Best Estimate Plus Uncertainty Analysis of LBLOCA for a Pickering B CANDU Reactor

2010 ◽  
Author(s):  
Larry Blake ◽  
George Gavrus ◽  
Jack Vecchiarelli ◽  
J. Stoklosa
Keyword(s):  
Power Plants ◽  
Safety Margin ◽  
Reactor Power ◽  
Natural Uranium ◽  
Nuclear Industry ◽  
Initial Increase ◽  
Power Pulse ◽  
Pulse Phase ◽  
Best Estimate ◽  
Safety Margins

The Pickering B Nuclear Generating Station consists of four CANDU reactors. These reactors are horizontal pressure tube, heavy water cooled and moderated reactors fuelled with natural uranium. Under a postulated large break loss of coolant accident (LOCA), positive reactivity results from coolant void formation. The transient is terminated by the operation of the safety systems within approximately 2 seconds of the start of the transient. The initial increase in reactor power, terminated by the action of the safety system, is termed the power pulse phase of the accident. In many instances the severity of an LBLOCA can be characterized by the adiabatic energy deposited to the fuel during this phase of the accident. Historically, Limit of Operating Envelope (LOE) calculations have been used to characterize the severity of the accident. LOE analyses are conservative analyses in which the key operational and safety related parameters are set to conservative or limiting values. Limit based analyses of this type result in calculated transient responses that will differ significantly from the actual expected response of the station. As well, while the results of limit calculations are conservative, safety margins and the degree of conservatism is generally not known. As a result of these factors, the use of Best Estimate Plus Uncertainty (BEPU) analyses in safety analyses for nuclear power plants has been increasing. In Canada, the nuclear industry has been pursuing best estimate analysis through the BEAU (Best Estimate Analysis and Uncertainty) methodology in order to obtain better characterization of the safety margins. This approach is generally consistent with those used internationally. Recently, a BEAU analysis of the Pickering B NGS was completed for the power pulse phase of a postulated Large Break LOCA. The analysis comprised identification of relevant phenomena through a Phenomena Identification and Ranking (PIRT) process, assessment of the code input uncertainties, sensitivity studies to quantify the significance of the input parameters, generation of a functional response surface and its validation, and determination of the safety margin. The results of the analysis clearly demonstrate that the Limit of Operating Envelope (LOE) results are significantly conservative relative to realistic analysis even when uncertainties are considered. In addition, the extensive sensitivity analysis performed to supplement the primary result provides insight into the primary contributors to the results.

Engineering Practices for Tokamak Window Assemblies

2009 ◽  
Author(s):  
C. Waldon ◽  
R. Morrell ◽  
D. Buckthorpe ◽  
M. Davies ◽  
P. Sherlock
Keyword(s):  
Power Plants ◽  
Reactor Power ◽  
Nuclear Industry ◽  
Potential Candidate ◽  
Metallic Materials ◽  
Plasma Diagnostic ◽  
Analysis Experiment ◽  
Technical Ceramics ◽  
Definition Of ◽  
Engineering Practices

For fusion tokamak reactors the diagnostics and RF heating systems require the use of components with parts made of non-metallic materials. These can form part of the vacuum boundary of the tokamak which is the primary safety boundary and have a function of containing tritium fuel or activated gases and particulate debris. The engineering practices for such components and non-metallic materials are in an early state of preparation and require development to enable systems to be used in a safety and licensing context. Such developments will have to reflect the brittle nature of the materials, and are likely to be based on established arguments developed within the nuclear industry, such as containment and defence in depth. Given these requirements this task is a major challenge. The window systems fall broadly into two categories: • Transmission windows for the input of high-power microwaves to drive and heat the plasma; • Diagnostic windows to monitor the plasma. Currently there are no established fusion design codes that can be used to assure nuclear safety and a consistent engineering approach for either application. This paper reviews the progress made in developing such practices for transmission and diagnostic windows made from ceramic materials. The investigations undertaken and the engineering practices addressed for the tokamak windows generally fall into the following areas: • reviews of potential candidate materials along with a summary of the available property data; • definition of the function of torus window assemblies and an outline of the complexity and variety of design considerations (including historical failures, and statutory requirements); • development of the design methodology for technical ceramics; • definition of the design routes considered and selected (rule, analysis, experiment); • consideration of the material data available (or lack of) for technical ceramics and their failure criteria; • qualification and design of metallic / ceramic joints; • definition of the requirements with regard to quality control, from manufacture to in-service inspection; • development and formation of a draft code procedure. The practices and procedures developed are considered to be an important contribution and significant step forward in the development of a fusion tokamak windows code. Important contributions have been made to the design, procurement and installation philosophies for windows, especially the development of design criteria and the application of pressure proof-testing. This paper provides a review of key requirements and issues, with recommendations to allow development of the code for acceptance by nuclear regulators for tokamaks such as the International Tokamak Experimental Reactor (ITER) and future fusion reactor power plants.

