Design basis accidents and accident analysis with particular reference to offshore platforms

The individual channel monitoring (ICM) proposal to improve the safety performance of RBMK

The purpose of this review is to determine the adequacy of the High Flux Beam Reactor (HFBR) Technical Specifications for 40 MW operation by comparison with the HFBR Final Safety Analysis Report, particularly the accident analyses chapter. Specifically, the Technical Specifications were compared against the Design Basis Accident (DBA) Analyses presented in the Addendum to the HFBR FSAR for 60 MW Operation. The 60 MW DBA analyses was used since it is more current and complete than the analyses presented in the original FSAR which is considered obsolete. A listing of the required systems and equipment was made for each of the accidents analyzed. Additionally, the Technical Specification instrument setpoints were compared to the DBA analyses parametric values. Also included in this review was a comparison of the Technical Specification Bases against the FSAR and the identification of any differences. The HFBR Operations Procedures Manual (OPM) was also reviewed for any inconsistencies between the FSAR or the Technical Specifications. Upon completion of this review it was determined that the Technical Specifications are well written and the items commented on should not delay the low power restart (40 MW). Additionally, the OPM is also well written and does not require further modification before restart.

Tubular joints are integral components of offshore structures and as such, many codes and standards address their design. The fatigue damage process of tubular joints in offshore structures is inherently variable and uncertain. In addition, potentially, there are important bias and uncertainties introduced by the analytical models used in design codes and standards. Therefore, in the development of reliability based fatigue design and requalification criteria, it is important to recognize these differences and to develop the best estimates of joint fatigue durability. This paper summarizes a study to identify the primary sources of bias and uncertainty associated with assessment of fatigue lives of tubular joints in offshore platforms. Based on the extensive screened test data, the important uncertainty sources of tubular joint fatigue durability has been examined, and the uncertainty models are developed to facilitate development of reliability based fatigue design and requalification criteria and fatigue inspection guidelines for tubular joints in offshore platforms. Introduction The development of the reliability based offshore platform design and requalification guidelines involves defining the scope (class of structure, load cases, etc.) and code format - including definition of characteristics values; selecting reliability measures; assessing uncertainties in loads and resistance; establishing target reliability levels; and determination of factors of safety, and load and resistance factors. Tubular joints are one of the main technical problems and high cost areas in the design, construction and maintenance of steel, template-type offshore structures. The uncertainty assessment of the tubular joints is therefore one of the most important components in the development of the reliability based design and requalification criteria for offshore platforms. The nature of uncertainties of the joint capacity and durability and methods for their modeling and analysis is a key in reliability based structural design of offshore platforms. Similarly, it is a key in development of realistic inspection and maintenance programs for these structures. In the context of development of criteria for design and requalification of platforms1 and floating facilities2 in the Bay of Campeche and development of inspection and maintenance programs for these facilities3,4, this study was intended to address a paradox that had developed from inspections of platforms in the Bay of Campeche.5 This paradox was that when inspections were made of joints in older platforms that were evaluated using detailed fatigue analysis methods6 to have fatigue lives that were in the range of months to years, no cracking could be detected. However, damage was found at other joints where it was not predicted or expected.5 This raised the question of: what was the confidence with the fatigue analyses? The objective of this paper is to summarize development of uncertainty models of the fatigue strength of tubular joints. Following a brief review of uncertainty of tubular joint fatigue durability, the sources of uncertainties in tubular joint fatigue durability are logically organized and quantitatively and qualitatively evaluated. Based on this evaluation, the uncertainty models are developed to facilitate development of reliability based fatigue design and requalification criteria, and guidelines for the inspection and maintenance of template-type offshore platforms.

