title: Ð Ð°Ð·ÑÐ°Ð±Ð¾ÑÐºÐ° Ð¼ÑÐ»ÑÑÐ¸ÑÐ¸Ð·Ð¸ÑÐµÑÐºÐ¾Ð³Ð¾ ÑÐµÑÐ°ÑÐµÐ»Ñ Ð´Ð»Ñ Ð°Ð²Ð°ÑÐ¸Ð¸ LOFA ÑÐµÐ°ÐºÑÐ¾ÑÐ° ÐÐÐ­Ð -1000 Ñ Ð¸ÑÐ¿Ð¾Ð»ÑÐ·Ð¾Ð²Ð°Ð½Ð¸ÐµÐ¼ ÑÐ¸ÑÐ»ÐµÐ½Ð½ÑÑ Ð¼ÐµÑÐ¾Ð´Ð¾Ð²
title: Ð²ÑÐ¿ÑÑÐºÐ½Ð°Ñ ÐºÐ²Ð°Ð»Ð¸ÑÐ¸ÐºÐ°ÑÐ¸Ð¾Ð½Ð½Ð°Ñ ÑÐ°Ð±Ð¾ÑÐ° Ð¼Ð°Ð³Ð¸ÑÑÑÐ°
titleType: Subtitle
title: Developing a multiphysics solver of LOFA accident for VVER1000 reactor using numerical methods
Creators: name: ÐÑÐ·ÑÐ°Ð½, ÐÐµÑÐ¼ÐµÑ
Publisher: name: Ð¡Ð°Ð½ÐºÑ-ÐÐµÑÐµÑÐ±ÑÑÐ³ÑÐºÐ¸Ð¹ Ð¿Ð¾Ð»Ð¸ÑÐµÑÐ½Ð¸ÑÐµÑÐºÐ¸Ð¹ ÑÐ½Ð¸Ð²ÐµÑÑÐ¸ÑÐµÑ ÐÐµÑÑÐ° ÐÐµÐ»Ð¸ÐºÐ¾Ð³Ð¾
Year: 2023
Subjects: Ð¯Ð´ÐµÑÐ½ÑÐµ ÑÐµÐ°ÐºÑÐ¾ÑÑ Ð²Ð¾Ð´Ð¾-Ð²Ð¾Ð´ÑÐ½ÑÐµ
Ð°ÐºÑÐ¸Ð²Ð½Ð°Ñ Ð·Ð¾Ð½Ð°
ÐºÐ¸Ð½ÐµÑÐ¸ÑÐµÑÐºÐ¸Ð¹ Ð¿Ð¾Ð´Ñ Ð¾Ð´
Ð·Ð°Ð²Ð¸ÑÐ¸Ð¼Ð¾ÑÑÑ Ð¾Ñ Ð²ÑÐµÐ¼ÐµÐ½Ð¸
ÐºÐ°Ð½Ð°Ð» Ñ Ð¾Ð´Ð½Ð¸Ð¼ Ð½Ð°Ð³ÑÐµÐ²Ð¾Ð¼
Ð¼ÑÐ»ÑÑÐ¸ÑÐ¸Ð·Ð¸ÑÐµÑÐºÐ¸Ð¹ ÑÐµÑÐ°ÑÐµÐ»Ñ
Ð°Ð²Ð°ÑÐ¸Ñ Ñ Ð¿Ð¾ÑÐµÑÐµÐ¹ Ð¿Ð¾ÑÐ¾ÐºÐ°
core
kinetic approach
time dependence
single heated channel
multiphysics solver
flux loss accident

description: ÐÐ²Ð°ÑÐ¸Ð¸ Ñ Ð¿Ð¾ÑÐµÑÐµÐ¹ Ð¿Ð¾ÑÐ¾ÐºÐ° (LOFA) - Ð¾Ð´Ð½Ð° Ð¸Ð· Ð¾ÑÐ½Ð¾Ð²Ð½ÑÑ Ð°Ð²Ð°ÑÐ¸Ð¹ Ð½Ð° ÑÐ´ÐµÑÐ½ÑÑ ÑÐµÐ°ÐºÑÐ¾ÑÐ°Ñ, ÐºÐ¾ÑÐ¾ÑÐ°Ñ Ð¼Ð¾Ð¶ÐµÑ Ð¿ÑÐ¾Ð¸Ð·Ð¾Ð¹ÑÐ¸ Ð¿Ð¾ Ð½ÐµÑÐºÐ¾Ð»ÑÐºÐ¸Ð¼ Ð¿ÑÐ¸ÑÐ¸Ð½Ð°Ð¼. ÐÐ½Ð°Ð»Ð¸Ð· Ð¿Ð¾Ð²ÐµÐ´ÐµÐ½Ð¸Ñ ÑÐ´ÐµÑÐ½ÑÑ ÑÐµÐ°ÐºÑÐ¾ÑÐ¾Ð² Ð² ÑÐ»ÑÑÐ°Ðµ ÑÐ°ÐºÐ¸Ñ ÑÐ¸ÑÑÐ°ÑÐ¸Ð¹ ÑÐ²Ð»ÑÐµÑÑÑ Ð¾Ð´Ð½Ð¾Ð¹ Ð¸Ð· ÑÑÐ½Ð´Ð°Ð¼ÐµÐ½ÑÐ°Ð»ÑÐ½ÑÑ ÑÐµÐ¼ Ð¸ÑÑÐ»ÐµÐ´Ð¾Ð²Ð°Ð½Ð¸Ð¹ Ð² Ð¾Ð±Ð»Ð°ÑÑÐ¸ ÑÐ´ÐµÑÐ½Ð¾Ð¹ ÑÐ½ÐµÑÐ³ÐµÑÐ¸ÐºÐ¸. Ð ÑÐ°Ð¼ÐºÐ°Ñ Ð´Ð°Ð½Ð½Ð¾Ð¹ Ð´Ð¸ÑÑÐµÑÑÐ°ÑÐ¸Ð¸ ÑÐ°Ð·ÑÐ°Ð±Ð¾ÑÐ°Ð½ Ð¼ÑÐ»ÑÑÐ¸ÑÐ¸Ð·Ð¸ÑÐµÑÐºÐ¸Ð¹ ÑÐµÑÐ°ÑÐµÐ»Ñ LOFA Ð´Ð»Ñ ÑÐµÐ°ÐºÑÐ¾ÑÐ° ÐÐÐ­Ð -1000, Ð² ÐºÐ¾ÑÐ¾ÑÐ¾Ð¼ Ð½ÐµÐ¹ÑÑÐ¾Ð½Ð½ÑÐµ Ð¸ ÑÐµÐ¿Ð»Ð¾Ð³Ð¸Ð´ÑÐ°Ð²Ð»Ð¸ÑÐµÑÐºÐ¸Ðµ ÑÐ°ÑÑÐµÑÑ Ð²ÑÐ¿Ð¾Ð»Ð½ÑÑÑÑÑ Ð² Ð¿ÐµÑÐµÑÐ¾Ð´Ð½Ð¾Ð¼ ÑÐµÐ¶Ð¸Ð¼Ðµ. ÐÐµÐ¹ÑÑÐ¾Ð½Ð½ÑÐµ ÑÐ°ÑÑÐµÑÑ Ð²ÑÐ¿Ð¾Ð»Ð½ÑÑÑÑÑ Ñ Ð¿Ð¾Ð¼Ð¾ÑÑÑ ÑÑÐ°Ð²Ð½ÐµÐ½Ð¸Ð¹ ÑÐ¾ÑÐµÑÐ½Ð¾Ð¹ ÐºÐ¸Ð½ÐµÑÐ¸ÐºÐ¸ (PKE), ÐºÐ¾ÑÐ¾ÑÑÐµ ÑÐ²Ð»ÑÑÑÑÑ ÑÐ¿ÑÐ¾ÑÐµÐ½Ð½ÑÐ¼Ð¸ Ð²ÐµÑÑÐ¸ÑÐ¼Ð¸ ÑÑÐ°Ð²Ð½ÐµÐ½Ð¸Ð¹ Ð´Ð¸Ð½Ð°Ð¼Ð¸ÐºÐ¸ ÑÐµÐ°ÐºÑÐ¾ÑÐ°. Ð ÑÐ°Ð¼ÐºÐ°Ñ Ð¿Ð¾Ð´ÑÐ¾Ð´Ð° PKE ÑÐµÐ°Ð»Ð¸Ð·Ð¾Ð²Ð°Ð½Ñ ÑÐµÑÑÑÐµ ÑÐ°Ð·Ð»Ð¸ÑÐ½ÑÑ ÑÐ¸ÑÐ»ÐµÐ½Ð½ÑÑ Ð¼ÐµÑÐ¾Ð´Ð°, Ð°Ð½Ð°Ð»Ð¸Ð· ÐºÐ¾ÑÐ¾ÑÑÑ Ð¿ÑÐ¾Ð²Ð¾Ð´Ð¸ÑÑÑ Ð¿ÑÐ¸ Ð½ÐµÑÐºÐ¾Ð»ÑÐºÐ¸Ñ Ð¿Ð°ÑÐ°Ð¼ÐµÑÑÐ°Ñ Ð´Ð»Ñ Ð¾Ð¿ÑÐµÐ´ÐµÐ»ÐµÐ½Ð¸Ñ Ð½Ð°Ð¸Ð±Ð¾Ð»ÐµÐµ ÑÑÐ°Ð±Ð¸Ð»ÑÐ½Ð¾Ð³Ð¾ Ð¸ ÑÐ¾ÑÐ½Ð¾Ð³Ð¾ Ð¼ÐµÑÐ¾Ð´Ð°. Ð Ð½ÐµÐ¹ÑÑÐ¾Ð½Ð½Ð¾Ð¼ ÑÐµÑÐ°ÑÐµÐ»Ðµ ÑÐ°ÑÑÑÐ¸ÑÑÐ²Ð°ÑÑÑÑ Ð¿Ð»Ð¾ÑÐ½Ð¾ÑÑÐ¸ Ð½ÐµÐ¹ÑÑÐ¾Ð½Ð¾Ð² Ð¸ Ð¿ÑÐµÐºÑÑÑÐ¾ÑÐ¾Ð², Ð° ÑÐ°ÐºÐ¶Ðµ Ð¼Ð¾ÑÐ½Ð¾ÑÑÑ ÑÐµÐ°ÐºÑÐ¾ÑÐ°. Ð ÑÐµÑÐ°ÑÐµÐ»Ðµ Ð¿Ð¾ ÑÐµÑÐ¼Ð¾Ð³Ð¸Ð´ÑÐ°Ð²Ð»Ð¸ÐºÐµ Ð´Ð»Ñ Ð¼Ð¾Ð´ÐµÐ»Ð¸ÑÐ¾Ð²Ð°Ð½Ð¸Ñ ÑÐµÐ°ÐºÑÐ¾ÑÐ° ÑÐ°ÑÑÐ¼Ð°ÑÑÐ¸Ð²Ð°ÐµÑÑÑ Ð¾Ð´Ð¸Ð½Ð¾ÑÐ½ÑÐ¹ Ð½Ð°Ð³ÑÐµÑÑÐ¹ ÐºÐ°Ð½Ð°Ð» (SHC). Ð¥Ð°ÑÐ°ÐºÑÐµÑÐ¸ÑÑÐ¸ÐºÐ¸ Ð¿Ð¾ÑÐ¾ÐºÐ° ÑÐµÐ¿Ð»Ð¾Ð½Ð¾ÑÐ¸ÑÐµÐ»Ñ Ð¸ Ð·Ð½Ð°ÑÐµÐ½Ð¸Ñ ÑÐµÐ¼Ð¿ÐµÑÐ°ÑÑÑÑ ÑÐ°ÑÑÑÐ¸ÑÑÐ²Ð°ÑÑÑÑ Ñ Ð¿Ð¾Ð¼Ð¾ÑÑÑ ÑÐ¸ÑÐ»ÐµÐ½Ð½Ð¾ Ð´Ð¸ÑÐºÑÐµÑÐ¸Ð·Ð¸ÑÐ¾Ð²Ð°Ð½Ð½ÑÑ Ð²ÐµÑÑÐ¸Ð¹ ÑÑÐ°Ð²Ð½ÐµÐ½Ð¸Ð¹ ÑÐ¾ÑÑÐ°Ð½ÐµÐ½Ð¸Ñ Ð¸ ÑÑÐ°Ð²Ð½ÐµÐ½Ð¸Ð¹ ÑÐµÐ¿Ð»Ð¾Ð¿ÐµÑÐµÐ´Ð°ÑÐ¸ ÑÐ¾Ð¾ÑÐ²ÐµÑÑÑÐ²ÐµÐ½Ð½Ð¾. Ð ÐµÐ°ÐºÑÐ¸Ð²Ð½Ð¾ÑÑÑ Ð¸ Ð¼Ð°ÑÑÐ¾Ð²ÑÐ¹ ÑÐ°ÑÑÐ¾Ð´ Ð¿Ð¾Ð´Ð°ÑÑÑÑ Ð² ÑÐµÑÐ°ÑÐµÐ»Ñ Ð´Ð»Ñ ÐºÐ°Ð¶Ð´Ð¾Ð³Ð¾ Ð²ÑÐµÐ¼ÐµÐ½Ð½Ð¾Ð³Ð¾ ÑÐ°Ð³Ð°. ÐÐ¾Ð»ÑÑÐµÐ½Ð½ÑÐµ ÑÐµÐ·ÑÐ»ÑÑÐ°ÑÑ ÑÑÐ°Ð²Ð½Ð¸Ð²Ð°ÑÑÑÑ Ñ ÑÑÐ°Ð»Ð¾Ð½Ð½ÑÐ¼Ð¸ Ð·Ð½Ð°ÑÐµÐ½Ð¸ÑÐ¼Ð¸. Ð¡ÑÐ°Ð²Ð½ÐµÐ½Ð¸Ðµ Ð¿Ð¾Ð´ÑÐ²ÐµÑÐ¶Ð´Ð°ÐµÑ ÑÐ¾ÑÐ½Ð¾ÑÑÑ ÑÐ°Ð·ÑÐ°Ð±Ð¾ÑÐ°Ð½Ð½Ð¾Ð³Ð¾ ÑÐµÑÐ°ÑÐµÐ»Ñ. Ð Ð°Ð±Ð¾ÑÐ° Ð¼Ð¾Ð¶ÐµÑ Ð±ÑÑÑ Ð¿ÑÐ¾Ð´Ð¾Ð»Ð¶ÐµÐ½Ð° Ð¿ÑÑÐµÐ¼ Ð²Ð½ÐµÐ´ÑÐµÐ½Ð¸Ñ Ð°Ð²ÑÐ¾Ð¼Ð°ÑÐ¸ÑÐµÑÐºÐ¾Ð³Ð¾ Ð²ÑÑÐ¸ÑÐ»ÐµÐ½Ð¸Ñ ÑÐµÐ°ÐºÑÐ¸Ð²Ð½Ð¾ÑÑÐ¸ Ð¼ÐµÐ¶Ð´Ñ Ð¿Ð¾ÑÐ»ÐµÐ´ÑÑÑÐ¸Ð¼Ð¸ Ð²ÑÐµÐ¼ÐµÐ½Ð½ÑÐ¼Ð¸ ÑÐ°Ð³Ð°Ð¼Ð¸.
description: Loss of flow accidents (LOFA) are one of the major accidents for nuclear reactors that can occur for several reasons. Analysing the behaviour of nuclear reactors in case of such sitation has been one of the fundamental reserch subject in nuclear energy field. In the scope of this thesis, a multi-physic solver of LOFA for VVER-1000 reactor is developed where neutronic and thermal hydraulics calculations are performed in transient mode. Neutronic calculations are made with point kinetics equations (PKE) which are simplified versions of reactor dynamics equations. Four different numerical methods are implemented to PKE approach, analysis are performed under several parameters to determine most stable and accurate method. Neutron and precursor densities are calculated as well as reactor power in the neutronic solver. In thermal hydraulics solver, single heated channel (SHC) is considered to model reactor. Coolant flow properties and temperature values are calculated with numerically discretised versions of conservation equations and heat transfer equations respectively. Reactivity and mass flow rate are provided to the solver for each time step. Obtained results are compared with reference values. Comparisons prove the accuracy of the developed solver. This work can be developed further by implementing automatic reactivity calculator between subsequent time steps.