COMPBRN IIIe—a computer code for probabilistic fire risk analysis

The COMPBRN code has been used extensively to predict deterministically the time-to-damage of critical components in nuclear power plant fire risk analyses. Because there is a significant amount of uncertainties in the input parameters used in room fire simulations, the assessment of the damage time of the specified components must be performed probabilistically. This paper presents an updated version of the code, called COMPBRN IIIe, which emphasizes the importance of parameter uncertainty propagation by incorporating capabilities to provide probability distributions for component damage times. COMPBRN IIIe eliminates several errors from its previous versions and incorporates a user-friendly environment to assist users in preparing input files. With these improvements, the code can significantly reduce the time and effort required in the performance of a probabilistic fire risk assessment. A compartment fire simulation is also provided to demonstrate the application of the code.     

Design features of a cable–bollard vehicle barrier system

The design features of the stand-alone cable–bollard vehicle barrier system (Cable-Bollard VBS) developed for the Vermont Yankee Nuclear Power Plant (VY) to meet the design goals of the recent 10 CFR Part 73 rule changes are discussed. The design is based on the application of fundamental engineering principles to a dynamic system, recognizing that vehicle impact on a cable system is fundamentally different from vehicle impact on a bollard or other hard barrier. As such, rigorous attention is paid to cable anchor design and performance.     

Development of a dimensionless parameter for characterization of dielectric barrier discharge devices with respect to geometrical features

Non-thermal plasma (NTP) devices produce excited and radical species that have higher energy levels than their ground state and are utilized for various applications.There are various types of NTP devices,with dielectric barrier discharge (DBD) reactors being widely used.These DBD devices vary in geometrical configuration and operating parameters,making a comparison of their performance in terms of discharge power characteristics difficult.Therefore,this study proposes a dimensionless parameter that is related to the geometrical features,and is a function of the discharge power with respect to the frequency,voltage,and capacitance of a DBD.The dimensionless parameter,in the form of a ratio of the discharge energy per cycle to the gap capacitive energy,will be useful for engineers and designers to compare the energy characteristics of devices systematically,and could also be used for scaling up DBD devices.From the results in this experiment and from the literature,different DBD devices are categorized into three separate groups according to different levels of the energy ratio.The larger DBD devices have lower energy ratios due to their lower estimated surface discharge areas and capacitive reactance.Therefore,the devices can be categorized according to the energy ratio due to the effects of the geometrical features of the DBD devices,since it affects the surface discharge area and capacitance of the DBD.The DBD devices are also categorized into three separate groups using the Kriegseis factor,but the categorization is different from that of the energy ratio.     

An evaluation of approximate methods for calculating anisotropic flux escape and transmission probabilities for cylindrical annular regions

In the interface current method for solving the integral transport equation, most of the computer time is spent in the precalculation of various region transmission, escape and collision probabilities. Therefore, any reduction in computer time used to calculate these probabilities will directly affect the efficiency of this method. The normal methods such as the Trapezoidal rule or Gauss-quadrature formula for calculating the various probabilities require the evaluation of many Bickley functions per integral, which is time consuming. This paper discusses efficient methods for calculating the various probabilities corresponding to anisotropic terms of the angular flux expansion at region interfaces for cylindrical annular geometry. The paper first discusses the application of Bonalumi's Pseudo-Linear (P-L) approximation and the second-order correction to it which requires the evaluation of one and two Bickley functions, respectively, per integral. The accuracy of the second-order approximation is of the order of 0.1% for all the probabilities except the transmission probability from outer to outer surface (P_ρv∞) of the region (where ρ and v denote the order of anisotropy of the flux entering and leaving the outer boundary of the annular region). The paper then discusses a Gauss-quadrature formula with exponential weighting and its modified form which is suitable for calculating P^ρv_∞ efficiently.     

Triple-humped fission barrier model for a new 238U neutron cross-section evaluation and first validations

A new neutron-induced cross-section evaluation of ²³⁸U from 1 keV up to 200 MeV has been performed using only nuclear reactions models. A new fission penetrability model taking into account a triple humped barrier has been developed. A clear improvement has been observed for K-effective validation tests (up to 30 MeV) with this new evaluation. This improvement is mainly due to a better treatment of the inelastic exit channel.     

An improved procedure for the determination of apparent strain

An improved procedure for theoretically determining the apparent strain is proposed. Based on the results of an experimental study into the behaviour of high temperature strain gages in a varying thermal environment a theoretical formula is developed. One of the significant features of the proposed formula lies in its ability to predict the apparent strain without actually carrying out the experiment, provided the relevant strain gage material parameters as also the coefficient of thermal expansion of the specimen material are known. A comparison study of the results obtained by using the proposed formula with the experimentally determined results shows excellent agreement. Its application to the development of improved temperature compensated gages is also indicated.     

