Preliminary Investigations of the Feasibility of In-Vessel Melt Retention Strategies for a Small Modular Reactor Concept

2019 ◽  
Vol 5 (2) ◽  
Author(s):  
Lena Andriolo ◽  
Clément Meriot ◽  
Nikolai Bakouta
Keyword(s):  
Heat Dissipation ◽  
Analytical Calculation ◽  
Heat Fluxes ◽  
Reference Case ◽  
Pressurized Water ◽  
Vessel Steel ◽  
Small Modular Reactor ◽  
Lower Head ◽  
Water Reactors ◽  
The Impact

The study presented in this paper is part of the technological surveillance performed at the Electricité De France (EDF) Research and Development (R&D) Center, in the Pericles department, and investigates the feasibility of modeling in-vessel melt retention (IVMR) phenomena for small modular reactors (SMR) with the modular accident analysis program version 5 in its EDF proprietary version (MAAP5_EDF), applying conservative hypotheses, such as constant decay heat after corium relocation to the lower head. The study takes advantage of a corium stratification model in the lower head of the vessel, developed by EDF R&D for large-sized prospective pressurized water reactors (PWRs). The analysis is based on a stepwise approach in order to evaluate the impact of various effects during IVMR conditions. First, an analytical calculation is performed in order to establish a reference case to which the MAAP5_EDF code results are compared. In a second step, the impact of the lower head geometry, vessel steel ablation, and subsequent relocation on the heat flux has been analyzed for cases where heat dissipation through radiation is neglected (in first approximation). Finally, the impact of heat losses through radiation as well as the crust formation around the pool has been assessed. The results demonstrate the applicability of the MAAP5_EDF code to SMRs, with heat fluxes lower than 1.1 MW/m2 for relevant cases, and identify modeling improvements.

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.

Crack-Opening Displacement and Leak-Rate Calculations for Full Structural Weld Overlays

2009 ◽  
Author(s):  
D.-J. Shim ◽  
E. Kurth ◽  
F. Brust ◽  
G. Wilkowski ◽  
A. Csontos ◽  
...  
Keyword(s):  
Crack Opening ◽  
Nuclear Industry ◽  
Leak Rate ◽  
Opening Displacement ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Weld Overlay ◽  
Water Reactors ◽  
The Impact ◽  
Weld Overlays

Full structural weld overlays have been used in the U.S. nuclear power industry for over twenty years in boiling water reactors (BWRs). Primary water stress corrosion cracking (PWSCC) in nickel-based dissimilar metal welds (DMWs) has been experienced in pressurized water reactors (PWRs) since the early 1990s. As a result, the nuclear industry is implementing full structural weld overlays (FSWOL) as a PWSCC mitigation technique that may be used on primary coolant lines previously approved for Leak-Before-Break (LBB). This work investigates the effect of the FSWOL on the leakage behavior of these lines with postulated defects. In this paper, finite element (FE) based crack-opening displacements (CODs) were developed for pipes with a FSWOL with postulated complex cracks. The COD solutions were then employed in standard leak-rate calculations, where equivalent crack morphology parameters were developed to consider a flow through two different crack morphologies, i.e., PWSCC through the DMW and corrosion fatigue through the weld overlay. The results of the sensitivity study and a discussion on the impact of the weld overlay on the leakage behavior concludes this paper.

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.

Regulatory Perspective on Advanced Finite Element Flaw Growth Analysis of Alloy 82/182 Butt Welds

2008 ◽  
Author(s):  
Edmund J. Sullivan ◽  
Aladar A. Csontos ◽  
Timothy R. Lupold ◽  
Chia-Fu Sheng
Keyword(s):  
Finite Element ◽  
Nuclear Power ◽  
Regulatory Review ◽  
Pressurized Water ◽  
Butt Weld ◽  
Butt Welds ◽  
Primary Water ◽  
Water Reactors ◽  
The Impact ◽  
Growth Analyses