Canadian Nuclear Safety Commission: Readiness to Regulate SMRs in Canada

2011 ◽  
Author(s):  
S. Herstead ◽  
M. de Vos ◽  
S. Cook
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Safety Analysis ◽  
Reactor Power ◽  
Nuclear Safety ◽  
Regulatory Framework ◽  
Nuclear Industry ◽  
Plant Design ◽  
Regulatory Documents

The success of any new build project is reliant upon all stakeholders — applicants, vendors, contractors and regulatory agencies — being ready to do their part. Over the past several years, the Canadian Nuclear Safety Commission (CNSC) has been working to ensure that it has the appropriate regulatory framework and internal processes in place for the timely and efficient licensing of all types of reactor, regardless of size. This effort has resulted in several new regulatory documents and internal processes including pre-project vendor design reviews. The CNSC’s general nuclear safety objective requires that nuclear facilities be designed and operated in a manner that will protect the health, safety and security of persons and the environment from unreasonable risk, and to implement Canada’s international commitments on the peaceful use of nuclear energy. To achieve this objective, the regulatory approach strikes a balance between pure performance-based regulation and prescriptive-based regulation. By utilizing this approach, CNSC seeks to ensure a regulatory environment exists that encourages innovation within the nuclear industry without compromising the high standards necessary for safety. The CNSC is applying a technology neutral approach as part of its continuing work to update its regulatory framework and achieve clarity of its requirements. A reactor power threshold of approximately 200 MW(th) has been chosen to distinguish between large and small reactors. It is recognized that some Small Modular Reactors (SMRs) will be larger than 200 MW(th), so a graded approach to achieving safety is still possible even though Nuclear Power Plant design and safety requirements will apply. Design requirements for large reactors are established through two main regulatory documents. These are RD-337 Design for New Nuclear Power Plants, and RD-310 Safety Analysis for Nuclear Power Plants. For reactors below 200 MW(th), the CNSC allows additional flexibility in the use of a graded approach to achieving safety in two new regulatory documents: RD-367 Design of Small Reactors and RD-308 Deterministic Safety Analysis for Small Reactors. The CNSC offers a pre-licensing vendor design review as an optional service for reactor facility designs. This review process is intended to provide early identification and resolution of potential regulatory or technical issues in the design process, particularly those that could result in significant changes to the design or analysis. The process aims to increase regulatory certainty and ultimately contribute to public safety. This paper outlines the CNSC’s expectations for applicant and vendor readiness and discusses the process for pre-licensing reviews which allows vendors and applicants to understand their readiness for licensing.

Best Estimate Simulation and Uncertainty Analysis of LB LOCA for KudanKulam VVER-1000 NPP

2009 ◽  
Author(s):  
Luben Sabotinov ◽  
Abhishek Srivastava ◽  
Pierre Probst
Keyword(s):  
Uncertainty Analysis ◽  
Nuclear Power ◽  
Input Data ◽  
Power Plants ◽  
Computer Code ◽  
Uncertainty Evaluation ◽  
Statistical Tool ◽  
Best Estimate ◽  
Safety Margins ◽  
Input Uncertainties

In the accident analysis of the Nuclear Power Plants (NPP) nowadays the international licensing practice considers several acceptable options for demonstrating the safety i.e. use of conservative computer codes with conservative assumptions, best estimate codes combined with conservative assumptions and conservative input data and application of best estimate codes with assumptions and realistic input data but associated with uncertainty evaluation of the results. The last option is particularly attractive because it allows for more precise prediction of safety margins with respect to safety criteria and their future use for power up-rating. The best estimate simulation with uncertainty analysis constitutes the framework of the present study which is to apply the last version of the French best estimate computer code CATHARE 2 in order to predict the thermal-hydraulic phenomena in the Indian KudanKulam Nuclear Power Plant (KK NPP) with VVER-1000 reactors during LB LOCA and to evaluate uncertainty along with sensitivity studies using the IRSN methodology. The paper first describes the modeling aspects of LB LOCA with CATHARE and then it presents the basic results. It highlights the use of SUNSET statistical tool developed by IRSN for sampling, management of several runs using CATHARE and further post treatment for uncertainty and sensitivity evaluation. The paper also deals with the difficulties associated with the selection of input uncertainties, code applicability and discusses the challenges in uncertainty evaluation.

