Computational Analysis of Transient 3D-Melt-Water Interactions

2000 ◽  
Author(s):  
Nikolay Ivanov Kolev
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Computational Analysis ◽  
Computer Code ◽  
3D Analysis ◽  
Damage Potential ◽  
Pressurized Water ◽  
Modeling Technology ◽  
Melt Water ◽  
Water Reactors

Abstract This work is part of the Siemens effort to estimate the damage potential of melt water interaction in future nuclear power plants with pressurized water reactors. After creating a modeling technology, the IVA5 computer code, verifying it by comparison with numerous separated effect tests, system tests and analytical benchmarks, performing many 2D computational analysis we present in this work complete 3D analysis of melt water interactions. Interesting conclusions for the systems analyzed are drown. Some of the limitations of the technology are also demonstrated.

Comparison of Thermal Hydraulic Simulations of Beyond Design Basis Events Using the MAAP4 and CENTS Computer Codes

2009 ◽  
Author(s):  
N. Reed LaBarge ◽  
Barbara R. Baron ◽  
Raymond E. Schneider ◽  
Mathew C. Jacob
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Computer Code ◽  
Hydraulic Modeling ◽  
Severe Accident ◽  
Pressurized Water ◽  
Design Basis ◽  
Best Estimate ◽  
Water Reactors ◽  
Thermal Hydraulic Modeling

The MAAP4 computer code (Reference 1) is often used to perform thermal hydraulic simulations of severe accident sequences for nuclear power plant Probabilistic Risk Assessments (PRAs). MAAP4 can be used to simulate accidents for both Boiling Water Reactors (BWRs) as well as Pressurized Water Reactors (PWRs). This assessment employs MAAP 4.0.6a for PWRs (References 1 and 5), which incorporates explicit thermal hydraulic modeling of the Reactor Coolant System (RCS) and Steam Generators (SGs), along with a nodalized integrated containment model. In the PRA environment, MAAP4 has been used for applications such as the development of PRA Level 1 and Level 2 success criteria and human action timings. The CENTS computer code (Reference 2) is a simulation tool that is typically used to analyze non-Loss of Coolant Accident (non-LOCA) events postulated to occur in nuclear power plants incorporating Combustion Engineering (CE) and Westinghouse Nuclear Steam Supply System (NSSS) designs. It is licensed by the NRC perform design basis non-LOCA safety analyses. It is a best estimate code which uses detailed thermal hydraulic modeling of the RCS and SGs; however, it does not model the containment performance. It is used to perform a wide spectrum of licensing and best estimate non-LOCA event analysis and has the capability to simulate operator actions. The CENTS models are the basis for several full scope simulators in the industry. The purpose of the analyses described in this paper is to compare MAAP4 and CENTS predictions for the Station Blackout (SBO) and Total Loss of Feedwater (TLOFW) scenarios for a representative PWR in the Westinghouse fleet that employs a CE NSSS design. The results of this comparison are used to highlight postulated MAAP4 user challenges and assist in developing guidance on selecting MAAP4 parameters for use in these scenarios. The results of the analyses presented in this paper indicate several useful insights. Overall, this paper shows that when care is taken to normalize the MAAP4 and CENTS primary side natural circulation flowrate and SG modeling, the trends of the MAAP4 and CENTS predictions of core uncovery agree reasonably well.

scholarly journals Reactor Vessel Internals Segmentation Experience using Mechanical Cutting Tools

2013 ◽  
Vol 10 (2) ◽  
pp. 6-10 ◽  
Author(s):  
Petr Pospíšil
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cutting Tools ◽  
Reactor Vessel ◽  
Boiling Water Reactors ◽  
Abrasive Waterjet ◽  
Pressurized Water ◽  
Water Reactors ◽  
Mechanical Cutting ◽  
Reactor Internals

Abstract Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive waterjet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Mechanical cutting has been used by Westinghouse since 1999 for both Pressurized Water Reactors (PWR’s) and Boiling Water Reactors (BWR’s) and its process has been continuously improved over the years. The complexity of the work requires well designed and reliable tools. Different band saws, disc saws, tube cutters and shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is very suitable for submerged applications. Westinghouse experience in mechanical cutting has demonstrated that it is an excellent technique for segmentation of internals. In summary, the purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation process, based on actual experience from the work that has been completed to date.

scholarly journals Plant Performance History of an Innovative Gate Valve in Critical Service Applications

2017 ◽  
Author(s):  
M. S. Kalsi ◽  
Patricio Alvarez ◽  
Thomas White ◽  
Micheal Green
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Gate Valve ◽  
Plant Performance ◽  
Flow Loop ◽  
Performance History ◽  
Pressurized Water ◽  
Maintenance Costs ◽  
History Of ◽  
Water Reactors

