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The Latest Results from source rerm Research: Overv¡ew and Outlook
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The chemistry of radiotoxic ruthenium in a severe nuclear power accident has actively been investigated
especially during the last decades. The Ru studies have covered the release from a fuel, the transport in the primary circuit
and the behaviour in the containment building. The gathered experimental data have been utilized to understand the key
parameters governing the Ru chemistry in a severe accident (SA) and to check the ability of the existing models of SA
analysis codes to explain the experimental results. To further increase the knowledge on Ru behaviour, the collaboration on
international level has been intensive. Lately, the widest and most active networks have been EU SARNET and EU SARNET2.
The valuable effort of these networks on sharing information of e.g. national programs and on interpreting the experimental
results is continued in EU NUGENIA program. More detailed studies on separate phenomena have been conducted e.g. as
part of OECD/NEA STEM/START and ISTP/VERDON programs. Furthermore, Phébus FP tests have produced valuable data
on integral phenomena.
The large-scale integral and semi-integral experiments have confirmed that Ru release depends strongly on carrier gas. Ru is
significantly released from an irradiated fuel sample under oxidizing conditions, in particular when air is involved. In
addition, the oxidation of UO2 fuel seems to lead to a higher Ru release than in case of MOX fuel. Ruthenium can be
transported to the containment atmosphere both in gaseous and particulate forms. The small-scale separate-effect
experiments gave a detailed view on Ru transport. A high fraction of ruthenium was detected as particles at the outlet of the
model primary circuit in an air atmosphere. However, the observed gaseous Ru fraction is higher than what could be
expected based on thermodynamic equilibrium calculations. Further studies on the effect of flow residence time in a
temperature gradient for the equilibrium of Ru oxides have been conducted.
The effect of other fission products in the gas phase, as well as FP deposits on the surface of primary circuit, on the Ru
transport has been investigated. For example, caesium containing deposits seemed to trap gaseous ruthenium effectively.
Similarly in case of control rod residues, silver particles in the gas phase of the circuit acted as a sink for gaseous Ru. In an
air ingress accident, the effect of air radiolysis products on the Ru chemistry becomes important. As the main air radiolysis
products can be considered as oxidizing agents, their ability to oxidize the lower oxides of Ru to higher oxidation state has
been examined.
Most of Ru in the containment building ends up as deposits on the containment surfaces and in the sump. Experiments on the
radiolytical revaporisation of ruthenium deposits on the epoxy paint surface indicated the release of gaseous ruthenium and
it was enhanced under humid atmosphere and elevated temperature. It appeared that the products of air radiolysis caused by
γ-radiation promoted the formation of gaseous ruthenium from Ru oxide deposits on paint in a higher amount than could be
expected by pure ozone action. Concerning the irradiation tests of perruthenate aqueous solutions, they indicated the
formation of gaseous Ru by γ-radiolysis products in solution.
A new major version of the European severe accident integral code ASTEC, developed by IRSN with some GRS support, was delivered in November 2015 to the ASTEC worldwide community.Main modelling features of this V2.1 version are summarised in this paper. In particular, the in-vessel coupling technique between the reactor coolant system thermal-hydraulics module and the core degradation module has been strongly re-engineered to remove some well-known weaknesses of the former V2.0 series. The V2.1 version also includes new core degradation models specifically addressing BWR and PHWR reactor types, as well as several other physical modelling improvements, notably on reflooding of severely damaged cores, Zircaloy oxidation under air atmosphere, corium coolability during corium concrete interaction and source term evaluation.Moreover, this V2.1 version constitutes the back-bone of the CESAM FP7 project, which final objective is to further improve ASTEC for use in Severe Accident Management analysis of the Gen.II-III nuclear power plants presently under operation or foreseen in near future in Europe. As part of this European project, IRSN efforts to continuously improve both code numerical robustness and computing performances at plant scale as well as users' tools are being intensified.Besides, ASTEC will continue capitalising the whole knowledge on severe accidents phenomenology by progressively keeping physical models at the state of the art through a regular feed-back from the interpretation of the current and future experimental programs performed in the international frame.