Limitations of traditional tools for beyond design basis external hazard PRA

Thermal-hydraulic phenomena for water cooled nuclear reactors

Increased attention is being given to the analysis of anticipated plant transients and abnormal plant operating conditions. This is in contrast to past regulatory activities devoted primarily to the analysis of design basis accidents (DBA), such as the large-break loss-of-coolant accident (LOCA), which were designed to represent worst or bounding cases. This change has come about due to the realization that anticipated transients with multiple failures (including operator actions) are the major contributors to risk from reactor operations and that the low probability DBAs are relatively small contributors to risk. The accident at Three Mile Island Unit 2 (TMI-2) has shown that anticipated transients, combined with multiple equipment and/or operator failures, can result in significant core damage.

Various questions can be examined when discussing safety in general. Among these, some key issues are the attitude toward risk and its acceptance, the ways of identifying, analyzing, and quantifying risks, and societal factors and public opinion toward risks. The identification and quantification of risks are central in the regulatory framework and decision making and will be the focus of this article. Various approaches have been used for safety analysis over the years. This paper will survey some of the central attitudes in the nuclear reactor regulation philosophy and discuss the historical background surrounding them. Among these, we mention the “defense-in-depth” approach, the design basis accident (DBA), and beyond design basis accident (BDBA) analyses and discuss the rather subjective nature of their associated decision making. We maintain that it has long been recognized that the natural approach that comes out of the scientific method of inquiry is the probabilistic one, which in today's tools is conducted through the probabilistic safety analysis (PSA) method. This approach unlike the deterministic one, which produced concepts like DBA and defense-in-depth, enables us to put risks into context and to compare different risks such as those posed by different activities in particular or by other industries in general. It has consequently been gaining wide acceptance in regulatory bodies around the world as an effective tool in the inspection and regulation of nuclear reactors. Yet, it is also recognized that despite significant development over the past few decades, PSA still suffers from some well-known deficiencies. Its main benefit at this point is its contribution to identification and prioritization of design features, maintenance, management, and quality assurance (QA) important to safety. PSA has mostly been used in the nuclear power industry, but in recent years it has also started to be incorporated in research reactor (RR) safety analysis, and we therefore cover the subject of PSA usage for this purpose as well.

This article, written by Assistant Technology Editor Karen Bybee, contains highlights of paper OTC 20208, "Sakhalin-1: Technology Development for Frontier Arctic Projects," by J.M. Hamilton, ExxonMobil Upstream Re - search, and J.E. Jones, ExxonMobil Development Company, originally prepared for the 2009 Offshore Technology Conference, Houston, 4–7 May. The paper has not been peer reviewed. As the oil and gas industry pushes into frontier Arctic areas, new technologies and design criteria will be required to ensure safety and cost effectiveness of increasingly challenging developments. Sakhalin-1 provides a good example of a development in a frontier subarctic area where location-specific design criteria had to be developed and innovative-technology solutions were required to deal with demanding environmental conditions including sea ice, large breaking waves, and earthquakes. Introduction First production from the Exxon Neftegas Limited (ENL) Sakhalin-1 Phase-1 development of Chayvo field began in October 2005, marking the culmination of nearly 10 years of design-criteria and technology development. The development concept consists of an offshore drilling platform, an onshore drilling and treating facility at Chayvo, an oil-export pipeline that runs east/west across the island and Tatar Strait to mainland Russia, and a storage and tanker export terminal at DeKastri. The oil is exported by ice-strengthened tankers south through the Tatar Strait, which is partially covered by up to medium first-year ice (70 to 120 cm) in the winter. The offshore drilling platform, Orlan, is approximately 10 km offshore in approximately 15-m water depth. The platform annually is exposed to approximately 4000 km of drifting first-year pack ice up to 1.5-m (level ice) thickness, which originates in the Sea of Okhotsk north of the island and drifts south along the eastern coast. The controlling ice load features are pressure ridges, with keels extending to depths greater than 30 m (2000-year design value). The design ice load on the 80-m-wide, vertical-sided structure is approximately 385 MN. Design wave loads are generated by late-summer and fall storms that can produce breaking waves at the platform. Global wave loads are on the order of the ice loads, and wave run-up and locally high slamming pressures represent design challenges. Sakhalin Island seismicity is moderate to high, with the active Piltun fault system running north/south on the island as close as 10 km from onshore facilities and 20 km from the offshore platform. The 1995 Neftegorsk earthquake, which registered moment magnitude 7.1, had its epicenter approximately 50 km from the Sakhalin-1 facilities. All facilities must be designed for earthquakes, and the export pipeline, which crosses multiple faults, must be able to withstand several meters of fault displacement.

