Integrating Safety, Operations, Security, and Safeguards Into the Design of Small Modular Reactors

Author(s):  
Bobby D. Middleton ◽  
Carmen Mendez

The existing regulatory structure for nuclear power plants impacts both the design and the operation of the facility [1]. The current structure has been known to be overly conservative in several instances. This overly conservative approach results in operational costs to the facility that decrease the profit margin for nuclear power companies. The current design and build process also results in expensive retrofitting and contributes excess costs to the operations of the facility [1]. The current fleet of nuclear reactors is composed mainly of large light water reactors (LWRs) that can, to some extent, counteract these operational costs by the sheer volume of energy produced. However, the deliberately small size of small modular reactors (SMRs) prevents them from benefitting from this economy of scale. In order to be built and operated economically, SMR vendors must find ways to bring the life cycle costs in line with the economic requirements of nuclear power companies. Sandia National Laboratories has developed a framework that allows vendors and operators to address many of the operational costs during the design and manufacture stages of the SMR life cycle. The framework allows certain operational costs to be addressed in the design stages, thereby decreasing the operational costs, especially those costs associated with staffing and retrofitting. The framework pulls together best practices that have been applied successfully in other industries. Concurrent Engineering (CE) frames the procedural stages, from defining the expectations of the facility deployment, through the identification of regulatory requirements, to the pre-conceptual, conceptual and detailed design stages. A Project Management Organization is critical to the time management and success of implementing CE. The use of Integrated Safety, Operations, Security, and Safeguards (ISOSS) will lead to achieve a more efficient, cost-effective, and reliable plant. The Balance Model is introduced as a tool to document conflicts between functional areas and identify balancing strategies for conflict resolution in the requirements. Life-Cycle Cost Analysis (LCCA) is proposed as a variable for decision making. Facility Lifecycle Management with Building Information Modeling (BIM) is encouraged to support the Build, Activation, Continued Operations and Decommissioning of the facility [1]. To ensure that the deployment of SMR is effective and cost efficient, the ideal time to implement the framework is now, before SMR designs reach the detailed stage. SMRs hold a lot of potential and this framework can help the nuclear industry realize that potential.

Author(s):  
Ronald C. Lippy

The nuclear industry is preparing for the licensing and construction of new nuclear power plants in the United States. Several new designs have been developed and approved, including the “traditional” reactor designs, the passive safe shutdown designs and the small modular reactors (SMRs). The American Society of Mechanical Engineers (ASME) provides specific Codes used to perform preservice inspection/testing and inservice inspection/testing for many of the components used in the new reactor designs. The U.S. Nuclear Regulatory Commission (NRC) reviews information provided by applicants related to inservice testing (IST) programs for Design Certifications and Combined Licenses (COLs) under Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” in Title 10 of the Code of Federal Regulations (10 CFR Part 52) (Reference 1). The 2012 Edition of the ASME OM Code defines a post-2000 plant as a nuclear power plant that was issued (or will be issued) its construction permit, or combined license for construction and operation, by the applicable regulatory authority on or following January 1, 2000. The New Reactors OM Code (NROMC) Task Group (TG) of the ASME Code for Operation and Maintenance of Nuclear Power Plants (NROMC TG) is assigned the task of ensuring that the preservice testing (PST) and IST provisions in the ASME OM Code to address pumps, valves, and dynamic restraints (snubbers) in post-2000 nuclear power plants are adequate to provide reasonable assurance that the components will operate as needed when called upon. Currently, the NROMC TG is preparing proposed guidance for the treatment of active pumps, valves, and dynamic restraints with high safety significance in non-safety systems in passive post-2000 reactors including SMRs.


