scholarly journals Dynamic PRA Prospects for the Nuclear Industry

2021 ◽  
Vol 9 ◽  
Author(s):  
Nathan E. Wiltbank ◽  
Camille J. Palmer
Keyword(s):  
Nuclear Industry ◽  
Dynamic Events ◽  
Safety Margins ◽  
Dynamic Probabilistic Risk Assessment ◽  
Regulatory Bodies ◽  
Event Trees ◽  
Historical Role ◽  
Passively Safe ◽  
Exciting Time ◽  
Gen Iv

This review paper highlights approaches and tools available to the nuclear industry for dynamic probabilistic risk assessment (DPRA) using dynamic event trees. DPRA is an emerging methodology that has advantages as compared to traditional, static PRA predominantly owing to the addition of time dependent modeling. Traditional PRAs predefine events and outcomes into Event Trees (ET) and Fault Trees (FT), that are coupled with various combinations of Initiating Events (IE), Top Events (TE), branches, end states and sequences. A more complete depiction of the system and accident progression behavior can be quantified using DPRA to account for dynamic events such as those involving human actions. This paper discusses the strengths and needs of existing DPRA tools to align with the risk informed methodology currently used in the nuclear industry. DPRA is evolving during an exciting time in the nuclear industry with emerging advanced reactor designs also coming on the scene. Advanced nuclear (Gen IV) designs often incorporate passively safe systems that have less readily available data for traditional PRA due to their limited operating history. DPRA is a promising methodology that can address this challenge and demonstrate to the regulatory bodies and public that advanced designs operate within safety margins. In this light, the paper considers the historical role of PRA in the nuclear industry and motivation for considering dynamic PRA models. An introduction to the differences inherent in DPRA and how it complements and enhances existing PRA approaches is discussed. Additionally, a review of research from U.S national laboratories and universities features recent DPRA tool advancements that could be applied in the nuclear industry. These DPRA approaches and tools are summarized and examined to thoughtfully provide a path forward to best leverage existing research and integrate DPRA into advanced reactor design and analysis.

A Heuristic for Designing Hybrid Compact Heat Exchangers for Nuclear Applications

2021 ◽  
Author(s):  
Venkata Rajesh Saranam ◽  
Peter Carter ◽  
Kyle Rozman ◽  
Ömer Dogan ◽  
Brian K. Paul
Keyword(s):  
Heat Exchanger ◽  
Diffusion Bonding ◽  
Heat Exchangers ◽  
Nuclear Power ◽  
Nuclear Industry ◽  
Compact Heat Exchangers ◽  
Bonding Parameters ◽  
Heat Exchanger Design ◽  
Creep Fatigue ◽  
Gen Iv

Abstract Hybrid compact heat exchangers (HCHEs) are a potential source of innovation for intermediate heat exchangers in nuclear industry, with HCHEs being designed for Gen-IV nuclear power applications. Compact heat exchangers are commonly fabricated using diffusion bonding, which can provide challenges for HCHEs due to resultant non-uniform stress distributions across hybrid structures during bonding, leading to variations in joint properties that can compromise performance and safety. In this paper, we introduce and evaluate a heuristic for determining whether a feasible set of diffusion bonding conditions exist for producing HCHE designs capable of meeting regulatory requirements under nuclear boiler and pressure vessel codes. A diffusion bonding model for predicting pore elimination and structural analyses are used to inform the heuristic and a heat exchanger design for 316 stainless steel is used to evaluate the efficacy of the heuristic to develop acceptable diffusion bonding parameters. A set of diffusion bonding conditions were identified and validated experimentally by producing various test coupons for evaluating bond strength, ductility, porosity, grain size, creep rupture, creep fatigue and channel deviation. A five-layer hybrid compact heat exchanger structure was fabricated and tensile tested demonstrating that the bonding parameters satisfy all criteria in this paper for diffusion bonding HCHEs with application to the nuclear industry.

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.

