scholarly journals An Overview of the Pressurized Thermal Shock Issue in the Context of the NURESIM Project

2009 ◽  
Vol 2009 ◽  
pp. 1-13 ◽  
Author(s):  
D. Lucas ◽  
D. Bestion ◽  
E. Bodèle ◽  
P. Coste ◽  
M. Scheuerer ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Cfd Simulation ◽  
Two Phase ◽  
Pressurized Water ◽  
Simulation Tools ◽  
Pressurized Thermal Shock ◽  
Qualitative Level ◽  
Water Reactors ◽  
Near Term ◽  
Accident Scenarios

Within the European Integrated Project NURESIM, the simulation of PTS is investigated. Some accident scenarios for Pressurized Water Reactors may cause Emergency Core Coolant injection into the cold leg leading to PTS situations. They imply the formation of temperature gradients in the thick vessel walls with consequent localized stresses and the potential for propagation of possible flaws present in the material. This paper focuses on two-phase conditions that are potentially at the origin of PTS. It summarizes recent advances in the understanding of the two-phase phenomena occurring within the geometric region of the nuclear reactor,that is, the cold leg and the downcomer, where the “PTS fluid-dynamics" is relevant. Available experimental data for validation of two-phase CFD simulation tools are reviewed and the capabilities of such tools to capture each basic phenomenon are discussed. Key conclusions show that several two-phase flow subphenomena are involved and can individually be simulated at least at a qualitative level, but the capability to simulate their interaction and the overall system performance is still limited. In the near term, one may envisage a simplified treatment of two-phase PTS transients by neglecting some effects which are not yet well controlled, leading to slightly conservative predictions.

Small Modular Reactor and Large Nuclear Reactor Fuel Cost Comparison

2014 ◽  
Author(s):  
Christopher P. Pannier ◽  
Radek Škoda
Keyword(s):  
Nuclear Reactor ◽  
Cost Comparison ◽  
Design Parameters ◽  
Fuel Cost ◽  
Uranium Enrichment ◽  
Pressurized Water ◽  
Nuclear Reactor Fuel ◽  
Water Reactors ◽  
Near Term ◽  
Fuel Costs

Small modular reactors (SMRs) offer simple, standardized, and safe modular designs for new nuclear reactor construction. Factory built SMRs promise competitive economy when compared with the current reactor fleet. Construction cost of a majority of the projects, which are mostly in their design stages, is not publicly available, but variable costs can be determined from fuel enrichment, average burn-up, and plant thermal efficiency, which are published design parameters for many near-term SMR projects. This paper gives a simulation of the fuel cost of electricity generation for selected SMRs and large reactors, including calculation of optimal tails assay in the uranium enrichment process. The fuel costs of several SMR designs are compared between one another and with current generation large reactor designs providing a rough comparison of the long-term economics of a new nuclear reactor project. SMRs are predicted to have higher fuel costs than large reactors. Particularly, integral pressurized water reactors (iPWRs) are shown to have from 15% to 60% higher fuel costs than large reactors. Fuel cost sensitivities to reactor design parameters are presented.

Risk-Informed Analysis for Accident Tolerant Fuel in Pressurized Water Reactors

2020 ◽  
Author(s):  
Zhegang Ma ◽  
Carlo Parisi ◽  
Cliff Davis ◽  
Sai Zhang ◽  
Hongbin Zhang
Keyword(s):  
System Analysis ◽  
Hydrogen Generation ◽  
Cooling System ◽  
Pressurized Water ◽  
Water Reactor ◽  
Core Damage ◽  
Systems Research ◽  
Research Activities ◽  
Near Term ◽  
Accident Scenarios

