Development of Modeling Transport of Graphite Dust in HTGR During Normal Operation Condition

2012 ◽  
Author(s):  
Zhipeng Chen ◽  
Fei Xie ◽  
Yanhua Zheng ◽  
Lei Shi ◽  
Fu Li
Keyword(s):  
Fission Products ◽  
Life Time ◽  
Normal Operation ◽  
Primary Circuit ◽  
Potential Hazard ◽  
Operation Condition ◽  
Modeling Transport ◽  
Full Power ◽  
Accident Conditions ◽  
Helium Coolant

High temperature gas-cooled reactor (HTGR), especially the pebble-bed core type reactor, will inevitably cause the wear the graphite components and generate graphite dust in the core. The graphite dust is taken away by helium coolant and deposited on the surface of the primary circuit, and the fission products may be absorbed on the dust. Since it is possible that the fission products are released with dust under the accident conditions such as depressurization events, they have a potential hazard of radiation exposure to the environment. The objective of this paper is to develop a code for calculating the behaviour of graphite dust in the primary circuit of HTGR. The paper is focused on development of models for predicting the deposition rates of the dust. The purpose of the work is to estimate the amount and distribution of deposited dust during plant life time, which was assumed to be 40 full-power years. The result will lay the foundation for further studies of fission products releasing and interaction with dust under accident conditions.

Design and Analysis of Testing Platform for Study of Graphite Dust Emission Behavior During Accident Conditions

2010 ◽  
Author(s):  
Zhipeng Chen ◽  
Suyuan Yu
Keyword(s):  
Dust Emission ◽  
Fission Products ◽  
Primary Circuit ◽  
Primary System ◽  
Potential Hazard ◽  
Dust Deposition ◽  
Pebble Bed ◽  
Testing Platform ◽  
Emission Behavior ◽  
Accident Conditions

Very High temperature gas-cooled reactor (VHTR), especially the pebble-bed core type reactor, is inevitable to take place the wear of graphite components and generate the graphite dust in the core. The graphite dust was taken away by helium coolant and deposited on the surface of the primary circuit, and the fission products may be absorbed on the dust. Since it is possible that the fission products are released with dust under the accident conditions such as depressurization events, they have a potential hazard of radiation exposure to the environment. In this paper, VHTR as the research object, a testing platform is to be built with the purpose of investigating the behavior of graphite dust emission during the accident conditions. Circuit loop design is used to simulate the primary system and nitrogen is used as the working substance. Experiments of graphite dust deposition and resuspension study, as well as the study of graphite dust emission behavior during the accident conditions in pebble-bed type design VHTR will be conducted on the testing platform. The experimental data will be used for the development of VHTR source term analysis modeling.

scholarly journals A Safety Re-Evaluation of the AVR Pebble Bed Reactor Operation and Its Consequences for Future HTR Concepts

2008 ◽  
Author(s):  
Rainer Moormann
Keyword(s):  
Fission Product ◽  
End Of Life ◽  
Fission Products ◽  
Normal Operation ◽  
Primary Circuit ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Hot Gas ◽  
Design Basis ◽  
Gas Tight

