A statistical approach to the prediction of pressure tube fracture toughness

The fracture toughness of the zirconium alloy (Zr-2.5Nb) is an important parameter in determining the flaw tolerance for operation of pressure tubes in a nuclear reactor. Fracture toughness data have been generated by performing rising pressure burst tests on sections of pressure tubes removed from operating reactors. The test data were used to generate a lower-bound fracture toughness curve, which is used in defining the operational limits of pressure tubes. The paper presents a comprehensive statistical analysis of burst test data and develops a multivariate statistical model to relate toughness with material chemistry, mechanical properties, and operational history. The proposed model can be useful in predicting fracture toughness of specific in-service pressure tubes, thereby minimizing conservatism associated with a generic lower-bound approach.     

A consistent approach to severe accident management

The approach adopted for severe accident management (SAM) at the Loviisa nuclear power plant (in Finland) is presented and discussed. The approach includes a number of significant hardware changes and procedures that allow lowering of the lower head thermal insulation and neutron shield assembly, opening of the ice condenser doors, and spraying (externally) of the steel shell of the containment. It is expected that with these changes we can assure in-vessel debris coolability and retention, gradual burning of the hydrogen with good access to the ice condenser, and long term stabilization of the containment pressure, even in the absence of the residual heat removal system. Methodological aspects of demonstrating these SAM objectives, and the status of work in support of related quantifications (of key phenomena), are included in sufficient detail to provide an integrated perspective of the approach taken. The detailed quantifications, separately on each task, will follow, as respective research and quantification programs come to completion.     

A new approach to the use of genetic algorithms to solve the pressurized water reactor's fuel management optimization problem

A Genetic Algorithm (GA) based system, coupling the computer codes GENESIS 5.0 and ANC through the interface ALGER has been developed aiming at pressurized water reactor's (PWR) fuel management optimization. An innovative codification, the List Model (LM), has been incorporated into the system. LM avoids the use of heuristic crossover operators and only generates valid nonrepetitive loading patterns in the reactor core. The LM has been used to solve the Traveling Salesman Problem (TSP). The results got for a benchmark problem were very satisfactory, in terms of precision and computational costs. The GENESIS/ALGER/ANC system has been successfully tested in optimization studies for Angra 1 power plant reloads.     

A probabilistic model of wall thinning in CANDU feeders due to flow-accelerated corrosion

Wall thinning due to flow-accelerated corrosion (FAC) is a pervasive form of degradation in feeder pipes of the primary heat transport system of CANDU reactors. Prediction of the end-of-life of a feeder from wall thickness measurement data is confounded by the sampling and temporal uncertainties associated with the FAC degradation phenomenon. This paper presents a probabilistic model of wall thinning due to FAC, and calibrates it with a set of feeder wall thickness measurements obtained from a CANDU plant. The proposed model derives the feeder lifetime distribution, which is useful in developing optimum strategies for life-cycle management of the feeder system.     

A systematic approach to the radiation damage problem

To assess the suitability of a material for use as a core component in a fast reactor or for the first wall in a fusion reactor, it is necessary to know the irradiation damage behaviour of the material outside the usual materials testing data domain. In the present paper we propose a strategy based on a closely co-ordinated programme of experimental and theoretical research. The aim of this strategy is the systematic construction of a physically based model of the evolving damage structures. This would then allow both the necessary extrapolations of the data to the desired conditions to be achieved in a reliable fashion and provide a rational basis for the development of low-swelling alloys for the two nuclear systems.     

A novel approach to correlation reflectometry

The use of microwave synthesized sources improved significantly the performance of correlation reflectometry diagnostics for fusion plasmas providing broadband operation with high quality signals and fast frequency switching times. Recent new developments of those sources made at IPFN opened an all-new way to build correlation reflectometry diagnostics with improved performance and measuring capability. Some of the key features now attainable are independent probing and local oscillator signals generated by separated frequency synthesizers; very good frequency control and synchronization of both the local and the radio-frequency oscillators providing very high stability. The technological advances make it possible to further reduce the switching time between adjacent probing frequencies while keeping the ability to operate in a broad frequency range.Here we present a multichannel correlation system based on those novel technologies using several independent and self-contained fully synthesized broadband channels. With this type of system various type of measurements are possible: (i) fast stepwise radial scans to extract the radial profile of turbulence parameters (spectra, RMS, rotation from Doppler shift, etc.); (ii) radial scans of the plasma using two identical systems in parallel where one system operates in frequency steps and the other makes multiple small steps around each selected frequency of the first system. The proposed diagnostic is very powerful and versatile. With a system having n identical channels, all transmitting and receiving simultaneously, the number of simultaneous correlations measurements that can be performed is given by n × (n ? 1)/2. The minimal configuration uses two channels thus providing one (time domain) correlation signal channels can be easily added at any time increasing the measuring capability of the diagnostic. Several examples are given that illustrate the wide range of measurements that can be made using the novel approach for correlation reflectometry diagnostics.     

