权益分类	功能权益	普通用户	{{item.name}}会员
{{category.name}}	{{benefitItem.name}}

Enhanced Methodologies for Advanced Nuclear System Safety (eMEANSS)

先进核系统安全增强方法 (eMEANSS)

基本信息

批准号：
EP/T016329/1
负责人：
Simon Charles Middleburgh
金额：
$ 108.93万
依托单位：
Bangor University
依托单位国家：
英国
项目类别：
Research Grant
财政年份：
2022
资助国家：
英国
起止时间：
2022 至无数据
项目状态：
未结题

来源：
https://gtr.ukri.org/projects?ref=EP%2FT016329%2F1
关键词：
Enhanced Methodologies Advanced Nuclear System

项目摘要

A re-assessment of the impact of uncertainties within the nuclear industry is of paramount importance, not only ensuring the continued safety of nuclear energy systems, but also to ensure the economic viability of nuclear power, allowing for continued reductions in CO2 emissions globally. Uncertainties are unavoidable, and complex systems such as nuclear reactors are designed to cope with them. A naive approach would be to consider worst cases scenarios individually without considering their dependencies. This approach can produce over-designed and expensive systems without guaranteeing their overall safety. Proper quantification and propagation of uncertainty across multi-physical components allows one to determine vulnerable componentry, prioritise investments, identify operational margins and adopt relevant measures to guarantee safety whilst at the same time reducing the overall cost of advanced nuclear design.Methods will be synthesised as part of this project to improve the estimation of uncertainty/safety, bringing together researchers specialising in reactor physics, fuel performance, structural materials and uncertainty quantification. Work package 1: In reactor physics the new methods will be tested by considering the uncertainties propagated through a severe nuclear reactor accident assessment, specifically a loss-of-coolant accident (LOCA). The project will attempt to target and reduce uncertainties related to properties including nuclear data associated with specific isotopes and temperature dependent effects corresponding to neutron capture cross-sections. Drawing on the expertise in the UK and India, the enhancements in the methods utilised will have far-reaching impacts.Work package 2: Fatigue failure of graphite components, especially at high service temperatures, is of serious concern for next generation reactors. A design tool is to be produced that can efficiently incorporate variances in the mechanical and thermal loading history, and material properties to quantify a probable component life. In addition to the simple uncertainties in boundary conditions, complications arise from both the load sequence and the temperatures at which loading occurs, coupled with the impacts arising from neutron irradiation, temperature and coolant interactions. The world-leading team in the UK and India will generate new knowledge on the high temperature cyclic response of advanced nuclear graphite and will utilise it in the development of a new probabilistic modelling framework.Work package 3: Nuclear fuel performance codes predict the behaviour of fuel in a reactor, allowing operating regimes to be tested that avoid fuel melting or fuel failure. The models improved over decades of experience in the UO2-Zr system remain highly empirical (i.e. not mechanistic) and large uncertainties exist that are to be quantified through the use of uncertainty modelling (depending on each model's impact) and reduced through the addition of mechanistic models. Novel fuels with greater uncertainties will also be considered. Here, uncertainty modelling will be used to target the most rapid reduction of uncertainty of behaviour possible to expedite licensing and commercial use of the fuel.Work package 4: The uncertainty models will be identified and commonalities will be linked to enable the overarching uncertainty methodology to be formulated. This is an important task that will ensure the outputs from the targeted examples (in work packages 1-3) have far reaching impact beyond themselves in other areas of nuclear engineering and beyond. In addition to linking the uncertainty modelling methods this work package will lead by communicating the results to the wider community through publications and workshops.

重新评估核工业内部不确定性的影响至关重要，不仅可以确保核能系统的持续安全，而且可以确保核电的经济可行性，从而继续减少全球二氧化碳排放量。不稳定是不可避免的，而核反应堆等复杂系统的设计就是为了科普不稳定。一种天真的方法是单独考虑最坏的情况，而不考虑它们的依赖性。这种方法可能会产生过度设计和昂贵的系统，而无法保证其整体安全性。正确量化和传播多物理组件的不确定性，可以确定脆弱的组件，优先投资，确定运营利润，并采取相关措施，以确保安全，同时降低先进核设计的总成本。方法将作为本项目的一部分进行综合，以提高不确定性/安全性的估计，汇集了专门从事反应堆物理的研究人员，燃料性能、结构材料和不确定性量化。工作包1：在反应堆物理学中，新方法将通过考虑严重核反应堆事故评估，特别是冷却剂损失事故（LOCA）传播的不确定性进行测试。该项目将试图针对和减少与属性有关的不确定性，包括与特定同位素有关的核数据和与中子俘获截面相应的温度相关效应。借鉴英国和印度的专业知识，改进所采用的方法将产生深远的影响。工作包2：石墨部件的疲劳失效，特别是在高工作温度下，是下一代反应堆的严重问题。一个设计工具是生产，可以有效地结合变化的机械和热负荷的历史，材料性能，以量化可能的组件寿命。除了边界条件中的简单不确定性之外，加载顺序和加载发生时的温度以及中子辐照、温度和冷却剂相互作用产生的影响也引起了复杂性。英国和印度的世界领先团队将产生关于先进核石墨高温循环响应的新知识，并将其用于开发新的概率建模框架。工作包3：核燃料性能代码预测反应堆中燃料的行为，允许测试操作制度，以避免燃料熔化或燃料故障。在二氧化铀-锆系统几十年的经验中改进的模型仍然是高度经验性的（即不是机械性的），存在很大的不确定性，这些不确定性将通过使用不确定性建模（取决于每个模型的影响）加以量化，并通过增加机械性模型加以减少。还将考虑具有更大不确定性的新型燃料。在此，将使用不确定性建模，以尽可能快速地减少行为的不确定性，从而加快燃料的许可和商业使用。工作包4：将确定不确定性模型，并将共性联系起来，以便能够制定总体不确定性方法。这是一项重要的任务，将确保目标示例（工作包1-3）的输出在核工程的其他领域和其他领域产生深远的影响。除了将不确定性建模方法联系起来之外，该工作包还将通过出版物和研讨会将结果传达给更广泛的社区。