High-Level-Waste and Spent Fuel Storage in Switzerland

2001 ◽  
Author(s):  
Udo Sach ◽  
Goswin Schreck ◽  
Max Ritter ◽  
Jean-Pierre Wenger
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Radioactive Wastes ◽  
High Level Waste ◽  
Storage Facility ◽  
High Level ◽  
Interim Storage ◽  
Fuel Elements ◽  
And Storage

Abstract At present, Switzerland has no final repository for radioactive wastes. Very early, the Swiss nuclear power plant operators were aware of the necessity to expand interim storage capacity for spent fuel elements and operational wastes. Already in 1991, Nordostschweizerische Kraftwerke AG (NOK) therefore started building a reactor-site interim storage facility (ZWIBEZ) at its Beznau power plant site. Moreover, as early as in 1990, “ZWILAG Zwischenlager Würenlingen AG”, a company established by the nuclear power plant operators had initiated the licensing procedure for a central interim storage facility in Switzerland. This central interim storage facility is designed for the storage of all categories of radioactive wastes and includes a conditioning facility for low-level and medium-level waste. Eleven years later, in July 2001, the first transport and storage cask loaded with irradiated fuel elements was stored in this facility. For both of the stores the concept of dry interim storage in suitable storage casks in a storage hall was chosen for the storage of irradiated fuel elements and vitrified high-level wastes from reprocessing. Cooling is established through natural circulation. Leaktightness of the casks is continuously monitored by means of a cask monitoring system. The other wastes arising from nuclear power plant operation and reprocessing are stored in a ventilated storage hall which provides shielding and — depending on the radioactive inventory — protection against external impact. The conditioned radioactive wastes, packaged in drums, are placed into open storage containers with identical base and having the same sling points as ISO containers. These containers are stacked up in free-standing stacks up to a height of 16 m. The storage concept varies, depending on the radioactive inventory; for the ZWIBEZ reactor-site interim store, a storage hall for low-level waste has been built without partition walls, whereas the store for the medium and high-level waste in the central interim store ZWILAG has been designed with partition walls dividing the hall into several storage shafts which are closed by shielding slabs. By including a hot cell into the ZWILAG facility, the purpose of this facility has been expanded beyond interim storage of radioactive waste to cover also the visual inspection of fuel elements and vitrified waste canisters as well as the reloading of fuel elements and canisters from smaller transport casks into combined transport and storage casks. Furthermore, the hot cell enables inspection and/or repair work to be performed in the cask lid area of loaded transport and storage casks, the replacement of the lid seals of storage casks and the conditioning of medium-level waste.

Below the Horizon

2010 ◽  
Vol 132 (12) ◽  
pp. 30-34
Author(s):  
Kellen M. Giraud ◽  
Jay F. Kunze ◽  
James M. Mahar ◽  
Carl W. Myers
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Heat Waves ◽  
High Level Waste ◽  
Physical Attack ◽  
Suitable Site ◽  
Underground Facility ◽  
High Level

This article elaborates the advantages of underground nuclear parks over conventional nuclear power plant designs. Locating the reactors a few hundred feet underground in bedrock at a suitable site eliminates the need for containment structures, and the site would be largely impervious to physical attack from terrorists. (Indeed, it would be far easier to secure the few access points to an underground nuclear park than it is to protect the large perimeter of an isolated nuclear power plant.) A properly constructed underground facility would also be less subject to weather-related construction delays or the effects of hurricanes, tornadoes, flooding, or heat waves. Also, if designers were careful in the site selection, an underground nuclear park could virtually eliminate the transportation of hazardous nuclear waste material. Spent nuclear fuel could be moved via tunnel from the reactors to an array of storage tunnels; high-level waste could be permanently stored in another set of tunnels. When the reactors reach the end of their productive life, they can be decommissioned in place.

Model of a Dry Storage Facility for a Medium Nuclear Power Plant

2001 ◽  
Author(s):  
Mile Bace ◽  
Kresimir Trontl ◽  
Dubravko Pevec
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Radiation Dose ◽  
Dose Rate ◽  
Nuclear Power ◽  
Storage Facility ◽  
Spent Fuel ◽  
Radiation Dose Rate ◽  
Dry Storage ◽  
Fuel Assemblies

Abstract The intention was to model a dry storage facility that could satisfy the needs of a medium nuclear power plant similar to the NPP Krsko. The attention has been focused on radiation dose rate analyses and criticality calculations. Using the SCALE 4.4 code package and modified QAD-CGGP code, we modeled a facility that satisfies the basic criteria for public radiation protection. The capacity of the storage is 1,400 spent fuel assemblies which is adequate for a forty years medium NPP lifetime.

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.

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.

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)

scholarly journals Establishment of cementation parameters of dried waste from evaporation coming from NPP operation

2019 ◽  
Vol 7 (2A) ◽  
Author(s):  
Érica Rodrigues de Faria ◽  
Clédola Cássia Oliveira de Tello ◽  
Bruna Silveira Costa
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Scale Up ◽  
Radioactive Wastes ◽  
Intermediate Level ◽  
Drying Process ◽  
Waste Products ◽  
Liquid Wastes ◽  
To Receive

The radioactive wastes generated in Brazil are treated and sent to initial and intermediate storages. The “Project RBMN” proposes the implantation of the Brazilian repository to receive and permanently dispose the low and intermediate level radioactive wastes. The CNEN NN 6.09 standard – Acceptance Criteria for Disposal of Low and Intermediate Radioactive Wastes (LIRW) – establishes the fundamental requirements to accept the wastes packages in the repository. The evaporator concentrate is one of liquid wastes generated in a Nuclear Power Plant (NPP) operation and usually it is cemented directly inside the packing. The objective of this research is to increase the amount of the incorporated waste in each package, using the drying process before the cementation, consequently reducing the volume of the waste to be disposed.  Drying and cementation parameters were established in order to scale-up the process aiming at waste products that comply with the requirements of CNEN standard. The cementation of the resulting dry wastes was carried out with different formulations, varying the amount of cement, dry waste and water. These tests were analyzed in order to select the best products, with higher waste incorporation than current process and its complying the requirements of the standard CNEN NN 6.09. 

Selecting HLR and SR Standards for PRA Applications

2008 ◽  
Author(s):  
Steven E. Farkas
Keyword(s):  
Risk Assessment ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
White Paper ◽  
Application Process ◽  
Technical Adequacy ◽  
Application Mapping ◽  
Regulatory Guidance ◽  
High Level

This white paper provides guidance for mapping risk informed applications to PRA Scope and Technical Adequacy requirements. The discussion considers regulatory guidance contained in SRP 19.1 “Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities” and supplements industry guidance on application mapping contained in Section 3 “Risk Assessment Application Process of the ASME Standard for the PRA for Nuclear Power Plant Applications,” of the ASME PRA Standard with Regulatory Guide 1.200 clarifications. The general processes identified in this paper should also be applicable to mapping applications to any of the forthcoming ANS standard requirements. The aim of this white paper is to help identify which ASME PRA Standard High Level Requirements (HLRs) fall in the scope of a particular “risk informed” application to the NRC that will, in part, justify the requested change with risk metrics. To obtain values for risk metrics, it is necessary to manipulate the plant PRA model. The question becomes what should be changed in the model to represent the “risk informed” topic at hand? Each HLR has a set of supporting requirements (SRs) graded by “Capability Categories.” For many supporting requirements (SRs) under each HLR, there is no difference between the three capability categories. At issue is what features of the model have to conform to Capability Category 2 when Capability Category 2 differs from Capability Category 1.

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