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.

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.

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 Dust Deposition on the Gulf of California Caused by Santa Ana Winds

Atmosphere ◽  
2020 ◽  
Vol 11 (3) ◽  
pp. 275 ◽  
Author(s):  
Christian A. Álvarez ◽  
José N. Carbajal ◽  
Luis F. Pineda-Martínez ◽  
José Tuxpan ◽  
David E. Flores
Keyword(s):  
Gulf Of California ◽  
Dust Emission ◽  
Positive Influence ◽  
Dust Particles ◽  
Weather Research And Forecasting ◽  
Western Coast ◽  
Emission Rates ◽  
Dust Deposition ◽  
Santa Ana ◽  
Santa Ana Winds

Numerical simulations revealed a profound interaction between the severe dust storm of 2007 caused by Santa Ana winds and the Gulf of California. The weather research and forecasting model coupled with a chemistry module (WRF-CHEM) and the hybrid single-particle Lagrangian integrated trajectory model (HYSPLIT) allowed for the estimation of the meteorological and dynamic aspects of the event and the dust deposition on the surface waters of the Gulf of California caused by the erosion and entrainment of dust particles from the surrounding desert regions. The dust emission rates from three chosen areas (Altar desert, Sonora coast, and a region between these two zones) and their contribution to dust deposition over the Gulf of California were analyzed. The Altar Desert had the highest dust emission rates and the highest contribution to dust deposition over the Gulf of California, i.e., it has the most critical influence with 96,879 tons of emission and 43,539 tons of dust deposition in the gulf. An increase of chlorophyll-a concentrations is observed coinciding with areas of high dust deposition in the northern and western coast of the gulf. This kind of event could have a significant positive influence over the mineralization and productivity processes in the Gulf of California, despite the soil loss in the eroded regions.

scholarly journals Dust emission size distribution impact on aerosol budget and radiative forcing over the Mediterranean region: a regional climate model approach

2012 ◽  
Vol 12 (21) ◽  
pp. 10545-10567 ◽  
Author(s):  
P. Nabat ◽  
F. Solmon ◽  
M. Mallet ◽  
J. F. Kok ◽  
S. Somot
Keyword(s):  
Size Distribution ◽  
Radiative Forcing ◽  
Mediterranean Basin ◽  
Climate Model ◽  
Regional Climate ◽  
Dust Emission ◽  
Dust Deposition ◽  
The Mediterranean ◽  
Dust Distribution ◽  
The Mediterranean Basin

Abstract. The present study investigates the dust emission and load over the Mediterranean basin using the coupled chemistry–aerosol–regional climate model RegCM-4. The first step of this work focuses on dust particle emission size distribution modeling. We compare a parameterization in which the emission is based on the individual kinetic energy of the aggregates striking the surface to a recent parameterization based on an analogy with the fragmentation of brittle materials. The main difference between the two dust schemes concerns the mass proportion of fine aerosol that is reduced in the case of the new dust parameterization, with consequences for optical properties. At the episodic scale, comparisons between RegCM-4 simulations, satellite and ground-based data show a clear improvement using the new dust distribution in terms of aerosol optical depth (AOD) values and geographic gradients. These results are confirmed at the seasonal scale for the investigated year 2008. This change of dust distribution has sensitive impacts on the simulated regional dust budget, notably dry dust deposition and the regional direct aerosol radiative forcing over the Mediterranean basin. In particular, we find that the new size distribution produces a higher dust deposition flux, and smaller top of atmosphere (TOA) dust radiative cooling. A multi-annual simulation is finally carried out using the new dust distribution over the period 2000–2009. The average SW radiative forcing over the Mediterranean Sea reaches −13.6 W m−2 at the surface, and −5.5 W m−2 at TOA. The LW radiative forcing is positive over the basin: 1.7 W m−2 on average over the Mediterranean Sea at the surface, and 0.6 W m−2 at TOA.

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.

Component Design Considerations for Gas Turbine HTGR Power Plant

1975 ◽  
Author(s):  
C. F. McDonald ◽  
R. G. Adams ◽  
F. R. Bell ◽  
P. Fortescue
Keyword(s):  
Power Plant ◽  
Gas Turbine ◽  
Heat Exchangers ◽  
Gas Turbines ◽  
Power Conversion ◽  
Working Fluid ◽  
Primary Circuit ◽  
Primary System ◽  
Design Considerations ◽  
Conversion System

The gas turbine high-temperature gas-cooled reactor (HTGR) power plant combines the existing design HTGR core with a closed-cycle helium gas turbine power conversion system directly in the reactor primary circuit. The high density helium working fluid results in a very compact power conversion system. While the geometries of the helium turbomachinery, heat exchangers, and internal gas flow paths differ from air breathing gas turbines because of the nature of the working fluid and the high degree of pressurization, many of the aerodynamic, heat transfer and dynamic analytical procedures used in the design are identical to conventional open-cycle industrial gas turbine practice. This paper outlines some of the preliminary design considerations for the rotating machinery, heat exchangers, and other major primary system components for an integrated type of plant embodying multiple gas turbine loops. The high potential for further improvement in plant efficiency and capacity, for both advanced dry-cooled and waste heat power cycle versions of the direct-cycle nuclear gas turbine, is also discussed.

Approach to Derivation of Waste Acceptance Criteria for Mochovce Disposal Facility

2003 ◽  
Author(s):  
V. Hanusˇi´k ◽  
Z. Kusovska´ ◽  
J. Bala´zˇ ◽  
A. Mrsˇkova´
Keyword(s):  
Fission Products ◽  
Total Activity ◽  
Safety Analysis ◽  
Radioactive Wastes ◽  
Primary Circuit ◽  
Radiological Protection ◽  
Waste Streams ◽  
Acceptance Criteria ◽  
Near Surface ◽  
Dose Limits

In Slovakia, low and intermediate level radioactive wastes are disposed in a near-surface repository at Mochovce site. The repository, which was commissioned in September 2001, has a disposal capacity 22,320 m3. It is a vault-type concrete structure repository with reinforced concrete containers as the final waste packages. The Mochovce repository is designed to receive acceptable radioactive wastes from decommissioned A-1 power plant at Jaslovske´ Bohunice, operational waste from NPPs V-1 and V-2 at Jaslovske´ Bohunice site and NPP Mochovce, as well as institutional radioactive wastes. Generally, calculation endpoint of disposal facilities performance assessment is radiological impact on humans and environment. In that case, starting points of assessment are the waste activity concentrations and inventory activity. The acceptance of radioactive waste in Mochovce repository is one of the many elements that directly contribute to the safety of the disposal system. In Mochovce repository safety analysis, end points are both the concentration per package and total activity values. On the other hand, radiological protection criteria are the starting points of the calculation. This approach was developed and applied because the actual inventory that will be disposed of is highly uncertain. As a result of the accidents, the primary circuit was contaminated by fission products. Some auxiliary circuits and facilities were also contaminated. The complicated problem is the relatively high content of long-lived radionuclides (inclusive transuranic elements) in some waste streams. After two technological incidents at NPP A-1 uncertainties in waste inventory are large because of variability in the types of waste streams and variability in the quality and completeness of the waste characterization data. This paper presents the philosophy of safety analysis, development of scenarios, their modelling and approach that have been used to derive waste acceptance criteria, specifically limits of activity. The approach consists of the determination of radionuclides important for safety, the use of relevant safety scenarios, the setting of dose limits associated with scenarios, the calculation of activity limits and application of the simple summation rule. Finally, information is provided about short operation of the repository.

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