Anticipated Transient Without Scram Event Research Articles

A study of the behavior of the floating absorber for safety at transient (FAST) in an innovative sodium-cooled fast reactor (iSFR)

Negative reactivity feedback is one of the most important inherent safety characteristics of reactors for suppressing changes in thermal power and component temperatures during accidents. The floating absorber for safety at transient (FAST) has been suggested to improve the coolant temperature coefficient (CTC) in a sodium-cooled fast reactor (SFR) for a better self-stabilization of the reactor system. Previous studies on the FAST application showed that the FAST causes power excursion and temperature oscillations under anticipated transient without scram (ATWS) conditions. This study explains the variables affecting the FAST behavior and oscillations with mathematical expressions. In addition, a damping FAST is proposed to passively remove oscillations without any mechanical system. The damping FAST increases the damping ratio by reducing the pin diameter at a height where the difference between the FAST bottom elevation and the equilibrium position is maximum. The performance evaluation for the damping FAST is conducted in an innovative sodium-cooled fast reactor (iSFR) under ATWS conditions. The 7-mm damping FAST unilaterally inserts the negative reactivity in unprotected transient overpower event, dramatically lowering the core power without oscillations. While the previous FAST shows oscillatory behavior repeatedly approaching the coolant boiling temperature, the damping FAST stably keeps the coolant temperature below 800 °C. Moreover, the damping FAST lowers the maximum coolant temperatures for other ATWS events. These findings show the potential of the FAST system to significantly improve the inherent safety of SFRs.

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

Abstract 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.

Enhancing the Plant's Capability for Design Basis and Design Extension Conditions Based on Time-Dependent Context Evaluation of Human Performance in ATWS Events

The experience of severe accidents shows that reliable determination of technological process parameters is necessary but not always sufficient to avoid catastrophic consequences. The accident measures should be considered in a broader context that includes the human factor, organization of the nuclear technology, external influences, and safety culture. The anticipated transient without scram (ATWS) events were not considered in the original water water energy reactor (WWER) (Russian pressurized water reactors (PWR)) design basis accidents (DBA). The design extension conditions (DEC) scenarios progress in a context which is very uncertain and highly stressful for the operators. If a specific scenario requires some operators' actions as measures to mitigate, delay, or distribute the accident consequences, then the dynamics of accident context seem of primary importance for “best estimate” evaluations and enhancing the plant's capability. The paper presents the capacities of the performance evaluation of teamwork (PET) procedure for enhancing plant's capability for DEC based on best estimate context evaluation of human performance in ATWS events. The PET procedure is based on a thorough description of symptoms of various timelines and their context quantification. It is exemplified for different ATWS scenarios of the nuclear power plant (NPP) with WWER-1000 based on thermal-hydraulic simulations with RELAP5/MOD3.2 code and models.

A Long-Term MAAP 3.0B Analysis of a Severe Anticipated Transient Without Scram Accident in a Boiling Water Reactor

This paper reports that the long-term responses of the reactor coolant system and the containment of a boiling water reactor (BWR) during an anticipated transient without scram (ATWS) initiated by an inadvertent closure of the main steam isolation valves (MSIVs) are analyzed with the MAAP 3.0B computer code. The plant used in the analysis is the Kuosheng nuclear power station, a General Electric BWR6 reactor with a Mark III containment. A comparison of near-term system responses to an MSIV closure ATWS event shows that the MAAP 3.0B code fails to reproduce the detailed thermal-hydraulic results of the sophisticated RETRAN-02 code. The MAAP code, however, makes a reasonable prediction of the timing of major events and core power in the ATWS accident analyzed. The results of a long-term MAAP ATWS analysis show that if the standby liquid control system fails in an MSIV closure ATWS event, the ATWS event might develop into a core melt accident with the containment failing before the core melts.

Passive Safety Features of a Bottom-Supported Fast Breeder Reactor Vessel

A system dynamics analysis is applied to a pool-type fast breeder reactor to examine the influence of a bottom-supported reactor vessel (BSRV) design on anticipated transient without scram (ATWS) events such as an unprotected loss of flow (ULOF), an unprotected loss of heat sink (ULOHS), and an unprotected transient overpower (UTOP) by using the ARGO safety analysis code. The BSRV enhances negative feedback because of the differential displacement between the core and the control rod as compared with a top-supported reactor vessel. In particular, the BSRV has the potential, especially in a mixed-oxide-fueled core, to mitigate the design requirements to prevent boiling of the coolant during an ULOF and ULOHS through the elongation of the primary flow coastdown and enhancement of the axial expansion of the control rod drive line. In the metallic-fueled core, the effects of the BSRV on the ATWS events are diminished by the limitation of the sodium temperature increase.

An Investigation of Decreasing Reactor Coolant Inventory as a Mechanism to Reduce Power During a Boiling Water Reactor Anticipated Transient Without Scram

Under certain anticipated transient without scram (ATWS) sequences for a boiling water reactor, it would be desirable to reduce system power, particularly where the primary system has been isolated by closure of all main steam isolation valves and is discharging steam through its safety/relief valve system to the suppression pool. Reducing reactor power increases the time available to shut down the reactor by minimizing the heat dumped to the suppression pool and by helping to keep the suppression pool temperature within limits. Under proposed emergency procedure guidelines for the ATWS event, the reactor water level would be lowered to reduce reactor power. The analyses provide an assessment of the power level that would be attained, assuming the reactor operators were to reduce the the downcomer level down to the top of the active fuel.