Wyniki wyszukiwania - Biblioteka Nauki

1

CFD modeling of passive autocatalytic recombiners*

100%

Orszulik M. , Fic A. , Bury T.

|

nr 2

347-353

EN

This study deals with numerical modeling of passive autocatalytic hydrogen recombiners (PARs). Such devices are installed within containments of many nuclear reactors in order to remove hydrogen and convert it to steam. The main purpose of this work is to develop a numerical model of passive autocatalytic recombiner (PAR) using the commercial computational fluid dynamics (CFD) software ANSYS-FLUENT and tuning the model using experimental results. The REKO 3 experiment was used for this purpose. Experiment was made in the Institute for Safety Research and Reactor Technology in Julich (Germany). It has been performed for different hydrogen concentrations, different flow rates, the presence of steam, and different initial temperatures of the inlet mixture. The model of this experimental recombiner was elaborated within the framework of this work. The influence of mesh, gas thermal conductivity coefficient, mass diffusivity coefficients, and turbulence model was investigated. The best results with a good agreement with REKO 3 data were received for k-ɛ model of turbulence, gas thermal conductivity dependent on the temperature and mass diffusivity coefficients taken from CHEMKIN program. The validated model of the PAR was next implemented into simple two-dimensional simulations of hydrogen behavior within a subcompartment of a containment building.

2

Evaluation of passive autocatalytic recombiners operation efficiency by means of the lumped parameter approach*

100%

Bury T.

|

nr 2

339-345

EN

The problem of hydrogen behavior in containment buildings of nuclear reactors belongs to thermal-hydraulic area. Taking into account the size of systems under consideration and, first of all, safety issues, such type of analyses cannot be done by means of full-scale experiments. Therefore, mathematical modeling and numerical simulations are widely used for these purposes. A lumped parameter approach based code HEPCAL has been elaborated in the Institute of Thermal Technology of the Silesian University of Technology for simulations of pressurized water reactor containment transient response. The VVER-440/213 and European pressurised water reactor (EPR) reactors containments are the subjects of analysis within the framework of this paper. Simulations have been realized for the loss-of-coolant accident scenarios with emergency core cooling system failure. These scenarios include core overheating and hydrogen generation. Passive autocatalytic recombiners installed for removal of hydrogen has been taken into account. The operational efficiency of the hydrogen removal system has been evaluated by comparing with an actual hydrogen concentration and flammability limit. This limit has been determined for the three-component mixture of air, steam and hydrogen. Some problems related to the lumped parameter approach application have been also identified.

3

A Novel Control-rod Drive Mechanism via Electromagnetic Levitation in MNSR

100%

Divandari M. , Hashemi-Tilehnoee M. , Khaleghi M. , Hosseinkhah M.

Nukleonika

|

2014

|

tom 59

|

nr 2

73-79

EN

In this paper, an electromagnetic levitation system was used with a synchronous motor to navigate the control rod of a small-type research reactor. The result from this prototype magnetic levitation system was in agreement with simulation results. The control system was programmed in MATLAB through open-loop system, closed-loop with state feedback and closed-loop with state feedback integral tracking. The final control system showed the highest performance with a low positioning error. Our results showed that the developed control system has the potential to be used as a reliable actuator in nuclear reactors to satisfy higher performance and safety.

4

Application of genetic algorithms in optimization of SFR nuclear reactor design

100%

Żurkowski W. , Sawicki P. , Kubiński W. , Darnowski P.

Nukleonika

|

2021

|

tom Vol. 66, No. 4

139--145

EN

This work presents a demonstrational application of genetic algorithms (GAs) to solve sample optimization problems in the generation IV nuclear reactor core design. The new software was developed implementing novel GAs, and it was applied to show their capabilities by presenting an example solution of two selected problems to check whether GAs can be used successfully in reactor engineering as an optimization tool. The 3600 MWth oxide core, which was based on the OECD/NEA sodium-cooled fast reactor (SFR) benchmark, was used a reference design [1]. The first problem was the optimization of the fuel isotopic inventory in terms of minimizing the volume share of long-lived actinides, while maximizing the effective neutron multiplication factor. The second task was the optimization of the boron shield distribution around the reactor core to minimize the sodium void reactivity effect (SVRE). Neutron transport and fuel depletion simulations were performed using Monte Carlo neutron transport code SERPENT2. The simulation resulted in an optimized fuel mixture composition for the selected parameters, which demonstrates the functionality of the algorithm. The results show the efficiency and universality of GAs in multidimensional optimization problems in nuclear engineering.

