Reliability, Availability, Maintainability, Inspectability Research Articles

ITER Framework for RAMI Engineering

The systems engineering process starts with the discovery of the real issues that need to be resolved and the identification of failures that are the most probable or/and have the highest negative impact during the life cycle of a project. Systems engineering involves finding mitigations to these most critical problems. This logic is fully followed in reliability, availability, maintainability, inspectability (RAMI) engineering. Although this area is at its beginning in fusion technologies, a few years ago the ITER Organization developed an approach to assess the RAMI requirement of systems. As an example of what a RAMI analysis can bring to the maintainability and thus operational availability of a nuclear fusion facility like ITER, the availability of the cask and plug remote handling system in charge of handling of port plugs and their moving between the port cells to the hot cell facility is addressed in the case of diagnostic equatorial port plugs.

RAMI comparison of two China test blanket modules

Tritium breeding blanket module is one of the most important parts of fusion reactor. One of the main goals of ITER is to test and validate design concepts of tritium breeding blankets. China has proposed two test blanket module (TBM) concepts, including Helium cooled ceramic breeder test blanket module system (HCCB TBS) and the dual functional lithium-lead test blanket module system (DFLL TBS), which will be tested in one of the three equatorial ports of ITER. A detailed comparative study on the design schemes of these two TBSs concepts was performed in this paper. In this study, the RAMI (Reliability, Availability, Maintainability, Inspectability) approach was applied on those two test blanket systems. The modeling and analyzing process were done by the Reliability and Probabilistic Safety Assessment Software System RiskA. In addition, comparative study in the phases of functions, capabilities and efficiencies of tritium breeding, heat removal and coolant purification is also performed. It is found that HCCB TBS is a better choice currently to fulfill the engineering demand of ITER construction for its lower difficulty of technical implementation. However, DFLL TBS possesses a high performance prospect from the viewpoint of long-term ITER development due to its inherent advantages compared to HCCB TBS. Discussions of priorities for future research needed for these two TBS are also proposed in the paper.

Preliminary RAMI analysis of DFLL TBS for ITER

ITER is the first fusion machine fully designed to prove the physics and technological basis for next fusion power plants. Among the main technical objectives of ITER is to test and validate design concepts of tritium breeding blankets relevant to the fusion power plants. To achieve this goal, China has proposed the dual functional lithium-lead test blanket module (DFLL TBM) concept design. The DFLL TBM and its associated ancillary system were called DFLL TBS. The DFLL TBS play a key role in next fusion reactor. In order to ensure reliable and available of DFLL TBS, the risk control project of DFLL TBS has been put on the schedule. As the stage of the ITER technical risk control policy, the RAMI (Reliability, Availability, Maintainability, Inspectability) approach was used to control the technical risk of ITER. In this paper, the RAMI approach was performed on the conceptual design of DFLL TBS. A functional breakdown was prepared on DFLL TBS, and the system was divided into 3 main functions and 72 basic functions. Based on the result of functional breakdown of DFLL TBS, the reliability block diagrams were prepared to estimate the reliability and availability of each function under the stipulated operating conditions. The inherent availability of the DFLL TBS expected after implementation of mitigation actions was calculated to be 98.57% over 2 years based on the ITER reliability database. A Failure Modes Effects and Criticality Analysis (FMECA) was performed with criticality charts highlighting the risk level of the different failure modes with regard to their probability of occurrence and their effects on the availability.

Failure rate modeling using fault tree analysis and Bayesian network: DEMO pulsed operation turbine study case

Availability will play an important role in the Demonstration Power Plant (DEMO) success from an economic and safety perspective. Availability performance is commonly assessed by Reliability Availability Maintainability Inspectability (RAMI) analysis, strongly relying on the accurate definition of system components failure modes (FM) and failure rates (FR). Little component experience is available in fusion application, therefore requiring the adaptation of literature FR to fusion plant operating conditions, which may differ in several aspects. As a possible solution to this problem, a new methodology to extrapolate/estimate components failure rate under different operating conditions is presented. The DEMO Balance of Plant nuclear steam turbine component operated in pulse mode is considered as study case. The methodology moves from the definition of a fault tree taking into account failure modes possibly enhanced by pulsed operation. The fault tree is then translated into a Bayesian network. A statistical model for the turbine system failure rate in terms of subcomponents’ FR is hence obtained, allowing for sensitivity analyses on the structured mixture of literature and unknown FR data for which plausible value intervals are investigated to assess their impact on the whole turbine system FR. Finally, the impact of resulting turbine system FR on plant availability is assessed exploiting a Reliability Block Diagram (RBD) model for a typical secondary cooling system implementing a Rankine cycle. Mean inherent availability for a period of 20 years plant mission range from 79% to 82% when varying turbine subcomponents FR from 1E-4/h to 1E-8/h, respectively, and much lower than the 97% performance expected for steady state operation.

