scholarly journals Assessment of the Shelter Structures to Be Dismantled after NSC Installation. Comparative Analysis of Dismantling Options

2019 ◽  
pp. 17-22
Author(s):  
Yu. Nemchynov ◽  
A. Bambura ◽  
I. Sazonova ◽  
K. Babik ◽  
V. Shcherbin ◽  
...  
Keyword(s):  
Bearing Capacity ◽  
Nuclear Power ◽  
Power Plants ◽  
Building Structures ◽  
Safety Measures ◽  
Safety Requirements ◽  
State Research Institute ◽  
Stability Bearing Capacity ◽  
Climatic Impacts ◽  
Two Stages

The surveys of the bearing and enclosing civil structures damaged after the Chornobyl Unit 4 accident that were conducted by the State Research Institute of Building Structures and Institute for Safety Problems of Nuclear Power Plants from 1995 to 2012 showed that a number of structures were in unstable condition. They include a group of structures that are especially dangerous in terms of the bearing capacity and are very likely to collapse. To ensure safe operation, immediate stabilization measures were developed and successfully implemented at the Shelter in 2005—2008. The justifying calculations show that the structures will comply with nuclear and radiation safety requirements (in terms of stability, bearing capacity etc.) over 15 years. The most unstable structures have to be dismantled by 2023, which is one of the conditions in the strategic plan for further transformation of the Shelter into a safe system. Two stages of safety measures have been defined for the unstable structures to be dismantled. The functional purpose of and climatic impacts on the bearing structures and, as a consequence, the Shelter lifetime are subject to change after stabilization of the unstable structures and installation of the New Safe Confinement (NSC) into the designed position. The paper analyzes scenarios for dismantling of unstable structures, stabilization measures and the probability of failure after implementation of the stabilization measures. A list of structures subject to early and deferred dismantling is provided. Shelter safety criteria and radiation protection objectives are considered.

Example of Graded and Lifecycle Phase-Specific Security Controls for Nuclear I&C and EPS Use Cases

2018 ◽  
Author(s):  
Venesa Watson ◽  
Edita Bajramovic ◽  
Xinxin Lou ◽  
Karl Waedt
Keyword(s):  
Computer Security ◽  
Nuclear Power ◽  
Power Plants ◽  
Supplier Relationships ◽  
Safety Requirements ◽  
Electrical Systems ◽  
Security Controls ◽  
Systems Security ◽  
Computer Based ◽  
Iec 27002

Working Group WGA9 of IEC SC45A (Nuclear I&C and ES), has recently completed a further working draft (WD) of the new IEC 63096 (unpublished) standard, aptly entitled Nuclear Power Plants – Instrumentation, Control and Electrical Systems – Security Controls. IEC 63096 specifically focuses on the selection and application of computer security controls for computer-based I&C and ES systems. This standard follows the commonly accepted ISO/IEC 27000 series security objectives of confidentiality, integrity and availability, and borrows and expands the objectives and implementation guidance from ISO/IEC 27002, while considering recommendations on sector-specific standards by ISO/IEC 27009. In addition, this guidance introduces a security grading, as well as lifecycle phase-specific controls. The grading aligns with the stringency of security controls, starting with Baseline Requirements (BR), Security Degree S3 and up to S1 (from lowest to highest degree). The lifecycle phase concerns the I&C development (D), project engineering (E) and operation and maintenance phases (O). This paper applies a sub-clause of IEC 63096 clause 15 (Supplier Relationships), to a programmable logic controller (PLC) that is typically used in power plants, to show the intended use of this standard and how it complements highest safety requirements in power plants. The Supplier Relationship clause concerns topics related to supply chain security, and is used to develop a use case example for the PLC. This example demonstrates how the controls and security degrees fits the implementation guidance from ISO/IEC 27002 and how they can be methodically applied to an I&C system.

Utilization of Micro-Electronic-Machine Systems (MEMS) to Possible Future Use in the Enhanced Analysis of Safety in Nuclear Power Plants

2002 ◽  
Author(s):  
Anmol Bhavnani
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Focal Point ◽  
Main Idea ◽  
Commercial Sector ◽  
Safety Measures ◽  
Machine Systems

The focal point of this paper is to go in-depth in to the potential utilization of MEMS to further enhance safety measures within nuclear power plants. Robots, which are being researched and developed in Sandia National Laboratories, sometimes built as small as the size of a pollen grain, can be utilized to constantly monitor the stress analysis within all aspects of running a Nuclear Power Plant. From cooling towers to detecting miniscule cracks within pipes, MEMS can be utilized to constantly detect and even possibly repair minor faults within the overall structure of a nuclear power plant. MEMS technologies provide the ability to reliably produce micro actuators and sensors to meet these mission requirements. MEMS technologies are also attracting an increasing interest from the commercial sector for various applications. Currently, Sandia National Laboratories has been developing MEMS technologies to support its core missions of weapon surety, stockpile maintenance, and national security interests. Already, the project has been responsible for numerous electromechanical systems in nuclear weapons, which ensure nuclear safety and provide reliable arming, fusing and firing. With these factors in consideration, the main idea of this paper is to present ideas for producing sensors and robots on a micro scale, which could be programmed to communicate and work within each other to have enhanced safety and efficiency within a nuclear power plant.