Abstract This paper presents the research activities performed by Idaho National Laboratory (INL) for the Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program, Risk-Informed System Analysis (RISA) Pathway, Enhanced Resilient Plant (ERP) Systems research, using the probabilistic risk assessment (PRA) tool SAPHIRE and the deterministic best estimate tool RELAP5-3D for risk-informed analysis. The ERP research supports DOE and industry initiatives by developing Accident Tolerant Fuel (ATF), the Diverse and Flexible Coping Strategy (FLEX), and passive cooling system designs to enhance existing reactors’ safety features (both active and passive) and to substantially reduce operating costs of nuclear power plants (NPPs) through risk-informed approaches to analyze the plant enhancements and their characterization. The risk-informed analysis used SAPHIRE and RELAP5-3D to evaluate the risk impacts from near-term ATF (FeCrAl and Chromium-coated clads) on a generic Westinghouse three-loop pressurized water reactor (PWR) under the following accident scenarios: station blackout (SBO), loss of feedwater (LOFW), steam generator tube rupture (SGTR), loss-of-coolant accidents (LOCAs), locked rotor transient, turbine trip transient, anticipated transient without scram (ATWS), and main steam line break (MSLB). The RELAP5-3D simulations included the time to core damage, time to 0.5 kilograms hydrogen generation, and total hydrogen generation. The simulation results show there are modest gains of coping time (delay of time to core damage) due to efficacy of the near-term ATF designs in various accident scenarios. The risk benefits on behalf of the core damage frequency (CDF) brought by the ATF designs would be small for most of the scenarios. However, results revealing much less hydrogen being produced at the time of core damage show a clear benefit in adopting ATFs.

Transient Thermal Behaviors of SBO Accident for a 200MW OFNP Under Heaving Motion Conditions

2018 ◽  
Author(s):  
Qiqi Yan ◽  
Simin Luo ◽  
Yapei Zhang ◽  
Limin Liu ◽  
Guanghui Su ◽  
...  
Keyword(s):  
System Analysis ◽  
Nuclear Reactor ◽  
Analysis Model ◽  
Pressurized Water ◽  
Thermal Behaviors ◽  
Severe Accidents ◽  
Nuclear Plants ◽  
Station Blackout ◽  
Water Reactors ◽  
Transient Thermal

For some Pressurized Water Reactors (PWR) operated on automobiles, boats or deep sea vessels, system characteristics is important for understanding their safety during severe accidents. The development of an analysis code and the transient thermal beaviors of a floating nuclear reactor under heaving motion are described in this paper. By modifying the control equations based on the mathematical models of ocean conditions, an ocean condition available system analysis code named RELAP5/GR was developed from RELAP5 MOD3.2 to simulate the transient thermal-hydraulic response of the nuclear reactor systems to the motion conditions in accidents, which is an advanced and independent node programming code. Using the code, the analysis model was established for a small 200MW offshore floating nuclear plants (OFNP). The transient thermal behaviors of the whole system were analyzed in the cases of the station blackout accident under heaving motion conditons. The analysis shows that all the results can be reasonably explained and the code development is successful at this stage.

Fast Neutron Reduction Techniques for Plant Lifetime Extension

2010 ◽  
Author(s):  
Choon Sung Yoo ◽  
Byoung Chul Kim ◽  
Tae Je Kwon
Keyword(s):  
Fast Neutron ◽  
Thermal Shock ◽  
Neutron Irradiation ◽  
Neutron Transport ◽  
Reactor Vessel ◽  
Pressurized Water ◽  
Reduction Techniques ◽  
Pressurized Thermal Shock ◽  
Water Reactors ◽  
Inner Vessel

A Pressurized Thermal Shock (PTS) Event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A PTS concern arises if one of these transients acts in beltline region of a reactor vessel where a reduced fracture resistance exists due to neutron irradiation. Such an event may produce a flaw or cause the propagation of a flaw postulated to exist near the inner vessel wall surface, thereby potentially affecting the integrity of the vessel. In this paper fast neutron flux reduction techniques were implemented to reduce the potential risk of PTS due to the neutron irradiation on the pressure vessel beltline region. And the RTPTS value for the end of life of the plant was projected using the fast neutron fluence obtained by neutron transport calculations according to the various core loading pattern and reduction program possible for the future cycles.

