Release Limits and Decontamination Efficacy for Tritium: Lessons Learned Outside Nuclear Power Operations

2007 ◽  
Vol 93 (suppl 3) ◽  
pp. S155-S159
Author(s):  
Edward J. Waller ◽  
David Cole ◽  
Terry Jamieson
Keyword(s):  
Nuclear Power ◽  
Lessons Learned ◽  
Power Operations ◽  
Release Limits

A Global Collaboration for Community-Based Disaster Preparation and Health Promotion: Fukushima to Zuunbayan in Mongolia

2020 ◽  
pp. 1-7
Author(s):  
Chieri Yamada ◽  
Bolormaa Tsedendamba ◽  
Amarbileg Shajbalidir ◽  
Teruko Horiuchi ◽  
Katsuko Suenaga ◽  
...  
Keyword(s):  
Emergency Preparedness ◽  
Nuclear Power ◽  
Information Dissemination ◽  
Good Health ◽  
Lessons Learned ◽  
Global Collaboration ◽  
Psychological Issues ◽  
Radiation Emergency ◽  
Disaster Preparation ◽  
And Behavior

Abstract Excessive radiation exposure has adverse effects on health. In Fukushima, psychological issues such as anxiety are still affecting people nine years after the Fukushima nuclear power plant accident in 2011. In light of the lessons learned from Fukushima communities, a joint Japanese and Mongolian research team introduced a community program to the Zuunbayan district in Mongolia, which is located near a uranium deposit, to promote good health by strengthening radiation emergency preparedness. The program, which commenced in 2017, aimed to increase community participation, education, information dissemination, and capacity of community preparedness. After two years a monitoring study showed that, out of 227 respondents, the proportions who thought that any level of radiation was dangerous decreased from 53.3% in 2017 to 33.9% in 2019. Moreover, half of the respondents knew that there were safe and unsafe radiation levels and that their community was safe. This global collaboration demonstrated that a lesson learned from a disaster can be applied to other countries and changed people’s recognition and behavior toward good health and disaster/emergency preparedness.

Expectations for Inservice Testing Programs at New Nuclear Power Plants

2017 ◽  
Author(s):  
Thomas G. Scarbrough
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Core ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Mechanical Equipment ◽  
American Society ◽  
Regulatory Commission ◽  
Safety Systems

In a series of Commission papers, the U.S. Nuclear Regulatory Commission (NRC) described its policy for inservice testing (IST) programs to be developed and implemented at nuclear power plants licensed under 10 CFR Part 52. This paper discusses the expectations for IST programs based on those Commission policy papers as applied in the NRC staff review of combined license (COL) applications for new reactors. For example, the design and qualification of pumps, valves, and dynamic restraints through implementation of American Society of Mechanical Engineers (ASME) Standard QME-1-2007, “Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,” as accepted in NRC Regulatory Guide (RG) 1.100 (Revision 3), “Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants,” will enable IST activities to assess the operational readiness of those components to perform their intended functions. ASME has updated the Operation and Maintenance of Nuclear Power Plants (OM Code) to improve the IST provisions for pumps, valves, and dynamic restraints that are incorporated by reference in the NRC regulations with applicable conditions. In addition, lessons learned from performance experience and testing of motor-operated valves (MOVs) will be implemented as part of the IST programs together with application of those lessons learned to other power-operated valves (POVs). Licensee programs for the Regulatory Treatment of Non-Safety Systems (RTNSS) will be implemented for components in active nonsafety-related systems that are the first line of defense in new reactors that rely on passive systems to provide reactor core and containment cooling in the event of a plant transient. This paper also discusses the overlapping testing provisions specified in ASME Standard QME-1-2007; plant-specific inspections, tests, analyses, and acceptance criteria; the applicable ASME OM Code as incorporated by reference in the NRC regulations; specific license conditions; and Initial Test Programs as described in the final safety analysis report and applicable RGs. Paper published with permission.

Identification of Failure Modes Under Design Extension Conditions

2015 ◽  
Author(s):  
Naoto Kasahara ◽  
Izumi Nakamura ◽  
Hideo Machida ◽  
Hitoshi Nakamura ◽  
Koji Okamoto
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Failure Modes ◽  
Low Cycle Fatigue ◽  
External Pressure ◽  
Lessons Learned ◽  
Local Failure ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Loading Mode

As the important lessons learned from the Fukushima-nuclear power plant accident, mitigation of failure consequences and prevention of catastrophic failure became essential against severe accident and excessive earthquake conditions. To improve mitigation measures and accident management, clarification of failure behaviors with locations is premise under design extension conditions such as severe accidents and earthquakes. Design extension conditions induce some different failure modes from design conditions. Furthermore, best estimation for these failure modes are required for preparing countermeasures and management. Therefore, this study focused on identification of failure modes under design extension conditions. To observe ultimate failure behaviors of structures under extreme loadings, new experimental techniques were adopted with simulation materials such as lead and lead-antimony alloy, which has very small yield stress. Postulated failure modes of main components under design extension conditions were investigated according three categories of loading modes. The first loading mode is high temperature and internal pressure. Under this mode, ductile fracture and local failure were investigated. At the structural discontinuities, local failure may become dominant. The second is high temperature and external pressure loading mode. Buckling and fracture were investigated. Buckling occurs however hardly break without additional loads or constraints. The last loading is excessive earthquake. Ratchet deformation, collapse, and fatigue were investigated. Among them, low-cycle fatigue is dominant.