5

A novel control-rod drive mechanism via electromagnetic levitation in MNSR

100%

Divandari M. , Hashemi-Tilehnoee M. , Khaleghi M. , Hosseinkhah M.

Nukleonika

|

2014

|

tom Vol. 59, No. 2

73--79

EN

In this paper, an electromagnetic levitation system was used with a synchronous motor to navigate the control rod of a small-type research reactor. The result from this prototype magnetic levitation system was in agreement with simulation results. The control system was programmed in MATLAB through open-loop system, closed-loop with state feedback and closed-loop with state feedback integral tracking. The final control system showed the highest performance with a low positioning error. Our results showed that the developed control system has the potential to be used as a reliable actuator in nuclear reactors to satisfy higher performance and safety.

6

CFD modeling of passive autocatalytic recombiners

100%

Orszulik M. , Fic A. , Bury T.

Nukleonika

|

2015

|

tom Vol. 60, No. 2

347--353

EN

This study deals with numerical modeling of passive autocatalytic hydrogen recombiners (PARs). Such devices are installed within containments of many nuclear reactors in order to remove hydrogen and convert it to steam. The main purpose of this work is to develop a numerical model of passive autocatalytic recombiner (PAR) using the commercial computational fluid dynamics (CFD) software ANSYS-FLUENT and tuning the model using experimental results. The REKO 3 experiment was used for this purpose. Experiment was made in the Institute for Safety Research and Reactor Technology in Julich (Germany). It has been performed for different hydrogen concentrations, different flow rates, the presence of steam, and different initial temperatures of the inlet mixture. The model of this experimental recombiner was elaborated within the framework of this work. The influence of mesh, gas thermal conductivity coefficient, mass diffusivity coefficients, and turbulence model was investigated. The best results with a good agreement with REKO 3 data were received for k-ε model of turbulence, gas thermal conductivity dependent on the temperature and mass diffusivity coefficients taken from CHEMKIN program. The validated model of the PAR was next implemented into simple two-dimensional simulations of hydrogen behavior within a subcompartment of a containment building.

7

Finding optimal of the Egyptian second nuclear reactor core patterns using genetic algorithm

100%

El-Sayed Wahed M. , Ibrahim W. Z.

|

tom Vol. 55, No. 3

331-338

EN

The second Egyptian research reactor ETRR-2 went critical on the 27th of November 1997. The National Center of Nuclear Safety and Radiation Control (NCNSRC) has the responsibility of the evaluation and the assessment of the safety of this reactor. Fuel management reloads for Egypt’s second research reactor have been carried out according to the fuel management scheme suggested by the reactor designer (INVAP). The start up core consists of three different fuel types, while the equilibrium core has only one fuel type called standard fuel. The fuel management scheme consists in considering the core as being partitioned into eight zones. Each zone will correspond to a chain of fuel movements. In each fuel cycle two of these chains will be involved, in which eight fuel elements will be moved, from them two spent fuel elements will be extracted and two fresh fuel elements will be inserted in the core. In this paper we solve a model as a one big nonlinear multi objective discrete optimization problem using genetic algorithm. Results are compared with INVAP values.

8

Evaluation of passive autocatalytic recombiners operation efficiency by means of the lumped parameter approach

100%

Bury T.

Nukleonika

|

2015

|

tom Vol. 60, No. 2

339--345

EN

The problem of hydrogen behavior in containment buildings of nuclear reactors belongs to thermal- -hydraulic area. Taking into account the size of systems under consideration and, first of all, safety issues, such type of analyses cannot be done by means of full-scale experiments. Therefore, mathematical modeling and numerical simulations are widely used for these purposes. A lumped parameter approach based code HEPCAL has been elaborated in the Institute of Thermal Technology of the Silesian University of Technology for simulations of pressurized water reactor containment transient response. The VVER-440/213 and European pressurised water reactor (EPR) reactors containments are the subjects of analysis within the framework of this paper. Simulations have been realized for the loss-of-coolant accident scenarios with emergency core cooling system failure. These scenarios include core overheating and hydrogen generation. Passive autocatalytic recombiners installed for removal of hydrogen has been taken into account. The operational effiiency of the hydrogen removal system has been evaluated by comparing with an actual hydrogen concentration and flammability limit. This limit has been determined for the three-component mixture of air, steam and hydrogen. Some problems related to the lumped parameter approach application have been also identified.

9

Numerical Solution of Fractional Neutron Point Kinetics Model in Nuclear Reactor

88%

Nowak T. K. , Duzinkiewicz K. , Piotrowski R.

Archives of Control Sciences

|

2014

|

tom Vol. 24, no. 2

129--154

EN

This paper presents results concerning solutions of the fractional neutron point kinetics model for a nuclear reactor. Proposed model consists of a bilinear system of fractional and ordinary differential equations. Three methods to solve the model are presented and compared. The first one entails application of discrete Grünwald-Letnikov definition of the fractional derivative in the model. Second involves building an analog scheme in the FOMCON Toolbox in MATLAB environment. Third is the method proposed by Edwards. The impact of selected parameters on the model’s response was examined. The results for typical input were discussed and compared.

10

Computations of fuel management in MARIA reactor with highly poisoned beryllium matrix

88%

Andrzejewski K. J. , Kulikowska T. A. , Marcinkowska Z. E.

|

tom Vol. 53, No. 4

173-179

EN

The high-flux research reactor MARIA has been operated in Poland since 1975. Its core consists of loop type fuel channels placed in a beryllium block matrix. Irradiation of beryllium by neutrons results in a build-up of 6Li, and 3He isotopes with large thermal neutron absorption cross-sections. In addition, tritium is formed and decays into 3He which complicates the transmutation chains. Thus, the fuel management of the reactor depends on the beryllium poisoning. The isotopic transmutations in beryllium have to be computed in parallel to the fuel depletion. In this paper a comparison of the measured and computed results is given during: reactor operation in the period up to January 2004, modernization break in the reactor operation from January 2004 to February 2005 and reactor operation from February 2005. The measured and computed effects comprise: reactivity effects due to the fuel burn-up and beryllium poisoning by 6Li, and 3He.

11

Hydrogen embrittlement and delayed failure

88%

Szklarska-Śmiałowska Z.

Ochrona przed Korozją

|

2005

|

tom nr 9

290-294

EN

Examples of hydrogen embrittlement (HE) at high and low temperatures are given. The hydrogen induced cracking of Alloy 600 in a condition simulating that existing in a pressurized water reactor environment (350oC) and delayed fracture of an alloy at ambient temperatures are reported. The HE of bolts used in the automotive industry at ambient temperatures is discussed. In particular, a note has been taken to correlate the content of hydrogen in metals with their propensity to failure.

PL

Podano przykłady kruchości wodorowej w wysokich i niskich temperaturach. Opisano pękanie wodorowe stopu Alloy 600 w warunkach symulu- jących środowisko ciśnieniowego wodnego reaktora jądrowego (350oC), a także opóźnione pękanie tego stopu w temperaturze otoczenia. Omówiono kruchość wodorową stalowych śrub używanych w przemyśle motoryzacyjnym. Zwrócono uwagę na zależność między zawartością wodoru w metalach a ich skłonnością do pękania.

12

Development of an embedded FPGA-based data acquisition system dedicated to zero power reactor noise experiments

75%

Arkani M. , Khalafi H. , Vosoughi N.

Metrology and Measurement Systems

|

2014

|

tom Vol. 21, nr 3

433--446

EN

An embedded time interval data acquisition system (DAS) is developed for zero power reactor (ZPR) noise experiments. The system is capable of measuring the correlation or probability distribution of a random process. The design is totally implemented on a single Field Programmable Gate Array (FPGA). The architecture is tested on different FPGA platforms with different speed grades and hardware resources. Generic experimental values for time resolution and inter-event dead time of the system are 2.22 ns and 6.67 ns respectively. The DAS can record around 48-bit x 790 kS/s utilizing its built-in fast memory. The system can measure very long time intervals due to its 48-bit timing structure design. As the architecture can work on a typical FPGA, this is a low cost experimental tool and needs little time to be established. In addition, revisions are easily possible through its reprogramming capability. The performance of the system is checked and verified experimentally.