Modal Analysis of a Triangular-Pitch Bundle Subjected to Two-Phase Cross-Flow

2017 ◽  
Author(s):  
Enrico Deri
Keyword(s):  
Modal Analysis ◽  
Fluid Dynamic ◽  
Cross Flow ◽  
Nuclear Industry ◽  
Mode Shapes ◽  
Physical Analysis ◽  
Maintenance Policy ◽  
Two Phase ◽  
Safety Margins ◽  
Lift And Drag

Flow-induced vibrations of tubes in two-phase heat exchangers are a concern for the nuclear industry. EDF has developed a numerical tool, which allows one to evaluate safety margins and thereafter to optimize the exchanger maintenance policy. The software is based on a semi analytical model of fluid-dynamic forces and dimensionless fluid force coefficients which need to be evaluated by experiment. A test rig was presented in previous PVP conferences with the aim of assessing parallel triangular tube arrangement submitted to a two-phase vertical cross-flow: a kernel of nine flexible tubes is set in the middle of a rigid bundle. These tubes vibrate as solid bodies (in translation) both in the lift and drag directions. This paper presents some extended physical analysis applied to some selected points of the aforementioned experiment series: the response modes are identified by means of operational modal analysis (i.e. under unmeasured flow excitation) and presented in terms of frequency, damping and mode shapes. Among all the modes theoretically possible in the bundle, it was found that some of them have a higher response depending on the flow velocity and the void fraction. Mode shapes allow to argue if lock-in is present and to clarify the role of lift and drag forces close to the fluidelastic instability.

scholarly journals Reliability Prediction of Acrylonitrile O-Ring for Nuclear Power Applications Based on Shore Hardness Measurements

Polymers ◽  
2021 ◽  
Vol 13 (6) ◽  
pp. 943
Author(s):  
Alvaro Rodríguez-Prieto ◽  
Ernesto Primera ◽  
Mariaenrica Frigione ◽  
Ana María Camacho
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Storage Time ◽  
Storage Temperature ◽  
Hardness Measurement ◽  
Nuclear Industry ◽  
Service Performance ◽  
Service Durability ◽  
Regulatory Bodies ◽  
Useful Life

The degradation of polymeric components is of considerable interest to the nuclear industry and its regulatory bodies. The objective of this work was the development of a methodology to determine the useful life—based on the storage temperature—of acrylonitrile O-rings used as mechanical sealing elements to prevent leakages in nuclear equipment. To this aim, a reliability-based approach that allows prediction of the use-suitability of different storage scenarios (that involve different storage times and temperatures) considering the further required in-service performance, is presented. Thus, experimental measurements of Shore A hardness have been correlated with storage variables (temperature and storage time). The storage (and its associated hardening) was proved to have a direct effect on in-service durability, reducing this by up to 60.40%. Based on this model, the in-service performance was predicted; after the first three years of operation the increase in probability of failure (POF) was practically insignificant. Nevertheless, from this point on, and especially, from 5 years of operation, the POF increased from 10% to 20% at approximately 6 years (for new and stored). From the study, it was verified that for any of the analysis scenarios, the limit established criterion was above that of the storage time premise considered in usual nuclear industry practices. The novelty of this work is that from a non-destructive test, like a Shore A hardness measurement, the useful life and reliability of O-rings can be estimated and be, accordingly, a decision tool that allows for improvement in the management of maintenance of safety-related equipment. Finally, it was proved that the storage strategies of our nuclear power plants are successful, perfectly meeting the expectations of suitability and functionality of the components when they are installed after storage.

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.

scholarly journals Siting Consideration for Nuclear Power Plant: A Review

2017 ◽  
Vol 2 (3) ◽  
Author(s):  
Muhammad Adil Khattak ◽  
Aishah Umairah ◽  
Mohamad Amirudin Mohamad Rosli ◽  
Syeheer Sabri ◽  
Mohammad Akmal Saad ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Site Selection ◽  
Nuclear Power ◽  
Selection Process ◽  
Economic Feasibility ◽  
Nuclear Industry ◽  
Geographical Environment ◽  
Detailed Evaluation ◽  
Regulatory Bodies

The purpose of this research is to study in detail about the site selection process in nuclear power plant (NPP) construction. There are various factors that contribute to the site selection which involves in-depth investigation and detailed evaluation before the site is being finalized and proposed. There are two main objectives in siting of NPP; ensuring the technical and economic feasibility of the plant and minimising potential adverse impacts on the community and environment. Geographical environment also plays an important role in siting of NPP where the source of water should be abundance. Country requirement for siting of NPP would be different for every country where they are controlled by their own regulatory bodies. About 64 published studies (1967-2017) are reviewed in this paper. It is marked from the literature survey articles that siting process is extremely crucial step in constructing NPP where public acceptance is one of the main factors, that need to be considered. Malaysia is one of the new countries embark in nuclear industry that still is in the planning phase to plan and construct its first NPP.

