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[en] Highlights: • A structural method for uncertainty evaluation of constitutive models is proposed. • Several tools for model uncertainty quantification are investigated and discussed. • Uncertainties of important models during the LBLOCA are quantified. • Best estimate plus uncertainty analysis is applied to the LP-02-6 experiment. - Abstract: Best estimate (BE) codes are developed to carry out realistic safety analysis of the nuclear reactor, and generally hundreds of constitutive models are comprised in a BE code. Nevertheless, uncertainties of these constitutive models are often not properly handled in the best estimate plus uncertainty (BEPU) analysis. It is not sufficient to fully evaluate the uncertainties of different sorts of models with only one method. Thereby a structural method for uncertainty quantification (UQ) of constitutive models is proposed and relevant description is presented. Based on the method, constitutive models will be classified into two categories according to the characteristics, namely the independent model and the dependent model, and different methods will be adopted for different sorts of model. Several statistical methods for UQ of constitutive models such as the non-parametric curve estimation method, the Bayesian calibration method and the coverage calibration method are evaluated and the characteristics of these methods are discussed. In addition, several methods for the construction of surrogate model are utilized to reduce the computational cost, and a model selection technique is adopted to opt the optimal model among all alternative models. The large break loss of coolant accident (LBLOCA) experiment LOFT LP-02-6 is utilized to verify the proposed structural method, and the BEPU analysis of the LP-02-6 experiment is carried out. The results show that uncertainty intervals of the identified models obtained through the structural method are reasonable, uncertainties of the peak cladding temperature (PCT) as well as the accumulator injection time (AIT) are quantified, and the sensitivity analysis is carried out to evaluate the influence of different input parameters on the 1st PCT, 2nd PCT and the AIT.
[en] Highlights: • Safety critical systems are designed to function in safe manner so that its failure should not lead to the catastrophic effects. • Due to safety significance of such systems, these have high performance requirements. • The strategy discussed for performance analysis of safety critical and control systems and to estimate performance based risk factor. • The technique elaborates Petri nets to estimate performability to ensure system dependability requirements. • The technique has been validated on 17 safety critical and control systems of Nuclear Power Plant. - Abstract: Non-functional requirements play a critical role in designing variety of applications domain ranging from safety-critical systems to simple gaming applications. Performance is one of the crucial non-functional requirements, especially in control and safety systems, which validates the design. System risk can be quantified as a product of probability of system failure and severity of its impact. In this paper, we devise a technique to do the performance analysis of safety critical and control systems that helps to estimate the risk. The technique elaborates Petri nets to estimate performability to ensure system dependability requirements. We illustrate the technique on a case study of Nuclear Power Plant. The technique has been validated on its 17 safety critical and control systems.
[en] Highlights: • ROP aging challenge related to power derating is addressed. • ROP detector layout optimization and HSP re-classification are the options explored to mitigate power derating challenge. • Significant improvement in the ROP TSP value for an aged CANDU reactor can be realized from implementing these options. - Abstract: Over the past few years, Candu Energy Inc. (a wholly owned subsidiary of SNC-Lavalin Inc., which acquired the assets of Atomic Energy of Canada Limited’s Commercial Reactor Division) has been continuously developing and evaluating various options to improve the regional overpower protection (ROP) margin in aged CANDU 600 MW (CANDU 6®) reactors. This paper presents results from applying a couple of margin improvement options to a generic aged CANDU 6 reactor, namely ROP detector layout optimization and application of a revised handswitch position designation. Application of these options requires no change to the ROP analysis methodology, statistical approach or acceptance criterion. As such, any increase in ROP margin associated with these options carries little or no licensing risk and are not expected to require more than one standard outage to implement
[en] Fuel management in PWR nuclear reactors is comprised of a collection of principles and practices required for the planning, scheduling, refueling, and safe operation of nuclear power plants to minimize the total plant and system energy costs to the extent possible. Despite remarkable advancements in optimization procedures, inherent complexities in nuclear reactor structure and strong inter-dependency among the fundamental parameters of the core make it necessary to evaluate the most efficient arrangement of the core. Several patterns have been presented so far to determine the best configuration of fuels in the reactor core by emphasis on minimizing the local power peaking factor (Pq). In this research, a new strategy for optimizing the fuel arrangements in a VVER-1000 reactor core is developed while lowering the Pq is considered as the main target. For this purpose, a Fuel Quality Factor, Z(r), served to depict the reactor core pattern. Mapping to ideal pattern is tracked over the optimization procedure in which the ideal pattern is prepared with considering the Z(r) constraints and their effects on flux and Pq uniformity. For finding the best configuration corresponding to the desired pattern, Cellular Automata (CA) is applied as a powerful and reliable tool on optimization procedure. To obtain the Z(r) constraints, the MCNP code was used and core calculations were performed by WIMS and CITATION codes. The results are compared with the predictions of a Neural Network as a smart optimization method, and the Final Safety Analysis Report (FSAR) as a reference proposed by the designer.