13

A study of the effects of changing burn-up and gap gaseous compound on the gap convection coefficient (in a hot fuel pin) in VVER-1000 reactor

75%

Rahgoshay M. , Rahmani Y.

Nukleonika

|

2007

|

tom Vol. 52, nr 3

93-95

EN

In this article we worked on the result and process of calculation of the gap heat transfer coefficient for a hot fuel pin in accordance with burn-up changes in the VVER-1000 reactor at the Bushehr nuclear power plant (Iran). With regard to the fact that in calculating the fuel gap heat transfer coefficient, various parameters are effective and the need for designing a model is being felt, therefore, in this article we used Ross and Stoute gap model to study impacts of different effective parameters such as thermal expansion and gaseous fission products on the hgap change rate. Over time and with changes in fuel burn-up some gaseous fission products such as xenon, argon and krypton gases are released to the gas mixture in the gap, which originally contained helium. In this study, the composition of gaseous elements in the gap volume during different times of reactor operation was found using ORIGEN code [3]. Considering that the thermal conduction of these gases is lower than that of helium, and by using the Ross and Stoute gap model, we find first that the changes in gaseous compounds in the gap reduce the values of gap thermal conductivity coefficient, but considering thermal expansion (due to burn-up alterations) of fuel and clad resulting in the reduction of gap thickness we find that the gap heat transfer coefficient will augment in a broad range of burn-up changes. These changes result in a higher rate of gap thickness reduction than the low rate of decrease of heat conduction coefficient of the gas in the gap during burn-up. Once these changes have been defined, we can proceed with the analysis of the results of calculations based on the Ross and Stoute model and compare the results obtained with the experimental results for a hot fuel pin as presented in the final safety analysis report of the VVER-1000 reactor at Bushehr [2]. It is noteworthy that the results of accomplished calculations based on the Ross and Stoute model correspond well with the existing experimental results for this reactor.

14

Zastosowanie modeli odwrotnych do diagnozowania stanów utraty chłodzenia w reaktorach jądrowych typu wrzącego

75%

Bednarski M. , Cholewa W. , Frid W.

Diagnostyka

|

2002

|

tom Vol. 27

105-110

PL

Celem referatu jest pokazanie przykładu zastosowania odwrotnych modeli diagnostycznych. Zapobieganie wypadkom w elektrowniach jądrowych, w szczególności przypadkom uszkodzenia rdzenia, gdzie ryzyko uwolnienia produktów radioaktywnych jest największe, jest sprawą priorytetową dla bezpieczeństwa. W celu analizy potencjalnie możliwych wypadków, jak również w celu ich zapobiegania oraz zarządzania nimi, stworzono wiele programów symulacyjnych oraz systemów wspomagających podejmowanie decyzji (Computerized Decision Support Systems - CDSS) przez operatorów. Bazują one na metodach deterministycznych i probablistycznych. W przypadku reaktorów jądrowych rozwój szybkich narzędzi symulacyjnych daje możliwość zastosowania metod diagnostycznych bazujących na przykładach. Przedstawione tutaj lokalne modele odwrotne są przykładem takiego właśnie podejścia.

EN

Aim of this paper is to present example of application inverse diagnostic models. Accident prevention in nuclear plants, in particular in case of the core damage, where the risk of release radioactive products is the highest, is the priority cause for safety. In order to analyze potentially possible accidents and also to prevent and to manage them, a lot of simulation codes and Computerized Decision Support Systems (CDSS) was implemented. They base on deterministic and probabilistic methods. In case of nuclear reactors, recent progress of very fast simulation tools opens possibility of applying case-based diagnostic methods. The method described in the paper, which uses local inverse models, is an example of such approach.

