Development and adoption of measures for increasing the safety of power-generating units of nuclear power plants with RBMK reactors in 1986–2005

The basic stages and directions of upgrading and reconstruction of the power-generating units of nuclear power plants with RBMK reactors over a period of 20 yr since the Chernobyl accident are reflected in this paper. The greatest amount of work was done in 1986–1990 on all units with RBMK-1000 and-1500 reactors which were operating at that time in the USSR. The purpose of the upgrade was to improve the neutron-physical characteristics of the reactor cores, increase the response speed of the safety and control system and decrease the positive effect of water outflow from the cooling loop of this system, increase the flow capacity of the systems performing emergency discharge of the steam-gas mixture from the reactor, and improve the basic operating documentation. The subsequent stages of the upgrading and reconstruction concerned mainly the first-generation power-generating units (Nos. 1 and 2 units of the Leningrad and Kursk nuclear power plants). These works were performed to make the power-generating units conform to the requirements of the modern normative documents on safety and allow for the possibility of these units to remain in operation after the nominal 30-yr service life has been exhausted. __________ Translated from Atomnaya énergiya, Vol. 100, No. 4, pp. 312–320, April, 2006.     

Improvement of regulatory documents on decommissioning power-generating units of nuclear power plants

Approaches to revising the documentation regulating the safety of decommissioning power-generating units of nuclear power plants are examined. It is proposed that the concept of planning decommissioning for operating and newly designed power-generating units of nuclear power plants at all stages of their lifecycle be introduced into the analysis. It is shown that the sections concerning the decommissioning of the power-generating units of nuclear power plants be revised in documents validating operation and decommissioning.     

Nuclear Power Plant Safety and Strength of Equipment Components with Service-Life Extension of First-Generation Power-Generating Units

The safety of nuclear power plants and the strength of the components of the first-loop equipment operating under pressure are examined together. The problem is analyzed for an actual situation with possible extension of the operational service life of the first-generation power-generating units. Since it is impossible to upgrade the first-generation units, in accordance with modern safety requirements, effective compensating measures at other stages of a multistage protection system must be formulated and implemented in the form of appropriate protection and containment safety systems. It is suggested that as one such measure special attention be focused on the diagnostics of the service life of equipment taking account of the design and real cyclic and other loads which occur. 9 references.     

Nominal Installed Capacity Utilization Factor (ICUF) as an Indicator of the Installed Capacity Utilization Efficiency of a Nuclear Power Plant

The nominal installed capacity utilization factor for power-generating units in different types of nuclear power plants is analyzed. It is shown that this indicator does not correlate with the operative regulation for servicing power-generating units, and it does not take into account the probability of unanticipated power decreases due to equipment failures and the present operating conditions of nuclear power plants in the common power system operating in Russia.     

核电站严重事故后果概率安全评价方法研究

核电站严重事故后果概率安全评价(PSA)是采用概率论的方法对核电站放射性后果进行分析,并定量给出放射性物质对核电站周围公众的健康效应影响。以国内某压水堆核电站为参考厂址,建立合适的场外后果分析模型。采用分层抽样方法对参考厂址1a的气象数据进行抽样,源项和释放特征等数据取自二级PSA的研究结果。利用事故后果评价程序对核电站严重事故后果进行计算,并用概率论方法对结果进行评估。通过计算将各事故和事故谱的场外个人剂量表示为CCDF曲线和总频率-剂量曲线,再用概率论方法得到不同距离处个人剂量超过指定剂量的条件概率;也可用此方法对确定烟羽应急计划区的安全准则中所描述的"大多数严重事故序列"进行量化。     

让切尔诺贝利核事故的警钟长鸣——纪念切尔诺贝利核事故25周年

一次严重事故发生后总引发业内人士颇有兴趣的讨论并相应地对核安全框架做出修改,但每次重大事故的起因并非相同,触及到的问题也不一样。25年前的切尔诺贝利核事故让世界震惊;25年中5个堆发生堆熔远远超出了国际上对严重事故发生概率的规定;25年中因堆熔事故11个机组报废、两个厂址变成了人们不敢前往的地方。面对这一切,我们需要有全新的核安全思维。     

Subcritical space nuclear system without most movable control systems

This paper describes the design and analysis of advanced space nuclear reactor (ASNR) whose design combines the advantages of radioisotope thermoelectric generator (RTG) and space nuclear reactor (SNR). As opposed to current SNRs designs, ASNR is a subcritical system driven by ²³²U–Be neutron source to generate thermal power continuously. Most movable control systems in the SNR design are removed. The detailed neutronic calculations by MCNPX (Monte Carlo N-Particle eXtended), including k_eff, flux, burn-up, loss-ratio of neutron source and immersion reactivity, show that ASNR has higher criticality safety and more compact structure to bear the risk of immersion accident compared with the past SNRs, and the new system can provide more thermal power than RTG. Furthermore, the neutron source efficiency is optimized to improve the utilization of ²³²U–Be neutron source with the improvement of criticality safety. Compared with the past designs of space nuclear power, ASNR could provide enough thermal power and avoid the occurrence of serious immersion accident in the case of total control system failure. ASNR has potential for future deep space missions.     

