Dynamic Correcting Dispersion Parameters of Lagrangian Puff Model in Atmospheric Tracer Experiments

In atmospheric dispersion models of nuclear accident, the empirical dispersion coefficients were obtained under certain experiment conditions, which is different from actual conditions. This deviation brought in the great model errors. A better estimation of the radioactive nuclide’s distribution could be done by correcting coefficients with real-time observed value. This reverse problem is nonlinear and sensitive to initial value. Genetic Algorithm (GA) is an appropriate method for this correction procedure. Fitness function is a particular type of objective function to achieving the set goals. To analysis the fitness functions’ influence on the correction procedure and the dispersion model’s forecast ability, four fitness functions were designed and tested by a numerical simulation. In the numerical simulation, GA, coupled with Lagrange dispersion model, try to estimate the coefficients with model errors taken into consideration. Result shows that the fitness functions, in which station is weighted by observed value and by distance far from release point, perform better when it exists significant model error. After performing the correcting procedure on the Kincaid experiment data, a significant boost was seen in the dispersion model’s forecast ability.

Molten Salt Reactor With Simplified Fuel Recycling and Delayed Carrier Salt Cleaning

The Molten Salt Reactor (MSR) represents an old concept, but its properties are qualifying it for the advanced utilization: inherent safety, excellent neutron economy, possibility of continuous or batch reprocessing without fuel fabrication. The aim of this paper is to characterize the MSR unique fuel cycle advantages in different neutron spectra using the results of ERANOS-based EQL3D and ECCO-MATLAB based EQL0D procedures. It also focuses on the low production of higher actinides in the Th-U cycle and based on the results, it proposes a simplified in situ recycling of the fuel and the delayed ex situ carrier salt cleaning or direct disposal by vitrification.

A Research Reactor Source Term and Dose Results in the Event of a LOCA

The present study discusses the source term derivation and dose result calculation for a hypothetical accident sequence in the Greek Research Reactor – 1 (GRR-1). A loss-of-coolant accident (LOCA) has been selected as a credible accident sequence. The source term derivation has been based on the GRR-1 confinement performance where the inventory has been computed assuming continuous reactor operation. A core damage fraction of 30% has been considered for the calculations while conservative core release fractions have been employed. The radionuclides released from the reactor core to the confinement atmosphere have been subjected to natural decay, deposition on and resuspension from various internal surfaces before being led to the release pathway. It has been assumed that an emergency shutdown is initiated immediately after the beginning of the accident sequence and the emergency ventilation system is also activated. Subsequently, the source term has been derived comprising of noble gases, iodine and aerosol. The JRODOS computational software for off-site nuclear emergency management has been utilized to estimate the dose results from the LOCA-initiated source term that is released in its entirety from the reactor stack at ambient temperature. The Local Scale Model Chain in conjunction with the DIPCOT atmospheric dispersion model that is embedded in JRODOS have been used with proper parameterization of the calculation settings. Five weather scenarios have been selected as representative of typical meteorological conditions at the reactor site. The scenarios have been assessed with the use of the Weather Research and Forecast model. Total effective, skin, thyroid, lung and inhalation doses downwind of the reactor building and up to a distance of 10 km have been calculated for each weather scenario and are presented. The total effective gamma dose rate at a fixed distance from the reactor building has been assessed. The radiological consequences of the dose results are discussed.

The AP1000 LBLOCA Analysis Using SCDAP/RELAP5

In this study, a transient performance simulation model of AP1000 using SCDAP/RELAP5 4.0 is developed. The reactor coolant system (RCS) and passive core cooling system (PXS) are modeled respectively. Various kinds of hydrodynamic component including Volume, Junction, Separator, Accumulator, Branch, Pipe, Valve and Pump are adopted to simulate the fluid system of AP1000. The DECLG (double-end rupture of cold leg) accident is simulated and analyzed. To study the effect of axial heat conduction, two kinds of heat structure with and without reflooding model are employed to simulate the fuel rod respectively. The comparison shows that the 2D heat conduction play important role in the reflooding process.