Reliability Data Analysis for Tritium Extraction System Pipes of CN HCCB TBM

A tritium extraction system (TES) is one of important ancillary systems of the China helium cooled ceramic breeder test blanket module (CN HCCB TBM). The high reliability and availability of TES will be needed for safety operation of circulation and processing of tritium purge gas. Probabilistic Safety Assessment (PSA) and reliability, availability, maintainability, inspectability (RAMI) analysis of the TES should be performed during the design phase according to the CN HCCB TBM plan. Accurate reliability data is required to provide meaningful results for PSA and RAMI analysis. Since there is no TES pipes failure rate data available from fusion operating experiences. Therefore, it is necessary to use the existing pipe operating experiences to calculate component failure rates of the TES pipes. In this paper, the k factor method was used to adjust the failure rates of the TES pipes under the different environments and conditions by combining physics of failure models and a conservative estimation technique. The failure rates of the TES can provide references for the optimization of TES pipes design and data support for PSA and RAMI analysis of the TES.

Results of the RAMI analyses performed for the IFMIF accelerator facility in the engineering design phase

This paper presents a summary of the RAMI (Reliability Availability Maintainability Inspectability) analyses done for the IFMIF (International Fusion Materials Irradiation Facility) Accelerator facility in the Engineering Design Phase. The methodology followed, the analyses performed, the results obtained and the conclusions drawn are described. Moreover, the consequences of the incorporation of the RAMI studies in the IFMIF design are presented and the main outcomes of these analyses are shown.

Reliability and safety analysis for systems of fusion device

Fusion energy or thermonuclear power is a promising, literally endless source of energy. Development of fusion power is still under investigation and experimental phase, and a number of fusion devices are under construction in Europe. Since fusion energy is innovative and fusion devices contain unique and expensive equipment, an issue of their reliability is very important from their efficiency perspective.A Reliability, Availability, Maintainability, Inspectability (RAMI) analysis is being performed or is going to be performed in the nearest future for such fusion devices as ITER and DEMO in order to ensure reliable and efficient operation for experiments (e.g., in ITER) or for energy production purposes (e.g., in DEMO). On the other hand, rich experience of the reliability and Probabilistic Safety Analysis (PSA) exists in nuclear industry for fission power plants and other nuclear installations.In this paper, the Wendelstein 7-X (W7-X) device is mainly considered. This stellarator device is in commissioning stage in the Max-Planck-Institut für Plasmaphysik, Greifswald, Germany (IPP). In the frame of cooperation between the IPP and the Lithuanian Energy Institute (LEI) under the European Fusion Development Agreement a pilot project of a reliability analysis of the W7-X systems was performed with a purpose to adopt Nuclear Power Plant (NPP) PSA experience for fusion device systems. During the project reliability and safety (risk) analysis of a Divertor Target Cooling Circuit, which is an important system for permanent and reliable operation of in-vessel components of the W7-X, was performed.Application of reliability and safety analysis for systems of fusion device enable to recommend and prioritize availability improvement measures for such systems and thus increase efficiency of fusion device. The methods and models described in this paper could be easily adopted for other systems of various fusion devices.

Technological and physics assessments on heating and current drive systems for DEMO

The physics requirements of the heating and current (H&CD) systems in a Demonstration Fusion Power Plant (DEMO) are often beyond the actual level of design maturity and technology readiness required. The recent EU fusion roadmap advocates a pragmatic approach and favours, for the initial design integration studies, systems to be as much as possible, extrapolated from the ITER experience. To reach the goal of demonstrating the production of electricity in DEMO with a closed fuel cycle by 2050, one must ensure reliability, availability, maintainability, inspectability (RAMI) as well as performance, efficiency and optimized design for the H&CD systems.In the recent Power Plant Physics & Technology (PPP&T) Work Programme, a number of H&CD studies were performed. The four H&CD systems Neutral Beam (NB) Injection, Electron Cyclotron (EC), Ion Cyclotron (IC) and Lower Hybrid (LH) were considered. First, a physics optimization study was made assuming all technologies are available and identifying which parameters are needed to optimize the performance for given plasma parameters. Separately, the (i) technological maturity was considered (e.g. 240GHz gyrotrons for EC) and (ii) technologies were adapted (e.g. multi-stage depressed collector for EC) or (iii) novel solutions (e.g. photo-neutralization for NB or new antennae concepts for IC) were studied to overcome the limitations of the present H&CD systems with respect to DEMO requirements. Further constraints imposed by remote maintenance or breeding blanket interactions were considered.