Over the last decades, several experimental programs devoted to the source term of fission products (FP) and actinides released from PWR fuel samples in severe accident (SA) conditions have been initiated throughout the world. In France, in this context, the Institute for Radiological Protection and Safety (IRSN) and Electricité de France (EDF) have supported the analytical VERCORS program which was performed by the "Commissariat à l'Energie Atomique" (CEA). The VERCORS facility at the LAMA-laboratory (CEA-Grenoble, France) was designed to heat up an irradiated fuel sample -taken from EDF's nuclear power reactors -to fuel relocation, and to capture the fission products released from the fuel and deposited downstream on a series of specific filters (impactors, bead-bed filter, …). On-line gamma detectors aimed at the fuel position, filters and gas capacity monitored the progress of FP release from the fuel, FP deposition on the filters and the fission gases emitted by the fuel (xenon and krypton). Before and after the test, a longitudinal gamma-scan of the fuel was conducted to measure the initial and final FP inventory in order to evaluate the quantitative fractions of FP emitted by the fuel during the test. All the components of the loop were then gamma-scanned to measure and locate the FPs released during the test and to draw up a mass balance of these FP.
Abstract Chemical revaporisation or physical resuspension of fission product deposits from the primary circuit is now recognised to be a major source term in the late phase of fuel degradation in a severe nuclear accident. These results come from tests carried out under different experimental projects in the European Commission (EC) Framework Programmes. These include the revaporisation tests carried out at the Transuranium Institute (ITU), Karlsruhe under the Fourth Framework Programme, the Phébus FP post-test analysis programme that examined FPT1, FPT3 and FPT4 deposits in separate-effect tests as well as EXSI-PC tests carried out at VTT, Espoo. The first tests at ITU and VTT concentrated on the behaviour of caesium as a very important fission product; this has helped detailed interpretation of the integral Phébus FP tests and has clarified some puzzling observations. Testing with Phébus FPT1 and FPT4 deposits at ITU demonstrated that revaporisation is a likely, rather than a possible, phenomenon with a severely degrading bundle. They have also shown that any changes in temperature (substrate or gas), flow rate or atmosphere composition or pressure can lead to the volatilisation or removal of the deposited caesium. Cs was particularly easy to follow given the high activity levels of Cs in the deposit. However further analysis of the deposits shows that other fission products are also subject to revaporisation. In the most recent FPT3 test, the chemical analysis of the filters has enabled examination of other fission products and demonstrated that these can be equally active in such conditions. Further separate effect tests in the EXSI-PC facility at VTT, Espoo have also given further insight as to the chemical reactions that major fission products (e.g. Cs, I) undergo under steam flows. One important result is the significant fraction of iodine that was released and transported in gaseous form at rather low circuit temperatures. In support of the experimental data, ‘ab initio’ theoretical approaches are being used at IRSN to demonstrate the interaction mechanisms of iodine and caesium vapours with typical primary circuit substrates under severe accident conditions. These approaches are expected to help interpret the Phébus FP experiments and VERCORS fission product tests as well as the CEA’s on-going ISTP-VERDON tests under mixed air and steam conditions. The combination of the three different research approaches will enable a much improved understanding of major chemical interactions in the primary circuit and so permit a more accurate simulation of a severe accident in primary circuits of water-cooled reactors with the ASTEC integral code, using improved thermodynamic data in the SOPHEAROS module. This, in turn will help to reduce the uncertainties in the anticipated source term to the environment.
Ruthenium is a semi-volatile element originating as a fission product in nuclear reactors that can be released in case of a severe nuclear accident. In this work, the impact of atmosphere composition on the transport of ruthenium through the primary circuit was examined. The effects of silver nanoparticles representing aerosols and NO2 gas as a product of air radiolysis were studied. Quantification of ruthenium transported both as gas and aerosol was performed. Chemical composition of ruthenium species was evaluated. The transport of gaseous ruthenium through the facility increased significantly when NO2 gas was fed into the atmosphere. When both silver aerosols and NO2 were fed into the atmosphere, the transport of ruthenium in gaseous and aerosol forms was promoted. It was concluded that the composition of atmosphere in the primary circuit will have a notable effect on the speciation of ruthenium transported into the containment building during a severe accident and thus on the potential radioactive release to the environment.
In this project the release, transport and speciation of ruthenium in conditions simulating an air ingress accident was studied. Ruthenium dioxide was exposed to oxidising environment at high temperature (1100-1700 K) in a tubular flow furnace. At these conditions volatile ruthenium species were formed. Majority of the released ruthenium was deposited in the tube as RuO2. Depending on the experimental conditions ∼1-26 wt-% of the released ruthenium was trapped in the outlet filter as RuO2 particles. In stainless steel tube -0-8.8 wt-% of the released ruthenium reached the trapping bottle as gaseous RuO4. A few experiments were carried out, in which revaporisation of ruthenium deposited on the tube walls was studied. In these experiments oxidation of RuO2 took place at a lower temperature. During revaporisation experiments 35-65% of ruthenium transported as gaseous RuO4. In order to close mass balance and achieve better time resolution four experiments were carried out using a radioactive tracer.
