Research Works on Iodine and Ruthenium Behavior in Severe Accident Conditions

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Laurent Cantrel ◽  
Thierry Albiol ◽  
Loïc Bosland ◽  
Juliette Colombani ◽  
Frédéric Cousin ◽  
...  
Keyword(s):  
Porous Materials ◽  
Nuclear Power ◽  
Cooling System ◽  
Fission Products ◽  
Simulation Software ◽  
Severe Accident ◽  
Counter Ions ◽  
Metal Organic ◽  
Nature Gas ◽  
Accident Conditions

This paper deals with near past, ongoing, and planned R&D works on fission products (FPs) behavior in reactor cooling system (RCS), containment building and in filtered containment venting systems (FCVS) for severe accident (SA) conditions. All the researches are collaborative works; the overall objective is to develop confident models to be implemented in simulation software. After being initiated in 2004, researches on iodine transport through the RCS are still ongoing and for containment, the last advances are linked to the source term (ST) evaluation and mitigation (STEM) OECD/NEA project. The objective is to improve the evaluation of ST for a SA on a nuclear power plant and to reduce uncertainties on specific phenomena dealing with the chemistry of two major FPs: iodine and ruthenium. For ruthenium attention has been paid to study the amount and nature (gas/aerosol partition) of ruthenium species along the RCS. A follow-up, called STEM2, has started to reduce some remaining issues and be closer to reactor conditions. For FCVS works, the efficiencies for trapping iodine covering scrubbers and dry filters are examined to get a clear view of their abilities in SA conditions. Another part is focused on specific porous materials able to trap volatile iodine. Influence of zeolite materials parameters (nature of the counter-ions, structure, Si/Al ratio…) are tested as well as new kind of porous materials constituted by Metal organic Frameworks will also be looked at.

IRSN R&D Actions on FP Behaviour for RCS, Containment and FCVS in Severe Accident Conditions

2016 ◽  
Author(s):  
Laurent Cantrel ◽  
Thierry Albiol ◽  
Loïc Bosland ◽  
Juliette Colombani ◽  
Frédéric Cousin ◽  
...  
Keyword(s):  
Porous Materials ◽  
Nuclear Power ◽  
Source Term ◽  
Cooling System ◽  
Fission Products ◽  
Simulation Software ◽  
Severe Accident ◽  
Status Report ◽  
Diagnosis And Prognosis ◽  
Accident Conditions

This paper deals with near past, ongoing and planned R&D works on fission products (FPs) behaviour in Reactor Cooling System (RCS), containment building and in Filtered Containment Venting Systems (FCVS) for severe accident (SA) conditions. For the last topic, in link with the Fukushima post-accident management and possible improvement of mitigation actions for such SA, the FCVS topic is again on the agenda (see Status Report on Filtered Containment Venting, OECD/NEA/CSNI, Report NEA/CSNI/R(2014)7, 2014.) with a large interest at the international scale. All the researches are collaborative works; the overall objective is to develop confident models to be implemented in ASTEC SA simulation software. After being initiated in the International Source Term Program (ISTP), researches devoted to the understanding of iodine transport through the RCS are still ongoing in the frame of a bilateral agreement between IRSN and EDF with promising results. In 2017, a synthesis report of the last 10 years of researches, which have combined experimental and fundamental works based on the use of theoretical chemistry tools, will be issued. For containment, the last advances are linked to the Source Term Evaluation and Mitigation (STEM) OECD/NEA project operated by IRSN. The objective of the STEM project was to improve the evaluation of Source Term (ST) for a SA on a nuclear power plant and to reduce uncertainties on specific phenomena dealing with the chemistry of two major fission products: iodine and ruthenium. More precisely, the STEM project provided additional knowledge and improvements for calculation tools in order to allow a more robust diagnosis and prognosis of radioactive releases in a SA. STEM data will be completed by a follow-up, called STEM2, to further the knowledge concerning some remaining issues and be closer to reactor conditions. Two additional programmes deal with FCVS issues: the MItigation of outside Releases in the Environment (MIRE) (2013–2019) French National Research Agency (NRA) programme and the Passive and Active Systems on Severe Accident source term Mitigation (PASSAM) (2013–2016) European project. For FCVS works, the efficiencies for trapping iodine with various FCVS, covering scrubbers and dry filters, are examined to get a clear view of their abilities in SA conditions. Another part, performed in collaboration with French universities (Lorraine and Lille 1), is focused on the enhancement of the performance of these filters with specific porous materials able to trap volatile iodine. For that, influence of zeolites materials parameters (nature of the counter-ions, structure, Si/Al ratio …) will be tested. New kind of porous materials constituted by Metal organic Frameworks (MOF) will also be looked at because they can constitute a promising way of trapping.

scholarly journals A Conceptual Approach to Eliminate Bypass Release of Fission Products by In-Containment Relief Valve under SGTR Accident

2018 ◽  
Vol 2018 ◽  
pp. 1-12
Author(s):  
Taeseok Kim ◽  
Wonjun Choi ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Hoichul Jung ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Base Case ◽  
Conceptual Approach ◽  
Relief Valve ◽  
Radioactive Nuclides