scholarly journals Technical Maturity Assessment of Risk-Informed Safety Analysis Tools

2020 ◽  
Author(s):  
Yong-Joon Choi
Keyword(s):  
Best Practice ◽  
Safety Margin ◽  
Assessment Method ◽  
Nuclear Industry ◽  
Analysis Tool ◽  
Program Optimization ◽  
Analysis Tools ◽  
Maturity Assessment ◽  
Safety Margins ◽  
Requirement Traceability

Abstract Ensuring maximum safety while enhancing economic benefit is one of most important goal of In the of US Light Water Reactor Sustainability (LWRS) program. Optimization of the safety margins will provide best practice to achieve this goal which can also lead to cost reduction. Under the LWRS framework, the Risk-Informed Systems Analysis (RISA) Pathway has been focusing on the optimization of safety margin and minimization of uncertainties to ensure both safety and economics at the highest level. One of the important activities of the pathway is to deploy risk-informed analysis tools to related nuclear industry to support precise representation of safety margins and factors that contribute to cost and safety. The tools therefore need highest technical maturity so that industry can use immediately with strong credibility. The tools should have a capability to support risk-informed decision making for both probabilistic and deterministic elements of safety. The RISA Pathway, therefore, have been performing a comprehensive assessment of technical maturity and verification and validation (V&V) status of selected tools to improve adaptability to the industry. The technical maturity assessment includes three work scope: (a) define requirements based on risk-informed concept; (b) investigate and review development and V&V status for technical maturity assessment; and (c) identify technical gap and propose improvement to meet RISA toolkit requirements. The Requirement Traceability Matrix (RTM) concept was used to capture the requirements from user and developer of the project and/or software. The importance of each requirements was evaluated by Phenomena Identification and Ranking Technology (PIRT) which systematically gathers information and ranks the importance of the information. Finally, degree of the maturity was measured by Technology Readiness Level (TRL) for estimating the maturity of the technologies during the development and acquisition phase of certain technology. This paper summarizes development of assessment method and technical evaluation of multi-purpose probabilistic risk analysis tool RAVEN.

An Assessment of Margin in Design Steam Hammer Forces for Combined Cycle Power Plants

2002 ◽  
Author(s):  
D. Zheng ◽  
A. T. Vieira ◽  
J. M. Jarvis
Keyword(s):  
Power Plants ◽  
Classical Method ◽  
Combined Cycle ◽  
Nuclear Industry ◽  
Piping System ◽  
Valve Closure ◽  
Velocity Correlation ◽  
Two Phase ◽  
Best Estimate ◽  
Main Steam

All combined cycle steam plants have rapid-closing stop valves in steam lines to protect the turbine. The rapid valve closure produces a steam hammer in the piping resulting in large forces for which the piping system and supporting structures need to be designed. These forces are typically calculated using the classical Method Of Characteristics (MOC) solution. An evaluation has been conducted which compares the forces computed using the classical methods with a best-estimate approach. This comparison has been done to define margin, and to benchmark and identify potential refinements in the techniques used for evaluating steam hammer loads. The best-estimate approach involves the use of the RELAP5 computer program. RELAP5 is used extensively in the Nuclear Industry to evaluate fast thermal hydraulic transients. It has the capability to analyze subcooled liquid, two-phase and saturated or superheated steam piping system. The models used in RELAP5 are best estimate results in comparison to the MOC solution which are mathematically derived from theory. The compressible flow program GAFT is used to obtain the MOC solution. The main steam line of a single Heat Recovery Steam Generator combined cycle plant is modeled with both the GAFT program and with a PC version of RELAP5. Identical piping lengths, mass flow rates, pressures are used in each model. Also, a stop valve closure time of 100 milliseconds is modeled. As RELAP5 output results are pressure, flow rate, velocity, and density, the resultant forces are generated using the R5FORCE program, a post-processor to compute associated transient forces on straight piping links. The GAFT program, which is specifically designed to compute steam hammer forces, computes the force history internally on straight piping lengths. A comparison of the peak force from GAFT and from RELAP for every piping link has been generated. Through the comparison, both RELAP5 and GAFT have been verified for the evaluation of rapid valve closure reaction loads. The comparison also shows that the classical method typically over-predicts the best-estimate solution by 15% to 20% for straight piping links. Although not confirmed, a better agreement between the two methods would be expected if a more accurate steam sonic velocity correlation and valve closure model are incorporated into the classical solution. Theis study helps to quantify the degree of conservatism inherent in the classical approach.