Development Trends in Nuclear Technology and Related Safety Aspects

2002 ◽  
Author(s):  
B. Kuczera ◽  
P. E. Juhn ◽  
K. Fukuda
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Fuel Cycle ◽  
Nuclear Reactors ◽  
Nuclear Technology ◽  
User Requirements ◽  
Safety Standards ◽  
Safety Requirements ◽  
Fuel Cycles ◽  
High Level

The IAEA Safety Standards Series include, in a hierarchical manner, the categories of Safety Fundamentals, Safety Requirements and Safety Guides, which define the elements necessary to ensure the safety of nuclear installations. In the same way as nuclear technology and scientific knowledge advance continuously, also safety requirements may change with these advances. Therefore, in the framework of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) one important aspect among others refers to user requirements on the safety of innovative nuclear installations, which may come into operation within the next fifty years. In this respect, the major objectives of the INPRO subtask “User Requirements and Nuclear Energy Development Criteria in the Area of Safety” have been: a. to overview existing national and international requirements in the safety area, b. to define high level user requirements in the area of safety of innovative nuclear technologies, c. to compile and to analyze existing innovative reactor and fuel cycle technology enhancement concepts and approaches intended to achieve a high degree of safety, and d. to identify the general areas of safety R&D needs for the establishment of these technologies. During the discussions it became evident that the application of the defence in depth strategy will continue to be the overriding approach for achieving the general safety objective in nuclear power plants and fuel cycle facilities, where the emphasis will be shifted from mitigation of accident consequences more towards prevention of accidents. In this context, four high level user requirements have been formulated for the safety of innovative nuclear reactors and fuel cycles. On this basis safety strategies for innovative reactor designs are highlighted in each of the five levels of defence in depth and specific requirements are discussed for the individual components of the fuel cycle.

Impact of Deviations Handling on Outages

2016 ◽  
Author(s):  
Naoëlle Matahri
Keyword(s):  
Power Plants ◽  
Technical Support ◽  
Safety Measures ◽  
Design Change ◽  
Safety Requirements ◽  
Support Organization ◽  
Corrective Measures ◽  
Nuclear Safety Authority ◽  
The Impact ◽  
The Relationship

Based on the information provided by the operators, IRSN experts select and analyze different deviations presenting a possible generic nature which could affect the safety of power plants. Some of these deviations result in non-compliance (NC) with the safety requirements. To maintain an acceptable level of safety, an operator has to implement corrective measures for any situation of non-compliance with safety requirements. IRSN, the Technical Support Organization of the French Nuclear Authority, analyzes the different deviations to assess the impacts on the concerned NPPs safety. Based on the impact on safety, measures should be applied immediately or during the next outages, on a reactor or on several of them. The permanent corrective measures schedule is defined taking into account the “NC” safety impact. However, for some of the “NCs”, it can be difficult to define and implement swift permanent corrective measures, especially when the lack of compliance affects several similar units and requires a design change. This paper explains the French approach of deviations treatment and specifically the relationship between the Nuclear Safety Authority, the Technical Support Organization, IRSN and the Licensee, EDF during an outage.

scholarly journals Regulatory Nuclear and Radiation Safety Requirements for NPP Power Supply Systems

2017 ◽  
pp. 46-49
Author(s):  
V. Levakin ◽  
K. Yefimova ◽  
S. Polyvoda ◽  
V. Iokst
Keyword(s):  
Power Supply ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Radiation Safety ◽  
Safety Requirements ◽  
Electrical Systems ◽  
Emergency Power ◽  
Nuclear And Radiation Safety ◽  
Supply Systems

The paper presents review of the requirements from the new regulation NP 306.2.205-2016 “Requirements for Power Supply Systems Important to Safety of Nuclear Power Plants” and recommendations of IAEA and WENRA for the construction of electrical systems important to safety of nuclear power plants. The research is focused on main differences of NP 306.2.205-2016 from standards that applied to NPP emergency power supply systems (PNAE G-9-026-90, PNAE G-9-027- 91) and which were cancelled in 2016.

