Fundamental Safety Functions Research Articles

Concept design, application and optimization of operational parameters for new forced safety injection tank system

Background The engineered safety systems are designed to execute fundamental safety functions encompassing reactivity confinement, reactivity control and decay heat removal. Failure of any one of these functions can result in severe accident conditions. Passive systems have been implemented as a better option for plant safety. This research proposes a modification of existing safety injection tanks to the new forced safety injection tanks (FSITs) that utilize the backpressure from the pressurizer or steam generator to drive coolant into the reactor core under high pressure conditions. Methods FSITs aim to extend the coping-time during accidents like Station Blackout, providing an extended timing window for deployment of FLEX systems. The mathematical design is proposed and implemented into the thermal-hydraulic input of a nuclear power plant to demonstrate the system’s applicability that was demonstrated by leveraging the conventional PRA approach and risk quantification. Results The suggested system useful when used in the accident scenario of Station black out in-coincident with turbine-driven pumps fail to run. As the proposed design is a conceptual design, the optimization of associated operational parameters and set points is necessary. This optimization is performed using the risk analysis and virtual control environment-based Dynamic Probabilistic Risk Assessment framework. The method and findings of this study affirm that the coping-time for Station Blackout can be significantly extended, ensuring a substantial margin for the effective deployment of the FLEX. Conclusions A concept design of a passive forced safety injection system was suggested and demonstrated by integrating the mathematical model into a thermal-hydraulic model of a nuclear power plant. The parameters were optimized and the results demonstrated that the new system was effective in recovering the nuclear power plant from the accidents such as SBO with turbine-driven pumps fail to operate.

Ensuring safety of new, advanced small modular reactors for fundamental safety and with an optimal main heat transport systems configuration

Abstract Many countries are considering Small and Modular Reactors as a viable alternative to counter the climate-change/global-warming with a quick deployment of green, carbon free nuclear energy option in the energy mix. Proponents of SMRs claim that these designs rely more on enhanced inherent/engineered safety and passive features with novel concepts. SMRs are being designed to be fabricated at a factory and then transported as ‘modules’ to the sites for installation either as a single module or multiple module plant. There are many variant of SMRs under considerations/design/construction/commissioning/operation states and majority of the, more than 70 odd SMRs are in the design stage. The paper focuses on safety aspects while addressing the fundamental safety requirement that are derived from fundamental safety principles, the acceptance criteria, the expected/envisaged safety targets and not only the economic impact/considerations. The assessment basis for requirements towards safety enhancements and their extent of assurance in the design are highlighted against the claims made. Ensuring SMR safety with respect to the fundamental safety functions will depend on the foreseen/predicted fission product releases, following overheating of the fuel, during the worst/credible accident conditions and likelihood of occurrence of these accidents. Innovations in the development of advanced fuel, deploying passive safety systems, novel concepts in main heat transport system configuration and advanced features in instrumentation can help in realising the goal of ensured enhanced safety in the SMRs, both in preventive and mitigation domains during severe accidents. Enhancements in the acceptance criteria and deterministic and probabilistic safety targets is also expected and may be envisaged. The paper brings out the challenges faced in the design and regulation of the new NPPs, while addressing fundamental safety principles implementation, generic, specific safety issues, and only genuine innovations can ensure and improve the safety. Aspects related to passive systems and the optimal main heat removal system configuration of the NPPs are also discussed. The aspects related to concurrent design and regulation of new NPPs including SMRs also has been brought out in the paper.

Model predictive and fuzzy logic controllers for reactor power control at Reaktor TRIGA PUSPATI

The Reaktor TRIGA PUSPATI (RTP) rely extensively on a core power control system for control reactivity to fulfilment the fundamental safety function for a nuclear reactor. It is technically challenging to operate within tight multiple parameter constraints and keep the core power output stable. At present, the power tracking performance of the system could be considered unsatisfactory which produces a relatively long settling time and high control effort. Hence, a study of a model predictive control (MPC) strategy by integrating the current Power Change Rate Constraint (PCRC) using fuzzy logic which is part of the core power control design is conducted. In this paper, the MPC design based on mathematical models of the reactor core included point kinetics model, thermal-hydraulic model, reactivity model and dynamic rod position model help to enhance core power control. The power tracking performance of the proposed control method and previous Feedback Control Algorithm-Fuzzy PCRC is compared via computer simulation with different power range operations. Overall, the results show the MPC-Fuzzy PCRC strategy provides a greater level of operational safety and optimum operation of the TRIGA reactor.