scholarly journals Institutional Conditions for the Development of the Nuclear-Industrial Complex and Radioactive Waste Management in Ukraine

2020 ◽  
pp. 49-62
Author(s):  
V. Shevchenko ◽  
◽  
А. Mukhachev ◽  
V. Lyashenko ◽  
N. Osadcha ◽  
...  
Keyword(s):  
Radioactive Waste ◽  
Waste Management ◽  
Power Plants ◽  
Process Development ◽  
Economic Security ◽  
Natural Uranium ◽  
Nuclear Industry ◽  
Industrial Complex ◽  
Radioactive Waste Management ◽  
The Creation

Trends in the development of the nuclear-industrial complex and radioactive waste management are analyzed. Among the main problems of development of the nuclear-industrial complex and radioactive waste management are the imperfection of the legal framework, lack of investment. The contribution of the nuclear industry of Ukraine to the creation of gross domestic product is not significant, but its role is important in ensuring economic security and achieving energy independence of the country. The state of the nuclear-industrial complex in other countries of the world has been studied. It is expedient to use an innovative approach for the development of the nuclear-industrial complex. This approach is presented as a set of three interrelated blocks, namely: methodological and informational; diagnostic and orientation; evaluation and procedural. Directly, the nuclear-industrial complex, which is a leading link in the nuclear-energy complex of Ukraine, can be considered a complex sector of the national economy, including: uranium production, which creates a basis for meeting the needs of nuclear power plants in natural uranium in the medium and long term; zirconium production, which involves the establishment of zirconium production. The results of the implementation of the regional program for the development of the nuclear-industrial complex should include the following: increasing the competitiveness of enterprises in key strategic industries: nuclear, mining, metallurgy, chemical and mechanical engineering; increasing the innovation of production through the development of scientific potential of the region, the commercialization of the scientific process; development of enterprises on the basis of the latest technologies of industrial waste processing, including for the development of the region's infrastructure; reduction of man-caused load on the environment; creating more attractive and diverse jobs; ensuring the stabilization of social processes in the mining regions. One of the directions of modernization of the nuclear-industrial complex of Ukraine is the creation of SMR reactors (Smallmodularreactor) and its installation instead of the existing ones. Their production must be carried out at Ukrainian enterprises. The necessity of increasing the volume of uranium production, the level of its enrichment and at the same time solving environmental issues on waste disposal is substantiated.

scholarly journals Risk-Informed Safety Margin Characterization

2009 ◽  
Author(s):  
Stephen M. Hess ◽  
Nam Dinh ◽  
John P. Gaertner ◽  
Ronaldo Szilard
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
National Laboratory ◽  
Safety Margin ◽  
The United States ◽  
Department Of Energy ◽  
United States Department ◽  
Plant Design ◽  
Time Horizons ◽  
Safety Margins

The concept of safety margins has served as a fundamental principle in the design and operation of commercial nuclear power plants (NPPs). Defined as the minimum distance between a system’s “loading” and its “capacity”, plant design and operation is predicated on ensuring an adequate safety margin for safety-significant parameters (e.g., fuel cladding temperature, containment pressure, etc.) is provided over the spectrum of anticipated plant operating, transient and accident conditions. To meet the anticipated challenges associated with extending the operational lifetimes of the current fleet of operating NPPs, the United States Department of Energy (USDOE), the Idaho National Laboratory (INL) and the Electric Power Research Institute (EPRI) have developed a collaboration to conduct coordinated research to identify and address the technological challenges and opportunities that likely would affect the safe and economic operation of the existing NPP fleet over the postulated long-term time horizons. In this paper we describe a framework for developing and implementing a Risk-Informed Safety Margin Characterization (RISMC) approach to evaluate and manage changes in plant safety margins over long time horizons.