On October 13, 2006, the Wolf Creek Nuclear Operating Corporation performed preweld overlay inspections using manual ultrasonic testing (UT) techniques on the surge, spray, relief, and safety nozzle-to-safe end dissimilar metal (DM) and safe end-to-pipe stainless steel butt welds. The inspection identified five circumferential indications in the surge, relief, and safety nozzle-to-safe end DM butt welds that the licensee attributed to primary water stress corrosion cracking (PWSCC). These indications were significantly larger and more extensive than previously seen for the case of circumferential indications in commercial pressurized water reactors. As a result of the NRC staff’s initial flaw growth analyses, the NRC staff obtained commitments from the nuclear power industry licensees to complete pressurizer nozzle DM butt weld inspections on an accelerated basis. In addition, the industry informed the NRC staff that it would undertake a task to refine the crack growth analyses using more realistic assumptions to address the NRC staff’s concerns regarding the potential for rupture without prior evidence of leakage from circumferentially oriented PWSCC in pressurizer nozzle welds. These new analyses are referred to as advanced finite element (AFE) analyses. This paper will discuss the regulatory review of the industry’s AFE analyses. This discussion will include the NRC staff’s approach to the review, the differences between the industry’s AFE analyses and the NRC staff’s confirmatory analyses, and the NRC staff’s acceptance criteria. The paper will discuss the impact of the AFE analyses on the regulatory process. Finally, the paper will discuss possible future regulatory and research applications for AFE analyses as well as additional NRC research projects intended to address some of the uncertainties in this type of analysis.

Small Modular Reactors: Learning From the Past

2021 ◽  
Author(s):  
Esam Hussein
Keyword(s):  
High Temperature ◽  
Pressurized Water Reactors ◽  
Nuclear Technology ◽  
Pressurized Water ◽  
The Past ◽  
Canadian Experience ◽  
Small Modular Reactor ◽  
Technology Experience ◽  
Water Reactors ◽  
High Temperature Gas

Abstract Most emerging small modular reactor (SMR) designs resemble older reactors that were designed in the early days of the nuclear technology. Experience with operating these old reactors can contribute to the licensing of new SMRs, by showing that some safety concepts were already proven, and their viability was demonstrated. This paper shows examples of older reactors in each of the reemerging SMR concepts of integrated pressurized water reactors, high temperature gas cooled reactors, molten salt reactors, and liquid metal cooled fast reactors. The Canadian experience with the WR-1 organic cooled reactor is also discussed to examine whether it can inform the development on an organic cooled SMR.

The Impact of Transient Behavior of Corium in the Lower Head of a Reactor Vessel for In-Vessel Melt Retention Strategies

2016 ◽  
Author(s):  
L. Carénini ◽  
F. Fichot
Keyword(s):  
Nuclear Power ◽  
Metal Layer ◽  
Heat Fluxes ◽  
Severe Accident ◽  
Total Power ◽  
Light Water Reactors ◽  
Light Water ◽  
Accident Management ◽  
Water Reactors ◽  
The Impact

One of the main goals of severe accident management strategies is to mitigate radiological releases to people and environment. To choose the most appropriate strategy, one needs to know the probability of its success taking into account the associated uncertainties. In the field of corium and debris behavior and coolability, research programs are still on going and the possibilities to efficiently cool and retain corium and debris inside the Reactor Pressure Vessel (RPV) then inside the containment are difficult to evaluate. This leads to uncertainties in safety assessments particularly when margins to RPV or containment failure are too weak. In Vessel Melt Retention (IVMR) strategies for Light Water Reactors (PWR, BWR, VVER) intend to stabilize and retain the core melt in the RPV (as it happened during the TMI-2 accident). This would reduce significantly the threats to the last barrier (the containment) and therefore reduce the risk of release of radioactive elements to the environment. This type of Severe Accident Management (SAM) strategy has already been incorporated recently in the SAM guidance (SAMG) of several operating medium size Light Water Reactors (reactor below 500MWe (like VVER440)) and is part of the SAMG strategies for some Gen III+ PWRs of higher power like the AP1000. A European project coordinated by IRSN and gathering 23 organizations (Utilities, Technical Support Organizations, Nuclear Power Plant vendors, Research Institutes…) has been launched in 2015 with as main objective the evaluation of feasibility of IVMR strategies for Light Water Reactors (PWR, VVER, BWR) of total power around 1000MWe (which represent a significant part of the European Nuclear Power Plants fleet). This paper intends to show how it is possible to introduce transient evolutions of the stratified corium pool in the evaluation of the heat flux profile along the vessel wall. Indeed, due to chemical reactions in the U–Zr–O–Fe molten pool, separation between non-miscible metallic and oxide phases may occur, modifying the thermal load applied to the RPV. If stabilized stratified corium configurations are well defined and modeled, transient evolutions of material layers in the corium pool are still difficult to predict. The evaluations presented are based on calculations performed with the severe accident integral code ASTEC (Accident Source Term Evaluation Code) for a typical PWR reactor. The modeling of transient evolution of corium layers leads to configurations with a thin light metal layer on top of the oxidic one, increasing the so called “focusing effect” (intense heat fluxes on the RPV walls adjacent to the top metal layer). A sensitivity study on some uncertain parameters is proposed to evaluate the impact on the kinetics of layers inversion. Depending on the duration of these transient heat fluxes, the mechanical strength of the RPV could be challenged.