Author(s):  
Ronald Lippy

The nuclear industry is preparing for the licensing and construction of new nuclear power plants in the United States. Several new designs have been developed, including more traditional evolutionary designs, passive reactor designs, and small modular reactors (SMRs). ASME (formerly the American Society of Mechanical Engineers) provides specific codes used to perform inspections and testing, both preservice and inservice, for many of the components used in the new reactor designs. The U.S. Nuclear Regulatory Commission (NRC) reviews information provided by applicants related to inservice testing (IST) programs for design certification (DC) and combined license (COL) applications under Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” of Title 10, “Energy,” of the Code of Federal Regulations (10 CFR Part 52) (Reference 1). The 2012 Edition of the ASME OM Code, Operation and Maintenance of Nuclear Power Plants, defines a post-2000 plant as a nuclear power plant that was issued (or will be issued) its construction permit, or combined license for construction and operation, by the applicable regulatory authority on or after January 1, 2000. The ASME New Reactors OM Code (NROMC) Task Group (TG) is assigned the task of ensuring that the preservice testing (PST) and inservice testing (IST) provisions in the ASME OM Code are adequate to provide reasonable assurance that pumps, valves, and dynamic restraints (snubbers) for post-2000 plants will operate when needed. Currently, the NROMC TG is preparing proposed guidance for the treatment of active pumps, valves, and dynamic restraints with high safety significance in nonsafety systems for passive post-2000 plants, including SMRs. (Note: For purposes of this paper, “post-2000 plant” and “new reactor” are used interchangeably throughout.) Paper published with permission.


2016 ◽  
Vol 5 (1) ◽  
pp. 143-153
Author(s):  
Horatio Sam-Aggrey

Small modular reactors (SMRs) are being touted as safer, more cost effective, and more flexible than traditional nuclear power plants. Consequently, it has been argued that SMR technology is pivotal to the revitalization of the nuclear industry at the national and global levels. Drawing mainly on previously published literature, this paper explores the opportunities and challenges related to the deployment of SMRs in the northern territories of Canada. The paper examines the potential role of SMRs in providing an opportunity for remote mines in northern Canada to reduce their vulnerability and dependence on costly, high-carbon diesel fuel. The paper also outlines and discusses some of the potential socio-economic barriers that could impede the successful introduction of SMRs in the territories. These issues include: economic factors (such as the price of primary minerals and economics of mineral exploration, and the cost of SMR deployment), the lack of infrastructure in the territories to support mining developments, and the issues pertaining to the social acceptance of nuclear power generation.


Author(s):  
Joseph Cluever ◽  
Thomas Esselman ◽  
Paul Bruck

Abstract The nuclear industry has recently been shifting to value-based maintenance in order to keep nuclear power competitive in the power generation market. A key challenge in value-based maintenance is the optimization of a maintenance schedule. With most components having ten to fifteen available maintenances, the complexity of the optimization grows quite quickly. This paper presents a methodology for combining maintenance effectiveness, cost estimates, failure impacts, and overall reliability data to estimate an expected life cycle cost (LCC) for a component. The maintenance types are categorized into three types: monitoring, wear-rate reducing (e.g. oil change), or life-restoring (e.g. refurbishment). Each maintenance type has a different effect on problem detection, degradation rate, and future life expectancy. A Markov model uses the maintenance effects to estimate the distribution of what state of degradation a component is in at a specific time and concurrently, the component failures, maintenance costs, and failure impacts are tallied up in order to provide an expected life cycle cost for a given maintenance schedule. Optimization of the maintenance schedule is performed using the genetic algorithm where multiple maintenance schedules are simultaneously calculated, compared, and evolved in order to find the lowest expected life cycle costs. The genetic algorithm was selected as a suitable optimization algorithm for its ability to find relatively close approximations to the global optimum with relative ease while concurrently being able to handle non-smooth objective functions.


Author(s):  
T. Cheng ◽  
M. D. Pandey ◽  
W. C. Xie

Degradation of systems and components operating in harsh environment has an adverse effect on safety and reliability of nuclear power plants. Condition-based maintenance (CBM) programs are used to preventively maintain degrading components, which minimize the risk of failure. However, maintenance programs can be costly due to frequent inspection, increased outage time, and redundant maintenance of functional components. The optimization of maintenance programs over the life cycle of systems is an important issue for the plant managers. The paper presents an advanced model for the evaluation of life cycle cost of degrading components in the nuclear plants, which can be used for the maintenance optimization. The proposed model is based on the more precise finite time horizon formulation, instead of using asymptotic formulae that may lead to inaccurate results in practical settings.


