Jovica R. Riznic

The Logical Framework Approach-Millennium:

The Logical Framework Approach (LFA) has proved to be a valuable tool for project approval, design, and evaluation. However, a few pitfalls make it hard to use within todays project management framework and to integrate with other project management tools. This article proposes an updated version of the LFA to improve its compatibility with todays corporate culture, project management framework, and tools. We propose to call the updated tool the Logical Framework Approach-Millennium (LFA-M). The LFA-M is a seven-step approach leading to the development of the Logframe-Millennium (LF-M), a five-column and four-line matrix describing major project commitments and providing an overall understanding of the project. It was successfully implemented at the Canadian Space Agency and the Canadian Nuclear Safety Commission. The LFA-M fits well within todays project management framework and corporate culture and leads easily to other project management tools.

Evaluation of the Integrity of Steam Generator Tubes Subjected to Flow Induced Vibrations

Flow-induced vibrations (FIV) continue to affect the operations of nuclear power plant components such as heat exchanger tube bundles. The negative effect of FIV is in the form of tube fatigue, cracking, and fretting wear at the supports. Fretting wear at the supports is the result of tube/support impact and friction. Fluidelastic and turbulence forces are the two main excitation mechanisms that feed energy into the system causing these violent vibrations. To minimize this effect all support clearances must be kept at a very small value. This paper investigates the consequences of losing the effectiveness of a particular support as a result of corrosion or excessive fretting wear. A full U-bend tube subjected to both fluidelastic and turbulence forces is utilized in this work. The performance of countermeasures such as the installation of additional flat bars in the U-bend region is thoroughly investigated. The investigation utilized both deterministic and probabilistic techniques.

ASSESSMENT OF STEAM GENERATOR TUBE FLAW SIZE AND LEAK RATE MODELS

Abstract The Canadian Nuclear Safety Commission recently developed the CANTIA (CANDUTM Tube Inspection Assessment) methodology for probabilistic assessment of inspection strategies for steam generator (SG) tubes as a direct effect on the early detection and prevention of tube failure and primary-to-secondary leak of reactor coolant. In an effort to improve CANTIA, an SG tube integrity assessment code, a relevant survey of the literature on the discharge of subcooled water from cracks and critical flow models, SG tube cracks, leakage, and probabilistic assessment methodologies was carried out. The original CANTIA and ANL/CANTIA code models for the flaw opening area and flow leakage rate were reviewed. The predictions from the crack opening area and the leakage flow rate models were compared with experimental measured data from cracked SG tubes.

An Experimental Model Study of Steam Generator Tube Loading During a Sudden Depressurization

This paper presents the results of an experimental model study of a postulated Main Steam Line Break (MSLB) accident on the transient loading of steam generator tubes in a nuclear plant. The problem involves complex transient two-phase flow dynamics and fluid-structural loading processes. A better understanding of this phenomenon will permit the development of improved design tools to ensure steam generator safety. The pressure and temperature were measured upstream and downstream of the tube bundle and the transient tube loads were directly measured using dynamic piezoelectric load cells. High- speed videos were taken to observe and better understand the flow phenomena causing the tube loading. The working fluid was R-134a and the tube bundle was a normal triangular array with a pitch ratio of 1.36. The flow through the bundle was choked for the majority of the transient. The transient tube loading is explained in terms of the associated fluid mechanics. An empirical model is developed that enables the prediction of the maximum tube loads once the pressure drop is known.

Instrumentation Development and Validation for an Experimental Two-Phase Blowdown Facility

An experimental facility was designed and built to study the loading on steam generator tubes during a blowdown. The facility used refrigerant R-134a and measurements were taken for static and dynamic pressures as well as tube loading and temperatures. Commissioning experiments indicated that the off-the-shelf dynamic pressure transducers and load cells could not take the mechanical and thermal shock loading caused by the blowdown and produced spurious results of no value. This paper presents the instrumentation problems found, explains why they occurred, describes the remedial procedures employed, and outlines the instrumentation validation methodologies developed. The success of the instrumentation development is demonstrated in a series of experiments designed to assess the rapid transient measurement system.

