Study on construction of loss of component cooling water system initiating event fault tree for fire event PSA

ABSTRACT

In this study, the construction of the loss of component cooling water system (LOCCWS) initiating event (IE) fault tree (FT) for an actual fire event probabilistic safety assessment (PSA) model of the Korean reference nuclear power plant considering only IE initiators was validated. The quantification results of the LOCCWS accident sequences obtained using an LOCCWS IE FT model with only initiators are similar to that with initiators and enabling events. This confirmed that the LOCCWS IE FT for an actual fire event PSA model could be constructed by considering only IE initiators. In addition, the same LOCCWS accident sequences were quantified assuming that fire triggering only the LOCCWS IE leads to reactor shutdown. Compared with the quantification result obtained based on the assumption that any fire included in the fire event PSA leads to reactor shutdown, the core damage frequency (CDF) can be reduced. Thus, it can be concluded that there is a possibility of underestimation of CDF when the LOCCWS IE FT model with only initiators is used and the assumption that fire triggering only the LOCCWS IE results in reactor shutdown is employed for the quantification of LOCCWS accident sequences.     

Development of the methods to change the fault tree logic structure according to the changing conditions of a system in a one top PSA model

Korea Atomic Energy Research Institute (KAERI) has developed two methods to handle the problems when a one top PSA model used in a risk monitor was developed that could not match the structure of house trees. These two methods are (1) the development of a special gate called a ‘condition gate’ and (2) the implementation of an If-Then-Else (ITE) logic in the PSA model structure. The proposed methods were applied to an example case, e.g., the station blackout (SBO) accident sequences of a one top PSA model. It was demonstrated that these methods could properly handle the complex condition changes of a system during the SBO accident sequences by using the proposed methods.     

Change in activity of catalysts for the oxidation of tritium during a fire event

The catalytic performance should be maintained in any off normal events. Fire accident is the typical off normal event. In the fusion plant, typical combustibles are evaluated to be polymeric low-halogen cables. Produced gases from burned low-halogen cable may affect the activity of catalysts for the oxidation of tritium. We experimentally demonstrated the influence of produced gases from burned low-halogen cable on the activity of catalyst using tritium gas. Our evaluation showed that ethylene, methane and benzene were major produced gases. The activity of catalysts for the oxidation of tritium during a fire event was evaluated using a commercial hydrophilic Pt/Al₂O₃ catalyst and a commercial hydrophobic Pt-catalyst. The temperature of catalytic reactor was selected to be 423 and 293 K. At 423 K, no considerable decrease in catalytic activity was observed for both catalysts even in the presence of produced gases from burned low-halogen cable. At 293 K, considerable increase in catalytic activity was initially observed for both catalysts due to the effect of produced hydrogen. Then the temporary decrease was observed, however the catalytic activity was gradually recovered to be the original activity. Consequently, the irreversible decrease in activity of the catalysts during a fire event was not observed.     

秦山二期扩建工程低温超压事故PSA分析

低温超压事故在电厂停堆期间发生频率较高,并有可能导致堆芯熔化,是停堆工况下一个重要的安全问题。本文对一回路发生低温超压事故进程进行研究和分析,参考相关资料建立事件树,进行定量化计算,得到低温超压事故导致的堆芯损坏频率,并进行简单的结果评价。     

Modelling of radioactive material release from waste packages in case of a fire event

Assumed incidents in the operational phase of the planned German repository Konrad for radioactive waste with negligible heat production were investigated in order to assess their possible radiological consequences. Release fractions of the radioactive substances contained in waste packages were assessed from experimental data obtained under thermal impact. They are given for halogens, tritium, ‘4C and other radionuclides and are classified according to the waste form groups and waste container classes.     

An integrated approach to the reliability of engineering systems

An integrated approach for the reliability of engineering systems consisting of interdependent structural, mechanical, and electrical subsystems is presented. The method is a blend of structural and systems reliability methods and is capable of accounting for physical or functional dependencies between subsystems and their components. Advantages of the method over the existing methods are discussed and some of the difficulties in applying it are pointed out. As an example application, a simple method for the reliability assessment of a general system subjected to seismic hazard is presented.     

An analytical solution to a two-dimensional model of the rewetting of a hot dry rod

The model considers a hot dry rod of infinite length cooled by a film of liquid moving along its surface. The heat transfer coefficient is assumed to be constant on the wet side and zero on the dry side of the rewetting front, and the liquid film is assumed to move at constant speed. We derive an analytical formula relating the temperature difference in the rod, the temperature at the rewetting front, the wet side heat transfer coefficient, and the rewetting speed. The formula is thought to apply to the rewetting of a fuel rod during emergency cooling by flooding.     