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise a conceptual approach of installing In-Containment Relief Valve (ICRV) from steam generator (SG) to the free space in the containment building and it was simulated by MELCOR code for numerical analysis. Simulation results show that the radioactive nuclides were not released to the environment in the ICRV case. However, the containment pressure increased more than the base case, which is a disadvantage of the ICRV. To minimize the negative effects of the ICRV, the ICRV linked to Reactor Drain Tank (RDT) and cavity flooding was performed. Because the overpressurization of containment is due to heat of ex-vessel corium, only cavity flooding was effective for depressurization. The conceptual design of the ICRV is effective in mitigating the SGTR accident.

scholarly journals Radiation-Tolerance Assessment of a Redundant Wireless Device

2018 ◽  
Vol 170 ◽  
pp. 03004
Author(s):  
Q. Huang ◽  
J. Jiang
Keyword(s):  
Nuclear Power ◽  
Evaluation Method ◽  
Radiation Condition ◽  
Severe Accident ◽  
Radiation Environment ◽  
Radiation Tolerance ◽  
Monitoring Device ◽  
Wireless Monitoring ◽  
Study Results ◽  
Accident Conditions

This paper presents a method to evaluate radiation-tolerance without physical tests for a commercial off-the-shelf (COTS)-based monitoring device for high level radiation fields, such as those found in post-accident conditions in a nuclear power plant (NPP). This paper specifically describes the analysis of radiation environment in a severe accident, radiation damages in electronics, and the redundant solution used to prolong the life of the system, as well as the evaluation method for radiation protection and the analysis method of system reliability. As a case study, a wireless monitoring device with redundant and diversified channels is evaluated by using the developed method. The study results and system assessment data show that, under the given radiation condition, performance of the redundant device is more reliable and more robust than those non-redundant devices. The developed redundant wireless monitoring device is therefore able to apply in those conditions (up to 10 M Rad (Si)) during a severe accident in a NPP.

scholarly journals Contribution to the study of fission products release from nuclear fuels in severe accident conditions: effect of the pO2 on Cs, Mo and Ba speciation

2020 ◽  
Vol 6 ◽  
pp. 2 ◽  
Author(s):  
Claire Le Gall ◽  
Fabienne Audubert ◽  
Jacques Léchelle ◽  
Yves Pontillon ◽  
Jean-Louis Hazemann
Keyword(s):  
Oxygen Potential ◽  
Fission Products ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Mox Fuel ◽  
Chemical Behaviour ◽  
Accident Conditions ◽  
The Impact

The objective of this work is to experimentally investigate the effect of the oxygen potential on the fuel and FP chemical behaviour in conditions representative of a severe accident. More specifically, the speciation of Cs, Mo and Ba is investigated. These three highly reactive FP are among the most abundant elements produced through 235U and 239Pu thermal fission and may have a significant impact on human health and environmental contamination in case of a light water reactor severe accident. This work has set out to contribute to the following three fields: providing experimental data on Pressurized Water Reactor (PWR) MOX fuel behaviour submitted to severe accident conditions and related FP speciation; going further in the understanding of FP speciation mechanisms at different stages of a severe accident; developing a method to study volatile FP behaviour, involving the investigation of SIMFuel samples manufactured at low temperature through SPS. In this paper, a focus is made on the impact of the oxygen potential towards the interaction between irradiated MOX fuels and the cladding, the interaction between Mo and Ba under oxidizing conditions and the assessment of the oxygen potential during sintering.

Modeling of Thermal Hydraulics Features of Top Water Reflood Experiment PARAMETER-SF3 Using SOCRAT/V2 Code

2009 ◽  
Author(s):  
Arcadii E. Kisselev ◽  
Valerii F. Strizhov ◽  
Alexander D. Vasiliev ◽  
Vladimir I. Nalivayev ◽  
Nikolay Ya. Parshin
Keyword(s):  
Experimental Data ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Numerical Code ◽  
Thermal Hydraulics ◽  
Vver Fuel ◽  
Design Basis ◽  
Best Estimate ◽  
Core Cooling

The PARAMETER-SF3 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700÷2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in October 31, 2008, and was the third of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. The test bundle was made up of 19 fuel rod simulators with a length of approximately 3.12 m (heated rod simulators) and 2.92 m (unheated rod simulator). Heating was carried out electrically using 4-mm-diameter tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVERs: Zr1%Nb, 9.13 mm outside diameter, 0.7 mm wall thickness. After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF3 test, the top flooding was initiated. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT/V2 was used for the calculation of PARAMETER-SF3 experiment. The counter-current flow limitation (CCFL) model was implemented to best estimate numerical code SOCRAT/V2 developed for modeling thermal hydraulics and severe accident phenomena in a reactor. Thermal hydraulics in PARAMETER-SF3 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V2 were compared with experimental data concerning different aspects of thermal hydraulics behavior including the CCFL phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF3 test.

