Thermohydraulic and neutron-physical calculation of the VVER-300 nuclear reactor

Contents

INTRODUCTION

1. DETERMINATION OF CORE GEOMETRICAL PARAMETERS

2. THERMAL HYDRAULIC CALCULATION

2.1 Distribution of linear thermal load by EP height per FA

2.2 Calculation of heat flux density distribution by EP height

2.3 Calculation of average coolant flow rate in the channel

2.4 Calculation of enthalpy and relative enthalpy distribution by EP height

2.5 Calculation of heat carrier temperature distribution by EP height

2.6 Calculation of TVEL shell temperature distribution by EP height

2.7 Calculation of temperature distribution in the fuel pellet center by EP height

2.8 Hydraulic calculation of EP

3. FOUR-GROUP NEUTRON-PHYSICAL CALCULATION

3.1 Determination of geometric characteristics

3.2 Concentration of substances in design areas

3.3 Calculation of the first neutron group

3.4 Calculation of the second neutron group

3.5 Calculation of the third neutron group

3.6 Calculation of the fourth neutron group

3.7. Total multiplication factor

LIST OF SOURCES USED

APPENDIX A

APPENDIX B

APPENDIX B

Introduction

Currently, the Russian Federation is one of the leaders in the development of the nuclear industry. Currently, 38 power units with a total installed capacity of 30,300 MW are operating at 11 nuclear power plants in Russia, including:

- 21 reactors with water under pressure (13 VBER1000, 3 VBER1200 and 5 VBER440);

- 13 channel boiling reactors (10 RBMK1000 and 3 EGP6);

- two fast neutron reactors (BN600 and BN800);

- two KLT40C type PATES reactors with an electric capacity of 35 MW each.

More than half of these units have already developed their resource and require replacement.

The need for medium-capacity nuclear power plants exists in countries and regions with poorly developed network infrastructure, in remote areas where the delivery of organic fuel is difficult. The use of equipment and design solutions used in projects of reactor plants of greater capacity allows to unify the material base necessary for construction.

The VVER-300 power reactor is a heterogeneous, hull type, operating on thermal neutrons. Water simultaneously acts as heat carrier and retarder. The primary coolant is water of high purity under pressure of 16.5 MPa, with the addition of boric acid. The reactor itself is a cylindrical vertical housing with an elliptical bottom. Core and other internals are located inside the housing.

The reactor uses two circulation loops, in which water circulates due to the operation of two main circulation pumps. The thermal capacity of the prototype reactor is 850 MW.

The purpose of the course design is to evaluate the thermophysical reliability of the reactor core. Determine distribution of coolant flow through the reactor channels, pressure along the circulation circuit, temperature in the reactor elements, as well as parameters of the primary circuit equipment.

Thermohydraulic calculation of gases

Using the values obtained during the geometric calculation, we determine the change in the temperature of the coolant passing through the core from the moment of its entry to the moment of exit from it. To do this, we consider one elementary cell, taking into account the fact that in the core of heterogeneous reactors of a cylindrical shape, there is a cosine wave law for distributing energy release over height.

Hydraulic calculation of EP

The purpose of the hydraulic calculation is to determine the pressure losses in the channels and the power costs for pumping the coolant. Pressure drops are calculated in each i-th section, then added up. Losses consist of friction losses, local resistances, leveling component and coolant acceleration losses.

Conclusion

In the process of performing this work, geometric, thermal-hydraulic and neutron-physical calculations were made.

In the geometric calculation, the volume (Vaz) and diameter (Daz) of EP were determined to be 14.17 m3 and 2.26 m, respectively. According to the specified parameters of the reactor power and specific energy release, the amount of FAs equal to 85 pcs was calculated.

All distributions were made in the core height calculation for medium and maximum load channels. As a result, during thermohydraulic calculation, distributions of linear thermal load, heat flow density, average flow rate of coolant in the channel, enthalpy, relative enthalpy, coolant temperature, TVEL shell temperature, as well as temperature in the center of the fuel pellet were obtained.

The result of the thermohydraulic calculation showed that the pressure losses are 8.63 * 104Pa and have little effect on the thermophysical properties of the coolant and its pumping through the core.

In the course of thermohydraulic calculation, an estimate of the safety factor before the heat exchange crisis was made, which showed that a reactor with such parameters can operate in a normal mode, since over the entire height of the core the value of heat flow from the surface does not approach the value of critical heat flow with a given safety factor before the heat exchange crisis. Even taking into account the error of the formula V.S. Osmachkin in 20%, the minimum critical coefficient is kk ≈ 1.5, which is less than the obtained coefficient in the calculations .

During the work, a four-group neutron-physical calculation was carried out, as a result, multiplication coefficient values ​ ​ for 1-4 groups were obtained: for the first group - 0.073, for the second group - 0.013, for the third group - 0.132, for the fourth group - 1.092. The total multiplication factor k = 1.31, which proves the operability of this reactor plant.

Reactor power plant VVER300 is designed for safe operation in nominal mode and is thermophysically reliable.

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