The Effect of Post-Irradiation Annealing on VVER-440 RPV Materials Mechanolocal Properties and Nano-Structure Under Re-Irradiation

2009 ◽  
Author(s):  
S. Rogozkin ◽  
A. Chernobaeva ◽  
A. Aleev ◽  
A. Nikitin ◽  
A. Zaluzhnyi ◽  
...  
Keyword(s):  
Transition Temperature ◽  
Power Plant ◽  
Atom Probe Tomography ◽  
Nuclear Power ◽  
Temperature Shift ◽  
Atom Probe ◽  
Structure Evolution ◽  
Nano Structure ◽  
High Level ◽  
Nuclear Power Plant Unit

The present work provides the analyses of embrittlement behavior and atom probe tomography study of nano-structure evolution of VVER-440 RPV materials under irradiation and re-irradiation. Specimens from VVER-440 weld with high level of cupper (0.16 wt.%) and phosphorus (0.027–0.038 wt.%) were irradiated in surveillance channels of Rovno Nuclear Power plant unit 1 (Ro-1). The embrittlement behavior has been assessed by transition temperature shift.

Steam Line Integrity Analyses for LTO: Enhancing Credibility With Comprehensive Measurement

2012 ◽  
Author(s):  
Horst Rothenhöfer ◽  
Andreas Manke
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Simulation Models ◽  
Proper Function ◽  
Blow Down ◽  
Term Operation ◽  
High Level ◽  
Comprehensive Measurement

The safety relevant components of nuclear power plant Neckarwestheim 1 — in service since 1976 — have been reviewed and updated for long-term operation (LTO). The actions included hardware retrofits as well as updates of analysis according to the latest state of the scientific and technical knowledge. For large piping such as the steam lines, the established pipes have been retained while the supports have been optimized. All shock absorbers (snubbers) including corresponding inertia have been eliminated resulting in a defined guidance and statically defined displacements. The integrity analyses for the optimized steam lines, including break preclusion, have been validated successfully with comprehensive measurements. The verification has delivered an extra high level of credibility, exceeding the “standard” requirements to achieve fitness for service in long-term operation. Measurement and validation, which are the main focus of this paper, range from monitoring of service loads to the static and dynamic measurements of pressure, local temperatures and displacements during initial start-up after implementation of the design modifications. The proper function of supports has been proved and the quality of the simulation models has been confirmed. Some expected and some unexpected dynamic events have been detected during blow-down tests. It was demonstrated that the amplitudes of all dynamic loads stay within limits. The validation of analyses with comprehensive measurement has been an important proof of quality and delivered the redundancy required for the integrity of a nuclear power plant in service, enhancing the high level of safety even more.

Risk Based Inspection Pilot Study of Ignalina Nuclear Power Plant, Unit 2

2002 ◽  
Author(s):  
Bjorn Brickstad ◽  
Adam Letzter ◽  
Arturas Klimasauskas ◽  
Robertas Alzbutas ◽  
Linas Nedzinskas ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Pilot Study ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Fracture Mechanics ◽  
Nuclear Power ◽  
Damage Mechanism ◽  
Core Damage ◽  
Steel Welds ◽  
Nuclear Power Plant Unit

A project with the acronym IRBIS (Ignalina Risk Based Inspection pilot Study) has been performed with the objective to perform a quantitative risk analysis of a total of 1240 stainless steel welds in Ignalina Nuclear Power Plant, unit 2 (INPP-2). The damage mechanism is IGSCC and the failure probabilities are quantified by using probabilistic fracture mechanics. The conditional core damage probabilities are taken from the plant PSA.