Accident Analysis of Fukushima Daiichi Nuclear Power Plant Unit 2 by the SAMPSON Severe Accident Code

2014 ◽  
Author(s):  
Atsuo Takahashi ◽  
Marco Pellegrini ◽  
Hideo Mizouchi ◽  
Hiroaki Suzuki ◽  
Masanori Naitoh
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Accident Analysis ◽  
Unknown Boundary ◽  
Long Time ◽  
Fukushima Daiichi ◽  
Nuclear Power Plant Unit

The transient process of the accident at the Fukushima Daiichi Nuclear Power Plant Unit 2 was analyzed by the severe accident analysis code, SAMPSON. One of the characteristic phenomena in Unit 2 is that the reactor core isolation cooling system (RCIC) worked for an unexpectedly long time (about 70 h) without batteries and consequently core damage was delayed when compared to Units 1 and 3. The mechanism of how the RCIC worked such a long time is thought to be due to balance between injected water from the RCIC pump and the supplied mixture of steam and water sent to the RCIC turbine. To confirm the RCIC working conditions and reproduce the measured plant properties, such as pressure and water level in the pressure vessel, we introduced a two-phase turbine driven pump model into SAMPSON. In the model, mass flow rate of water injected by the RCIC was calculated through turbine efficiency degradation the originated from the mixture of steam and water flowing to the RCIC turbine. To reproduce the drywell pressure, we assumed that the torus room was flooded by the tsunami and heat was removed from the suppression chamber to the sea water. Although uncertainties, mainly regarding behavior of debris, still remain because of unknown boundary conditions, such as alternative water injection by fire trucks, simulation results by SAMPSON agreed well with the measured values for several days after the scram.

scholarly journals Analysis of Release Model Effect in the Transport of Fission Products Simulating the FPT3 Test Using MELCOR 2.1 and MELCOR 2.2

2021 ◽  
Vol 13 (14) ◽  
pp. 7964
Author(s):  
Alain Flores y Flores ◽  
Danilo Ferretto ◽  
Tereza Marková ◽  
Guido Mazzini
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Nuclear Industry ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Accident Simulation ◽  
Experimental Facilities ◽  
The Difference

The severe accident integral codes such as Methods for Estimation of Leakages and Consequences of Releases (MELCOR) are complex tools used to simulate and analyse the progression of a severe accident from the onset of the accident up to the release from the containment. For this reason, these tools are developed in order to simulate different phenomena coupling models which can simulate simultaneously the ThermoHydraulic (TH), the physics and the chemistry. In order to evaluate the performance in the prediction of those complicated phenomena, several experimental facilities were built in Europe and all around the world. One of these facilities is the PHEBUS built by Institut de Radioprotection et de Sûrete Nucléaire (IRSN) in Cadarache. The facility reproduces the severe accident phenomena for a pressurized water reactor (PWR) on a volumetric scale of 1:5000. This paper aims to continue the assessment of the MELCOR code from version 2.1 up to version 2.2 underlying the difference in the fission product transport. The assessment of severe accident is an important step to the sustainability of the nuclear energy production in this period where the old nuclear power plants are more than the new reactors. The analyses presented in this paper focuses on models assessment with attention on the influence of B4C oxidation on the release and transport of fission products. Such phenomenon is a concern point in the nuclear industry, as was highlighted during the Fukushima Daiichi accident. Simulation of the source term is a key point to evaluate the severe accident hazard along with other safety aspects.

scholarly journals Steam Oxidation of Silicon Carbide at High Temperatures for the Application as Accident Tolerant Fuel Cladding, an Overview

Thermo ◽  
2021 ◽  
Vol 1 (2) ◽  
pp. 151-167
Author(s):  
Hai V. Pham ◽  
Masaki Kurata ◽  
Martin Steinbrueck
Keyword(s):  
Silicon Carbide ◽  
High Temperatures ◽  
Nuclear Power ◽  
Steam Oxidation ◽  
Severe Accident ◽  
Light Water Reactors ◽  
Severe Accidents ◽  
Potential Benefits ◽  
Water Reactors ◽  
Accident Conditions

Since the nuclear accident at Fukushima Daiichi Nuclear Power Station in 2011, a considerable number of studies have been conducted to develop accident tolerant fuel (ATF) claddings for safety enhancement of light water reactors. Among many potential ATF claddings, silicon carbide is one of the most promising candidates with many superior features suitable for nuclear applications. In spite of many potential benefits of SiC cladding, there are some concerns over the oxidation/corrosion resistance of the cladding, especially at extreme temperatures (up to 2000 °C) in severe accidents. However, the study of SiC steam oxidation in conventional test facilities in water vapor atmospheres at temperatures above 1600 °C is very challenging. In recent years, several efforts have been made to modify existing or to develop new advanced test facilities to perform material oxidation tests in steam environments typical of severe accident conditions. In this article, the authors outline the features of SiC oxidation/corrosion at high temperatures, as well as the developments of advanced test facilities in their laboratories, and, finally, give some of the current advances in understanding based on recent data obtained from those advanced test facilities.

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