Failure Criteria for Unirradiated PWR Cladding Subjected to Ring Compression Tests

The ring compression test applied to nuclear fuel cladding is relatively easy to perform but difficult to interpret. It can be representative of the loading state associated to a hypothetical spent fuel assembly drop accident. This is particularly important for spent fuel cladding subjected to drying operations previous to storage and transportation, because they may produce hydride reorientation along the radial direction of cladding. In this paper, experimental testing and numerical simulations are combined to obtain operative failure criteria from the results of the ring compression tests on unirradiated pre-hydrided samples with radial hydrides, simulating drying, storage and subsequent transport conditions.

Computational Methods to Support Brittle Fracture Assessment of PSVs Involving Autorefrigeration

Pressure safety relief valve (PSV) operation generally leads to cooling of the valve itself and the piping connected to the PSV. The temperatures may reach values below the minimum design metal temperature (MDMT) of the valve, and therefore the valve needs to be assessed for brittle fracture susceptibility. Simplistic determination of the minimum metal temperature in the valve may disqualify these valves during the brittle fracture assessments (BFA). Replacement may be time consuming and may not be cost effective. In such circumstances, a sophisticated and more representative BFA approach involving the use of computational fluid dynamics (CFD) followed by finite element method (FEM) based stress analysis which may be further followed by fracture mechanics can be adopted based on the concepts defined in ASME/API 579. The accuracy of the BFA depends on the accuracy of each of the computational method involved in the assessment. Among all the computational methods, CFD poses significant challenge. The low temperature may have been caused due to Joule-Thompson effect or auto-refrigeration. While Joule-Thompson effect can be best captured with easy to implement and robust CFD methods, auto-refrigeration involving adiabatic flashing which causes additional complexity and requires multiple sensitivity studies performed to determine the accuracy of the CFD approach. In this paper, an overview of the computational methods used in the brittle fracture assessment of PSVs is presented. Specific CFD method details are provided for PSV involving the flashing of liquid hydrocarbon to vapor is presented in the form of a case study derived from downstream industry application.

Effective Outage Planning and Implementation for Nuclear Power Plants in the UK

Effective outage planning and implementation is critical to the efficient and safe operation of commercial nuclear power plants in the UK. Statutory outages are necessary for refuelling, for preventive and corrective maintenance when shutdown conditions are required, and for major modification and improvement projects. Outages involve the support of many companies and individuals working together and, as such, require high levels of coordination. Planning of activities before the outage is critical to the overall success of the outage. Establishing the integrity of power plant piping and pressure vessels is a key objective as part of any outage and the methodology and processes involved are the subject of this paper. Establishing the integrity of piping and pressure vessels requires an understanding of the specific threats, their relationship to the overall condition of the system, and the mitigating measures required to assure safe operation. Understanding the specific threats allows the engineering function of an organisation to advise on pipework and pressure vessel ‘Minimum Acceptable Thicknesses’ which can be used to assure integrity via comparison with thicknesses measured during outage inspections. Minimum Acceptable Thicknesses should be recorded in the outage management documentation so they are accessible during the outage implementation phase. Historically a variety of different methodologies have been used to advise on Minimum Acceptable Thickness requirements including design drawing specified minimum thicknesses, design code based required thicknesses and thicknesses calculated based on Fitness for Purpose methods. It is important that a robust procedure be applied to promote consistency of approach as regards the calculation of pipework and pressure vessel Minimum Acceptable Thickness requirements across all power station assets. An additional consideration is that of ensuring that the approach adopted is consistent with high level safety case guidance, i.e., the assessment is appropriate for the failure tolerability of the plant item. This paper provides an overview of the strategy, methodologies and processes employed to determine Minimum Acceptable Thicknesses for pipework components. These ensure that, over a specified inspection interval, were the weld/component to be defect free, it would not fail due to any of the relevant failure mechanisms, which typically are plastic collapse, creep rupture, fatigue, incremental collapse (ratcheting) or buckling. Readers of this paper will gain a valuable insight into the statutory outage process applicable to nuclear power plants in the UK. A particular focus of this paper is on the structural integrity assessments applied in a non-traditional sense prior to, during and after the statutory outage. As well as sharing a valuable insight into the assessment methodologies this paper highlights best industrial practice.

Application of Visual Beam Ultrasonic Phased Array in Inspection of Weld in Tubular Node

Responding to complexity and particularity welding on the geometry of TKY tubular node, this paper constructs mathematical model of tubular joint weld of arbitrary section by simplifying the geometry structure, and draws welded joints and ultrasonic sound beam lines based on the actual specifications in order to solve the problems of low efficiency, positioning difficulty, missing inspection and etc. The computeraided simulation technology can realize the visualization in the beam coverage model of welded joints, which can commendably guide the design of ultrasonic phased array inspection and overcome the blindness of the instrument detection parameters, thus improving the effectiveness and pertinence of the actual detection. Study shows that it is beneficial to enhance the effectiveness of the detection tubular joint weld by employment of Visual beam and ultrasonic phased array technology.

