydraulic feedback) is an under development Monte Carlo code for simulating both GEN II/III reactors as well as innovative nuclear reactor designs, based on the high energy physics code GEANT3.21 of CERN. ANET is built through continuous GEANT3.21 applicability amplifications, comprising the simu...
Conference Paper
Full-text available
Nowadays, Monte-Carlo criticality analysis is performed utilizing the power iteration that calculates the fundamental eigenpair of the steady-state/k-eigenvalue form of the neutron transport equation. Whereas this method guarantees the convergence to the fundamental eigenmode, very often the convergence is slow. Consequently, it is of high interest...
Article
Off-site early emergency management actions due to a potential release from the Greek Research Reactor-1 are assessed with the aid of a modern decision support tool. To this end, a hypothetical loss-of-coolant accident has been assumed under two distinct ventilation schemes of the reactor building. Five, site-typical, meteorological scenarios have...
Conference Paper
Τhe dynamic analysis of the reactor behavior is crucial regarding design safety issues. Especially for a nuclear reactor core conversion, the study of the behavior of the “new” core configuration under various transient circumstances, normal or accidental, is a major requirement. So far, the transient analysis of the reactor core is performed with...
Article
The neutron flux trap effect was experimentally studied in the subcritical assembly of the Atomic and Nuclear Physics Laboratory of the Aristotle University of Thessaloniki, using delayed gamma neutron activation analysis. Measurements were taken within the natural uranium fuel grid, in vertical levels symmetrical to the Am–Be neutron source, befor...
Conference Paper
In the field of reactor physics the transient behavior of the reactor core is mainly analyzed using deterministic algorithms. However, deterministic algorithms make use of various approximations mainly in geometric and energetic domain which may induce inaccuracy. On the other hand Monte-Carlo analysis, which generally does not require significant...
Conference Paper
ANET is being developed targeting to a multiple capabilities code which can inherently, dynamically and accurately simulate GEN II/III reactors as well as Accelerator Driven Systems (ADSs). ANET is oriented towards an open-source pure Monte-Carlo transient code with Thermal-Hydraulics (T-H) feedback. It incorporates the treatment of all types of pa...
Conference Paper
Full-text available
In the field of nuclear reactor analysis, multiphysics calculations that account for the bonded nature of the neutronic and thermal-hydraulic phenomena are of major importance for both reactor safety and design. So far in the context of Monte-Carlo neutronic analysis, the Gauss-Seidel iterative scheme, in a version for individual single-physics sol...
Conference Paper
In the upcoming decades huge quantities of energy will be needed worldwide, especially in the form of electricity generated in an environmentally friendly manner. Nuclear power is the most environmentally propitious way of producing electricity at a large scale ensuring, at the same time, stability of supply and considerable reduction of the carbon...
Conference Paper
Instrumentation and control (I&C) is a key element to research reactor (RR) safety responsible for control actions involving startup, power regulation and shutdown as well as abnormal condition management. Unfortunately, system obsolescence that is frequently accompanied by spare part unavailability may result in operational problems and extended r...
Conference Paper
Full-text available
The neutron flux trap effect was experimentally studied in the sub-critical assembly of the Atomic and Nuclear Physics Laboratory of the Aristotle University of Thessaloniki, using delayed gamma neutron activation analysis (DGNAA). Measurements were taken within the fuel grid, in vertical levels symmetrical to the Am-Be neutron source, before and a...
Article
Full-text available
Research reactors are used for many applications: material testing; radioisotope production; beam-line applications for material research; nuclear transmutation doping; neutron activation analysis; neutron radiography experiments; fuel waste management; and other neutron and nuclear material related quantities, features, and research areas of inter...
Conference Paper
The presence of fast neutron spectra in new reactor concepts (such as Gas Cooled Fast Reactor, new generation Sodium Cooled Fast Reactor, Lead Fast Reactor, Accelerator Driven System and nuclear Fusion Reactors) is expected to induce a strong impact on the contained materials, including structural materials (e.g. steels), nuclear fuels, neutron ref...