Author(s):  
Amritpal S. Agar ◽  
Andy J. Fry ◽  
Martin J. Goodfellow ◽  
Yee M. Goh ◽  
Linda B. Newnes

Life cycle cost is an important consideration for the development and selection of new power generation technology. Large nuclear power plants (NPPs) have been subject to capital cost escalation, stemming from delays related to late design changes, procurement issues for major components, and regulatory enforced changes. These factors have contributed to the significant risk premium associated with gigawatt scale “Gen III+” designs, which have incurred significant financing costs. Large NPPs have become prohibitively expensive for many utility investors in liberalized markets and smaller economies. The challenge of reducing upfront capital costs is one of the requirements that have driven the development of innovative Small Modular Reactors (SMRs). These designs are said to offer reduced unit cost and reduced risk due to certainty of delivery, which could lead to a lower cost of capital for a utility customer. By offering a product with more cost certainty the SMR could restore investor confidence in nuclear power. The life cycle cost estimates associated with the different SMR designs are uncertain at the early stage of development. However, designers need to understand, with some confidence, the impact of technical decisions at the early development phase on the life cycle cost. This study presents an overview of cost uncertainty associated with the early design stage of the SMR. The types of cost estimating approaches available at the concept design phase are identified and categorized in terms of their expected accuracy ranges. The Overnight Cost of Construction (OCC) is an important driver of the life cycle cost of a power generation project. The expected accuracy ranges from each estimating method are used to illustrate the sensitivity of cost uncertainty to the level of design maturity. By understanding the sources and impact of cost uncertainty decision making during product development can be optimized to meet both technical and commercial requirements.


Energies ◽  
2021 ◽  
Vol 14 (13) ◽  
pp. 3832
Author(s):  
Awwal Mohammed Arigi ◽  
Gayoung Park ◽  
Jonghyun Kim

Advancements in the nuclear industry have led to the development of fully digitized main control rooms (MCRs)—often termed advanced MCRs—for newly built nuclear power plants (NPPs). Diagnosis is a major part of the cognitive activity in NPP MCRs. Advanced MCRs are expected to improve the working environment and reduce human error, especially during the diagnosis of unexpected scenarios. However, with the introduction of new types of tasks and errors by digital MCRs, a new method to analyze the diagnosis errors in these new types of MCRs is required. Task analysis for operator diagnosis in an advanced MCR based on emergency operation was performed to determine the error modes. The cause-based decision tree (CBDT) method—originally developed for analog control rooms—was then revised to a modified CBDT (MCBDT) based on the error mode categorizations. This work examines the possible adoption of the MCBDT method for the evaluation of diagnosis errors in advanced MCRs. We have also provided examples of the application of the proposed method to some common human failure events in emergency operations. The results show that with some modifications of the CBDT method, the human reliability in advanced MCRs can be reasonably estimated.


2021 ◽  
Vol 96 ◽  
pp. 105173
Author(s):  
Bo Yang ◽  
Yi-Ming Wei ◽  
Lan-Cui Liu ◽  
Yun-Bing Hou ◽  
Kun Zhang ◽  
...  

Author(s):  
Koichi Tsumori ◽  
Yoshizumi Fukuhara ◽  
Hiroyuki Terunuma ◽  
Koji Yamamoto ◽  
Satoshi Momiyama

A new inspection standard that enhanced quality of operating /maintenance management of the nuclear power plant was introduced in 2009. After the Fukushima Daiichi nuclear disaster (Mar. 11th 2011), the situation surrounding the nuclear industry has dramatically changed, and the requirement for maintenance management of nuclear power plants is pushed for more stringent nuclear safety regulations. The new inspection standard requires enhancing equipment maintenance. It is necessary to enhance maintenance of not only equipment but also piping and pipe support. In this paper, we built the methodology for enhancing maintenance plan by rationalizing and visualizing of piping and pipe support based on the “Maintenance Program” in cooperating with 3D-CAD system.


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