A preliminary approach to the ALFRED reactor control strategy

In this paper, a preliminary approach to the definition of a suitable control strategy for the Advanced Lead Fast Reactor European Demonstrator (ALFRED), developed within the European 7th Framework Program, has been undertaken. The Generation IV reactors offer new challenges for what concerns the nuclear power plant control since several constraints both on primary and secondary loops have to be faced, differently from the conventional Light Water Reactors. A simulator of the ALFRED plant has been developed in a previous work (Ponciroli et al., 2014) with the main purpose of studying the system free dynamics and stability features in a control-oriented perspective. Based on the outcomes of these investigations, in the present work, the possibility of adopting decentralized control schemes has been investigated. Accordingly, Single Input Single Output control laws have been applied directly to the selected couples of input–output variables, which have been identified first on the basis of the preliminary plant dynamics analyses, and then confirmed by the indications of the Relative Gain Array method. Afterwards, two different control schemes have been studied depending on the number of available inputs, and then implemented and compared in order to evaluate the effect of each control action on the associated potential control strategy effectiveness. As a last step, the ALFRED control system has been finalized. The regulator design has been set up based on a simultaneous feedforward-feedback scheme incorporating four closed feedback loops. A controlled power reduction and a controlled overpower transient have been simulated in order to assess the performance of the two proposed control schemes. Results show that both the adopted control strategies can assure an efficient control of the thermal power while guaranteeing an effective control of lead and steam temperatures as well. In addition, some non-negligible differences between the two schemes have been observed and discussed in the simulation results of control and controlled variables.     

Effect of length of thinning area on the failure behavior of carbon steel pipe containing a defect of wall thinning

The present study performed pipe failure tests using 102 mm-Sch. 80 carbon steel pipe with various simulated wall thinning defects, to investigate the effect of axial length of wall thinning and internal pressure on the failure behavior of pipe thinned by flow accelerated corrosion (FAC). The tests were conducted under loading conditions of four-point bending with and without internal pressure. The results showed that a failure mode of pipe with a defect depended on the magnitude of internal pressure and axial thinning length as well as stress type and thinning depth and circumferential angle. Both load carrying capability (LCC) and deformation capability (DC) were depended on stress type in the thinning area and dimensions of thinning defect. For applying tensile stress to the thinned area, the dependence of LCC on the axial length of wall thinning was determined by circumferential thinning angle, and the DC was proportionally increased with increase in axial length of wall thinning regardless of the circumferential angle. For applying compressive stress to thinned area, however, the LCC was decreased with increase in axial length of the thinned area. Also, the effect of internal pressure on failure behavior was characterized by failure mode of thinned pipe, and it promoted crack occurrence and mitigated a local buckling of the thinned area.     

Canadian regulatory approach to ensuring the implementation of effective ageing management programs for nuclear power plants

Effective ageing management is a key element of the strategy for maintaining long-term safety and reliability of nuclear power plants (NPP). This paper presents the Canadian regulatory perspective on managing the ageing of CANDU NPP. This paper looks introspectively at the current level of ageing management effort at Canadian reactor sites, and discusses the need to increase efforts in order to account for the increasing effects of degradation mechanisms. It examines accepted international practice for ageing management, making reference to International Atomic Energy Agency (IAEA) suggested practice, and goes on to describe the overall Canadian regulatory approach to ageing management, including the development of regulatory requirements for ageing management programs. In conclusion, the paper indicates a path forward involving a proactive ageing management approach.     