Incorporation of Risk Insights in the Regulatory Treatment of Nuclear Power Plant Structures, Systems, and Components

2002 ◽  
Author(s):  
Eugene Imbro ◽  
Thomas G. Scarbrough
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Special Treatment ◽  
Regulatory Commission ◽  
High Level

The U.S. Nuclear Regulatory Commission (NRC) has established an initiative to risk-inform the requirements in Title 10 of the Code of Federal Regulations (10 CFR) for the regulatory treatment of structures, systems, and components (SSCs) used in commercial nuclear power plants. As discussed in several Commission papers (e.g., SECY-99-256 and SECY-00-0194), Option 2 of this initiative involves categorizing plant SSCs based on their safety significance, and specifying treatment that would provide an appropriate level of confidence in the capability of those SSCs to perform their design functions in accordance with their risk categorization. The NRC has initiated a rulemaking effort to allow licensees of nuclear power plants in the United States to implement the Option 2 approach in lieu of the “special treatment requirements” of the NRC regulations. In a proof-of-concept effort, the NRC recently granted exemptions from the special treatment requirements for safety-related SSCs categorized as having low risk significance by the licensee of the South Texas Project (STP) Units 1 and 2 nuclear power plant, based on a review of the licensee’s high-level objectives of the planned treatment for safety-related and high-risk nonsafety-related SSCs. This paper discusses the NRC staff’s views regarding the treatment of SSCs at STP described by the licensee in its updated Final Safety Analysis Report (FSAR) in support of the exemption request, and provides the status of rulemaking that would incorporate risk insights into the treatment of SSCs at nuclear power plants.

Accident Analysis of Fukushima Daiichi Nuclear Power Plant Unit 2 by the SAMPSON Severe Accident Code

2014 ◽  
Author(s):  
Atsuo Takahashi ◽  
Marco Pellegrini ◽  
Hideo Mizouchi ◽  
Hiroaki Suzuki ◽  
Masanori Naitoh
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Accident Analysis ◽  
Unknown Boundary ◽  
Long Time ◽  
Fukushima Daiichi ◽  
Nuclear Power Plant Unit

The transient process of the accident at the Fukushima Daiichi Nuclear Power Plant Unit 2 was analyzed by the severe accident analysis code, SAMPSON. One of the characteristic phenomena in Unit 2 is that the reactor core isolation cooling system (RCIC) worked for an unexpectedly long time (about 70 h) without batteries and consequently core damage was delayed when compared to Units 1 and 3. The mechanism of how the RCIC worked such a long time is thought to be due to balance between injected water from the RCIC pump and the supplied mixture of steam and water sent to the RCIC turbine. To confirm the RCIC working conditions and reproduce the measured plant properties, such as pressure and water level in the pressure vessel, we introduced a two-phase turbine driven pump model into SAMPSON. In the model, mass flow rate of water injected by the RCIC was calculated through turbine efficiency degradation the originated from the mixture of steam and water flowing to the RCIC turbine. To reproduce the drywell pressure, we assumed that the torus room was flooded by the tsunami and heat was removed from the suppression chamber to the sea water. Although uncertainties, mainly regarding behavior of debris, still remain because of unknown boundary conditions, such as alternative water injection by fire trucks, simulation results by SAMPSON agreed well with the measured values for several days after the scram.

Recovery and Resilience After a Nuclear Power Plant Disaster: A Medical Decision Model for Managing an Effective, Timely, and Balanced Response

2013 ◽  
Vol 7 (2) ◽  
pp. 136-145 ◽  
Author(s):  
C. Norman Coleman ◽  
Daniel J. Blumenthal ◽  
Charles A. Casto ◽  
Michael Alfant ◽  
Steven L. Simon ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Decision Model ◽  
Large Scale ◽  
Medical Decision ◽  
Incident Management ◽  
Additional Information ◽  
Medical Issues ◽  
High Level

AbstractResilience after a nuclear power plant or other radiation emergency requires response and recovery activities that are appropriately safe, timely, effective, and well organized. Timely informed decisions must be made, and the logic behind them communicated during the evolution of the incident before the final outcome is known. Based on our experiences in Tokyo responding to the Fukushima Daiichi nuclear power plant crisis, we propose a real-time, medical decision model by which to make key health-related decisions that are central drivers to the overall incident management. Using this approach, on-site decision makers empowered to make interim decisions can act without undue delay using readily available and high-level scientific, medical, communication, and policy expertise. Ongoing assessment, consultation, and adaption to the changing conditions and additional information are additional key features. Given the central role of health and medical issues in all disasters, we propose that this medical decision model, which is compatible with the existing US National Response Framework structure, be considered for effective management of complex, large-scale, and large-consequence incidents. (Disaster Med Public Health Preparedness. 2012;0:1-10)

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