Assessment of neutronic safety parameters of VVER-1000 core under accident conditions

Abstract

The major issue in designing safety and control systems for power reactors is preventing accidents or reducing the hazard imposed. Any change in reactor power causes a change in temperature and material densities of the system that will affect the reactivity, which in turn, causes transients/accident conditions. The analysis of main safety parameters is essential for any change to ensure the safe operation of the reactor in both steady and transient situations. In this paper, the impact of temperature, moderator density, boron concentration and control rods position changes on the safety parameters is studied to assess the behavior of the reactor under accident conditions. The main safety parameters as temperature reactivity coefficients, power peaking factor, effective delayed neutron fraction and generation time are evaluated under accident conditions for VVER-1000 core using MCNP6 code and ENDF/B-VII.1library. The results evidenced that the reactivity coefficients are still negative, under accident conditions, which guarantee the safe operation of the reactor. The effective delayed neutron fraction increased as a result of increased temperature, decreased moderator density and addition of the soluble boron. The worth of the control rods is sufficient to make the reactor into a subcritical state either in emergency situations or with regular shutdown.

Introduction

The analysis of the possible changes in reactivity is an essential part of the overall safety assessment of nuclear reactors. Increasing temperature will create feedback mechanisms in the reactor; Doppler broadening, thermal expansion and density changes which will induce spectral shifts. These changes will influence the reactivity, thus causing transients. The reactivity changes are generally referred to as reactivity feedback and are characterized by reactivity coefficients. Reactivity coefficients are the important parameters which influence the safety and controllability of the nuclear reactors. It's defined as the change of reactivity per unit change in some operating parameter of the reactor. The sign and magnitude of these coefficients are very important in determining the behavior of the core operational and accident conditions. Fuel temperature coefficient (FTC) is a major parameter in the assessment of a nuclear reactor core; it's called the prompt fuel coefficients. The fuel temperature effect is prompt where the power is generated in the fuel region instantaneously while the moderator temperature effect is delayed since the heat transfers to the moderator after seconds. Mainly, for all reactivity initiated accidents (RIA) it will be the primary and most significant feedback that will compensate the inserted positive reactivity. In power reactors, the Doppler coefficient is always negative for the reactor control purpose. The moderator temperature doesn't change immediately, but evaluation of moderator temperature coefficient is important to ensure the stability during normal operation and accident conditions (Rokhmadi and Zuhair, 2015; Thilagam et al., 2007; Lamarsh et al., 2001).

Also, two controllable conditions; control rod position and soluble poison concentration, significantly affect the reactivity balance in the nuclear reactor system. Boron concentration coefficient BCC, is defined as the reactivity changes per change in boron concentration. Primarily, it is a function of the ratio of boron absorption to the total absorption. The reactivity worth of control rod is a key parameter in many accident analyses. The control rods worth is essential in the evaluation of the shutdown margins of the reactor in addition to the reactivity insertion rates in response to a reactor scram. The reactivity worth of the control rod is a measurement of the efficiency of the control rod to absorb excess reactivity. Various conditions in reactor core affect the reactivity worth of the control rods. The determination of the control rod worth, in any change, is important to assure the safe and reliable operation of the reactor system. The Shutdown margin (SDM) defines the degree of sub-criticality that would be obtained immediately following the insertion of all rods, assuming that all control rods are fully inserted except for the single rod with the highest integral worth (so called stuck control rod), which is assumed to be fully withdrawn. It is defined as a reactivity difference in two states; when CRs inserted in the core and fully withdrawn of the core (Shiraniet.al, 2014; DOE FUNDAMENTALS HANDBOOK NUCLEAR PHYSICS, 1993).

Existing of delayed neutrons is essential in nuclear reactors, since its control the speed at which the reactor can increase its power. Without the delayed neutrons, the power of the reactor would increase to a high degree in a short period of time cause significant damage. Accident analysis and the control of a nuclear reactor as well the conversion of a reactor period into reactivity require the information for the effective delayed neutron parameters in addition to their decay constants. The delayed neutron fraction defined as the number of fission induced by delay neutrons divided by the total number of fission induced in the same system (Ott and Neuhold, 1985; Chae et al., 1998).

The reactivity coefficients, the delayed neutron fraction, the neutron generation time, and the power peaking factor are the most important neutronic parameters for determining the state of the reactor. Many research studies are performed to determine the essential safety parameters at ordinary operating conditions for pressurized water reactors (PWR) and research reactors using several computer codes. In the following paragraphs some of these studies are mentioned.

Surian Pinem et al., have performed the calculations of the temperature reactivity coefficients of fuel and moderator, the moderator density and the boron concentration of the AP1000 core at the hot full power condition (HFP) using NODAL3 Code (Pinemet al., 2018).

Tayfun AKYUREK, has performed the calculation of reactivity and FTC (Doppler coefficient) with various weight percent of burnable poisons at different temperatures for Westinghouse type pressurized water reactor (PWR) assembly using the MCNP (Monte Carlo N-Particle) code (Tayfun, 2018).

Phan Thi Thuy Giang and Do Quang Binh, have performed the calculations of fuel and moderator temperature coefficients of reactivity for the DaLat nuclear research reactor (DNRR), using Highly Enriched Uranium fuels HEU (36%) and Low Enriched Uranium Fuels LEU (19.75%) using the WIMSD code and CITATION code (Phan and Do Quang, 2015).

In this paper, main safety parameters of power reactors such as reactivity coefficients, reactivity worth of control rods, effective delayed neutron fraction, generation time, power density peaking factor and radial flux distribution are calculated. The assessment of reactor safety parameters under accident conditions is performed for a VVER-1000 reactor core using the MCNP6 code (Pelowitz, 2013) and ENDF/B-VII.1 library (Chadwicket al., 2011). The data of VVER-1000 were taken from the IAEA benchmark (IAEA-TECDOC-847) (IAEA, 1995).