Conference Paper
Full-text available
In this work the new Monte Carlo code ANET is tested on criticality calculations. ANET is developed based on the high energy physics code GEANT of CERN and aims at progressively satisfying several requirements regarding both simulations of GEN II/III reactors, as well as of innovative nuclear reactor designs such as the Accelerator Driven Systems (...
Article
In an operating nuclear reactor core, various physical phenomena of different nature are interrelated. Multi-physics calculations that account for the interrelated nature of the neutronic and thermal–hydraulic phenomena are of major importance in reactor safety and design and as a result a special effort is developed within the nuclear engineering...
Conference Paper
Full-text available
Within the context of an operating nuclear reactor core, multi-physics calculations that account for the bonded nature of the neutronic and thermal-hydraulic phenomena are of major importance in reactor safety and design. In this work, the ongoing development of a tool for neutronic/thermal-hydraulic coupled calculations is presented. OpenMC, a Mon...
Conference Paper
The upcoming development of the IV generation fast reactor systems requires the examination of the behaviour exhibited by the structural and fuel materials under the irradiation conditions prevailing in these reactors. The lack of operating fast reactors and, hence, the lack of capability to perform experiments in such irradiation environments can...
Conference Paper
Full-text available
Commercial nuclear power plants produce an excess of plutonium during their operation that could be utilized as mixed-oxide (MOX) uranium and plutonium fuel in existing or advanced nuclear reactors. This creates the requirement of reliably simulating such MOX fuelled reactor cores and, at the same time, the necessity to validate neutronic codes and...
Conference Paper
ANET is a new stochastic neutronics code which is being developed based on the high energy physics code GEANT of CERN, for simulating both GEN II/III reactors as well as innovative nuclear reactor designs. ANET has already been successfully tested with respect to criticality computations, while in this work its reliability in computing neutron flue...
Article
Full-text available
The presence of fast neutron spectra in new reactors is expected to induce a strong impact on the contained materials, including structural materials, nuclear fuels, neutron reflecting materials, and tritium breeding materials. Therefore, introduction of these reactors into operation will require extensive testing of their components, which must be...
Conference Paper
The present study discusses the source term derivation and dose result calculation for a hypothetical accident sequence in the Greek Research Reactor – 1 (GRR-1). A loss-of-coolant accident (LOCA) has been selected as a credible accident sequence. The source term derivation has been based on the GRR-1 confinement performance where the inventory has...
Article
Full-text available
The estimated source term, dose results and radiological consequences of selected accident sequences in the Greek Research Reactor – 1 are presented and discussed. A systematic approach has been adopted to perform the necessary calculations in accordance with the latest computational developments and IAEA recommendations. Loss-of-coolant, reactivit...
Article
The reliability and verification of numerical solutions derived from neutronic codes and the use of nuclear data libraries is a very important issue in nuclear technology. To this purpose, computational benchmarks based on well-defined problems with a complete set of input and a unique solution, are often used. The OECD/NEA VENUS-2 is a widely used...
Article
In this paper the relative performance of different simulation approaches is examined, focusing on the neutron fluence rate distribution in a nuclear reactor core. The main scope of the work is to benchmark and validate the neutronics code systems utilized in the Greek Research Reactor (GRR-1) for a highdensity Low Enriched Uranium (LEU) core of co...
Conference Paper
The Monte Carlo code ANET is under development based on the high energy physics code GEANT3.21. It can simulate ADS systems, accounting for the high-energy proton beam interaction with spallation target, neutron trajectories, and the reaction rate distribution in the core. At the same time, it is structured to account for core materials modificatio...
Article
Nuclear energy industry asks for an optimized exploitation of available natural resources and a safe operation of reactors. A closed fuel cycle requires the mass of fissile material depleted in a reactor to be equal to or less than the fissile mass produced in the same or in other reactors. In this work, a simple closed cycle scheme is investigated...
Article
the “candle” reactors, etc. These reactor designs introduce computational nuclear technology problems the solution of which necessitates a new, global and dynamic computational approach of the system. A continuous feedback procedure must be established between the main inter-related parameters of the system such as the chemical, physical and isotop...
Article
One important parameter in the design and the analysis of a nuclear reactor core is the reactivity worth of the control rods, i.e. their efficiency to absorb excess reactivity. The control rod worth is affected by parameters such as the fuel burnup in the rod vicinity, the Xe concentration in the core, the operational time of the rod and its positi...
Article
Τhe capability of achieving optimum utilization of the deterministic neutronic codes is very important, since, although elaborate tools, they are still widely used for nuclear reactor core analyses, due to specific advantages that they present compared to Monte Carlo codes. The user of a deterministic neutronic code system has to make some signific...
Article
The reliable estimate of gamma heating of materials in a research reactor core, including irradiated samples or core compartments, is recognized as directly related to reactor safety. A review of related reports indicates that the capability of accurate assessment of gamma heating of materials in nuclear reactor cores remains a matter of investigat...
Article
Assessment of the gamma heating deposited on samples irradiated in material testing reactors is basic safety issue. The GHRRC (Gamma Heating in Research Reactor Cores) code developed in NCSR “Demokritos” is used here to estimate the relative importance of the mechanisms contributing to the total gamma heating of the irradiated material, for a core...
Article
Knowledge of the efficiency of a control rod to absorb excess reactivity in a nuclear reactor, i.e. knowledge of its reactivity worth, is very important from many points of view. These include the analysis and the assessment of the shutdown margin of new core configurations (upgrade, conversion, refueling, etc.) as well as several operational needs...
Article
When depleted control rods are planned to be used in new core configurations, their worth has to be accurately predicted in order to deduce key design and safety parameters such as the available shutdown margin. In this work a methodology is suggested for the derivation of the distributed absorbing capacity of a depleted rod, useful in the case tha...
Article
In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal–hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the “Accelerator part” of the installation, i.e. the computation of the spect...
Article
The design or modification and in general the analysis and control of nuclear reactors require complex calculations, which are carried out by numerical codes including neutronic and thermal-hydraulic components. Among the neutronic codes, the deterministic ones which solve the neutron transport/diffusion equation simulate the reactor core by dividi...
Article
A basic safety requirement for a research reactor is the reliable estimation of the gamma heating of samples irradiated in the reactor core. A three-dimensional numerical code of gamma heating using a point kernel parameterization is developed. The heating due to γ-rays, produced from U235 fission and from (n, γ) reactions with the core materials i...
Article
Α simple analytical model is presented which indicates that the ratio of heating power densities of two different materials, irradiated under the same conditions inside a reactor core, can be estimated from material properties only. The developed approximate method allows simplifying the measurement technique of a number of samples from different m...
Article
A Monte Carlo simulation of the Greek Research Reactor was carried out using MCNP-4C2 code and continuous energy cross-section data from ENDF/B-VI library. A detailed model of the reactor core was employed including standard and control fuel assemblies, reflectors and irradiation devices. The model predicted neutron flux distributions within the co...
Conference Paper
For a five-year transitional period the Greek Research Reactor (GRR-1) was operating with a mixed core, containing both Low Enrichment (LEU) and High Enrichment (HEU) Uranium MTR- type fuel assemblies. The neutronic study of the GRR-1 conversion to LEU has been performed using a code system comprising the core-analysis code CITATION-LDI2 and the ce...
Conference Paper
The thermal-hydraulic analysis performed for the needs of the conversion of the open pool 5MW Greek Research Reactor (GRR-1) to a pure Low Enrichment (LEU) configuration is presented. The methodology was based on a complete set of neutronic calculations performed for the new core configuration, in compliance with pre-defined Operation Limiting Cond...