In support of the analyses for the Rivne Nuclear Power Plant (RNPP) VVER-440/213 (Ukraine) Safety Analysis Report (SAR), detailed MELCOR and CONTAIN models of the containment were developed. The RNPP containment features a bubble condenser tower with air locks and active and passive spray systems. Code input models were developed to accurately represent the containment volumes, room interconnections, structural masses, and the engineering safety features. Although MELCOR 1.8.3 [1] was the primary tool for the SAR containment analysis, comparison calculations were performed using CONTAIN Version 1.12 [2]. Consequently, both the response of the VVER-440 containment to limiting design conditions as well as a comparison of the two codes is presented. In the context of SAR requirements, the present application was performed for design basis accidents with conservative assumptions to compare the containment temperature and pressure with design criteria. The peak containment pressure and temperature were evaluated using the most intensive release of the primary and secondary coolant into the hermetic compartments, in particular, for the large break loss of coolant accident and main steam line break. Conservative coolant release data were evaluated using the RELAP5/Mod3.2 SAR model. The selection of the accident scenario, initial and boundary conditions, and the major results are presented. The results of the analyses will be included in the design basis accident analysis chapter of the RNPP SAR.

Companies involved in the nuclear energy domain, like component and platform manufacturers, system integrators, and utilities, have well-established yearly trainings on Nuclear Safety Culture. These trainings are typically covered as part of the annual quality assurance-related refresher trainings, introductory courses for new employees, or indoctrinations of temporary staff. Gradually, security awareness trainings are also addressed on a regular basis, typically with a focus on information technology, the daily office work, test bay, or construction site work environment, and some data protection and privacy-related topics. Due to emerging national nuclear regulation, steadily but surely, specialized cybersecurity trainings are foreseen for integrators and utilities. Beyond these safety, physical security and cybersecurity specific trainings, there is a need to address the joint part of these disciplines, starting from the planning phase of a new nuclear power plant (NPP). The engineers working on safety, physical protection, and cybersecurity must be aware of these interrelations to jointly elaborate a robust instrumentation and control architecture (defense-in-depth, design basis events, functional categorization and systems classification) and a resilient security architecture (security by design, security grading, zone model or infrastructure domain, security conduits, forensic readiness, security information, and event management). This paper provides more in-depth justification of when and where additional training is needed, due to the ubiquitous deployment of digital technology in new NPPs. Additionally, for existing NPPs, the benefits of conveying knowledge by training on specific interfaces between the involved disciplines will be discussed. Furthermore, the paper will address the need of focused training of management stakeholders, as eventually, they must agree on the residual risk. The decision-makers are in charge of facilitating the interdisciplinary cooperation in parallel to the allocation of resources, e.g., on security certifications of products, extended modeling-based safety and security analyses and security testing coverage.

