Methods for reliability of structures under multiple time varying loads

Current probabilistic studies on load combination have been concentrating on the modeling and linear combination of loads and load effects. Herein, a methodology for the assessment of life-time reliability of structural systems under multiple time varying loads is proposed. The method is based on considerations of load occurrences (including coincidence) and conditional probability of failure given occurrence whereby current methods of structural reliability and random vibration analysis can be incorporated. It has a wide range of applications. The emphasis is on the time domain load behavior, structural capacity uncertainty and non-linear dynamic load effects. The accuracy and computational advantage of this method as compared with other methods are examined by a numerical example of a simple frame under loads modeled as Poisson renewal pulse processes. Also, the errors implied in the widely used ‘Turkstra’ rule, a recently proposed modification of this rule, the load reduction factor method, and a SRSS (square root of sum of squares) rule for load combinations are quantified for linear combination and the implications in risk evaluation mentioned.     

A simple mass and heat balance model for estimating plant conditions during the Fukushima Dai-ichi NPP accident

A simple evaluation method for the analysis of thermal-hydraulic transients in reactor pressure vessel (RPV) and primary containment vessel (PCV) is proposed to support understanding the accident behaviors of the Fukushima Dai-ichi nuclear power plant (NPP). Since most of the measurements of the plants were unavailable especially in the early stage of the accident, and the accessibility to the plants had been limited by radiation, analytical investigation for the plant was required to understand the plant conditions such as the magnitude of the damages. In order to provide easy-to-use technical tools to support the analytical investigation, we developed a simplified analysis code, named “HOTCB”, based on total mass and heat balances in a lamped parameter system. The HOTCB code has capabilities to treat two-phase fluid including water, steam, and non-condensable gas in a wide range of temperatures up to highly superheated conditions, and to consider heat structures, i.e. heat capacities and heat transfer to the fluid. The code was provided to Tokyo Electric Power Company (TEPCO) and was practically used for the analysis on the accident. This paper provides the details of the code and simulations of Unit 1 and Unit 2 reactors of Fukushima Dai-ichi nuclear power plant (NPP) as examples to show the usefulness of the code.     

Methods of estimating high-boiling channel reactor (RBMK) fuel-pin and cassette reliability

Conclusions These formulas for the point and interval fault-free probability estimators for RBMK cassettes and fuel pins have not been derived before, nor have the formulas for the estimators for the mean working life in terms of statistical working data, and experience with using them has shown that they are reasonably accurate and convenient in engineering applications. They have been incorporated into the sectional TSD. The formulas for estimating the mean service lives of cassettes and pins are analogous to those given for the mean working life. To derive these estimators requires one to split up the initial statistical information into several types, since the range of cassette and pin reliability parameters for the RBMK is dependent on the type of initial data available, as are the algorithms for calculating them.Translated from Atomnaya Énergiya, Vol. 58, No. 6, pp. 404–409, June, 1985.     

Methods of calculating radial boundary conditions for flux trap control assemblies

Some methods of calculating a two- or four-group albedo matrix for the control assemblies of VV'ER-440 reactors are discussed. All of these methods use the S_N program DTF-IV to calculate partial neutron currents, but the group constants for the DTF-IV calculations were obtained by different methods. The influence of the boron steel absorber on the thermalization cross section of the water in the control assembly turned out to affect the albedo matrix significantly. The control assembly itself does not, however, have any significant effect on the group constants of adjacent fuel assemblies.Published in Atomnaya Energiya, Vol. 52, No. 1, pp. 39–44, January, 1982.     

The prediction of failure strain for Zircaloy-4 subjected to non-stationary loading conditions

A modified Monkman-Grant relationship (MMG) is applied to predict — as far as the life time is known — the failure strain of Zry-4 subjected to tensile rupture test at load as well as temperature ramp conditions, respectively. As the analysis has shown in the first case, a simple relationship exists between the minimum creep rate and the stress rate. Thus, this quantity appearing in the MMG is phenomenologically connected with the test conditions. For failure strain predictions in temperature ramp tests the introduction of an effective temperature has shown to be advantageous. As compared to the peculiarities of the problem, the agreement between experiments and calculations is encouraging.     