Design Improvement Studies for Future SFR

Prototype Fast Breeder Reactor (PFBR), a 500 MWe, (U-Pu)O2 fuelled, sodium cooled, pool type fast reactor, is in advanced stage of construction at Kalpakkam, India. Based on the experience gained during the design, manufacture and erection of various reactor components of PFBR, it is planned to construct Sodium cooled Fast Reactors (SFR) by adopting twin unit (2×500 MWe reactors) concept. The future Fast Breeder Reactor (FBR) – 1 & 2 have three main heat transport circuits, namely primary sodium, secondary sodium and steam-water systems. All the reactor internals including core and primary heat transport circuit systems are contained in a single vessel called main vessel and it is closed with top shield. Reactor assembly forms the heart of the Nuclear Steam Supply System. A detailed and exhaustive design / optimisation exercise was initiated towards improving the economic competitiveness and enhancing the safety of future FBRs. It is observed that the overall dimensions of the reactor assembly contribute immensely to the capital cost. In this context, detailed studies were carried out towards optimizing the overall dimensions of the reactor assembly. Further, the reactor assembly design in particular has been engineered to favour manufacture of integrated assembly and erection of the same, as a single unit, in reactor vault to reduce construction time. Various activities undertaken towards technology development of critical components have enhanced the confidence level in the improved design concepts and reducing time for manufacture and erection. In addition to the reactor assembly, specific improvements have been made in decay heat removal systems and sodium purification system. The layout incorporates a twin unit concept in which the ex-vessel fuel handling system and fuel storage building are shared. This paper discusses the basis for undertaking the review exercise and experience gained during construction of PFBR and highlights the design studies and technology development carried out for future SFR.

Analysis of Heterogeneous Cores Using SN Code With Discontinuity Factor

Solving diffusion equations with the equivalence homogenization theory is the common method in reactor neutronics. But for some case, as for stronger absorbers, the diffusion equations will bring great errors and the transport method will be more suitable. The discontinuity factor theory has been successfully used in core diffusion computation programs and effectively reduced the homogenization error. The method of using the discontinuity factor in the transport method were studied. The result shows that higher accuracy was obtained from the discrete ordinates core transport computation program with discontinued factor.

To Analyze the Core Damage Scenario via the RISMC-Based Metatheory and Reliability Theory

With “theory of nuclear safety (TONS)”, this paper intends to explain the Core Damage (CD) scenario of a Nuclear Power Plant (NPP) with the systematic methodology, many notions introduced here can be extended to other types of nuclear installations, as well. This systematic methodology combines the Risk-Informed Safety Margin Characterization (RISMC) Metatheory of TONS, and the basic reliability theory. A “metatheory” of such theories, here, is a theory to analyze the Theory of Nuclear Safety (TONS); in its own theory system, it is designed to summarize the safety of a NPP. Meanwhile, the basic reliability theory, which is decided by the authors, is focus on the mission reliability model (a model can be established by Reliability Block Diagram (RBD)); then the related basic concepts, is simple and clear, and quite mature in NPP field. The present work outlines the traditional reliability theory and the RISMC-based Metatheory, and these two concepts here are taken as the appropriate TONS to analyze the CD Scenario, after that, a renovate or renew TONS, from these two sides, can be introduced to analyze the fundamental safety of NPP.

Methodology to Evaluate Nuclear Power Plant Vulnerability to Events in the Transmission Grid

This paper presents a general approach to evaluate the risk of trip or Loss of Off-site Power (LOOP) events in nuclear power plants due to contingencies in the power grid. The proposed methodology is based on the Zone of Vulnerability concept for nuclear plants introduced by EPRI in previous work. The proposed methodology is intended to be part of an integrated probabilistic risk assessment tool that is being developed under ongoing EPRI R&D programs. A detailed analysis of many events occurred in actual nuclear plants has been performed in order to identify, classify and characterize the various vulnerability and type of failures that may affect a nuclear plant. Based the outcome of that analysis, a methodology for evaluating the impact of off-site transmission system events on nuclear plants has been outlined. It includes description of the type of contingencies and conditions that need to be included in the analysis, as well as provisions regarding the simulation tools and models that should be used in each case. The methodology is illustrated in a simplified representation of the Western Electricity Coordinating Council (WECC) system in the U.S.

On Features of Thermal Design of Passive Evaporation-and-Condensation Systems of Reactor Thermal Shielding

Further development of nuclear engineering is inseparably linked with the requirement of vast application of the passive systems of heat removal running without human intervention. Creation of such systems is impossible, if only conventional engineering solutions are used. As known, to prevent propagation of the fission products into the environment there are three safety barriers. To provide operation of the third safety barrier (containment shell), in particular, of the reactor cavities both in operational and emergency modes a passive evaporation-and-condensation (EC) system of heat removal is proposed. The features of thermal design of the EC systems for thermal shielding of the reactor cavities are considered. They make it possible to determine the optimal main design variables of the EC systems and prove reasonability and efficiency of their application. The performed study validates engineering feasibility of an efficient EC system for thermal shielding of the reactor equipment.