Iodine is one of the most radiotoxic fission product released from fuel during a severe nuclear power plant accident. Within the containment building, iodine compounds can react e.g. on the painted surfaces and form gaseous organic iodides.
In this study, it was found out that gaseous methyl iodide (CH 3 I) is oxidised when exposed to beta radiation in an oxygen containing atmosphere. As a result, nucleation of aerosol particles takes place and the formation of iodine oxide particles is suggested. These particles are highly hygroscopic. They take up water from the air humidity and iodine oxides dissolve within the droplets. In order to mitigate the possible source term, it is of interest to understand the effect of beta radiation on the speciation of iodine.
The SARNET network (Severe Accident Research NETwork of excellence), co-funded by the European Commission from 2004 to 2013, has allowed to significantly improve the knowledge on severe accidents and to disseminate it through courses and ERMSAR conferences. The major investigated topics, involving more than 250 researchers from 22 countries, were in- and ex-vessel corium/debris coolability, molten-core–concrete-interaction, steam explosion, hydrogen combustion and mitigation in containment, impact of oxidising conditions on source term, and iodine chemistry. The ranking of the high priority issues was updated to account for the results of recent international research and for the impact of Fukushima nuclear accidents in Japan. In addition, the ASTEC integral code was further developed to capitalize the new knowledge. The network has reached self-sustainability by integration in mid-2013 into the NUGENIA Association. The main activities and outcomes of the network are presented.
An Organisation for Economic Co-operation and Development (OECD) status report on iodine behaviour published in February 2007 concluded that although the understanding of iodine behaviour in containment had advanced considerably over the past several decades, there were still areas where further investigation was warranted. The OECD initiated the Behaviour of Iodine Project (BIP) to investigate two of these areas:
Source Term has been one of the main research areas addressed within the SARNET network during the 7th EC Framework Programme of EURATOM. The entire source term domain was split into three major areas: oxidising impact on source term, iodine chemistry in the reactor coolant system and containment and data and code assessment. The present paper synthesises the main technical outcome stemming from the SARNET FWP7 project in the area of source term and includes an extensive list of references in which deeper insights on specific issues may be found. Besides, based on the analysis of the current state of the art, an outlook of future source term research is outlined, where major changes in research environment are discussed (i.e., the end of the Phébus FP project; the end of the SARNET projects; and the launch of HORIZON 2020). Most probably research projects will be streamlined towards: release and transport under oxidising conditions, containment chemistry, existing and innovative filtered venting systems and others. These will be in addition to a number of projects that have been completed or are ongoing under different national and international frameworks, like VERDON, CHIP and EPICUR started under the International Source Term Programme (ISTP), the OECD/CSNI programmes BIP, BIP2, STEM, THAI and THAI2, and the French national programme MIRE. The experimental PASSAM project under the 7th EC Framework programme, focused on source term mitigation systems, is highlighted as a good example of a project addressing potential enhancement of safety systems based on already available knowledge.
The objectives of the SARNET network of excellence are to define and work on common research programs in the field of severe accidents in Gen. II-III nuclear power plants and to further develop common tools and methodologies for safety assessment in this area. In order to ensure that the research conducted on severe accidents is efficient and well-focused, it is necessary to periodically evaluate and rank the priorities of research. This was done at the end of 2008 by the Severe Accident Research Priority (SARP) group at the end of the SARNET project of the 6th Framework Programme of European Commission (FP6). This group has updated this work in the FP7 SARNET2 project by accounting for the recent experimental results, the remaining safety issues as e.g. highlighted by Level 2 PSA national studies and the results of the recent ASAMPSA2 FP7 project. These evaluation activities were conducted in close relation with the work performed under the auspices of international organizations like OECD or IAEA. The Fukushima-Daiichi severe accidents, which occurred while SARNET2 was running, had some effects on the prioritization and definition of new research topics. Although significant progress has been gained and simulation models (e.g. the ASTEC integral code, jointly developed by IRSN and GRS) were improved, leading to an increased confidence in the predictive capabilities for assessing the success potential of countermeasures and/or mitigation measures, most of the selected research topics in 2008 are still of high priority. But the Fukushima-Daiichi accidents underlined that research efforts had to focus still more to improve severe accident management efficiency.