Non reflecting boundary conditions and their use in estimating the effects of coolant channel failure in a PHWR

Non reflecting boundary conditions are applied to the truncation boundary of large domains to restrict the domain to a small region of interest from which waves travel only outwards. In this paper, this procedure is used to calculate the pressure field in the moderator of a pressurized heavy water reactor after a coolant channel has failed. Non reflecting boundary conditions are applied to the moderator boundary while a perfectly reflecting boundary is assumed for the channels.     

Preservice and inservice requirements for containment structures in the United States - A status report

Assuring the lifetime integrity of containment structures for nuclear power plants is becoming increasingly important as existing design criteria are reexamined, as new requirements for containment inspection and testing are formulated, and as today's operating nuclear plants are growing older.The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code contains requirements for the design and construction and for the preservice and inservice requirements for nuclear power plant systems and components in the United States. Section III of the ASME Code contains the rules for design and construction of nuclear systems and components. Rules for the preservice examination, inservice inspection, system pressure testing, repair, modification and replacement of nuclear systems and components are contained in Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components. Compliance with the rules of the ASME Code in the United States is mandated by the federal government in Title 10, Part 50 of the Code of Federal Regulations (10CFR50).Section XI of the ASME Code contains separate rules for metal (Class MC) and concrete (Class CC) containments. Requirements for Class MC containments have been published in Subsection IWE, Requirements for Class MC Components of Light-Water Cooled Power Plants, of Section XI. Rules for Class CC containments are currently being developed and will be published in Subsection IWL, Requirements for Class CC Components of Light-Water Cooled Power Plants, of Section XI.First published in 1981, Subsection IWE has been adopted by a number of state jurisdictions in the United States and is presently being reviewed by the United States Nuclear Regulatory Commission. Federal regulations that will require mandatory compliance by nuclear plant owners are forthcoming. When implemented, the requirements in Subsection IWE and Subsection IWL will provide a reasonable and systematic basis for assuring the integrity of metal and concrete containment structures during their service lifetime.This paper presents an overview of the preservice and inservice requirements for containment structures in Section XI of the ASME Code with consideration of the practical factors that should accompany user compliance.     

级联衰变核素活度测量方法的比较和探讨

符合法和计数相加法是级联衰变核素活度的两种测量方法。文章分析了两种方法的原理、联系和区别。在此基础上，提出了相加法和直接计数相加法，并指出，相加法和符合法在数学上是等价的，直接计数相加法和计数相加法是等价的。     

On the multiplication factor and reactivity definitions for subcritical reactor systems

An analysis is made on the merit of different functions adopted for weighting neutron processes in subcritical nuclear reactor systems, as it appears in expressions of relevant integral quantities, such as reactivity worths, prompt neutron lifetimes, etc. All weight functions may be shown to depend on some sort of explicit or implicit, real or fictitious, system control. Associated with the importance function relevant to the reactor power control, the multiplication factor k_sub and generalized reactivity values ρ_gen are defined. The difference (1-k_sub)/k_sub is shown to be the more appropriate index for generally representing the ADS subcriticality. However, in certain circumstances, when an accidental event is studied in which the criticality condition may be surpassed during the transient, it appears more appropriate to take into account the standard multiplication factor (k_eff) and the reactivity values to which the transient is associated and for the definition of which the standard adjoint flux is adopted.     

Darlington GS vacuum building - Internal structures and foundation

This paper describes the evolution of the internal structure from initial concept to final design. Fundamental changes to the original configuration were precipitated by the action of large seismic forces acting on a top-heavy configuration. Prestressing was eliminated in deference to high humidity. Aspects of the elevated water tank's peripheral support beam are discussed vis-à-vis an adjacent slipforming operation, and practical construction limitations on steel placement. Also reviewed are the shortening of peripheral columns due to shrinkage and creep, and considerations of crack control for purposes of water-tightness. The authors justify the choice of stainless steel for fabrication of the siphon system's riser pipes. The foundation slab must resist the combined effects of vacuum pressure, hydrostatic uplift, and the seismic reactions of the internal structure and perimeter wall. The dependency of a key foundation component, the gallery roof slab, on the dome tendon layout is high-lighted; and aspects of its constructability are reviewed in light of congestion of vertical tendon anchorages, and of reinforcement. The design of the air-tight slab liner is reviewed, attention focusing on weld design under vacuum and accident temperature loads; on corrosion protection; and on the related construction access bulkhead - its ASME requirements and fabrication tolerances.     