A study of risk reduction for some mitigation features for PWRs

The risk reduction attainable with mitigation features in a large-dry pressurized water nuclear reactor (PWR) is evaluated. The calculations are made in a probabilistic risk analysis framework, and they are based on Zion Probabilistic Safety Study (ZPSS). Some of the modifications made to this study are also taken into account.The mitigation designs considered consist of features for simultaneously controlling late containment overpressure, containment basemat penetration, and hydrogen burning. The individual mitigation features include: a passive containment heat removal system (PCHRS), a filtered-vented containment system (FVCS), a core ladle, and controlled hydrogen burning. Emphasis is placed on comparison of PCHRS and FVCS design options. The results include calculations of the sensitivity to several failure mode probabilities and to the probability of core meltdowns with containment bypass.     

A probabilistic seismic risk assessment procedure for nuclear power plants: (I) Methodology

A new procedure for probabilistic seismic risk assessment of nuclear power plants (NPPs) is proposed. This procedure modifies the current procedures using tools developed recently for performance-based earthquake engineering of buildings. The proposed procedure uses (a) response-based fragility curves to represent the capacity of structural and nonstructural components of NPPs, (b) nonlinear response-history analysis to characterize the demands on those components, and (c) Monte Carlo simulations to determine the damage state of the components. The use of response-rather than ground-motion-based fragility curves enables the curves to be independent of seismic hazard and closely related to component capacity. The use of Monte Carlo procedure enables the correlation in the responses of components to be directly included in the risk assessment. An example of the methodology is presented in a companion paper to demonstrate its use and provide the technical basis for aspects of the methodology.     

An evaluation of containment inerting and air dilution systems as methods for post-accident hydrogen control in BWRS

Hydrogen control is important in post-accident situations because of possibilities for containment rupture due to hydrogen deflagration or detonation. Post-accident hydrogen generation in BWR containments is analyzed as a function of engineered hydrogen control system, assumed either nitrogen inerting or air dilution. Fault tree analysis was applied to assess the failure probability per demand of each system. These failure rates were then combined with the probability of accidents producing various hydrogen generation rates to calculate the overall system hydrogen control probability. Results indicate that both systems render approximately the same overall hydrogen control failure rate (air dilution: 8.3 × 10⁻²−1.1 × 10⁻²; nitrogen inerting: 1.3 × 10⁻²−2 × 10⁻³). Drywell entries and unscheduled shutdowns were also analyzed to determine the impact on the total BWR accident risk as it relates to the decay heat removal system. Results indicate that inerting may increase the overall risk due to a possible increase in the number of unscheduled shutdowns due to a lessened operator ability to correct and identify ‘unidentified’ leakage from the primary coolant system. Further, possible benefits of inerting due to reduced torus corrosion and fire risk in containment appear to be dominated by the possible operations-related disadvantages.     

Hazard curve evaluation method development for a forest fire as an external hazard on nuclear power plants

A method to obtain a hazard curve of a forest fire was developed. The method has four steps: a logic tree formulation, a response surface evaluation, a Monte Carlo simulation, and an annual exceedance frequency calculation. The logic tree consists domains of “forest fire breakout and spread conditions”, “weather conditions”, “vegetation conditions”, and “forest fire simulation conditions.” Condition parameters of the logic boxes are static if stable during a forest fire or not sensitive to a forest fire intensity, and non-static parameters are variables whose frequency/probability is given based on existing databases or evaluations. Response surfaces of a reaction intensity and a fireline intensity were prepared by interpolating outputs from a number of forest fire propagation simulations by fire area simulator (FARSITE). The Monte Carlo simulation was performed where one sample represented a set of variable parameters of the logic boxes and a corresponding intensity was evaluated from the response surface. The hazard curve, i.e. an annual exceedance frequency of the intensity, was therefore calculated from the histogram of the Monte Carlo simulation outputs. The new method was applied to evaluate hazard curves of a reaction intensity and a fireline intensity for a typical location around a sodium-cooled fast reactor in Japan.     