15

Energetyka jądrowa w Polsce i na świecie

75%

Skolik K.

|

tom nr 1

55--60

PL

Obecnie (stan na 8.01.2020) na świecie pracuje 448 reaktorów jądrowych, pokrywając ok. 10% zapotrzebowania na energię elektryczną. Najwięcej (96) reaktorów znajduje się w USA, na kolejnych miejscach są Francja (58) i Chiny (48). Rozwój energetyki jądrowej na całym świecie jest koniecznością i warunkiem pozwalającym na ograniczenie emisji gazów cieplarnianych w świetle obecnego kryzysu klimatycznego. Niestety, trendy nie są optymistyczne, a kraje europejskie wręcz odwracają się od energetyki jądrowej. Sztandarowym przykładem są Niemcy, gdzie zamknięto już większość elektrowni jądrowych, a pozostałe 6 bloków ma zostać wyłączonych do końca 2022 r.

16

Analysis of the BEAVRS PWR benchmark using SCALE and PARCS

75%

Darnowski P. , Pawluczyk M.

Nukleonika

|

2019

|

tom Vol. 64, No. 3

87--96

EN

This paper presents an analysis of the Benchmark for Evaluation And Validation of Reactor Simulations (BEAVRS) performed using SCALE 6.1.2 and PARCS 3.2 computer codes. The benchmark specifi cation contains a detailed design, operational data and measurements for a real 4-loop Westinghouse pressurized water reactor (PWR). The lattice physics simulations were prepared using TRITON depletion sequence and NEWT neutron transport solver (SCALE package). The 238-neutron group library based on evaluated nuclear data fi le – ENDF/B-VII nuclear data libraries was applied. A set of branch and burnup calculations was prepared, and group constants in the form of PMAXS fi les were generated with GenPMAXS. The full-core models were prepared using the PARCS nodal-diffusion core simulator. The PMAXS libraries were used with PARCS to investigate the core operation. The hot zero power measurement data, including control rod worths and critical boron concentrations, were compared using simulations, and satisfactory results were achieved. The fi rst fuel cycle was simulated, and acceptable agreement with boron letdown curve and measurements were obtained. Finally, conclusions and recommendations for future research were presented.

17

Wyznaczenie charakterystyk eksploatacyjnych siłowni turbinowej o mocy 400 MW z reaktorem wysokotemperaturowym w zmiennych warunkach pracy

75%

Drosińska-Komor M. , Szewczuk-Krypa N. , Głuch J.

Przegląd Mechaniczny

|

2017

|

tom nr 6

45--47

PL

W artykule przedstawiono charakterystyki sprawnościowe dla siłowni turbinowej o mocy 400 MW zasilanej przez reaktor wysokotemperaturowy. Przedstawione charakterystyki stworzone zostały dla zmiennych warunków pracy opisanego obiegu. Obliczenia numeryczne symulujące działanie siłowni wykonano w programie DIAGAR. Jako zmienne warunki ruchu obiegu jądrowego przyjęto zmianę ciśnienia w skraplaczu. Na tej podstawie określono zmianę sprawności oraz jednostkowego zużycia ciepła omawianego układu.

EN

In the article performance characteristic for the 400 MW nuclear steam power plant are presented. Characteristics were created for steam turbine variable working conditions. Numerical calculations that simulate system operation in DIAGAR have been prepared. As a variable working conditions of nuclear steam power plant changes of condenser pressure were adopted.

18

Comparative analysis of thermodynamic cycles of selected nuclear ship power plants with hightemperature helium- cooled nuclear reactor

75%

Szewczuk-Krypa N. , Grzymkowska A. , Głuch J.

|

tom S 1

218--224

EN

This paper presents a comparative analysis of thermodynamic cycles of two ship power plant systems with a hightemperature helium- cooled nuclear reactor. The first of them is a gas system with recuperator , in which classical gas chamber is substituted for a HTGR reactor (High Temperature Gas-cooled Reactor) . The second of the considered cycles is a combined gas-steam system where working medium flux from gas turbine outlet is directed into waste heat boiler and its heat is utilized for production of superheated steam to drive steam turbine. Preliminary calculations of the combine cycles showed that it is necessary to expand the system by adding to its steam part an inter-stage overheat for secondary steam, owing to that a required degree of steam dryness at outlet from the turbine can be reached, ensuring its correct operational conditions. The analyzed power systems were compared to each other with regard to efficiency of their thermodynamic cycles. Also, efficiency of particular cycles were subjected to optimization in respect to such parameters as : working gas temperature at outlet from reactor in gas system as well as steam pressure at outlet from waste heat boiler and partition pressure in steam part of combined system. Advantages of nuclear power plants compared with the classical power systems dominating currently in sea transport were also discussed.