This paper presents a procedure a coupling of the finite element method and normal mode technique for obtaining the elastic dynamic response of offshore platforms subjected to arbitrary time and space-dependent forces. The finite element method deals with the complexities of the actual problem by reducing it to an assembly of structural elements, with known elastic and inertial properties, connected together at a discrete number of points on their boundaries. The normal mode technique is then used to reduce the resulting matrix equation of order N to the solution of N single degree-of-freedom equations with certain constraints on the assumed form of damping. An example analysis, utilizing this procedure and an actual offshore platform under the action of two different types of dynamic forces, is presented. These forcing functions are: wave forces as determined through the use of a fifth-order gravity-wave theory and seismic forces as generated by the north-south component of the May 18, 1940, El Centro, California, accelerogram. The platform is assumed to respond according to the assumptions associated with the Bernoulli-Euler beam and bar theory. INTRODUCTION In the field of offshore platform analysis and design, as in many other fields, the tools of analysis are derived through an evolutionary process. As knowledge and ability increase for more accurate prediction of the environmental forces on platforms, the methods of analysis must improve to maintain consistent accuracy. The American Petroleum Institute(l) in its Recommended Practice for Planning, Designing, and Constructing Fixed Offshore Platforms recognizes this evolutionary process and encourages designers of offshore platforms to use "all research advances available to them" The advent of high-capacity, high-speed digital computers has revolutionized the field of structural dynamic analysis. In the past the approach was to reduce the structural model to the simplest form, usually an "equivalent"single-or dual-degree-of-freedom system or "equivalent" beam or bar model for which known closed-form solutions to the dynamic equations were available. While this approach gave valuable insights into the nature of the dynamic response problem, it did not give the detailed elastic response information that is possible today. This paper presents a method which the authors believe is the next step in the evolutionary process of offshore-platform dynamic analysis. It is a combination of the finite element method and the normal mode technique. The finite element method more accurately takes into account the geometry of the system than the so-called "equivalent" methods. Although the finite element method and the normal 'mode technique have been known for a number of years, it is only with advanced computer technology that these approaches have been economically feasible for the analysis of large complex structures. This economic feasibility, the increased accuracy of predictions of dynamic forces on platforms in the operating environment, and the desire of operators to construct offshore platforms in deeper water and use them under more hostile operating conditions, will force designers to use more advanced methods of dynamic analysis.

Since the Three Mile Island Unit 2 (TMI-2), accident in 1979 which led to the meltdown of about one half of the reactor core and to limited releases of radioactive materials to the environment, an important international effort has been made on severe accident research. The present work aims to investigate the behaviour of a Small Modular Reactor during severe accident conditions. In order to perform these analyses, a SMR has been studied for the European reference severe accident analysis code ASTEC, developed by IRSN and GRS. In the thesis will be described in detail the IRIS Small Modular Reactor; the reference reactor chosen to develop the ASTEC input deck. The IRIS model was developed in the framework of a research collaboration with the IRSN development team. In the thesis will be described systematically the creation of the ASTEC IRIS input deck: the nodalization scheme adopted, the solution used to simulate the passive safety systems and the strong interaction between the reactor vessel and the containment. The ASTEC SMR model will be tested against the RELAP-GOTHIC coupled code model, with respect to a Design Basis Accident, to evaluate the capability of the ASTEC code on reproducing correctly the behaviour of the nuclear system. Once the model has been validated, a severe accident scenario will be simulated and the obtained results along with the nuclear system response will be analysed.

In-place analysis for offshore platforms is essentially required to make proper design for new structures and true assessment for existing structures. The structural integrity of platform components under the maximum and minimum operating loads of environmental conditions is required for risk assessment and inspection plan development. In-place analyses have been executed to check that the structural member with all appurtenances robustness and capability to support the applied loads in either storm condition or operating condition. A nonlinear finite element analysis is adopted for the platform structure above the seabed and the pile-soil interaction to estimate the in-place behavior of a typical fixed offshore platform. The analysis includes interpretation of dynamic design parameters based on the available site-specific data, together with foundation design recommendations for in-place loading conditions. The SACS software is utilized to calculate the natural frequencies of the model and to obtain the response of platform joints according to in-place analysis then the stresses at selected members, as well as their nodal displacements. The directions of environmental loads and water depth variations have important effects on the results of the in-place analysis behavior. The result shows that the in-place analysis is quite crucial for safe design and operation of offshore platform and assessment for existing offshore structures.

Statistical analysis of ship accidents and review of safety level

The responsiveness of motor vehicle death rates of youths aged fifteen through twenty-four to variations in the cost of beers as reflected by differences in State excise tax rates on beer is investigated. Evidence for this age group is provided on the extent to which declining real beer excise taxes have contributed to increases in fatal motor vehicle crashes and on the extent to which increases in real beer taxes can serve as a potent instrument in the antidrinking campaign. The effect of an increase in the legal drinking age on youth motor vehicle deaths is also examined.