Choice of the unit power-generating capacity of a nuclear power plant

A characteristic of the present status of nuclear power in Russia is that in the next few years it will be necessary to make basic technological and economic decisions that will have long-term consequences. These decisions must concern all aspects of the nuclear-power complex. Specifically, at the present time there is no validation of the present and future requirements for the capacity of serially manufactured power-generating units of nuclear power plants with VVER or fast reactors. The problem of choosing the unit capacity of a nuclear power plant must be examined taking account of different factors and not solely from the standpoint of minimizing the capital and operational components of the cost of electricity. The main objective of this work was to develop recommendations for validating the optimal capacity of powergenerating units in nuclear power plants (capital costs, construction time, harm due to unanticipated stoppage of the power-generating units, unification, manufacturing quality, harm due to accidents, and so forth), the possibilities of electric grids, and the regional demand for electricity. Translated from Atomnaya énergiya, Vol. 105, No. 5, pp. 243–248, November, 2008.     

Experience in operating and upgrading the No. 5 unit of the Novovoronezh nuclear power plant – practical base for developing a reliable source of nuclear energy

The No. 5 unit of the Novovoronezh nuclear power plant, starting commercial operations on September 26, 1980, is the first power-generating unit with a 1000 MW VVER in our country. The assimilation of its power gave invaluable experience to designers, builders, and equipment manufacturers; this experience was taken into account in the design solutions for next-generation power-generating units. A large volume of work on increasing the efficiency, reliability, and safety was performed over a 30-year service life. At present, the power-generating unit has been shut down for a major overhaul for upgrading according a program for extending the service life by 25–30 years.     

Probabilistic Service-Life Assessment for Safety Analysis for Preparing a Power-Generating Unit of a Nuclear Power Plant for Service-Life Extension

Information about the actual history of cyclic loading in the actual state of the metal in equipment components must be used for analyzing the safety of power-generating units in operating nuclear power plants, in preparation for service-life extension, and to determine the moment of onset of rapid aging of the metal and the end of the period of stable operation.The properties of the exponential distribution function and probability functions which are constructed for various loading scenarios using the hypothesis of probabilistic summation of fatigue damage for estimating the -percentage residual service life of equipment components are examined.Probabilistic estimates of the service life for operating nuclear power plants make it possible to control effectively the residual service life of the components of a nuclear power plant on the basis of information provided by the diagnostics systems and by the systems monitoring the state of the metal and the data on loading parameters from the control systems.     

Failure assessment of RPV nozzle under loss of coolant accident

In case of a postulated loss of coolant accident (LOCA) of a reactor pressure vessel (RPV), the nozzle region experiences higher stresses and lower temperatures than the remaining part of the RPV. Thus, the nozzle is to be considered in the RPV safety assessment. For a LOCA event, three-dimensional elastic–plastic finite element calculations of stresses and strains in the intact RPV were performed. Using the substructure technique, fracture mechanics analyses were then carried out for several postulated cracks in the nozzle corner and in the circumferential weld below the nozzle. For different crack geometries and locations, the J-integral and the stress intensity factor were calculated as functions of the crack tip temperature. Based on the K_IC-reference curve and the J_R curve, both brittle and ductile instability of the postulated cracks were excluded. In order to reduce the expenses of three-dimensional finite element analyses for various crack geometries, an analytical procedure for calculating stress intensity factors of subclad cracks in cylindrical components was extended for cracks in the nozzle corner.     