Design Procedure of Spring Washers for Flanged Joints

Flanged joints may experience leaks due to rapid increases or decreases in process temperatures. The mode of failure of a flanged joint during a rapid temperature increase is usually manifested in the form of gasket crushing or bolt rupture. On the other hand, the mode of failure of a flanged joint during a rapid cooling event occurs as bolt loosening causing the bolts to lose the necessary pretention required for joint tightness. In the present work, a systematic procedure has been proposed to design/select spring washers to prevent gasket leaks during thermal upsets of flanged joints. The procedure is based on the maximum allowable bolt load which is equal to the minimum of the load that will overstress the bolt, the load that will overstress the flange, or the load that will crush the gasket. The procedure considers the minimum allowable bolt load which is equal to the maximum of the required bolt load for operating conditions and required bolt load for gasket seating conditions. Moreover, the proposed procedure considers the design for rapid heating or rapid cooling events of flanged joints. The proposed procedure is based on the load-deflection characteristics of spring washer assemblies (series, parallel and series-parallel arrangements). The procedure has been applied successfully to design spring washer assemblies for flanged joints that suffer critical temperature excursions.

Determination of High-Temperature Creep Property of High-Cr Steel Based Upon Indentation Test

In this paper, the procedure which can estimate creep exponent and coefficient in Norton’s law from the impression size rather than the penetration depth is discussed based upon a high-temperature creep indentation test. Firstly, an analytical solution related to the change in impression size with dwelling time at an indentation load is formulated by solving problem of infinite creeping media embedding spherical cavity subjected to an inner pressure which characterizes an indentation load. The applicability of the formula to elastic-plastic-creeping model resembling an actual response is checked by conducting non-linear finite-element analysis combined with contact option. Finally, creep indentation tests are conducted for a high-Cr ferritic heat-resisting steel. It is shown that the creep parameters at a lower stress level can be estimated at temperature 873K.

Unit Dose Factors for Transportation of Radioactive Materials

It will be necessary in the future to transport spent nuclear fuel on a large-scale basis from nuclear power plant sites to interim storage and/or a repository. Shipments of radioactive material are required to comply with regulations limiting the dose rate to no more than 0.1 mSv (10 mrem) per hour at 2 meters from the sides of the vehicle transporting the package. Determining the resulting dose to the public will be necessary for a number of reasons (e.g., stakeholder concerns, environmental impact statements). In order to understand the dose consequence of such a transportation system, this paper describes a method for determining unit dose factors. These are defined as the dose to the public per unit distance traveled along a road, rail, or waterway from one shipment assuming unit values for the other route specific parameters. The actual dose to the public is calculated using unit dose factors, the dose rate due to the radiation field emanating from the package, and characteristics of the route itself. Route specific parameters include the speed of the conveyance, the population density, and characteristics of the environment surrounding the route; these are provided by a routing tool. Using these unit dose factors, in conjunction with a routing tool, it will be possible to quantify the collective dose to the public and understand the ramifications of choosing specific routes.

Fatigue Monitoring in the Context of Long-Term Operation of the Goesgen Nuclear Power Plant Using AREVA’s FAMOSi

In the context of long-term operation or lifetime extension most regulatory bodies demand from utilities and operators of nuclear power plants to monitor and evaluate the fatigue of system, structures and components systematically. As does the Swiss Federal Nuclear Safety Inspectorate ENSI. The nuclear power plant Goesgen started its commercial operation in 1979 and will go into long-term operation in 2019. The increased demand for monitoring and evaluating fatigue due to the pending long-term operation led the Goesgen nuclear power plant to expand the scope of their surveillance and therefore to install AREVA’s fatigue monitoring system FAMOSi in the 2014 outage. The system consists of 39 measurement sections positioned at the primary circuit and the feed-water nozzles of the steam generators. The locations were chosen due to their sensitivity for fatigue. The installed FAMOSi system consists of a total of 173 thermocouples which were mounted in order to get the necessary input data for load evaluation. The advantage of FAMOSi is the possibility to obtain real data of transients near places with highest fatigue usage factors. Examples of steam generator feed-in during heating-up and cooling-down will be given. In addition, spray events before and after the installation of closed loop controlled spray valves will be compared. The measurements and the results of the load evaluation are not only of interest for internal use e.g. in regard to optimization of operation modes (e.g. load-following), but must also be reported to ENSI annually. In addition, by evaluation of stresses and determination of usage factors combined with an optimization of operation modes an early exchange of components can be avoided.

Design of an Experiment to Measure the Thermal Accommodation Coefficient Between Helium and Stainless-Steel in Concentric Cylinders

Heat transfer through a 1 mm gap between two concentric cylinders representing the gap between a fuel support basket and a canister is experimentally and numerically investigated. The objective of this work is to study rarefied gas heat transfer in a simple geometry, and to measure the thermal accommodation coefficient at the interface between stainless steel and rarefied helium. The thermal accommodation coefficient is used to characterize the interaction between gas molecules and wall at the molecular level. It is important to determine its value with precision for better determination of heat transfer at low pressure. The experimental procedure consists of measuring the temperature difference between the inner and outer cylinders as the pressure is decreased in the gap. By knowing the heat flux across the gap the thermal accommodation coefficient can be extracted from the theoretical expression relating the temperature difference to the radial heat flux. Three-dimensional simulations using the ANSYS/Fluent commercial code are conducted to assess on the design of the experimental apparatus. These simulations confirmed that the apparatus design is effective to study the heat transfer across rarefied gas and to determine the thermal accommodation coefficient for helium on stainless steel surface.