Effect of hydrocarbons on the efficiency of catalytic reactor of detritiation system in an event of fire

Detritiation system of a nuclear fusion plant is mandatory to be designed and qualified taking carefully into consideration all the possible extraordinary situations in addition to that in a normal condition. We focused on the change in the efficiency of tritium oxidation of a catalytic reactor in an event of fire where the air accompanied with hydrocarbons, water vapor, and tritium is fed into a catalytic reactor at the same time. Our test results on the effect of these gases on the efficiency of tritium oxidation of the catalytic reactor indicated; (1) tritiated hydrocarbon produces significantly by reaction between tritium and hydrocarbons in a catalytic reactor; (2) there is little possibility of degradation in the detritiation performance because the tritiated hydrocarbons produced in the catalyst reactor are combusted; (3) there is no possibility of uncontrollable rise in the temperature of the catalytic reactor by heat of reactions; and (4) saturated water vapor could temporarily poison the catalyst and degrades the detritiation performance. Our investigation indicated a saturated water vapor condition without hydrocarbons would be the dominant scenario to determine the amount of catalyst for the design of catalytic reactor of the detritiation system.     

Clogging of industrial pleated high efficiency particulate air (HEPA) filters in the event of fire

The IRSN and AREVA NC are currently conducting a common interest fire research programme with the aim, among other things, of improving knowledge of clogging of high efficiency particulate air (HEPA) filters and developing an empirical model for clogging of such filters by combustion aerosols. This model must - insofar as possible - be independent of the nature of the fuel and be able to be integrated in a calculation code covering the interaction between the ventilation and the fire. This paper discusses the influence of various “direct” factors such as the filtration velocity, the mass of deposited aerosol per filter area, the diameter and morphology of the combustion particles, the condensate content of the aerosols, and “indirect” factors such as the air flow feeding the fire and its oxygen content, which influence the evolution of the aeraulic resistance of a clogged filter.     

An evaluation of risk methods for prioritizing fire protection features: a procedure for fire barrier penetration seals

This paper generally evaluates risk methods available for prioritizing fire protection features. Risk methods involving both the use of qualitative insights, and quantitative results from a fire probabilistic risk analysis are reviewed. The applicability of these methods to develop a prioritized list of fire barrier penetration seals in a plant based on risk significance is presented as a procedure to illustrate the benefits of the methods. The paper concludes that current fire risk assessment methods can be confidently used to prioritize plant fire protection features, specifically fire barrier penetration seals. Simple prioritization schemes, using qualitative assessments and insights from fire PRA methodology may be implemented without the need for quantitative results. More elaborate prioritization schemes that allow further refinements to the categorization process may be implemented using the quantitative results of the screening processes in good fire PRAs. The use of the quantitative results from good fire PRAs provide several benefits for risk prioritization of fire protection features at plants, mainly from the plant systems analyses conducted for a fire PRA.     

An analytical solution to fuel-and-cladding model of the rewetting of a nuclear fuel rod

An exact solution of the quasi-steady two-dimensional conduction equation for the rewetting of a nuclear fuel rod in water reactor emergency core cooling is obtained for a fuel-and-cladding model. A method of solving non-separable differential equations is presented, which is used in the present analysis. The recently developed theorem of orthogonality of piecewise continuous eigenfunctions is also used to handle the composite rod in the present model. The present analysis reveals that the wet front velocity increases with the increase of the gap resistance between the fuel and the cladding, and approaches a limiting value, which is equal to the wet front velocity of the tube of cladding alone, as the gap resistance becomes infinite. For convenience in practical application, the results of the present analysis are correlated in simple expressions.     

A design-phase PSA of a nuclear-powered hydrogen plant

A probabilistic safety assessment (PSA) is being developed for a steam-methane reforming hydrogen production plant linked to a high-temperature gas-cooled nuclear reactor (HTGR). This work is based on the Japan Atomic Energy Research Institute's (JAERI) High Temperature Engineering Test Reactor (HTTR) prototype in Japan. The objective of this paper is to show how the PSA can be used for improving the design of the coupled plants. A simplified HAZOP study was performed to identify initiating events, based on existing studies. The results of the PSA show that the average frequency of an accident at this complex that could affect the population is 7 × 10⁻⁸ year⁻¹ which is divided into the various end states. The dominant sequences are those that result in a methane explosion and occur with a frequency of 6.5 × 10⁻⁸ year⁻¹, while the other sequences are much less frequent. The health risk presents itself if there are people in the vicinity who could be affected by the explosion. This analysis also demonstrates that an accident in one of the plants has little effect on the other. This is true given the design base distance between the plants, the fact that the reactor is underground, as well as other safety characteristics of the HTGR.     

