Пахниц А.В.

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Count articles: 8

Author articles

Nuclear reactor corium properties obtained at IGR research reactor

Authors: Скаков Мажын Канапинович, Мухамедов Н.Е., Пахниц А.В., Дерявко И.И.

Keywords: ИГР, расплав

Organisation: Филиал «Институт атомной энергии» РГП НЯЦ РК

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In the paper for the first time thermophysical properties (specific heat capacity, thermal diffusivity and thermal conductivity) of natural corium of a fast nuclear power reactor were determined in the temperature range from room one up to ~400 °С. The obtained data is oriented at use in temperature field calculations when modeling the processes of corium melt retention in fast nuclear reactor vessel.

In-reactor experiment for the testing of a fast-reactor pile in the contidions of loss-of-coolant accident

Authors: Пахниц А.В., Сураев А.С., Сулейменов Н.А.

Keywords: ИГР, внутриреакторный эксперимент, твэл

Organisation: Филиал «Институт атомной энергии» РГП НЯЦ РК

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In this paper, the behavior of a model nuclear fuel element of a fast neutron reactor was studied in order to study the effect of blanket zones on the distribution of fuel melt throughout the reactor core. A computational model of the experimental device has been developed, thermophysical calculations have been carried out to substantiate the safety of the in-reactor experiment, a diagram of reactor power has been determined, the implementation of which reached the initial stage of melting zones with high fuel enrichment. An in-reactor experiment on a IGR RRC with an experimental device equipped with a model fast-neutron fuel element was prepared and carried out.

Experimental studies in substantiation of sodium cooled fast reactors safety

Authors: Скаков М.К., Батырбеков Э.Г., Витюк В.А., Пахниц А.В., Вурим А.Д., Бакланов В.В., Камияма К., Мацуба К.

Keywords: ИГР, реактор на быстрых нейтронах, тяжелая авария, экспериментальная программа

Organisation: Филиал «Институт атомной энергии» РГП НЯЦ РК

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A large volume of scientific and research activities is currently carried out in substantiation of new projects on nuclear reactor safety, in which high emphasis is placed on “beyond-design basis” accidents, in spite of the fact that applied technical solutions reduce the probability of such accidents occurrence. To study parameters of severe accidents and develop measures to mitigate consequences an experimental base is necessary, which enables emergency situation modelling maximally close to real conditions. The National Nuclear Center (NNC) – the leading research organization in the atomic sphere in the Republic of Kazakhstan – possesses such an experimental base, providing scientific and technical support for peaceful uses of nuclear power, involving studies oriented towards the improvement of nuclear reactor safety [1]. The article presents the concept and general results of researches carried out on NNC’s experimental base in substantiation of nuclear reactors safety, realized in collaboration with JAEA, which is a key foreign partner in this area, as well as current plans on new experimental program implementation.

Researches of IGR characteristics with low enriched fuel

Authors: Пахниц А.В., Иркимбеков Р.А., Жагипарова Л.К., Вурим А.Д., Котов В.М., Бекмагамбетова Б.Е., Байгожина А.А., Мурзагалиева А.А.

Keywords: ИГР, уран, поток нейтронов, низкообогащенное топливо, высокообогащенное топливо

Organisation: Филиал «Институт атомной энергии» РГП НЯЦ РК

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The paper represents research results of possible conversion of IGR reactor from high-enriched uranium fuel to lowenriched uranium fuel. It is shown that on examined configurations in a «hot» reactor; thermal neutron flow in CEC, is significantly reduced, permissible temperature excess of water is possible on an output of a CEC fixed ampoule, which indicates necessity to continue of reactor core packaging for keeping its characteristics.

Modeling of IVG.1M reactor

Authors: Пахниц А.В., Иркимбеков Р.А., Жагипарова Л.К., Вурим А.Д., Котов В.М., Витюк Г.А., Байгожина А.А., Мурзагалиева А.А.

Keywords: ИВГ 1.M, топливо, низкообогащенное топливо

Organisation: Филиал «Институт атомной энергии» РГП НЯЦ РК

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The paper presents the models of IVG.1M reactor for neutron and physical and thermohydraulic computations. The models are used for computation of characteristics of IVG.1M reactor and advanced reactor with low-enriched uranium fuel. 

Main outcomes and future plan of the EAGLE project

Authors: Пахниц А.В., Гайдайчук В.А., Витюк В.А., Вурим А.Д., Зуев В.А., Колодешников А.А., Васильев Ю.С., Кубо Ш., Сато И., Камияма К., Матсуба К., Тоёока Ж., Тобита Е., Кониши К., Котаке С., Эндо Х., Кояма К.

Keywords: EAGLE, быстрые нейтроны, безопасность реактора, безопасные режимы работы

Organisation: Филиал «Институт атомной энергии» РГП НЯЦ РК

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Main outcomes and contributions to exclude the recriticality issue in sodium-cooled fast reactors (SFRs) accomplished through the EAGLE project, which is an experimental program conducted under the cooperation of Japan and the Republic of Kazakhstan since 1998, are summarized. Future plan, i.e., EAGLE3, to enhance the safety of SFRs is also described.

R&D and experimental programs for mitigating severe accidents consequences in GENIV SFRS and in the ASTRID technology demonstrator

Authors: Батырбеков Э.Г., Пахниц А.В., Витюк В.А., Вурим А.Д., Серр Ф., Пайо Ф., Сюто К., Тротиньон Л., Камияма К., Мацуба К., Кубо Ш., Тобита Е., Като А., Тоёока Д.

Keywords: кориум, внереакторные эксперименты, прототип технологического реактора ASTRID

Organisation: РГП «Национальный ядерный центр Республики Казахстан»

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The ASTRID technological prototype, designed by CEA with industrial partners, has high level of requirements, in particular for a robust safety demonstration. Despite high level of severe accident prevention with an innovative core, complementary safety devices are added to mitigate the severe accident consequences. A large R&D program is under way in cooperation with JAEA and NNC-RK. To extend programs carried-out in France in CABRI and SCARABEE reactors, and in IGR in Kazakhstan, the feasibility of the SAIGA program in IGR reactor was performed, and a new PLINIUS-2 platform for out-of-pile corium experiments is under development.

Computational validation of FA test modes of advanced reactor

Authors: Пахниц А.В., Жагипарова Л.К.

Keywords: ТВС, корпус ампулы ЭУ, экспериментальное устройство (ЭУ)

Organisation: Филиал «Институт атомной энергии» РГП НЯЦ РК

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The paper presents analysis of safe testing of the experimental device (ED). The calculation data of ED preheating are presented, the sequence of research start-up with fuel melting and movement of the melt throughout the discharge duct into the lower melt trap filled with sodium is analyzed. It is shown that the structural integrity of FA jacket and ED ampule casing will not be failed if accident is progressed.