Three-dimensional interpretation of cleavage fracture tests of cladded specimens with local approach to cleavage fracture

Electricité de France has conducted during these last years an experimental and numerical research programme in order to evaluate fracture mechanics analyses used in nuclear reactor pressure vessels integrity assessment, regarding the risk of brittle fracture. Two cladded specimens made of ferritic steel A508 Cl3 with stainless steel cladding, and containing shallow subclad flaws, have been tested in four point bending at very low temperature to obtain cleavage failure. The crack instability was obtained in base metal by cleavage fracture, without crack arrest. The tests have been interpreted by local approach to cleavage fracture (Beremin model) using three-dimensional finite element computations. After the elastic–plastic computation of stress intensity factor K_J along the crack front, the probability of cleavage failure of each specimen is evaluated using m, σ_u Beremin model parameters identified on the same material. The failure of two specimens is conservatively predicted by both analyses. The elastic–plastic stress intensity factor K_J in base metal is always greater than base metal fracture toughness K_1c. The calculated probabilities of cleavage failure are in agreement with experimental results. The sensitivity of Beremin model to numerical aspects is finally exposed.     

Probabilistic seismic safety study of an existing nuclear power plant

This study was conducted as part of an overall safety study of the Oyster Creek nuclear power plant. The earthquake hazard was considered as an initiating event that could result in radioactive release from the site as a result of core melt. The probability of earthquake initiated releases were compared with the probability of releases due to other initiating events.Three steps are necessary to evaluate the probability of earthquake initiated core melt.

1. (1) estimate the ground motion (peak ground acceleration) and uncertainty in this estimate as functions of annual probability of occurrence;

2. (2) estimate the conditional probability of failure and its uncertainty for structures, equipment, piping, controls, etc., as functions of ground acceleration; and

3. (3) combine these estimates to obtain probabilities of earthquake induced failure and uncertainties in such estimates to be used in event trees, system models, and fault trees for evaluating the probability of earthquake induced core melt.

This paper concentrates on the first two steps with emphasis on step 2. The major difference between the work presented and previous papers is the development and use of uncertainty estimates for both the ground motion probability estimates and the conditional probability of failure estimates.The ground motion capacity of a structure, component, etc., is treated for simplicity and clarity as a product random variable A given by , where is the best estimate of the median ground acceleration capacity, _R and _U are lognormal random variables with unit median and logarithmic standard deviation β_R and β_U, respectively. β_R expresses the dispersion in the ground acceleration capacity due to underlying randomness from such sources as (1) the variability of an earthquake time-history and thus of structural response when the earthquake is only defined in terms of the peak ground acceleration; and (2) the variability of structural material properties associated with strength, inelastic energy absorption and damping. Essentially, β_R represents those sources of dispersion which cannot be reduced by more detailed evaluation or more data. Uncertainty concerning the ground motion capacity is expressed by β_U which results from such things as (1) lack of complete knowledge of structural material properties; and (2) errors in calculating response due to approximate modelling. This paper presents a methodology (with examples) for estimating , β_R, and β_U for structures and components. These estimates are then used to estimate conditional probabilities of failure with confidence bounds on these estimates.The conclusion is that a rational approach exists for estimating earthquake induced probabilities of failure. Confidence bounds on such estimates can be developed to express uncertainty in the parameters used. Such an approach is preferable over one in which dispersion due to underlying randomness, and due to uncertainty in the data are combined into a single probability of failure estimate with no estimate of the uncertainty in this probability.     

Effect of thermal radiation on rewetting of top spray emergency coolant

A general physical model for top spray rewetting during an emrgency core cooling (ECC) transient is proposed which takes into account thermal radiation in the dry region. The model is employed to study the effect of thermal radiation on rewetting a single rod and a 3 × 3 rod bundle up to 2100°F. The results show that rewetting in a bundle is slower than for an isolated rod, due to reduced thermal radiation heat transfer in the dry region. Also, there is a definite correlation between the decreased radiation heat flux Δq_R and the corresponding decrease in rewetting velocity Δu. Values of Δu are not significant unless Δq_R is larger than 6000 Btu/hr ft², where Δq_R cannot exceed a value of 6000 Btu/hr ft² below a temperature of 1100°F, even in the most adverse conditions. Hence, it is concluded that radiant heat transfer does not significantly affect rewetting velocities up to an initial rod temperature of 1100°F. Beyond this temperature, the rewetting velocities change by more than 1.5% and hence radiation must be included in the model for top spray rewetting.     

VBÉR-300 reactor system and power-generating units for regional nuclear power production

The basic design solutions and characteristics of the VBéR-300 reactor system for the power-generating units of 150–300 MW(e) nuclear power plants and regional nuclear heat-and-electricity plants are described. The reactor system implemented as a unit is based on the technologies and solutions used for marine nuclear power systems, which have been corroborated by experience in operating nuclear-powered icebreakers. The technical-economic advantages of floating power-generating units are substantiated. __________ Translated from Atomnaya énergiya, Vol. 102, No. 1, pp. 35–39, January, 2007.     