316L不锈钢阻氢或氢同位素渗透搪瓷壁垒层的制备及性能研究

为解决316L不锈钢防氢及其同位素渗透问题,采用两搪两烧方法在不锈钢基体表面制备厚90~110μm的搪瓷涂层。利用X射线衍射仪、光学和扫面电子显微镜表征搪瓷涂层的显微结构、界面形貌,通过EDS线扫描分析界面处主要元素分布,结果显示搪瓷涂层结构致密,与基体形成化学结合,抗落球冲击和热震性能优异。西华特装置气相充氢试验与维氏显微硬度试验结果表明搪瓷涂层是一种有效地阻止氢及其同位素渗透的壁垒层。     

An evaluation of a direct cooling method for the ex-vessel corium stabilization

An evaluation of the ex-vessel core catcher system of a sample advanced light water reactor was presented. The core catcher was designed to cool down the molten corium through a combined injection of water and gas from the bottom of the molten corium, which could be effective in the reduction of rapid steam generation. By using the MELCOR code, a scenario analysis was performed for a representative severe accident scenario of the ALWR, that is, the 6-in. large break loss of coolant accident without safe injection. The spreading characteristics of ejected corium at vessel breach were asymptotically evaluated on the core catcher horizontal surface. The composition of the molten corium, the decay power level, and the sacrificial concrete ablation depth with time were obtained by a sacrificial concrete ablation analysis. The corium cooling history in the core catcher during the coolant injection was evaluated to calculate the temporal steam generation rate by considering an energy conservation equation. These were used as the major inputs for the temporal calculations of containment pressure which was performed by using the GASFLOW code. Several cases with change of water and gas injection rates were calculated. It was confirmed that the bottom water/gas injection system was an effective corium cooling method in the ex-vessel core catcher to suppress the quick release of steam.     

An approach to the construction of a one top fire event PSA model

An internal fire event probabilistic safety assessment (PSA) model has been generally quantified by modifications of a pre-developed internal events PSA model. New accident sequence logics not covered in the internal events PSA model are separately developed to incorporate them into the fire PSA model. Previous studies on the changes of the one top internal event PSA model for the one top fire event PSA model have been limited to the equipment failures affected by the fire. In addition, they assumed that the probabilities of basic events associated with equipment or cables impacted by the fire are one. However, the probabilities of spurious operation events and human failure events affected by the fire might not be estimated as one. In this study, new modification rules were proposed for the construction of a one top PSA model for fire events by using a one top internal event PSA model. The proposed new modification rules can be applied to all the fire damage events for the fire-induced equipment failure events and spurious operation events, human error events impacted by a fire, regardless of whether they are estimated as one or not. Applications of the proposed modification rules to the compartment and scenario-level fires for the hypothetical plants were performed for demonstrating their appropriateness to the changes of the one top internal event PSA model to the one top fire event PSA model. In addition, quantification procedure with the one top fire event PSA model was presented and discussed.     

某工厂夜光车间环境放射性监测、评价及防护

在黑暗环境中显示仪表指示常采用荧光技术.某工厂夜光车间是利用掺人放射性物质的荧光粉对玻璃制品进行涂抹、描绘等工艺加工的生产车间.由于荧光粉中掺人放射性物质226Ra、40K,226Ra、40K在自发衰变过程中会放出α、β、γ射线,对人体会产生不同程度的损伤,大剂量或累计照射会对人类造成一定辐射影响.因此,工作时应严防人为等意外因素可能造成超安全计量的放射性损害.为确保夜光车间工作人员和公众人身健康与安全,必须定期对夜光车间及周围环境进行放射性监测与评价.通过环境放射性监测,分析环境辐射水平,寻找可能存在的问题,进而提出保障人体健康的防护措施.     

An evaluation of grazing-incidence optics for neutron imaging

The focusing capabilities of neutron imaging optics based on the Wolter-1 geometry have been successfully demonstrated with a beam of long wavelength neutrons with low angular divergence. A test mirror was fabricated using an electroformed nickel replication process at Marshall Space Flight Center. The neutron current density gain at the focal spot of the mirror is found to be at least 8 for neutron wavelengths in the range from 6 to 20 Å. Possible applications of the optics are briefly discussed.     

Analysis of risk in taking measures for radiation and social protection of the public

The necessity of risk analysis for support of radiation and public protection measures and substantiated application of risk analysis for the development of a scientific-methodological and normative-legal base is examined. Recommendations are given for the development and use of risk analysis. For radiation accidents, a priori and postfactum risk assessment and harm must be distinguished (before and after an accident has occurred). In the latter case, detailed risk assessment is needed, especially to establish a link between sickness and the accident as well as for compensation of harm to the health of victims. A collection of risk indicators for different applications is given. A new additional risk indicator is proposed. This indicator is the relative loss of duration of life (lost years of life, referred to the year during which the source of risk acted). It is intended for establishing a universal limit on risk, suitable for any source of risk. The value of this basic limit is proposed and values of dose and risk limits derived from it for ionizing radiation and other sources of risk are presented. Requirements for methods of dose and risk assessment are formulated. 3 figures, 1 table, 16 references. Translated from Atomnaya énergiya, Vol. 87, No. 5, pp. 384–395, November, 1999.