Since the very beginning, the Phebus FP programme of integral experiments was considered as a necessary complement to the qualification "one by one" of physical models through separate effects tests. Small-scale analytical experiments are obliged to introduce hypotheses on the additivity of phenomena and do not allow to be sure that no important phenomenon has been omitted. Also the physico-chemical nature of a number of species can best be determined in integral type of experiments. For all those purposes, a series of five in-pile integral experiments has been performed. The facility provided prototypic reactor conditions which allowed the study of basic phenomena governing core degradation through to the late phase (melt pool formation), hydrogen production, fission product (FP) release and transport, circuit and containment phenomena, and iodine chemistry. For each of these topics, key lessons have been learnt and are described. Amongst the most important, one can cite: - The need to revisit cladding oxidation modelling, that impacts the hydrogen production kinetics. - The fuel collapse (transition from rod-like geometry towards a molten pool) at temperatures far below what was expected. - The fission product release from degrading fuel. - The chemical form of fission products when transported in the Reactor Coolant System, especially for iodine and caesium, the most important radionuclides. - The in-containment behaviour of iodine especially the reactions between iodine and paints and the trapping of iodine by silver under certain conditions. From these findings, the physical models implemented in simulation tools used for safety studies have been improved. Simulation tools have recently been extensively used for the understanding of the Fukushima accident events.
FPT3 was the last of the five in-pile integral experiments in the Phébus FP programme, whose overall purpose was to investigate fuel rod degradation and behaviour of fission products (FPs) released via the primary coolant circuit into the containment building. The results contribute to validation of models and computer codes used to calculate the source term for a severe accident with core meltdown in light water reactors. Unlike the previous tests, FPT3 used B4C as absorber material in the pre-irradiated (24.5 GWd/tU) fuel bundle, while featuring a steam-poor period as in FPT2, which used Ag/In/Cd absorber. The main FPT3 containment results are summarised: the source term of FPs, fuel and structural materials from the experimental circuit into the containment; the composition, morphology and deposition processes of aerosols in the containment atmosphere; the specific behaviour of the radiologically significant FP iodine; and finally the performance of passive autocatalytic recombiner (PAR) coupons exposed to the containment atmosphere just after the transient.
a b s t r a c t Plant assessments have shown that iodine contributes significantly to the source term for a range of acci-dent scenarios. Iodine has a complex chemistry that determines its chemical form and, consequently, its volatility in the containment. If volatile iodine species are formed by reactions in the containment, they will be subject to radiolytic reactions in the atmosphere, resulting in the conversion of the gaseous spe-cies into involatile iodine oxides, which may deposit on surfaces or re-dissolve in water pools. The con-centration of airborne iodine in the containment will, therefore, be determined by the balance between the reactions contributing to the formation and destruction of volatile species, as well as by the physico-chemical properties of the iodine oxide aerosols which will influence their longevity in the atmosphere. This paper summarises the work that has been done in the framework of the EC SARNET (Severe Accident Research Network) to develop a greater understanding of the reactions of gaseous iodine species in irradiated air/steam atmospheres, and the nature and behaviour of the reaction products. This work has mainly been focussed on investigating the nature and behaviour of iodine oxide aerosols, but earlier work by members of the SARNET group on gaseous reaction rates is also discussed to place the more recent work into context.
Ruthenium species, volatilized from damaged fuel during a severe accident in a nuclear power plant, are
radiotoxic and can be transported to the containment atmosphere in gaseous form. To limit the possible
source term to the environment, it is of interest to understand the behaviour of Ru after it has been
released from fuel and the phenomena taking place within the decreasing temperature section of the
reactor coolant system. This was investigated in the framework of EC SARNET and SARNET2 projects,
as a part of the Source Term work package, with several separate-effect tests on the transport and speciation
of Ru in primary circuit conditions considering the influence of other fission products as well.
The source of Ru was metallic Ru, RuO2 powder or gaseous RuO4. The large-scale integral tests of the Phébus
FP program were conducted with real irradiated fuel, and more realistic analysis on the release and
transport of Ru could be performed. Experimental studies proved that the transport of ruthenium to the
containment atmosphere took mainly place as RuO2 particles when Ru source was oxidized above
1250 �C. The fraction of Ru transported in gaseous form was at its highest when ruthenium was oxidized
at approx. 1000–1100 �C. A major part of the released Ru was deposited at the decreasing temperature
area of the circuit as RuO2. Revaporisation of the deposited Ru at low temperature was a significant
source of gaseous ruthenium. In order to understand the behaviour of ruthenium in these tests, the analysis
work was extensive and several simulations were carried out. As an outcome, the observed transport
and deposition of ruthenium was explained. The simulation studies gave also an insight into the performance
of the ASTEC code and some model improvements for Ru transport through the reactor coolant
system have been identified.