19

Ocena szczelności prętów paliwowych w reaktorach lekkowodnych na podstawie detekcji radioizotopów gazów szlachetnych

75%

Chotkowski M. , Siporska A. , Rogulski Z. , Szydłowski J. , Czerwiński A.

Przemysł Chemiczny

|

2014

|

tom T. 93, nr 12

2259-2262

PL

Przedstawiono dane dotyczące możliwości wykorzystania detekcji radioizotopów gazów szlachetnych do oceny stanu prętów paliwowych lekkowodnych reaktorów jądrowych. Ze względów pomiarowych, a także wydajności generowania poszczególnych gazowych produktów rozszczepienia jąder uranu najczęściej wykorzystywanymi radioizotopami z grupy helowców są izotopy ksenonu oraz kryptonu (85mKr, 133Xe, 135Xe). W celu detekcji tych gazów wykorzystuje się układy scyntylacyjne lub półprzewodnikowe (HPGe) w połączeniu z kolumnami adsorpcyjnymi wypełnionymi sorbentem węglowym. Przeprowadzone badania porównawcze materiałów węglowych oraz jednego zeolitycznego wykazały, że różną się one w sposób zasadniczy w stopniu adsorpcji 133Xe.

EN

133Xe was adsorbed from its mixts. with N2, Ar and air on activated C (grain size 1-3 mm or 3 mm) and on a moi. sieve 3 Å at 0, -78 and -196°C under lab. conditions (gas flow 0,1-0,4 L/min) and then desorbed in N2 at 40-150°C. The adsorption degree on C was 99-100% in the whole temp, range. On the mol. sieve, the adsorption degree decreased from 100% at -196°C down to 0% at 0°C.

20

Ocena szerokości szczeliny dylatacyjnej w blokach grafitowych reaktora Maria

75%

Zając B. , Olszewski G. , Piotrowska N.

Badania Nieniszczące i Diagnostyka

|

2019

|

tom nr 3

42--45

PL

Przedmiotem pracy była ocena szerokości szczeliny dylatacyjnej w blokach grafitowych reaktora MARIA NCBJ po ich długotrwałej eksploatacji w reaktorze. Nominalna szerokość szczeliny dylatacyjnej bloku grafitowego przed eksploatacją w reaktorze wynosiła ok. 30 mm. W czasie długotrwałej pracy reaktora, w wyniku interakcji pomiędzy szybkimi neutronami a grafitem następuje powolne pęcznienie grafitu powodujące porowatość i pogorszenie jego własności fizycznych. Efektem pęcznienia grafitu pod wpływem oddziaływania z neutronami następuje zmniejszenie się szczeliny dylatacyjnej. Badania szerokości szczeliny dylatacyjnej przeprowadzono metodą radiograficzną. Ze względu na trudności z zastosowaniem radiografii klasycznej (bloki grafitowe były napromieniowane) do badań wykorzystano płyty obrazowe i aparaturę do badań radiografią cyfrową. Badania bloków były przeprowadzone w tzw. Komorach Gorących nad reaktorem MARIA.

EN

The paper focuses on the assessment of the width of the expansion provision within the graphite blocks placed in MARIA nuclear reactor of the National Centre for Nuclear Research in Świerk, after their long-lasting operation in the reactor. The nominal width of the expansion provision before placing in the reactor was approximately 30mm. As a result of an interaction between the fast neutrons and the graphite within the reactor, there occurs an incremental swelling of the graphite, causing its porosity and deterioration of its physical properties. This results in narrowing of the expansion provision. Tests aimed at assessing the expansion provision were conducted with the use of radiographic testing. Due to the difficulties arising from applying classical radiographic testing (irradtiation of the graphite blocks),digital radiography instruments were employed and the testing was only possible in so called Hot Cells placed above the reactor.