The AVR pebble bed reactor (46 MWth) was operated 1967–1988 at coolant outlet temperatures up to 990°C. Also because of a lack of other experience the AVR operation is a basis for future HTRs. This paper deals with insufficiently published unresolved safety problems of AVR and of pebble bed HTRs. The AVR primary circuit is heavily contaminated with dust bound and mobile metallic fission products (Sr-90, Cs-137) which create problems in current dismantling. The evaluation of fission product deposition experiments indicates that the end of life contamination reached several percent of a single core inventory. A re-evaluation of the AVR contamination is performed in order to quantify consequences for future HTRs: The AVR contamination was mainly caused by inadmissible high core temperatures, and not — as presumed in the past — by inadequate fuel quality only. The high AVR core temperatures were detected not earlier than one year before final AVR shut-down, because a pebble bed core cannot be equipped with instruments. The maximum core temperatures were more than 200 K higher than precalculated. Further, azimuthal temperature differences at the active core margin were observed, as unpredictable hot gas currents with temperatures > 1100°C. Despite of remarkable effort these problems are not yet understood. Having the black box character of the AVR core in mind it remains uncertain whether convincing explanations can be found without major experimental R&D. After detection of the inadmissible core temperatures, the AVR hot gas temperatures were strongly reduced for safety reasons. Metallic fission products diffuse in fuel kernel, coatings and graphite and their break through takes place in long term normal operation, if fission product specific temperature limits are exceeded. This is an unresolved weak point of HTRs in contrast to other reactors and is particularly problematic in pebble bed systems with their large dust content. Another disadvantage, responsible for the pronounced AVR contamination, lies in the fact that activity released from fuel elements is distributed in HTRs all over the coolant circuit surfaces and on graphitic dust and accumulates there. Consequences of AVR experience on future reactors are discussed. As long as pebble bed intrinsic reasons for the high AVR temperatures cannot be excluded they have to be conservatively considered in operation and design basis accidents. For an HTR of 400 MWth, 900°C hot gas temperature, modern fuel and 32 fpy the contaminations are expected to approach at least the same order as in AVR end of life. This creates major problems in design basis accidents, for maintenance and dismantling. Application of German dose criteria on advanced pebble bed reactors leads to the conclusion that a pebble bed HTR needs a gas tight containment even if inadmissible high temperatures as observed in AVR are not considered. However, a gas tight containment does not diminish the consequences of the primary circuit contamination on maintenance and dismantling. Thus complementary measures are discussed. A reduction of demands on future reactors (hot gas temperatures, fuel burn-up) is one option; another one is an elaborate R&D program for solution of unresolved problems related to operation and design basis accidents. These problems are listed in the paper.

scholarly journals Assessment of occupational internal exposure to beta emitters from the nuclear reactor primary coolant circuit

2012 ◽  
Vol 18 (2) ◽  
pp. 41-47 ◽  
Author(s):  
Tomasz Pliszczyński ◽  
Katarzyna Ciszewska ◽  
Małgorzata Dymecka ◽  
Jakub Ośko ◽  
Zbigniew Haratym
Keyword(s):  
Nuclear Reactor ◽  
Cooling Water ◽  
Fission Products ◽  
Nuclear Research ◽  
Radioactive Isotopes ◽  
Internal Exposure ◽  
Normal Operation ◽  
Potential Hazard ◽  
Primary Coolant ◽  
Beta Emitters

Fission products of 235U or isotopes from activation may appear in technological waters at normal operation of a research reactor. Therefore, reactor cooling water may contain a number of beta radioactive isotopes including yttrium and strontium isotopes, which can pose potential hazard to reactor personnel. In order to asses internal exposure urinalysis is carried out. This work presents the method of sample preparation and measurement used by Radiation Protection Measurements Laboratory (RPLM) of the National Centre for Nuclear Research (NCNR). Method of various isotopes of yttrium and Sr-90 activity calculation is also shown. Determination of these isotopes activities in urine sample allows estimating the effective doses

scholarly journals Chemical Forms of Important Fission Products in Primary Circuit of HTR-PM under Conditions of Normal Operation and Overpressure and Water Ingress Accidents: A Study with a Chemical Thermodynamics Approach

2019 ◽  
Vol 2019 ◽  
pp. 1-12
Author(s):  
Chuan Li ◽  
Wenqian Li ◽  
Lifeng Sun ◽  
Haoyu Xing ◽  
Chao Fang
Keyword(s):  
High Temperature ◽  
Fission Products ◽  
Point Of View ◽  
Normal Operation ◽  
Primary Circuit ◽  
Chemical Thermodynamics ◽  
Chemical Forms ◽  
Operational Conditions ◽  
Water Ingress ◽  
High Temperature Gas

The chemical forms of important fission products (FPs) in the primary circuit are essential to the source term analysis of high-temperature gas-cooled reactors because the volatility, transfer, and diffusion of these radionuclides are significantly influenced by their chemical forms. Through chemical reactions with gaseous impurities in the primary circuit, these FPs exist in diverse chemical forms, which vary under different operational conditions. In this paper, the chemical forms of cesium (Cs), strontium (Sr), silver (Ag), iodine (I), and tritium in the primary circuit of the Chinese pebble-bed modular high-temperature gas-cooled reactor (HTR-PM) under normal conditions and accident conditions (overpressure and water ingress accident) are studied with chemical thermodynamics. The results under normal conditions show that Cs exists mainly in the form of Cs2CO3 at 250°C and gaseous form at 750°C, and for I and Ag, Ag3I3 and Ag convert to gaseous CsI and AgO, respectively, with increasing temperature, while SrCO3 is the only main kind of compound for Sr. It is also observed that new compounds are generated under accidents: I exists in HI form when a water ingress accident occurs. Regarding tritium, the chemical forms of FPs change little, but compounds need higher temperature to convert. Furthermore, hazard of some FPs in different chemical forms is also discussed comprehensively in this paper. This study is significant for understanding the chemical reaction mechanisms of FPs in an HTR-PM, and furthermore it may provide a new point of view to analyze the interaction between FPs and structural materials in reactor as well as their hazards.

