Analysis of the anticipated transient without scram (ATWS) initiated by emergency power mode through the full scope simulator

The integrity of the reactor coolant system is severely challenged as a result of an Emergency Power Mode – ATWS event. The purpose of this paper is to simulate the Anticipated Transient without Scram (ATWS) using the full scope simulator of Angra 2 Nuclear Power Plant with the Emergency Power Case as a precursor event. The results are discussed and will be used to examine the integrity of the reactor coolant system. In addition, the results were compared with the data presented in Final Safety Analysis Report (FSAR – Angra 2) in order to guarantee the validation of the methodology and from there analyze other precursor events of ATWS which presented only plausibility studies in FSAR – Angra 2. In this way, the aim is to provide and develop the knowledge and skill necessaries for control room operating personnel to ensure safe and reliable plant operation and stimulate information in the nuclear area through the academic training of new engineers. In the presented paper the most severe scenario is analyzed in which the Reactor Coolant System reaches its highest level of coolant pressure. This scenario is initiated by the turbine trip jointly with the loss of electric power systems (Emergency Power Mode). In addition, the failure of the reactor shutdown system occurs, i.e., control rods fail to drop into the reactor core. The reactor power is safely reduced through the inherent reactivity feedback of the moderator and fuel, together with an automatic boron injection. Several operational variables were analyzed and their profiles over time are shown in order to provide data and benchmarking references. At the end of the event, it was noted that Reactor shutdown is assured, as is the maintenance of subcriticality. Residual heat removal is ensured.

Reactor Shutdown and Reactor Restart

The power referred to most frequently in reactor physics is neutron power. Neutron power is essentially the fission rate multiplied by the average prompt energy released and recovered per fission (see Section 2.1.2). It is also called “prompt” power, as it appears very quickly following fission. We cannot measure neutron power directly, but we do monitor the neutron flux with ion chambers located outside the calandria and in-core flux detectors. These neutronic signals are calibrated to the thermal-power measurement which allows neutron power to be derived.

Evaluation of the Passive Reactor Shutdown System Performance in European Sodium Fast Reactor

The ESFR-SMART European project (Contract number: 754501) focuses on the development of innovative safety design options for European Sodium-cooled Fast Reactor (ESFR). The task of Work Package 1.3 is to assess the impact of the new safety measures on the reactor behaviour in the transients protected by either active or passive reactor shutdown systems. The aim of Task 4 in this Work Package is to evaluate the passive reactor shutdown system performance in the ESFR core. This paper deals with the results of this evaluation, which is based on the analysis of four transients passively protected by 12 Diversified Shutdown Device (DSD) rods. Simulations have been done with the SIM-SFR system code and demonstrated that DSD rods are capable to shutdown ESFR in a timely manner, in order to avoid the negative consequences of the analyzed transients. Although a total loss of heat sink transient is a practically eliminated event, it was included in the analysis to estimate the grace time before the core meltdown.

CANDU-Like Features in Emerging Small Modular Reactor Designs

Several small modular reactor (SMR) designs are emerging, but only the CANDU Small Modular Reactor and a couple of Indian designs incorporate the familiar features of the larger CANDU-reactors. This paper shows that while the CANDU concept did not seem to receive wider attention among SMR designers, it has influenced a few. The paper discusses how the CANDU operating experience can aid in the construction and operation of some SMRs. For example, the concept of passive reactor shutdown by draining the moderator, which was utilized in the early Pickering A units, is adopted in the Copenhagen Atomics Waste Burner; a molten slat (LiF-ThF$ _4 $) heavy-water moderated reactor. The heavy-water and lithium in this salt produce tritium and can benefit from the CANDU experience in handling tritium. The online refueling of CANDU reactors, their large heat sinks and seamless configuration are also reflected in SMR designs.

Accounting of the vver-1200 overload influence for fission products activities calculating

Current work is aimed at the analysis of the fission products decay influence during fuel reloading, when calculating the accumulated fission products activity for the VVER-1200 reactor fuel campaign. The Bateman problem solution based technique was used for calculations, within the framework of the two fissile nuclides approximation. The fission products producing process for the VVER-1200 reactor stationary campaign is considered, taking into account the reactor shutdown periods for refueling and without taking them into account (instant reload approximation). It was shown, that the instant reload approximation for fission products activity calculations gives the similar accurate result, as calculations with taking into account the shutdown periods. The results can be used to significantly simplify the calculations of fission product activity accumulation in nuclear power reactors.