scholarly journals Alternatif Fondasi Tiang Pancang Pada Gedung Pengendalian Penduduk dan Keluarga Berencana (P2KB) Palu

2021 ◽  
pp. 69-78
Author(s):  
R. Hidayat ◽  
Irdhiani Irdhiani
Keyword(s):  
Shear Strength ◽  
Bearing Capacity ◽  
Lateral Force ◽  
Empirical Method ◽  
Strength Parameter ◽  
Single Pile ◽  
Building Structures ◽  
Foundation Design ◽  
Safety Requirements ◽  
Shear Strength Parameter

The foundation design of Pengendalian Penduduk dan Keluarga Berencana is located on Jl. R.A. Kartini No. 100, Palu City, Central Sulawesi. In field testing using 2-point of (CPT), the value of hard-ground support at 8 – 9 m below the surface and the loads of building structures of 2 (two) floor buildings that work on the foundations are quite large, this is a consideration in choosing the type of used foundation. The point of this design is to obtain the dimensions of the pile foundation and calculate the bearing capacity of the foundation permit and settlement that meet the safety requirements. The calculation for bearing capacity of the foundation is calculated using (CPT) data and soil shear strength parameter data (c and tetha). Single pile is calculation by using the Semi-Empirical method and using Brooms Method for calculating lateral force on the driven piles. The dimensions of the foundation are planned based on the load (Qv) acting on the entire foundation. Calculation of bearing capacity of a single pile with a penetration depth of 8,20 m and varying dimensions are used in the planning. Based on the calculation of the bearing capacity of a single pile using (CPT) data on the load that works on the foundation, was obtained 25 cm and 30 cm diameter of pile. while the calculation uses ground shear strength parameter data (c and tetha) obtained pile diameter of 30 cm and diameter of 35 cm. Based on the calculation of the bearing capacity of a single pile, the dimensions of the foundation and the settlement in permits has reached the safety requirements.

A Critical Comparison of Seismic Qualification Requirements for Safety Equipment in Various Codes and Standards

2003 ◽  
Author(s):  
F. G. Abatt ◽  
Quazi Hossain ◽  
Milon Meyer
Keyword(s):  
Seismic Performance ◽  
Seismic Design ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Lower Class ◽  
Department Of Energy ◽  
Building Structures ◽  
Performance Category ◽  
Code Provisions

Evaluation of life safety risks to facility occupants, public, and the environment that may result from earthquake events involves both building structures and equipment supported from these structures. But, it is the seismic design of building structures that typically receive the bulk of the attention from the code committees of the national professional organizations and the regulatory authorities. For safety related equipment in nuclear facilities (e.g., Seismic Category I equipment in nuclear power plants and Seismic Performance Category 3 and 4 equipment in the Department of Energy facilities), the seismic design and analysis guidelines and acceptance criteria are well established. But, for Nonseismic Category equipment in nuclear power plants and Seismic Performance Category 1 and 2 equipment in Department of Energy facilities, these have not yet been developed to the same level of completeness and rigor. The code provisions and guidelines available today for these lower class/categories of equipment are briefly, but critically discussed here, along with a comparison of the results of the application of these code provisions.

scholarly journals Safety Assessment of Low-Contaminated Equipment Dismantling at Nuclear Power Plants

2015 ◽  
Vol 2015 ◽  
pp. 1-11 ◽  
Author(s):  
Egidijus Babilas ◽  
Eugenijus Ušpuras ◽  
Sigitas Rimkevičius ◽  
Gintautas Dundulis ◽  
Mindaugas Vaišnoras
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Structural Integrity ◽  
Building Structures ◽  
Nuclear Facilities ◽  
Energy Agency ◽  
Force Calculation

The decommissioning of nuclear facilities requires adequate planning and demonstration that dismantling and decontamination activities can be conducted safely. Existing safety standards require that an appropriate safety assessment be performed to support the decommissioning plan for each facility (International Atomic Energy Agency, 2006). This paper presents safety assessment approach used in Lithuania during the development of the first dismantling and decontamination project for Ignalina NPP. The paper will mainly focus on the identification and assessment of the hazards raised due to dismantling and decontamination activities at Ignalina Nuclear Power Plant and on the assessment of the nonradiological and radiological consequences of the indicated most dangerous initiating event. The drop of heavy item was indicated as one of most dangerous initiating events for the discussed Ignalina Nuclear Power Plant dismantling and decontamination project. For the analysis of the nonradiological impact the finite element model for the load drop force calculation was developed. The radiological impact was evaluated in those accident cases which would lead to the worst radiological consequences. The assessments results show that structural integrity of the building and supporting columns of building structures will be maintained and radiological consequences are lower than the annual regulatory operator dose limit.

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