Development of Fuel Product Barrier Monitoring System Based on State Functions in State-Oriented Emergency Operating Procedure

For pressurized water reactor nuclear power plants, in order to prevent the release of radioactive substances into environment, fission product barriers (FPBs) are constructed based on the concept of defense-in-depth, including fuel clad, reactor coolant system (RCS), and containment; the status of these FPBs is then acting as an important dimension to decision-making of emergency action levels (EALs). For CPR1000 nuclear power plants, state functions defined in state-oriented emergency operating procedure (SOP) are used to characterize postaccident physical conditions; their degradation substantially represents the challenges on fundamental safety functions and then on the integrity of FPBs in like manner, so degradation of these state functions is referred to as determining initial conditions of each FPB, by which the link between SOP and EALs is established. Then, an intelligent FPB monitoring system (FPBMS) aiming to automatically monitor states of FPBs is developed, verified, and validated. The pioneering work, by building bridges between state functions and initial conditions of FPBs and then computerizing them innovatively, proves that dynamical monitoring of states of FPBs during accident evolvement and real-time indication of loss or potential loss of FPBs can be achieved, which is most helpful in decision-making of EALs.

Safety of Nuclear Power Plants with Respect to the Fault Displacement Hazard

The hazard of permanent ground displacements/deformations can challenge the safety of the nuclear power plants. Increasing knowledge of the hazard and development of methods for structure–fault–displacement interaction motivates the designing of nuclear power plants for permanent ground displacement instead of abandoning the sites that could be affected by this kind of hazard. For the design basis, permanent ground displacement should be defined at the hazard level that complies with the probabilistic criteria for accounting for the natural hazards in the design that also ensure compliance with probabilistic safety criteria. In this paper, a procedure is proposed for the definition of the design basis permanent ground displacement that is based on the deaggregation of seismic design basis hazard. The definition of the displacement for the margin evaluation is also proposed. The feasibility of safe design is also demonstrated for the proposed definition of design basis hazard via qualitative judgement on the sensitivity of the structures, systems and components ensuring the fundamental safety functions with respect to the permanent ground displacement that is supported by existing case studies.

The Evaluation of the High Temperature Gas Cooled Reactor Safety to Fulfill the Requirement of the Next Generation Nuclear

High temperature gas cooled reactor (HTGR) has been considered to be the most promising option to meet energy demands in the future. It has also been selected as the next generation nuclear plant. The primary safety requirement of the next generation nuclear plant design is to limit radioactive material releases to practically eliminate the need for public evacuation or sheltering beyond the exclusion area boundary. The purpose of this study is to evaluate the safety design of HTGRs in order to fulfill the requirement of the next generation nuclear plant. To achieve this objective, inherent safety features, fundamental safety functions, and confinement functions realized into the design of HTGRs are comprehensively evaluated. It is found that design provisions of HTGRs can fulfill the intention of keeping radionuclides at their original sources. The layers of the coated fuel particles are very robust to retain nuclear fission products for all foreseeable reactivity events. There will be no possibility of radioactive materials to be released even though related safety systems and operator intervention are not involved in the recovery actions. This design has complied with the requirement of the next generation nuclear plant, which is to practically eliminate the need for public evacuation or sheltering beyond the exclusion area boundary.Keywords: High temperature gas cooled reactor, inherent safety features, fundamental safety functions, confinement functions, next generation nuclear plant

Probabilistic safety Analysis for Assessing the Failure of Heat Removal Control of AP1000

The safety level of nuclear power plant can be evaluated by using the adequacy of defence in depth that is implemented in the design. The defence in depth is to fulfill requirements of the fundamental safety function, which are reactivity control, heat removal control from the reactor, and confine control of radioactive material. Those controls shall be analyzed by probabilistic safety analysis. It is important to analyze the second fundamental safety function because the first fundamental safety function has a chance to fail. The purpose of this study is to calculate of the failure probability for the second fundamental safety function that is the heat removal control. The analysis is carried out by determining top event that is the change of heat removal control. Hereinafter, the intermediate events and the basic events are selected and the failure probabilistic is calculated using failure rate of component and human error. The AP1000 is used as the object of study. The generic data is collected from IAEA and other published documents. The results show that the probability of intermediate events are 5.52E-06 to 9.11E-04 per year and the total failure probability of the heat removal control is 1.89E-03 per year. Based on the assessment, it is concluded that the failure probability of the heat removal control is small enough and still within IAEA criteria.

Advanced micro-reactor concepts

Nuclear power systems capable of outputting low powers (<100 MWth) are increasingly receiving interest internationally for deployment not only as electricity production systems, capable of operating off-grid, but also as systems able to provide industrial process heat. These ‘micro-reactor’ concepts must demonstrate economic competitiveness with other potential solutions capable of providing similar power outputs. With this in mind, reactor technologies that offer inherent advantages associated with improved power density and simplified operation, both of which are important attributes that determine economic competitiveness, are reviewed in the context of the fundamental safety functions provided by the IAEA.The reactor technology chosen based on the results of the review were: low vapour pressure coolants like molten salt or liquid metal; solid moderator material; and conventional solid UO2 fuel. Initial infinite lattice neutronic studies indicated a series of positive reactivity coefficients. A finite system was also modelled using a molten salt as the coolant. When modelling the finite system the coolant temperature reactivity coefficient became negative, the void coefficient strongly negative and moderator temperature coefficient negative to weakly positive. Given that a number of reactivity coefficients were negative to strongly negative in the finite system, the weakly positive moderator temperature coefficient is not thought to be prohibitive. Thus the design should exhibit acceptable safety performance.Whilst the importance of leakage in fast reactor cores is well known, a key outcome from this study is the strong influence of leakage on all safety related parameters for the thermal reactor designs considered here with solid moderator material. Thus it seems that safety studies for such small cores should be based on full core calculations instead of the traditional infinite lattice studies for fuel assemblies.