低能N+注入油脂降解菌株的诱变选育及降解条件的初步研究

通过N^＋注入解脂耶罗威亚酵母（Yarrowia lipolytica）DC-3-2,筛选出一株油脂降解率提高了11．09％的高效菌株DC．3．2．50,经10次传代实验表明该菌株遗传稳定向良好。对DC-3-2-50的降解条件进行了初步研究,结果表明,在最佳条件：初始pH9．0、接种量3．0％、温度为25--28℃、摇床转速为180-200rpm、碳源为大豆油时,其降解率最高,可达87．7％。     

Verification method for durability of reinforced concrete structures subjected to salt attack under high temperature conditions

Immersion tests of concrete specimens, corrosion tests of reinforced concrete specimens and combination tests of carbonation and chloride ion penetration of concrete specimens were conducted under high temperature. Verification method of durability against salt attack to reinforced concrete structures under high temperature was discussed. The obtained results were summarized as follows. (a) Diffusion coefficient of chloride ion in concrete increased as temperature rose. The relationship between the logarithm of diffusion coefficient and the reciprocal of temperature showed linearity. (b) The chloride ion concentration of reinforcing steel corrosion initiation did not decrease at high temperature. (c) Diffusion coefficient of chloride ion might be larger in carbonated concrete. (d) Based on the test results, a verification method of durability against salt attack on reinforced concrete structures under high temperature up to 65 °C, for avoiding steel corrosion, was proposed.     

A new method for estimating heat flux in superheater and reheater tubes

In this paper a procedure on how to estimate the heat flux in superheater and reheater tubes utilizing the empirical formula and the finite element modeling is proposed. An iterative procedure consisting of empirical formulae and numerical simulation is used to determine heat flux as both temperature and scale thickness increase over period of time. Estimation results of the heat flux over period of time for two different design temperatures of the steam and different heat transfer parameters are presented.     

Thermo-mechanical loadings of primary structures in accidental conditions, and an approach to solution by limit design

We first describe the static thermo-mechanical loadings of some important structures due to accidental situations. We calculate the behaviour of these structures by several methods; elastic and plastic analysis, and limit design. We show the interest of this last method.     

Dynamic strain aging in stress controlled creep-fatigue tests of 316L stainless steel under different loading conditions

Stress controlled fatigue-creep tests were carried out for 316L stainless steel under different loading conditions, i.e. different loading levels at the fixed temperature (loading condition 1, LC1) and different temperatures at the fixed loading level (loading condition 2, LC2). Cyclic deformation behaviors were investigated with respect to the evolutions of strain amplitude and mean strain. Abrupt mean strain jumps were found during cyclic deformation, which was in response to the dynamic strain aging effect. Moreover, as to LC1, when the minimum stress is negative at 550 °C, abrupt mean strain jumps occur at the early stage of cyclic deformation and there are many jumps during the whole process. While the minimum stress is positive, mean strain only jumps once at the end of deformation. Similar results were also found in LC2, when the loading level is fixed at −100 to 385 MPa, at higher temperatures (560, 575 °C), abrupt mean strain jumps occur at the early stage of cyclic deformation and there are many jumps during the whole process. While at lower temperature (540 °C), mean strain only jumps once at the end of deformation.     

Methods in Environmental Sampling for Radionuclides

This paper reviews methods of environmental sampling for radionuclides around operational and preoperational nuclear power plants. We examine in detail the implications of the established radiation standards and their effect on sampling procedures. Transport mechanisms of radionuclides in liquid effluent, and the deposition of airborne radionuclides onto soil and vegetation are discussed. We evaluate water- and soil-sampling procedures. The Lawrence Livermore Laboratory program of terrestrial gamma-ray surveys at preoperational nuclear power plants is described.