Level 2 analysis to estimate LERF risk from a thermally induced rupture of a steam generator tube defect

This paper outlines the Level 2 portion of a methodology for determining the incremental induced steam generator tube rupture large early release fraction caused by an actual through-wall defect. This defect was responsible for the minor steam generator tube leak that occurred in September 2002 at the Comanche Peak Steam Electric Station Unit 1. In order to quantify the performance of the defect over the operating cycle, a range of defect lengths were input to the PROBFAIL computer code [Kenton, M., 2001. PROBFAIL: A Computer Code for Evaluating the Likelihood of Steam Generator Tube Rupture in Severe Nuclear Power Plant Accidents, CREARE TM-2138], using appropriate boundary conditions derived from MAAP4 [Henry, R., et al., May 1994. MAAP4—Modular Accident Analysis Program for LWR Power Plants, Computer Code Manual, EPRI Research Project 3131-02] runs. From the analysis of the calculated times of burst for each assumed defect length, the minimum through-wall defect length necessary for tube burst to occur prior to hot leg or surge line creep rupture was calculated. The probability that the defect would actually have this length was then estimated by determining the fraction of the cycle for which the defect would be at least that long. The methodology development and implementation relied on MAAP4 runs, which are discussed extensively in connection with their role in: (1) guiding the construction of the accident progression event tree, (2) generating relevant information for probability assignments in the various underlying fault trees and (3) obtaining boundary conditions of pressure and temperature for use in PROBFAIL. The overall increment in LERF due to the existence of the defect was calculated to be approximately 4E−08.     

A probability approach to estimating the safety of a reactor installation

Krasnaya Zvezda Scientific-Designer Bureau. Central Institute of Atomic Energy, All-Union Research Institute of Atomic Energy. Translated from Atomnaya Énergiya, Vol. 74, No. 1, pp. 38–42, January, 1993.     

A finite transform approach to the solution of radiation transport equation

Starting from the integral form of transport equation and defining a new transform G(x,μ,μ_m), the integration over space is carried out analytically resulting in an integral equation for F in μ. Use of Legendre polynomial for the solution of this equation is explored. The method is applicable to practical problems of radiation transport in multiregion, energy dependent systems with arbitrary degree of anisotropy. Its simplification for idealised systems and comparison with earlier work are indicated.     

A modular approach to lead-cooled reactors modelling

After an overview of the lego plant simulation tools (LegoPST), the paper gives some details about the ongoing LegoPST extension for modelling lead fast reactor plants. It refers to a simple mathematical model of the liquid lead channel dynamic process and shows the preliminary results of its application in dynamic simulation of the BREST 300 liquid lead steam generator. Steady state results agree with reference data [IAEA-TECDOC 1531, Fast Reactor Database, 2006 Update] both for water and lead.     

A reliability program approach to operational safety

A Reliability Program (RP) model based on proven reliability techniques is being formulated for potential application in the nuclear power industry. Methods employed under NASA and military direction, commercial airline and related FAA programs were surveyed and a review of current nuclear risk-dominant issues conducted. The need for a reliability approach to address dependent system failures, operating and emergency procedures and human performance, and develop a plant-specific performance data base for safety decision making is demonstrated.Current research has concentrated on developing a Reliability Program approach for the operating phase of a nuclear plant's lifecycle. The approach incorporates performance monitoring and evaluation activities with dedicated tasks that integrate these activities with operation, surveillance, and maintenance of the plant. The detection, root-cause evaluation and before-the-fact correction of incipient or actual systems failures as a mechanism for maintaining plant safety is a major objective of the Reliability Program.Embodied within the approach are (1) determination of acceptable safety system performance criteria and associated alert levels: (2) tracing and/or trending of in-plant and industry systems performance and management of the associated surveillance, maintenance, and reportable event data base; (3) determination of risk-importance prioritized systems, components, and root-causes and ad hoc response to inplant safety problems or potentially applicable industry problems identified by NRC or INPO; and (4) a closed-loop failure reporting and corrective action program for correcting performance criteria violations or identified problems either through changes in operation or maintenance or through changes in utility management practices.     

A thin film approach to protein crystallography

A novel approach for the investigation of proteins or macromolecules is outlined in this conceptual study. The preparation of grapho-epitaxial layers on nanotemplated substrates is proposed as an alternative to the preparation of single crystals by vapour diffusion techniques. Crystal structure investigations of such layers may then be performed via grazing-incidence diffraction (GIXRD) in the Laue mode. Quantitative expressions for the position and intensities of XRD peaks in this geometry are presented that fully consider the effects of refraction and absorption. A simulation of the Laue-GIXRD pattern of single-crystalline layers of Concanavalin A is given. The main challenges of the approach are concluded to relate to the preparation of single-crystalline protein layers. However, if those obstacles could be overcome, a 10–100-fold faster sample throughput would become possible.