[en] AMEBA is an Italian acronym which stands for 'alta moderazione e basso arricchimento' (high moderation and low enrichment). The AMEBA reactor is nothing more than a PWR which possesses very unusual values of both volumetric ratio moderator/fuel and U-235 enrichment of UO2. The possibility is shown of the technical realisation of a nuclear power plant equipped with an AMEBA PWR reactor. Among the most enticing properties of AMEBA are the following: self-shut-down in any abnormal condition, elimination of all need for control rods and boric acid dissolution in the water, absolute impossibility of reaching values of reactivity greater than a fraction of a dollar, intrinsic subcriticality, attaining to several dollars, in non-operative condition when the water is at ambient temperature, normal operation with a very small-sized pressurizer, self-start-up
[en] A passive safety injection system (PSIS) is proposed for Chashma nuclear power plant-1 (CHASNUPP-1) type nuclear power plants, for the simplification of their safety systems. This system is based upon passive components and is proposed in place of the existing safety injection system, for safety enhancement. The functionality of the proposed system is analyzed using reactor simulation. For this purpose an intermediate size break LOCA is simulated using the simulation software APROS. For this transient, different thermal-hydraulic parameters of the proposed and other safety related systems are presented and discussed. The results obtained show that the proposed system works properly by performing its role in the transient, leading to cold shutdown conditions
[en] Among the new failure modes introduced by computer into safety systems, the process interaction error is the most unpredictable and complicated failure mode, which may cause disastrous consequences. This paper presents safety analysis and constraint detection techniques for process interaction errors among hardware, software, and human processes. Among interaction errors, the most dreadful ones are those that involve run-time misinterpretation from a logic process. We call them the 'semantic interaction errors'. Such abnormal interaction is not adequately emphasized in current research. In our static analysis, we provide a fault tree template focusing on semantic interaction errors by checking conflicting pre-conditions and post-conditions among interacting processes. Thus, far-fetched, but highly risky, interaction scenarios involve interpretation errors can be identified. For run-time monitoring, a range of constraint types is proposed for checking abnormal signs at run time. We extend current constraints to a broader relational level and a global level, considering process/device dependencies and physical conservation rules in order to detect process interaction errors. The proposed techniques can reduce abnormal interactions; they can also be used to assist in safety-case construction.
[en] Highlights: • A fully-implicit high-order system thermal-hydraulics model for advanced non-LWR. • Applied fully-coupled JFNK solution scheme to avoid the operator-splitting errors. • The effects of different spatial and temporal discretization schemes are investigated. • High-order convergence rates of discretization schemes are verified. - Abstract: An advanced system analysis tool is being developed for advanced reactor safety analysis. This paper describes the underlying physics and numerical models used in the code, including the governing equations, the stabilization schemes, the high-order spatial and temporal discretization schemes, and the Jacobian Free Newton Krylov solution method. The effects of the spatial and temporal discretization schemes are investigated. Additionally, a series of verification test problems are presented to confirm the high-order schemes. It is demonstrated that the developed system thermal-hydraulics model can be strictly verified with the theoretical convergence rates, and that it performs very well for a wide range of flow problems with high accuracy, efficiency, and minimal numerical diffusions.
[en] Highlights: • An efficient MC method for the sensitivity calculation of reactivity coefficients is developed. • The sensitivity of reactivity coefficient is calculated by MC second-order perturbation techniques. • Its effectiveness is examined in a two-group homogeneous problem and Godiva. • S/U analyses are performed for MDC of a LWR pin cell and FTC of a CANDU 6 lattice model. - Abstract: The uncertainty quantification of the reactivity coefficients such as the fuel temperature coefficient (FTC) and the moderator density coefficient (MDC) is crucial for the nuclear reactor safety margin evaluation. This paper proposes a continuous-energy MC second-order perturbation (MC2P) method as a new way to estimate efficiently the sensitivity of reactivity coefficients to nuclear cross section data. The proposed MC2P method takes into account the second-order effects of the fission operator and the fission source distribution. The effectiveness of the MC2P method implemented in a Seoul National University MC code, McCARD, is demonstrated in a Godiva 235U density coefficient problem via comparison of its results with direct subtraction MC calculation. It is shown that the new method can predict the cross section sensitivities of the reactivity coefficient more accurately even with much smaller number of MC history simulations than the direct subtraction MC method. It is also shown that the proposed method is applicable for quantifying the uncertainties of the MDC of a LWR pin cell problem and the FTC of a CANDU 6 lattice cell problem due to the uncertainties of the nuclear cross section input data represented by nuclear cross section covariance data.
[en] The gravity-driven boron injection system (GDBIS), designed by the Institute of Nuclear Energy Technology (INET) of the Tsinghua University, PR China, is a new type of passive system to be applied in the 200 MW nuclear heating reactor (NHR-200), also designed by INET. The function of this system is to shut down the reactor in an emergency, in case control rods do not operate properly. A borate water tank is located 10 m above the top of the pressure vessel. When the pressure of the reactor and the boron tank balances, the borate water will be driven by gravity to flow into the reactor, and thus shut down the reactor. The thermal hydraulic performances of the system for cold (room temperature nitrogen) and hot (mixture of hot steam and nitrogen) operating conditions, especially the response time of pressure and water injection, have been researched under different initial conditions. Firstly, several factors, e.g. orifice on steam lines, and the volume ratio of the gas-steam spaces of the reactor and the boron tank, have effects on the pressure and water injection response time and other thermal hydraulic performance of the system. Secondly, the steam and liquid communication modes, namely the acting time and sequence of the action of valves connecting steam and liquid lines, have great influences on the performance of the system. Thirdly, the limited pressure balance time (about 1.0 s) can be achieved under the cold condition. This investigation shows that GDBIS can be properly used in the 200 MW nuclear heating reactor