Westinghouse Advanced Loop Tester (WALT) Update

2008 ◽  
Author(s):  
G. Wang ◽  
W. A. Byers ◽  
M. Y. Young ◽  
Z. E. Karoutas
Keyword(s):  
Nuclear Power ◽  
Hot Spot ◽  
Test Facility ◽  
Hydraulic Test ◽  
Pressurized Water ◽  
System Pressure ◽  
Test Loop ◽  
Zinc Addition ◽  
Water Reactors ◽  
The Impact

In order to understand crud formation on the fuel rod cladding surfaces of pressurized water reactors (PWRs), a crud Thermal-Hydraulic test facility referred to as the Westinghouse Advanced Loop Tester (WALT) was built at the Westinghouse Science and Technology Department Laboratories in October 2005. Since then, a number of updates have been made and are described here. These updates include heater rod improvements, system pressure stabilization, and more effective protection systems. After these updates were made, the WALT system has been operated with higher stability and fewer failures. In this test loop, crud can be deposited on the heater rod surface and the character of the crud is similar to what has been observed in the PWRs. In addition, chemistry in the WALT loop can be varied to study its impact on crud morphology and associated parameters. The WALT loop has been successful in generating crud and measuring its thermal impact as a function of crud thickness. Currently, this test facility is supporting an Electric Power Research Institute (EPRI) program to assess the impact of zinc addition to PWR reactor coolant. Meanwhile, the WALT system is also being utilized by Westinghouse to perform dry-out and hot spot tests. These tests support the industry goal of 0 fuel failures by 2010 set by Institute of Nuclear Power Operations (INPO). Another major goal of the Westinghouse tests is to gain a better understanding of unexpected changes in core power distributions in operating reactors known as crud induced power shifts (CIPS) or axial offset anomalies (AOA).

A Systematic Review of PCTRAN-Based Pressurized Water Reactor Transient Analysis

2021 ◽  
Author(s):  
Suubi Racheal ◽  
Yongkuo Liu ◽  
Miyombo Ernest ◽  
Abiodun Ayodeji
Keyword(s):  
Nuclear Reactor ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Research Education ◽  
User Expertise ◽  
Scale Down ◽  
Water Reactors ◽  
The Impact ◽  
And Training

Abstract The impact of nuclear accidents has been a topic of debate since the construction of the first nuclear reactor, and still stands as a key issue of public concern. Several codes and simulators have been used to study the transient progression in pressurized water reactors, and to evaluate the technical measures adopted to scale down the risk of accidents. However, some of these codes are not suitable for multipurpose research and training as they require significant user expertise, leading to analysis uncertainties largely from the code user effect. This paper presents a bird-eye view of one of the most widely used nuclear reactor transient analyzer — the Personal Computer Transient Analyzer (PCTRAN). This paper discusses the comparative advantages of the simulator from the users’ perspective, with specific attention to its utilization both for research and training. The paper also demonstrates the ease of usage by simulating common transient in a pressurized water reactor. Finally, observations and possible improvements to the code to increase its usability in research, education and training are discussed. This work aims to evaluate the robustness of the simulator towards better utilization for research and training, especially in nuclear newcomer countries.

A Methodology to Deduce the Microstructural Spatial Deformation of Polycrystalline Structures: Application to the Alloy 600

2009 ◽  
Vol 289-292 ◽  
pp. 137-144 ◽  
Author(s):  
Aurélie Clair ◽  
Marc Foucault ◽  
J. Marcos Salazar ◽  
Vincent Vignal ◽  
Eric Finot ◽  
...  
Keyword(s):  
Water Pressure ◽  
Surface Damage ◽  
Local Deformation ◽  
Alloy 600 ◽  
Pressurized Water ◽  
Spatial Deformation ◽  
Plastic Domain ◽  
Water Reactors ◽  
Polycrystalline Structures ◽  
The Impact

Studying the corrosion of the alloy 600, under water pressure, is of high importance to understand the ageing process of pressurized water reactors. Today, the impact of the oxide growth on the mechanical properties of nickel alloys is a challenge. The surface analysis and the quantification of the local deformation are key factors to deduce the surface damage of the substrate produced by corrosion. Here, we introduce a new methodology to determine the deformation distribution of the alloy 600 by using polycrystalline samples. The method is based on nanopads disposed on the surface samples, which allow a mapping, at the microscopic scale, of the spatial deformation. We applied to the samples a tensile loading in the plastic domain from 0 to 4%. The obtained asymmetric deformations distribution reflects the polycrystallinity of the system.

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