Application of probabilistic safety and reliability analysis for a system of fusion facility

Fusion or thermonuclear power is a promising, literally endless source of energy. Development of fusion power is still in investigation and experimentation phases and a number of fusion facilities are under construction in Europe. Since fusion energy is innovative and fusion facilities contain unique and expensive equipment an issue of their reliability is very important from their efficiency perspective. A Reliability, Availability, Maintainability, Inspectability (RAMI) Analysis is being performed or is going to be performed in the nearest future for such fusion facili­ties as ITER and DEMO in order to ensure reliable and efficient operation for experi­ments (e. g. in ITER) or for energy production purposes (e. g. in DEMO). On the other hand, rich experience of the Reliability and Probabilistic Safety Analysis (PSA) exists in nuclear industry for fission power plants and other nuclear installations. In this article, the Wendelstein 7-X (W7-X) facility is mainly considered. It is a stel­larator type fusion facility under construction in the Max-Planck-Institut fur Plasmaphysik, Greifswald, Germany (IPP). In the frame of cooperation between the IPP and the Lithuanian Energy Institute (LEI) under the European Fusion Development Agreement a pilot project of a reliability analysis of the W7-X systems was performed with a purpose to adopt NPP PSA experience for fusion facility systems. During the project a reliability analysis of a divertor target cooling circuit, which is an important system for permanent and reliable operation of in-vessel components of the W7-X, was performed.

Availability simulation software adaptation to the IFMIF accelerator facility RAMI analyses

Several problems were found when using generic reliability tools to perform RAMI (Reliability Availability Maintainability Inspectability) studies for the IFMIF (International Fusion Materials Irradiation Facility) accelerator. A dedicated simulation tool was necessary to model properly the complexity of the accelerator facility.AvailSim, the availability simulation software used for the International Linear Collider (ILC) became an excellent option to fulfill RAMI analyses needs. Nevertheless, this software needed to be adapted and modified to simulate the IFMIF accelerator facility in a useful way for the RAMI analyses in the current design phase. Furthermore, some improvements and new features have been added to the software. This software has become a great tool to simulate the peculiarities of the IFMIF accelerator facility allowing obtaining a realistic availability simulation. Degraded operation simulation and maintenance strategies are the main relevant features.In this paper, the necessity of this software, main modifications to improve it and its adaptation to IFMIF RAMI analysis are described. Moreover, first results obtained with AvailSim 2.0 and a comparison with previous results is shown.

On the problem of the elastic stability of a locally loaded cylindrical shell (Supplement)

The breeding blanket with integrated first wall (FW) is the key nuclear component for power extraction, tritium fuel sustainability, and radiation shielding in fusion reactors. The ITER device will address plasma burn physics and plasma support technology, but it does not have a breeding blanket. Current activities to develop “roadmaps” for realizing fusion power recognize the blanket/FW as one of the principal remaining challenges. Therefore, a central element of the current planning activities is focused on the question: what are the research and major facilities required to develop the blanket/FW to a level which enables the design, construction and successful operation of a fusion DEMO? The principal challenges in the development of the blanket/FW are: (1) the Fusion Nuclear Environment – a multiple-field environment (neutrons, heat/particle fluxes, magnetic field, etc.) with high magnitudes and steep gradients and transients; (2) Nuclear Heating in a large volume with sharp gradients – the nuclear heating drives most blanket phenomena, but accurate simulation of this nuclear heating can be done only in a DT-plasma based facility; and (3) Complex Configuration with blanket/first wall/divertor inside the vacuum vessel – the consequence is low fault tolerance and long repair/replacement time.These blanket/FW development challenges result in critical consequences: (a) non-fusion facilities (laboratory experiments) need to be substantial to simulate multiple fields/multiple effects and must be accompanied by extensive modeling; (b) results from non-fusion facilities will be limited and will not fully resolve key technical issues. A DT-plasma based fusion nuclear science facility (FNSF) is required to perform “multiple effects” and “integrated” experiments in the fusion nuclear environment; and (c) the Reliability/Availability/Maintainability/Inspectability (RAMI) of fusion nuclear components is a major challenge and is one of the primary reasons why the blanket/FW will pace fusion development toward a DEMO.This paper summarizes the top technical issues and elucidates the primary challenges in developing the blanket/first wall and identifies the key R&D needs in non-fusion and fusion facilities on the path to DEMO.