A previous paper [1] describes the key features of an innovative gate valve design that was developed to overcome seat leakage problems, high maintenance costs as well as issues identified in the Nuclear Regulatory Commission (NRC) Generic Letters 89-10, 95-07 and 96-05 with conventional gate valves [2,3,4]. The earlier paper was published within a year after the new design valves were installed at the Pilgrim Nuclear Plant — the plant that took the initiative to form a teaming arrangement as described in [1] which facilitated this innovative development. The current paper documents the successful performance history of 22 years at the Pilgrim plant, as well as performance history at several other nuclear power plants where these valves have been installed for many years in containment isolation service that requires operation under pipe rupture conditions and require tight shut-off in both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The performance history of the new valve has shown to provide significant performance advantage by eliminating the chronic leakage problems and high maintenance costs in these critical service applications. This paper includes a summary of the design, analysis and separate effects testing described in detail in the earlier paper. Flow loop testing was performed on these valves under normal plant operation, various thermal binding and pressure locking scenarios, and accident/pipe rupture conditions. The valve was designed, analyzed and tested to satisfy the requirements of ANSI B16.41 [9]; it also satisfies the requirements of ASME QME 1-2012 [10]. The results of the long-term performance history including any degradation observed and its root cause are summarized in the paper. Paper published with permission.

Power Maximization of Fully Passive Lead-Bismuth Eutectic Cooled Small Modular Reactor

2014 ◽  
Author(s):  
Jaehyun Cho ◽  
Yong-Hoon Shin ◽  
Il Soon Hwang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Natural Circulation ◽  
Reactor Core ◽  
Design Stage ◽  
Pressurized Water ◽  
Small Modular Reactor ◽  
Power Maximization ◽  
Lead Bismuth Eutectic ◽  
Water Reactors

Although the current Pressurized Water Reactors (PWRs) have significantly contributed to the global energy supply, PWRs have not been considered as a trustworthy energy solution owing to its several problems; spent nuclear fuels (SNFs), nuclear safety, and nuclear economy. In order to overcome these problems, lead-bismuth eutectic (LBE) fully passive cooling Small Modular Reactor (SMR) system is suggested. It is possible to not only provide the solution of the problem of SNFs through the transmutation feature of LBE coolant, but also increase the safety and economy through the concepts of the natural circulation cooling SMRs. It is necessary to maximize the advantages (safety and economy) of this type of Nuclear Power Plants for several applications in future. Accordingly, objective of the study is to maximize the reactor core power while the limitations of shipping size, materials endurance, long-burning criticality as well as safety under Beyond Design Basis Events must be satisfied. Design limitations of natural circulating LBE-cooling SMRs are researched and power maximization method is developed based on obtained design limitations. It is expected that the results are contributed to reactor design stage with providing several insights to designers as well as the methods for design optimization of other type of SMRs.

Background on Low Temperature Overpressure Protection System Setpoints for Pressure-Temperature Curves

2020 ◽  
Author(s):  
Jeffrey C. Poehler ◽  
Gary L. Stevens ◽  
Anees A. Udyawar ◽  
Amy Freed
Keyword(s):  
Low Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Pressure Vessels ◽  
Pressurized Water Reactors ◽  
Documentation System ◽  
Pressurized Water ◽  
System Pressure ◽  
Water Reactors

Abstract ASME Code, Section XI, Nonmandatory Appendix G (ASME-G) provides a methodology for determining pressure and temperature (P-T) limits to prevent non-ductile failure of nuclear reactor pressure vessels (RPVs). Low-Temperature Overpressure Protection (LTOP) refers to systems in nuclear power plants that are designed to prevent inadvertent challenges to the established P-T limits due to operational events such as unexpected mass or temperature additions to the reactor coolant system (RCS). These systems were generally added to commercial nuclear power plants in the 1970s and 1980s to address regulatory concerns related to LTOP events. LTOP systems typically limit the allowable system pressure to below a certain value during plant operation below the LTOP system enabling temperature. Major overpressurization of the RCS, if combined with a critical size crack, could result in a brittle failure of the RPV. Failure of the RPV could make it impossible to provide adequate coolant to the reactor core and result in a major core damage or core melt accident. This issue affected the design and operation of all pressurized water reactors (PWRs). This paper provides a description of an investigation and technical evaluation regarding LTOP setpoints that was performed to review the basis of ASME-G, Paragraph G-2215, “Allowable Pressure,” which includes provisions to address pressure and temperature limitations in the development of P-T curves that incorporate LTOP limits. First, high-level summaries of the LTOP issue and its resolution are provided. LTOP was a significant issue for pressurized water reactors (PWRs) starting in the 1970s, and there are many reports available within the U.S. Nuclear Regulatory Commission’s (NRC’s) documentation system for this topic, including Information Notices, Generic Letters, and NUREGs. Second, a particular aspect of LTOP as related to ASME-G requirements for LTOP is discussed. Lastly, a basis is provided to update Appendix G-2215 to state that LTOP setpoints are based on isothermal (steady-state) conditions. This paper was developed as part of a larger effort to document the technical bases behind ASME-G.