The Sensitivity of Risk-Informed Reactor Structural Integrity Analysis Results to Various Interpretations of Warm Pre-Stress

2009 ◽  
Author(s):  
Mark T. EricksonKirk ◽  
Terry L. Dickson
Keyword(s):  
Structural Integrity ◽  
The United States ◽  
Integrated Analysis ◽  
Total Risk ◽  
Pressurized Water ◽  
Best Estimate ◽  
Water Reactors ◽  
Using Data ◽  
Accident Scenarios ◽  
Reactor Operating

Warm pre-stress, or WPS, is a phenomenon by which the apparent fracture toughness of ferritic steel can be elevated in the fracture mode transition if crack is first “pre-stressed” at an elevated temperature. Taking proper account of WPS is important to the accurate modeling of the postulated accident scenarios that, collectively, are referred to as pressurized thermal shock, and to the accurate modeling of routine cool-down transients. For both accident and routine cool-downs the transients begin at the reactor operating temperature (approximately 290°C for pressurized water reactors in the United States) and proceed to colder temperatures as time advances. The probabilistic fracture mechanics code FAVOR, which is being used by the NRC to provide the technical basis for risk-informed revisions of 10 CFR 50.61 and 10 CFR 50 Appendix G, adopts a model of WPS as part of its fracture driving force module. In this paper we assess the conservatism inherent to the FAVOR WPS model relative to a best-estimate WPS model constructed using data recently produced by the European Commission “SMILE” project and published by Moinereau and colleagues. Assessments of the conservatisms inherent to the so-called “conservative principle” WPS model, and also to a classic LEFM model that does not credit WPS are also made. The data presented herein demonstrate that, for an integrated analysis of PTS risk, the FAVOR and conservative principle WPS models both over-estimate the vessel failure risk by a factor of between 2 and 3× relative to the best estimate model. Our examination of the effect of WPS models on the predictions of individual transients reveals that for the severe transients that dominate risk there is little difference (usually less than 4×) between the conditional probabilities of crack initiation and of through wall cracking predicted by the different WPS models. There are considerable differences in the predictions of the various WPS and non-WPS models for low severity transients, however, the contribution of these transients to the total risk of vessel failure is small.

How to Simulate the Microstructure Induced by a Nuclear Reactor with an Ion Beam Facility : DART

2009 ◽  
Vol 1215 ◽  
Author(s):  
Laurence Luneville ◽  
David Simeone ◽  
Gianguido Baldinozzi ◽  
Dominique Gosset ◽  
yves serruys
Keyword(s):  
Cross Sections ◽  
Nuclear Reactor ◽  
Nuclear Reactions ◽  
Ion Beam ◽  
Fast Neutrons ◽  
Relaxation Processes ◽  
Pressurized Water ◽  
Water Reactors ◽  
Displacement Per Atom ◽  
The Impact

AbstractEven if the Binary Collision Approximation does not take into account relaxation processes at the end of the displacement cascade, the amount of displaced atoms calculated within this framework can be used to compare damages induced by different facilities like pressurized water reactors (PWR), fast breeder reactors (FBR), high temperature reactors (HTR) and ion beam facilities on a defined material. In this paper, a formalism is presented to evaluate the displacement cross-sections pointing out the effect of the anisotropy of nuclear reactions. From this formalism, the impact of fast neutrons (with a kinetic energy En superior to 1 MeV) is accurately described. This point allows calculating accurately the displacement per atom rates as well as primary and weighted recoil spectra. Such spectra provide useful information to select masses and energies of ions to perform realistic experiments in ion beam facilities.

Use of PRA in the Design of the Westinghouse AP1000 Plant

2009 ◽  
Author(s):  
Robert J. Lutz ◽  
James H. Scobel ◽  
Richard G. Anderson ◽  
Terry Schulz
Keyword(s):  
Cooling Water ◽  
Heat Removal ◽  
Accident Prevention ◽  
Severe Accident ◽  
Pressurized Water ◽  
Core Damage ◽  
Service Water ◽  
Containment Shell ◽  
Water Reactors ◽  
Accident Scenarios

Probabilistic Risk Assessment (PRA) has been an integral part of the Westinghouse AP1000, and the former AP600, development programs from its inception. The design of the AP1000 plant is based on engineering solutions to reduce or eliminate many of the dominant risk contributors found in the existing generation of Pressurized Water Reactors (PWRs). Additional risk reduction features were identified from insights gained from the AP1000 PRA as it evolved with the design of the plant. These engineered solutions include severe accident prevention features that resulted in a significant reduction in the predicted core damage frequency. Examples include the removal of dependencies on electric power (both offsite power and diesel generators) and cooling water (service water and component cooling water), removal of common cause dependencies by using diverse components on parallel trains and reducing dependence on operator actions for key accident scenarios. Engineered solutions to severe accident consequence mitigation were also used in the AP1000 design based on PRA insights. Examples include in-vessel retention of molten core debris to eliminate the potential for ex-vessel phenomena challenges to containment integrity and passive containment heat removal through the containment shell to eliminate the potential for containment failure due to steam overpressure. Additionally, because the accident prevention and mitigation features of the AP1000 are engineered solutions, the traditional uncertainties associated with the core damage and release frequency are directly addressed.