Analysis of Radioactive Source Term for Modular HTGR During Normal Operation

2010 ◽  
Author(s):  
Jianzhu Cao ◽  
Tao Liu ◽  
Yuanyu Wu ◽  
Hong Li ◽  
Yuanzhong Liu
Keyword(s):  
Source Term ◽  
Noble Gas ◽  
Fission Products ◽  
Radioactive Source ◽  
Normal Operation ◽  
Design Parameters ◽  
Primary Circuit ◽  
Secondary Circuit ◽  
Activation Products ◽  
Basic Prediction

The methods of radioactive source term analysis are introduced in detail for the modular high temperature gas cooled reactor in China. Radioactive fission products and activation products produced in the reactor are described. For fission products, the emphasis is on the process from production through release to the environment for noble gas, iodine and long-lived metallic nuclides. For activation products, it mainly introduces the behaviors of H-3 and C-14. Especially the permeation process from primary circuit to secondary circuit is described for H-3. Using the preliminary design parameters of demonstration HTGR in China, basic prediction of radioactive source term is done and the results are given.

Characterization of Advanced Steels as Accident Tolerant Cladding for Light Water Reactor Nuclear Fuel

2015 ◽  
Author(s):  
Raul B. Rebak
Keyword(s):  
Zirconium Alloy ◽  
Fission Products ◽  
Alloy System ◽  
Department Of Energy ◽  
Normal Operation ◽  
Operation Conditions ◽  
Time Period ◽  
Environmental Test ◽  
Near Term ◽  
Accident Conditions

After the tsunami incident in Fukushima in March 2011, the international community is set to identify appropriate nuclear materials with increased accident tolerance with respect to the traditional UO2-zirconium alloy fuel system permitting loss of active cooling for a considerably longer time period, while maintaining or improving the fuel performance during normal operations. The researched safety characteristics of these advanced fuels are mainly: (a) Improved reaction kinetics with steam; (b) Slower hydrogen production rate; and (c) Enhanced retention of fission products. In the US the Department of Energy is supporting the development of an improved cladding using advanced steels such as the iron-chromium-aluminum (FeCrAl) alloy system. Environmental test results show that FeCrAl alloys are highly resistant to corrosion and environmental cracking under normal operation conditions and extremely resistant to attack by steam under accident conditions. That is, the replacement of a zirconium alloy using a ferritic material containing chromium and aluminum appears to be the most near term implementation for accident tolerant fuels.

scholarly journals Recent Advances in Protective Coatings for Accident Tolerant Zr-Based Fuel Claddings

Coatings ◽  
2021 ◽  
Vol 11 (5) ◽  
pp. 557
Author(s):  
Egor Kashkarov ◽  
Bright Afornu ◽  
Dmitrii Sidelev ◽  
Maksim Krinitcyn ◽  
Veronica Gouws ◽  
...  
Keyword(s):  
Protective Coatings ◽  
Zirconium Alloys ◽  
Nuclear Industry ◽  
Normal Operation ◽  
Severe Accidents ◽  
Recent Developments ◽  
Water Reactors ◽  
Accident Scenarios ◽  
Accident Conditions ◽  
Directed Research

Zirconium-based alloys have served the nuclear industry for several decades due to their acceptable properties for nuclear cores of light water reactors (LWRs). However, severe accidents in LWRs have directed research and development of accident tolerant fuel (ATF) concepts that aim to improve nuclear fuel safety during normal operation, operational transients and possible accident scenarios. This review introduces the latest results in the development of protective coatings for ATF claddings based on Zr alloys, involving their behavior under normal and accident conditions in LWRs. Great attention has been paid to the protection and oxidation mechanisms of coated claddings, as well as to the mutual interdiffusion between coatings and zirconium alloys. An overview of recent developments in barrier coatings is introduced, and possible barrier layers and structure designs for suppressing mutual diffusion are proposed.

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.

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