An analytical solution for the point reactor kinetics equations with one group of delayed neutrons and the adiabatic feedback model

The point reactor kinetics equations with one group of delayed neutrons and the adiabatic feedback model are solved analytically. The analytical solution is based on an expansion of the neutrons density in powers of the small parameter, the prompt neutrons generation time, into the second order differential equation in the neutron density. The relation between the time and the reactivity for reactor excursions near prompt critical is derived. Also, the neutron density and the average density of delayed neutron precursors as functions of reactivity are presented. The relations of reactivity, neutron density and temperature with time are calculated, drawn, and compared with other analytic method.     

An atomistic approach to self-diffusion in uranium dioxide

The formation and mobility of point defects in UO₂ have been studied within the framework of the Density Functional Theory. The ab initio Projector Augmented Wave method is used to determine the formation and migration energies of defects. The results relative to intrinsic point defect formation energies using the Generalized Gradient Approximation (GGA) and GGA+U approximations for the exchange-correlation interactions are reported and compared to experimental data. The GGA and GGA+U approximations yield different formation energies for both Frenkel pairs and Schottky trios, showing that the 5f electron correlations have a strong influence on the defect formation energies. Using GGA, various migration mechanisms were investigated for oxygen and uranium defects. For oxygen defects, the calculations show that both a vacancy and an indirect interstitial mechanism have the lowest associated migration energies, 1.2 and 1.1 eV respectively. As regards uranium defects, a vacancy mechanism appears energetically more favourable with a migration energy of 4.4 eV, confirming that oxygen atoms are much more mobile in UO₂ than uranium atoms. Those results are discussed in the light of experimentally determined activation energies for diffusion.     

An analytical approach to Monte Carlo spectrum unfolding

The Monte Carlo spectrum unfolding method offers model-independent means for extracting the maximum information from the measurement data together with their uncertainties. The uncertainty of the unfolded spectrum, including correlations, can also be obtained. Its main shortcoming is the huge number of calculations which must be carried out. The analytical unfolding method proposed reduces the amount of calculations by orders of magnitude. It essentially consists in replacing the most appropriate a priori probability distribution required for drawing sample spectra from it at random for the Monte Carlo statistics with a certain exponential distribution from which the desired spectrum and its uncertainty covariance matrix can be analytically calculated. By taking into account the principle of maximum entropy, only a few-parameter, non-linear optimization problem remains to be solved. The results are equivalent to those obtained by the Monte Carlo unfolding in the limiting case of an infinite number of energy groups.     

An object-oriented approach to simulation of IRIS dynamic response

In this paper the development of an adequate modelling and simulation tool for Dynamics and Control tasks is presented. The key features of the developed simulator are: “Modularity” - the system model is built by connecting the models of its components, which are written independently of their boundary conditions; “Openness” - the code of each component model is clearly readable and close to the original equations and easily customised by the experienced user; “Efficiency” - the simulation code is fast; “Tool support” - the simulation tool is based on reliable, tested and well-documented software.To achieve these objectives, the Modelica language was used as a basis for the development of the simulator. The Modelica language is the result of recent advances in the field of object-oriented, multi-physics, dynamic system modelling. The language definition is open-source and it has already been successfully adopted in several industrial fields.The test bed for the application of the object-oriented approach has been the new generation, integral type, IRIS nuclear reactor. IRIS (International Reactor Innovative and Secure) is a pressurized light water cooled, small/medium power (335 MWe) reactor reactor, under development by an international consortium of nineteen organizations from ten countries. The preliminary design has been completed and the pre-application licensing process with the US-Nuclear Regulatory Commission (NRC) is underway.To provide the required capabilities for the analysis, specific models for the nuclear reactor components have been developed, to be applied for the dynamic simulation of the IRIS integral reactor, albeit keeping general validity for PWR plants. The following Modelica models have been written to satisfy the IRIS modelling requirements and are presented in this paper: point reactor kinetic, fuel heat transfer, control rods model, and a once-through type steam generator, thus obtaining a specific library of nuclear models and components. As far as other classical power generation plant components are concerned, the Thermo Power open library, developed at Politecnico di Milano as well, has been adopted and is briefly presented in the paper. Originally conceived for conventional, fossil-fired plants, the highly modular approach allowed to effectively reuse the models of the balance of plant systems, which have been connected to the models of the nuclear power generation process, to obtain a system simulator for the IRIS reactor.Finally, preliminary results of the code validation process and the reactor dynamics are presented.     

An advanced approach to calculation of neutron resonance self-shielding

Based on the combination of subgroup method and characteristics method, a resonance self-shielding calculation code SGMOC is programmed. SGMOC code can handle the complex (both in geometry and resonant components) resonance problems. The numerical results are in good agreement with those of MCNP. In order to improve the SGMOC calculation accuracy, two techniques are utilized, i.e., the resonance interference effects between resonant nuclides are considered, and on the other hand, the elastic scattering resonance is taken into account. These two techniques can enhance the accuracy remarkably.