Instrumentation Development and Validation for an Experimental Study of Steam Generator Tube Loading During Blowdown

A Main-Steam-Line-Break accident in a CANDU nuclear steam generator produces a blowdown in which all the pressurised water is boiled off in a few seconds. The resulting high transient loading on the heat exchanger tubing could lead to their rupture, resulting in the release of radioactive materials out of containment. A better understanding of this phenomenon will permit the development of improved design tools to ensure steam generator safety in the event of such an accident. The paper presents a commissioned experimental rig and instrumentation system, for which a two-phase experimental program has been developed. Using R134a as the working fluid, measurements of temperature, pressure, and tube loading, as well as simultaneous high-speed flow visualisation, have been taken at conditions simulating a full-scale operating steam generator. The experimental results will be used to develop theoretical modelling tools for single and two-phase blowdown, such that an estimate of tube loading during Main-Steam-Line Break can be predicted from initial conditions.Copyright

COMMISSIONING TESTS FOR AN EXPERIMENTAL STUDY OF STEAM GENERATOR TUBE LOADING DURING BLOWDOWN

The goal of this research is to improve our understanding of the effects of a postulated Main Steam-Line Break on the transient loading of nuclear steam generator tubes. The analysis of this problem deals with the complex coupling of rapid transient two-phase flow dynamics and fluid-structural loading processes. A main concern of nuclear reactor safety is to ensure that radioactive materials produced by nuclear fission are safely contained. This type of accident produces a ‘blowdown’ in which the pressurised water in the steam generator is boiled off in a few seconds. The resulting transient loading on the steam generator tubing could lead to their rupture, resulting in the release of radioactive materials out of containment. A better understanding of this phenomenon will permit the development of improved design tools to ensure steam generator safety in the event of such an accident. This paper presents a work in progress, describing the purpose-built experimental facility and a summary of commissioning results, including an evaluation of the instrumentation and data collection methodology. The final results of this research will provide physical insights and guidance for the development of predictive modelling tools.

Experimental and Theoretical Analysis of Subcooled Water Discharge Through Simulated Steam Generator Tube Cracks

The steam generator (SG) tubes represent a major fraction of the reactor primary coolant pressure boundary surface area in both Canadian pressurized heavy water reactor (CANDU) reactors and pressurized water reactor (PWR). There is very limited data on the steam generator tube leak rate measurement. Most studies of subcooled choking flow are related to long tubes with L/D greater than 15. Also, all of those data have a channel length greater than 10 mm, which is not indicative of steam generator tubing. Steam generator tubes have a wall thickness typically less than 3 mm. Experiments were conducted on choking flow for various simulated crack geometries for vessel pressures up to 7 MPa with various subcoolings. Measurements were done on subcooled flashing flow rate through well defined simulated crack geometries with L/D at ∼2 and 5–6. Both homogeneous equilibrium and non-equilibrium mechanistic models were developed to model two-phase choking flow through slits. A comparison of the model results with experimental data shows that the homogeneous equilibrium based models grossly under predict choking flow rates in such geometries, while homogeneous non-equilibrium models greatly increase the accuracy of the predictions.Copyright

Assessment of Choking Flow Models in RELAP5 for Flashing Flow Through Small Cracks

The estimation of leak rates through steam generator tube crack is an important safety parameter. An assessment of the choking flow models in thermal-hydraulics code RELAP5 is performed and its applicability to predict choking flow rates through steam generator tube cracks is addressed. A RELAP5 nodalization was created to model experimental data from literature. It is found that both the Henry-Fauske and Ransom-Trapp models better predict choking mass flux for longer channels. As the length of a channel decreases the both models’ predictions diverge from each other. While RELAP5 has been shown to predict choking flow in large scale geometries, it is not suited well for small channel lengths. In the case of a more conservative approach, where over prediction of mass flux through short channels is best, the Henry-Fauske model would be most appropriate.Copyright

Experiments on Choking Flow of Subcooled Liquid Through a Simulated Steam Generator Tube Crack

The ability to estimate the leak rates from the through wall cracks in a steam generator tube is important in terms of radiological source terms and overall operational management of steam generators as well as demonstration of the leak-before-break condition. A literature survey showed that there are few data sets available on crack geometries related to steam generator tubing. In this study an experimental program was developed to measure the choking flow rate of subcooled water through well defined simulated steam generator tube crack geometries with L/D < 5.5, and results are compared with models in literature. Two types of test specimens were used in the experimental program. One, a round orifice like hole is created to simulate a pitting type flaw. The others are laser cut slits representing axial cracks. A pressure differential across the crack/break of 6.8 MPa is achieved. As subcooling increasing the flashing discharge rate also increases. A modified Burnel model was identified which predicts experimentally observed choking discharge rates from such geometries well.Copyright