Probabilistic analysis on the failure of reactivity control for the PWR

The fundamental safety function of the power reactor is to control reactivity, to remove heat from the reactor, and to confine radioactive material. The safety analysis is used to ensure that each parameter is fulfilled during the design and is done by deterministic and probabilistic method. The analysis of reactivity control is important to be done because it will affect the other of fundamental safety functions. The purpose of this research is to determine the failure probability of the reactivity control and its failure contribution on a PWR design. The analysis is carried out by determining intermediate events, which cause the failure of reactivity control. Furthermore, the basic event is determined by deductive method using the fault tree analysis. The AP1000 is used as the object of research. The probability data of component failure or human error, which is used in the analysis, is collected from IAEA, Westinghouse, NRC and other published documents. The results show that there are six intermediate events, which can cause the failure of the reactivity control. These intermediate events are uncontrolled rod bank withdrawal at low power or full power, malfunction of boron dilution, misalignment of control rod withdrawal, malfunction of improper position of fuel assembly and ejection of control rod. The failure probability of reactivity control is 1.49E-03 per year. The causes of failures which are affected by human factor are boron dilution, misalignment of control rod withdrawal and malfunction of improper position for fuel assembly. Based on the assessment, it is concluded that the failure probability of reactivity control on the PWR is still within the IAEA criteria.

Proposal of the confinement strategy of radioactive and hazardous materials for the European DEMO

Confinement of radioactive and hazardous materials is one of the fundamental safety functions in a nuclear fusion facility, which has to limit the mobilisation and dispersion of sources and hazards during normal, abnormal and accidental situations. In a first step energy sources and radioactive source have been assessed for a conceptual DEMO configuration. The confinement study for the European DEMO has been investigated for the main systems at the plant breakdown structure (PBS) level 1 taking a bottom-up approach. Based on the identification of the systems possessing a confinement function, a confinement strategy has been proposed, in which DEMO confinement systems and barriers have been defined. In addition, confinement for the maintenance has been issued as well. The assignment of confinement barriers to the identified sources under abnormal and accidental conditions has been performed, and the DEMO main safety systems have been proposed as well. Finally, confinement related open issues have been pointed out, which need to be resolved in parallel with DEMO development.

Development of a PWR-W GOTHIC 3D model for containment accident analysis

The confinement of radioactive material in a nuclear power plant, including the discharge control and the release minimization, is a fundamental safety function to be ensured in a design basis accident (DBA).For plant licensing analysis, the containment is usually modeled with a lumped parameter approach. Inherent to the lumped parameter approach is the assumption that within each region the fluid is well mixed. However, the containment is a large building with a complex configuration and it is distributed in several compartments that avoid the well mixing of the fluid and could have three-dimensional effects that affect the thermal–hydraulic behavior. Therefore, the commonly used lumped parameter approach may not be enough to capture these effects.In order to study these assumptions, four generic PWR containment models have been developed for Mass and Energy (M&E) release analysis with GOTHIC 8.0 (QA) code, three of them being subdivided and the fourth one is a lumped parameter model. A Large Break LOCA is simulated in order to compare the thermal–hydraulic behavior of the different models. The results show a high dependence on the three-dimensional phenomena, especially the temperature and velocity distribution. In contrast, the pressure evolution is qualitatively similar in all models with small quantitative differences.

Modelling of Heat Transfer Phenomena for Vertical and Horizontal Configurations of In-Pool Condensers and Comparison with Experimental Findings

Decay Heat Removal (DHR) is a fundamental safety function which is often accomplished in the advanced LWRs relying on natural phenomena. A typical passive DHR system is the two-phase flow, natural circulation, closed loop system, where heat is removed by means of a steam generator or heat exchanger, a condenser, and a pool. Different condenser tube arrangements have been developed for applications to the next generation NPPs. The two most used configurations, namely, horizontal and vertical tube condensers, are thoroughly investigated in this paper. Several thermal-hydraulic features were explored, being the analysis mainly devoted to the description of the best-estimate correlations and models for heat transfer coefficient prediction. In spite of a more critical behaviour concerning thermal expansion issues, vertical tube condensers offer remarkably better thermal-hydraulic performances. An experimental validation of the vertical tube correlations is provided by PERSEO facility (SIET labs, Piacenza), showing a fairly good agreement.