Calculation of the Reliability Indicators of the System Protected Object–Control and Protection System

A mathematical model is proposed for the reliability of an automated technological system consisting of a protected object, an automatic control system, and a safety system. The model is a superposition of alternating renewal processes. Relations are obtained for the probability of the first accident in the interval [0, t] and the probability of an accident both for nonrenewable and renewable components of the automated technological system. A study of the asymptotic properties of the mathematical model made it possible to write down the indicated reliability indicators in a simple form, specifically, in terms of stationary coefficients of readiness of the system components. Particular cases are examined, and the corresponding relations are presented for them.     

Probabilistic assessment of seismic risk for nuclear power plants

Probabilistic approaches to the design, siting, and safety analysis of nuclear power plants have been proposed by Farmer, Wall, and Garrick. Farmer and Wall established a limit line which delineates between acceptable and unacceptable risks. To implement the method, all accidental chains are systematically analyzed to determine their probability and associated fission product release magnitude; the combination is compared to the limit line. For proper implementation, the seismic risk should be evaluated in a quantified manner. Conceptually, this evaluation is made in two stages: the probability of an earthquake occurrence as a function of its intensity and, given a seismic intensity, the conditional probability of damage. This paper reports on an initial study into the latter aspect.The effect of uncertainty in several parameters which determine the response of a nuclear reactor building to earthquake forces is assessed. Probability distributions for material properties were determined from site measurements and these distributions were utilized for determining the building response and the damage criterion. A subjective probability density function for damping was assigned from the available information and the judgment of experienced engineers. Four accelerograms, El Centro N---S 1940, and three artificial earthquakes were used to represent the variability in the forcing functions. The uncertainty in the model idealization was assessed by evaluating three alternate models. A versatile computer program was developed to compute the response of the mathematical model to the forcing functions using matrix formulation and modal method of analysis. An exact solution, rather than numerical integration, was used to obtain the dynamic response of the system in generalized coordinates.The stresses within the reactor building are similar for different earthquakes considered in this study when they are normalized to ground acceleration, indicating that the shape of the accelerogram and its frequency content are less significant than the magnitude of the maximum ground acceleration for the reactor building. The variation in modulus of elasticity for concrete had a significant effect on the building response. Damping, in general, reduced the response, but in cases where the duration of an earthquake is short the effect was not very significant.A simple failure criteria for ultimate shear stress in shear walls, τ_ult = 4.75 √ƒ′_c, where ƒ′_c is the ultimate compressive strength of concrete, is used to estimate the initiation of cracking in the walls. The normal design of the reactor building is deterministic and is based on a 0.2 g design basis earthquake. Using the results obtained by the parametric study on the variation of response, the probability of damage was estimated by a Monte Carlo analysis. It was estimated that, given the occurrence of a design basis earthquake, there is less than approximately 10⁻³ probability of cracking in the shear walls of the reactor building. The initiation of cracking in the concrete should not lead to a significant release of contained fission products.     

核燃料元件制造厂UF_6气化工序风险分析

多种危险并存于核燃料元件制造厂,因此有必要对核燃料元件厂进行风险分析。目前有多种风险评估方法适用于核燃料元件制造厂风险评估,本文选取HAZOP和LOPA方法,对核燃料元件制造厂风险评估中的最重要工艺UF6气化工序进行了分析。HAZOP分析得到了可能产生严重后果的工艺偏离。LOPA分析得到了针对工艺偏离所采取的独立保护层措施所降低的风险和UF6气化工序的残余风险。     

Verification of rod safety via confidence intervals for a binomial proportion

The probabilistic safety assessed to a set of N fuel rods assembled in one core of a nuclear power reactor is commonly modelled by ∑_i≤N X_i, where X₁, …, X_N are independent Bernoulli random variables (rv) with individual probability p_i = P (X_i = 1) that the ith rod shows no failure during one cycle. This is the probability of the event that the ith rod will not exceed the failure limit during one cycle. The safety standard presently set by the German Reaktor-Sicherheitskommission (Reactor Safety Commission) requires that the expected number of unfailed rods in the core during one cycle is at least N − 1, i.e., E(∑_i≤N X_i) = ∑_i≤N p_i ≥ N − 1, whereby a confidence level of 0.95 for the verification of this condition is demanded. In this paper, we provide an approach, based on the Clopper–Pearson confidence interval for the proportion p of a binomial B(n, p) distribution, how to verify this condition with a confidence level of at least 0.95. We extend our approach to the case, where the set of N fuel rods is arranged in strata, possibly due to different design in each stratum.