A Comparative Study on Severe Accident Phenomena Related to Melt Progression in Sodium Fast Reactors and Pressurized Water Reactors

2020 ◽  
Vol 7 (3) ◽  
Author(s):  
Andrea Bachrata ◽  
Fréderic Bertrand ◽  
Nathalie Marie ◽  
Fréderic Serre
Keyword(s):  
Mitigation Strategies ◽  
Severe Accident ◽  
Pressurized Water Reactors ◽  
Physical Parameters ◽  
Fast Reactors ◽  
Pressurized Water ◽  
The Core ◽  
Sodium Fast Reactors ◽  
Water Reactors ◽  
Accident Scenarios

Abstract The nuclear safety approach has to cover accident sequences involving core degradation in order to develop reliable mitigation strategies for both existing and future reactors. In particular, the long-term stabilization of the degraded core materials and their coolability has to be ensured after a severe accident. This paper focuses on severe accident phenomena in pressurized water reactors (PWR) compared to those potentially occurring in future GenIV-type sodium fast reactors (SFR). First, the two considered reactor concepts are introduced by focusing on safety aspects. The severe accident scenarios leading to core melting are presented and the initiating events are highlighted. This paper focuses on in-vessel severe accident phenomena, including the chronology of core damage, major changes in the core configuration and molten core progression. Regarding the mitigation means, the in-vessel retention phenomena and the core catcher characteristics are reviewed for these different nuclear generation concepts (II, III, and IV). A comparison between the PWR and SFR severe accident evolution is provided as well as the relation between governing physical parameters and the adopted mitigation provisions for each reactor concept. Finally, it is highlighted how the robustness of the safety demonstration is established by means of a combined probabilistic and deterministic approach.

scholarly journals Modelling of Boiling Flows for Nuclear Thermal Hydraulics Applications—A Brief Review

Inventions ◽  
2020 ◽  
Vol 5 (3) ◽  
pp. 47
Author(s):  
Giovanni Giustini
Keyword(s):  
Heat Transfer ◽  
Liquid Water ◽  
Cfd Simulation ◽  
High Surface ◽  
Computational Modelling ◽  
Nucleate Boiling ◽  
Heat Fluxes ◽  
Thermal Hydraulics ◽  
Two Phase ◽  
Water Reactors

The boiling process is utterly fundamental to the design and safety of water-cooled fission reactors. Both boiling water reactors and pressurised water reactors use boiling under high-pressure subcooled liquid flow conditions to achieve high surface heat fluxes required for their operation. Liquid water is an excellent coolant, which is why water-cooled reactors can have such small sizes and high-power densities, yet also have relatively low component temperatures. Steam is in contrast a very poor coolant. A good understanding of how liquid water coolant turns into steam is correspondingly vital. This need is particularly pressing because heat transfer by water when it is only partially steam (‘nucleate boiling’ regime) is particularly effective, providing a great incentive to operate a plant in this regime. Computational modelling of boiling, using computational fluid dynamics (CFD) simulation at the ‘component scale’ typical of nuclear subchannel analysis and at the scale of the single bubbles, is a core activity of current nuclear thermal hydraulics research. This paper gives an overview of recent literature on computational modelling of boiling. The knowledge and capabilities embodied in the surveyed literature entail theoretical, experimental and modelling work, and enabled the scientific community to improve its current understanding of the fundamental heat transfer phenomena in boiling fluids and to develop more accurate tools for the prediction of two-phase cooling in nuclear systems. Data and insights gathered on the fundamental heat transfer processes associated with the behaviour of single bubbles enabled us to develop and apply more capable modelling tools for engineering simulation and to obtain reliable estimates of the heat transfer rates associated with the growth and departure of steam bubbles from heated surfaces. While results so far are promising, much work is still needed in terms of development of fundamental understanding of the physical processes and application of improved modelling capabilities to industrially relevant flows.

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