Coordinated Research Projects at IAEA Concerning RPV and Structural Integrity

2006 ◽  
Author(s):  
William Server ◽  
Timothy Hardin ◽  
Milan Brumovsky´
Keyword(s):  
Ferritic Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Current Status ◽  
Research Projects ◽  
Energy Agency ◽  
Pressurized Water ◽  
Toughness Testing ◽  
Water Reactors

The International Atomic Energy Agency (IAEA) has had a series of reactor pressure vessel (RPV) structural integrity programs that started back in the 1970s. These Coordinated Research Projects most recently have focused on use of the Master Curve fracture toughness testing approach for RPV and other ferritic steel components and on the issue of pressurized thermal shock (PTS) in operating pressurized water reactors. This paper will provide the current status for these projects and discuss the implications for improved safety of key ferritic steel components in nuclear power plants (NPPs).

Comparison of Nuclear Fitness for Service Codes: ASME Section XI and RSE-M AFCEN Codes

2014 ◽  
Author(s):  
Claude Faidy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Light Water Reactors ◽  
Pressurized Water ◽  
Term Operation ◽  
Mechanical Components ◽  
Water Reactors ◽  
Test Requirements ◽  
Service Inspection

Two major Codes are used for Fitness for Service of Nuclear Power Plants: one is the ASME B&PV Code Section XI and the other one is the French RSE-M Code. Both of them are largely used in many countries, partially or totally. The last 2013 RSE-M covers “Mechanical Components of Pressurized Water Reactors (PWRs): - Pre-service and In-service inspection - Surveillance in operation or during shutdown - Flaw evaluation - Repairs-Replacements parts for plant in operation - Pressure tests The last 2013 ASME Section XI covers “Mechanical components and containment of Light Water Reactors (LWRs)” and has a larger scope with similar topics: more types of plants (PWR and Boiling Water Reactor-BWR), other components like metallic and concrete containments… The paper is a first comparison covering the scope, the jurisdiction, the general organization of each section, the major principles to develop In Service Inspection, Repair-Replacement activities, the flaw evaluation rules, the pressure test requirements, the surveillance procedures (monitoring…) and the connections with Design Codes… These Codes are extremely important for In-service inspection programs in particular and essential tools to justify long term operation of Nuclear Power Plants.

The Next EDF Tracking System

2009 ◽  
Author(s):  
Emmanuelle Julli ◽  
Bertrand Lantes
Keyword(s):  
Radioactive Waste ◽  
Nuclear Power ◽  
Power Plants ◽  
Tracking System ◽  
Treatment Plant ◽  
Computer Application ◽  
Pressurized Water ◽  
Waste Package ◽  
Additional Processing ◽  
Water Reactors

EDF’s network of nuclear power plants (NPP) comprises 58 pressurized water reactors. Solid waste arising during plant operation (mainly VLLW, LLW and ILW) are conditioned and sent either to interim storage, an off site treatment plant for additional processing (e.g. the Centraco incinerator or the melting facilities of SOCODEI) or directly to one of the two final repositories operated by ANDRA, the French national radioactive waste management agency. The tracking system allows: - the checking of waste package characteristics against acceptance criteria of the final disposal facilities or off site treatment facilities; and - the transmission of the waste package data to ANDRA and SOCODEI. Since 1992, the EDF computer application DRA has been run on networked computers at EDF and ANDRA, and more recently at SOCODEI. DRA is also a key element in the management of radioactive waste. It allows a large range of inter comparisons to be made between the NPPs in operation and is thus the principal tool used optimize technical and economic performance. After 15 years of use, DRA was technically obsolete and could no longer be successfully developed to meet evolving regulatory requirements. It was therefore decided to completely replace the DRA system and in so doing to introduce new functionality.

Development of a Thermal-Hydraulic Analysis Code for Helical Coiled Once Through Steam Generator

2018 ◽  
Author(s):  
Jun Huang ◽  
Junli Gou ◽  
Haifu Ma ◽  
Jie Fan ◽  
Jianqiang Shan
Keyword(s):  
Heat Transfer ◽  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
High Efficiency ◽  
Fluid Model ◽  
Computer Code ◽  
Medium Size ◽  
Hydraulic Characteristics ◽  
Pressurized Water

Due to their advantages, such as compactness and high efficiency in heat transfer, helically coiled heat exchangers have been widely used by different type of nuclear power plants, especially by small and medium size reactors (SMRs). In order to analyze the thermal-hydraulic characteristics of a helical coiled once through steam generator (OTSG) for a small integral pressurized water reactor, a computer code is developed in this paper. The code is based on two-fluid model. The constitutive correlations are recommended based on the assessments with the compiled databases from the reviewed literatures. NUSOL SG is validated and verified against heat transfer in helical coiled tubes, and the calculation results agree well with the experiment data. The present study could provide references for the investigators to perform further investigations on the thermal hydraulic characteristics of helical coiled OTSGs.

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