Evaluation of Representative Piping Systems Designed by Implicit Fatigue Concept

2008 ◽  
Author(s):  
Shin-Beom Choi ◽  
Sun-Hye Kim ◽  
Yoon-Suk Chang ◽  
Jae-Boong Choi ◽  
Young-Jin Kim ◽  
...  
Keyword(s):  
Fatigue Life ◽  
Nuclear Power ◽  
Heat Removal ◽  
Environmental Water ◽  
Lessons Learned ◽  
Metal Fatigue ◽  
Aging Effects ◽  
Age Related ◽  
Piping Systems ◽  
Surge Line

NUREG-1801 provides generic aging lessons learned to manage aging effects that may occur during continued operation beyond the design life of nuclear power plant. According to this report, the metal fatigue, among several age-related degradation mechanisms, is identified as one of time-limited aging analysis item. The objective of this paper is to introduce fatigue life evaluation of representative surge line and residual heat removal system piping which was designed by implicit fatigue concept. For the back-fitting evaluation employing explicit fatigue concept, detailed parametric CFD as well as FE analyses results are used. The well-known ASME Section III NB-3600 procedure is adopted for the metal fatigue and NUREG/CR-5704 procedure is further investigated to deal with additional environmental water effects. With regard to the environmental effect evaluation, two types of fatigue life correction factors are considered, such as maximum Fen and individual Fen. As a result, it was proven that a thermal stratification phenomenon is the governing factor in metal fatigue life of the surge line and strain rate is the most important parameter affecting the environmental fatigue life of both piping. The evaluation results will be used as technical bases for continued operation of OPR 1000 plant.

scholarly journals Using and Explaining Individual Dosimetry Data

2017 ◽  
Vol 29 (2_suppl) ◽  
pp. 110S-119S ◽  
Author(s):  
Makoto Miyazaki
Keyword(s):  
Nuclear Power ◽  
Large Scale ◽  
Measurement Data ◽  
Lessons Learned ◽  
Individual Dose ◽  
Protection Measures ◽  
Measurement Results ◽  
Individual Dosimetry ◽  
Dose Measurements ◽  
Self Protection

Measurement of individual radiation dose is crucial for planning protective measures after nuclear accidents. The purpose of this article is to explain the various initiatives taken after the TEPCO Fukushima Daiichi Nuclear Power Plant accident, including the D-shuttle project wherein residents from affected areas wore a personal dosimeter to measure their own external exposure. The experience in Fukushima revealed several issues such as gaining residents’ trust and ensuring appropriate communication of the measured data. The D-shuttle project also revealed that obtaining individual dose measurement data had 2 purposes, as the information obtained was to be utilized by the residents for self-protection and by the authorities for deriving the dose distribution of the population to aid in designing large-scale protection measures. The lessons learned are that both the residents and the authorities need to understand and share the meaning of individual dose measurements and the measurement results must be used with due respect for the residents’ privacy and other concerns.

An Advanced Method for Evaluating Risk From Seismically-Induced Internal Fires and Floods

2017 ◽  
Author(s):  
Paul J. Amico ◽  
Pierre Macheret ◽  
Robert P. Kassawara
Keyword(s):  
Nuclear Power ◽  
Tohoku Earthquake ◽  
Lessons Learned ◽  
Advanced Method ◽  
Plant Safety ◽  
Nuclear Plants ◽  
Significant Interest ◽  
Realistic Representation ◽  
Nuclear Power Plant Safety ◽  
Technical Guidance

It has been traditional in assessment of nuclear power plant safety that both deterministic safety analyses and probabilistic safety analyses treat the potential effects of various hazards individually. That is, the safety implications of internal events (e.g., randomly occurring transients and LOCAs), internal hazards (e.g., internal fire and flood), and external hazards (e.g., earthquakes, tornados) are treated as independent occurrences. With the occurrence of the Great Tohoku earthquake and the effects observed at nuclear plants in Japan, it was realized that this approach failed to provide a realistic representation of risk, and now there is a significant interest in correlated hazards. As a result, EPRI embarked on the development of an improved methodology focusing on seismically-induced internal fires and internal floods. All the technical work on the methodology has been completed and draft technical guidance developed. This guidance has been provided to some plants that are interested in piloting the methodology. As of the date of paper submittal, two pilots are underway and three more are under consideration. Upon completion of the pilots, the methodology will be updated to incorporate the lessons-learned and published.

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