Development and Application of Irradiation Technology in HANARO

2008 ◽  
Vol 135 ◽  
pp. 93-98 ◽  
Author(s):  
Kee Nam Choo ◽  
Bong Goo Kim
Keyword(s):  
Functional Materials ◽  
Nuclear Industry ◽  
Precise Control ◽  
Research Institutes ◽  
Irradiation Test ◽  
Temperature Irradiation ◽  
Gen Iv ◽  
Electrical Materials ◽  
Irradiation Technology ◽  
Material Irradiation

A material capsule system including a main capsule, a fixing, a control, a cutting, and a transport system was developed for an irradiation test of non-fissile materials in HANARO. This capsule system has been actively utilized for the various material irradiation tests requested by users from research institutes, universities, and the industries. Based on the accumulated irradiation experience and the user’s sophisticated requirements, the instrumented capsule technologies for a more precise control of the irradiation temperature and fluence of a specimen were developed in HANARO. 6,000 specimens from domestic research institutes, nuclear industry companies and universities, were irradiated in HANARO for 54,000 hours by using the developed capsule irradiation system. Recently, development of new instrumented capsule technologies for an IP/OR irradiation test and a high temperature irradiation test has been performed in HANARO for relevant ongoing R&D for the international Gen IV Program. New irradiation technologies are also under development for the production of new functional materials including superconductor and optical/electrical materials.

CNSC Review of Safety Improvements: CANDU Fuel Modification in Canada

2013 ◽  
Author(s):  
Yujun Guo
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Safety ◽  
Nuclear Industry ◽  
Fuel Bundle ◽  
Safety Margins ◽  
Fuel Design

Aging of a CANDU nuclear power plant affects various safety margins of the plant. Margin to fuel sheath dryout is one of the safety margins that have been detrimentally affected, leading to a reduced margin to dryout with time. If no proactive actions are taken, the plant will have to de-rate its operation at an earlier time. To postpone the de-rating, the Canadian nuclear Industry has taken multi-initiatives to restore, or partially restore the safety margins that have been eroded due to plant aging. One of the initiatives is modification/re-optimization of the current fuel design, in order to improve the fuel thermalhydraulic performance, i.e., to suppress fuel sheath dryout, whereby offset partially the erosion of margin to fuel sheath dryout. In response to the initiative of fuel bundle modification, the Canadian Nuclear Safety Commission (CNSC) — the nuclear safety regulator — has set clear requirements and expectations and followed rigours processes and procedures for reviewing and licensing the modified fuel. This paper summarizes the fuel modification program in Canada, and the CNSC requirements, expectations, and review processes associated with licensing review of fuel modifications in Canada.

ACR-1000™ Design and Safety Performance

2010 ◽  
Author(s):  
Stephen Yu ◽  
Zoran Bilanovic ◽  
Patrick Reid ◽  
Paul Santamaura ◽  
Mike Soulard
Keyword(s):  
Phase 1 ◽  
Safety Margin ◽  
Safety Analysis ◽  
Operational Performance ◽  
Inherent Safety ◽  
Regulatory Requirements ◽  
Candu Reactors ◽  
Safety Margins ◽  
Safety Case ◽  
Regulatory Bodies

The ACR-1000™ design has evolved from AECL’s in depth knowledge of CANDU® systems, components and materials based on the CANDU 6 design, as well as the experience and feedback received from owners and operators of CANDU plants. The ACR® design retains the proven strengths and features of CANDU reactors, while incorporating innovations and state-of-the-art technology. It also features major improvements in inherent safety characteristics, safety margin and operational performance. The ACR design has been reviewed by domestic and international regulatory bodies, and has been given a positive regulatory opinion about its licensability in Canada and internationally. The Canadian regulator, the Canadian Nuclear Safety Commission (CNSC) completed the Phase 1 [1] and Phase 2 [2] pre-project design reviews in December 2008 and August 2009 respectively, and concluded that there are no fundamental barriers to licensing the ACR-1000 design in Canada. The generic PSAR for ACR-1000 was completed in September 2009. The PSAR contains the ACR-1000 design details, the safety and design methodology, and the safety analysis that demonstrate ACR-1000 safety case and compliance with Canadian and international regulatory requirements and expectations. The final stage of the ACR-1000 design is currently underway including documentation and additional confirmatory analysis, and the basic engineering will be completed in 2011. This paper provides an overview of the ACR-1000 design including a summary of the safety methodology used and compliance with regulatory and customer requirements, along with a demonstration of how modern expectations on safety margins and operational performance (i.e., typically characterized as Generation III+) are met. It also provides a summary of the safety analysis results (both deterministic and probabilistic) from the generic safety analysis that has been completed.

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