Article
The upper limits for the rate of release of radionuclides into the atmosphere, i.e., the "derived release limits," are calculated for the Greek Research Reactor (GRR-1) in order to determine possible operational schemes compatible with the effective dose limits for the general population. GRR-1 is located at the northwestern foot of Hymettos Mounta...
Article
Experiments in research reactors often pose the necessity of void insertion in several core positions. Also, sub-cooled boiling may introduce fluctuating quantities of bubbles. The reactivity change and its effect on the safety are important matters that must be assessed. In the present work, the reactivity induced by void insertion in various loca...
Article
Full-text available
The transport and diffusion code system of the Greek Research Reactor (GRR-1) consists of the computational packages SCAMPI and CITATION-LDI2. The methodology as well as the results obtained for the GRR-1 with respect to criticality and neutron flux calculations have been analytically described in previous reports [1, 2]. In the present report, the...
Article
Full-text available
PREFACE The transport diffusion code system of the Greek Research Reactor (GRR-1), consisting of the computational packages SCAMPI and CITATION-LDI2 has been analytically described in a previous report [1]. The application of the code system for criticality calculations is also presented in [1]. In the present report, the system is applied for the...
Article
Full-text available
The neutronics code system of the Greek Research Reactor (GRR-1) is based on the well-documented computational packages SCAMPI and CITATION-LDI2. The SCAMPI code system (collection of codes for manipulating multigroup cross section libraries) is developed in Unix environment and consists primarily of modules derived from the SCALE and AMPX code sys...
Article
The land-surface parameterization scheme BATS was incorporated into a diagnostic atmospheric modeling system, using an unstructured prismatic grid, to investigate the capability of a diagnostic tool, designed for emergency response, to assign the land-cover impact on the boundary layer structure and pollutant dispersion. In this framework, two appl...
Article
The advantageous utilization of triangular prismatic grid for flow simulation over irregular geometries is widely recognized. Such grid is here utilized to diagnose the atmospheric conditions and pollutant dispersion over complex inhomogeneous surfaces. One case with an extremely complex surface is resolved with regular prismatic grid. A second cas...
Article
An estimation of SO2 dispersion over complex terrain is attempted, using the diagnostic wind field generated from limited meteorological data. This effort aims to face the realistic requirement of emergency-response in case of industrial plants settled in poorly instrumented areas. The combination of three numerical procedures is utilized: (1) deta...
Article
A simulation of the thermally driven flow field prevailing over Attiki peninsula in Greece was made, by using the DEMOKRITOS Transport Code System applicable to terrains of high complexity. Two particular days included in the MEDCAPHOT experimental campaign over Attiki, i.e. 14 and 15 September 1994, dominated by mesoscale flow systems and characte...
Article
The numerical study of mesoscale wind field and dispersion over terrains belonging to large mountain structures (e.g. Alps) poses a number of difficulties. Narrow valleys and steep slopes can considerably determine the temperature and pressure gradients, thus affecting the mesoscale systems in a quite complex manner. The adequate description of the...
Article
A reliable method for meteorological pre-processing over complex topographies has been developed and implemented into the DETRACT code system, which is integrated into the RODOS decision support system. Shortcomings with respect to topography aspects, limited number of available stations and existing methods which depend on pure mathematics, have b...
Article
Full-text available
In a terrain of high complexity, such as mountain regions or local scale building structures, the topography is characterized by irregularities covering a wide range of dimensions and/or sizes. The irregularities can influence considerably the pressure gradients and induce mechanical effects on the wind systems. The air/ground interaction in terms...
Article
Geographical and meteorological data input tools used with the Demokritos atmospheric transport dispersion code system are specifically designed for dispersion computations over complex terrain. The code initialisation depends mainly on the topography simulator (DELTA), which provides detailed information with re