The Application of Best Estimate and Uncertainty Analysis Methodology to Large LOCA Power Pulse in a CANDU 6 Reactor

2002 ◽  
Author(s):  
A. Abdul-Razzak ◽  
J. Zhang ◽  
H. E. Sills ◽  
L. Flatt ◽  
D. Jenkins ◽  
...  
Keyword(s):  
Statistical Analysis ◽  
Uncertainty Analysis ◽  
Large Loss ◽  
Power Pulse ◽  
Analysis Methodology ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Safety Margins ◽  
Loss Of Coolant ◽  
Two Phases

The paper describes briefly a best estimate plus uncertainty analysis (BE+UA) methodology and presents its prototyping application to the power pulse phase of a limiting large Loss-of-Coolant Accident (LOCA) for a CANDU 6 reactor fuelled with CANFLEX® fuel. The methodology is consistent with and builds on world practice [1], [2]. The analysis is divided into two phases to focus on the dominant parameters for each phase and to allow for the consideration of all identified highly ranked parameters in the statistical analysis and response surface fits for margin parameters. The objective of this analysis is to quantify improvements in predicted safety margins under best estimate conditions.

A Simplified Methodology for Extremely Low-Probability Rupture Prediction for PWR LBB Applications

2011 ◽  
Author(s):  
Heqin Xu ◽  
Ashok Nana ◽  
Samer Mahmoud ◽  
Doug Killian
Keyword(s):  
Conditional Probability ◽  
Assessment Tool ◽  
Safety Margin ◽  
Nuclear Industry ◽  
Leak Rate ◽  
Fluid System ◽  
Design Basis ◽  
Safety Margins ◽  
Low Probability ◽  
The U.S

The leak-before-break (LBB) applicability is stated in General Design Criterion 4 (GDC-4) of Title 10 of the Code of Federal Regulation Part 50 (10 CFR 50). GDC-4 requires that analyses reviewed and approved by the U.S. Nuclear Regulatory Commission (NRC) demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping, in order that dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis. Standard review plan 3.6.3 (SRP-3.6.3) further requires a simultaneous safety margin of two and ten on the flaw size and leak rate detectability, respectively, for deterministic analyses, believing that the very conservative and restrictive safety margins would lead to extremely low probability of fluid system piping rupture. The technology advancements of recent years make it possible to numerically quantify the probability of rupture with confidence. Planned for completion within the next six years, a long-term, large-scale assessment tool, xLPR, is currently being developed by the U.S. NRC, in cooperation with the nuclear industry, to assess the extremely low probability of rupture. The tool will include comprehensive evaluations both before and after through-wall cracks are developed in the degraded components. In this study, we are going to utilize a simplified methodology to investigate the probability of piping rupture for a postulated through-wall crack. The conditional probability, when multiplied by the probability of having a through-wall crack during the life time of plant service, produces an overall probability of piping rupture. The major quantifiable uncertainties, such as the uncertainties associated with the material tensile properties and fracture toughness, and flow-path crack morphology parameters will be modeled as correlated random variables in this paper. Efficient Dimension-Reduction methods will be applied to predict this conditional probability and the results will be compared with the Monte Carlo simulation method. As a sample application of the proposed method, the relationship between the magnitude of the conditional probabilities and the required leak rate detection capability will be established.

Core Protection Calculator System: Past, Present, and Future

2010 ◽  
Author(s):  
Peter L. Hung
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Plant Protection ◽  
Power Level ◽  
Safety Margin ◽  
Operating Experience ◽  
Nucleate Boiling ◽  
The United States ◽  
Reactor Power ◽  
Safe Condition

The Core Protection Calculator System (CPCS) was the first implementation of digital computers in a nuclear power plant safety protection system. The system was based on first principles to calculate the specified acceptable fuel design limit (SAFDL) online. This approach provides the theoretical optimum safety margin. The first-of-its-kind system was installed in the United States at Arkansas Nuclear One Unit 2 (ANO-2) in 1980. Extensive efforts were made by Combustion Engineering and U.S. Nuclear Regulatory Commission (NRC) staff to gain licensing approval of the CPCS. Based on accumulated operating experience, numerous improvements were made to enhance the performance of the CPCS. The CPCS software provided the flexibility to readily accommodate these design changes. Currently, CPCS is implemented in 21 nuclear power plants in operation or under construction in the U.S.A. and Asia. The next generation CPCS will focus on optimizing the plant protection by improving the SAFDL calculation. By taking advantage of the advances in digital computer technology, the comprehensive safety analysis code will be used online. A more detailed core power map using the incore detector signals will be used as the basis of the departure from nucleate boiling ratio (DNBR) and local power density (LPD) calculation. A quick power reduction will provide adequate margin for most of the design basis events. For these events, CPCS will initiate a reactor power cutback as opposed to a reactor trip, which will maintain the plant at a safe condition with a reduced power level.

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