One of the most important areas of research concerning a hypothetical severe accident in a light water reactor (LWR) is determining the source term, i.e. quantifying the nature, release kinetics and global released fraction of the fission products (FPs) and other radioactive materials. In line with the former VERCORS programme to improve source term estimates, the new VERDON laboratory has recently been implemented at the CEA Cadarache Centre in the LECA-STAR facility. The present paper deals with the evaluation of the experimental equipment of this new VERDON laboratory (furnace, release and transport loops) and demonstrates its capability to perform experimental sequences representative of LWR severe accidents and to supply the databases necessary for source term assessments and FP behaviour modelling.
The objective of this work was to examine the chemical reactions taking place on primary circuit surfaces and their effect on fission product transport in a severe nuclear reactor accident. Especially transport of gaseous and aerosol phase iodine was studied. Caesium iodide (CsI) was used as precursor material for iodine species. Also, effects of molybdenum and boron on transport of iodine were investigated.
The experimental work showed that when CsI alone was used as a precursor, as much as 20% of the released iodine was in gaseous form and the rest as aerosol particles. Aerosol particles were most likely CsI. When the amount of hydrogen in the carrier gas was increased, the fraction of gaseous iodine decreased. When Boron was added to the precursor, a glassy caesium borate surface was formed on the crucible. Boron trapped most of the caesium and also a fraction of iodine, causing almost all released iodine to be in gaseous form. When Mo was introduced in the precursor, most of the iodine was again released in gaseous form. Oxidised Mo reacted with caesium releasing iodine from CsI. The effect of Mo on iodine transport depended much on H2 concentration and was observed to be substantially greater on stainless steel surface. When stainless steel crucible was used, Mo was found in small amounts from aerosol particles, indicating that it was probably released as caesium molybdate or as molybdenum oxide.
The Maus fission product (FP) programme studies the phenomenology of severe accidents in water-cooled nuclear reactors. Five tests were performed in the frame of the programme covering fuel-rod degradation and FP behaviour released via the coolant system into the containment. To model FP transport and behaviour in the coolant system, numerous physical and chemical phenomena have to be taken into account. In the vapour phase, for example, FP speciation, vapour condensation and vapour/surface or vapour/aerosol reactions have to be considered. The aerosol phase has to be modelled with nucleation, growth and deposition processes. Finally, remobilisation phenomena like resuspension and revaporisation have to be taken into account for delayed release into the containment. Four Phebus FP tests (FTP0, FPT1, FPT2, FPT3) have been modelled with the ASTEC/SOPHAEROS code. Modelling shows an overall good estimation of retention for the main FPs (e.g., I, Cs, Mo). Furthermore, a strong connection is revealed in the gaseous phase chemistry between I, Cs, Cd and Mo which has a great impact on gaseous iodine release into the containment. The Maus FP test modelling also exposes disagreement on FP retention when laminar gaseous flow is not well developed. Finally, probably the most significant shortcoming in modelling that Phebus-FP tests highlighted concerns vapour-phase iodine-chemistry modelling at low temperature. The study of this latter point is on going with the experimental programme ISTP/CHIP.
The Phebus and VERCORS data have played an important role in contemporary understanding and modeling of fission product release and transport from damaged light water reactor fuel. The data from these test programs have allowed improvement of MELCOR modeling of release and transport processes for both low enrichment uranium fuel as well as high burnup and mixed oxide (MOX) fuels. This paper discusses the synthesis of these findings in the MELCOR severe accident code. Based on recent assessments of MELCOR 1.8.5 fission product release modeling against the Phebus FPT-1 test and on observations from the ISP-46 exercise, modifications to the default MELCOR 1.8.5 release models are recommended. The assessments identified an alternative set of Booth diffusion parameters recommended by ORNL (ORNL-Booth), which produced significantly improved release predictions for cesium and other fission product groups. Some adjustments to the scaling factors in the ORNL-Booth model were made for selected fission product groups, including UO, Mo and Ru in order to obtain better comparisons with the FPT-1 data. The adjusted model, referred to as 'Modified ORNL-Booth,' was subsequently compared to original ORNL VI fission product release experiments and to more recently performed French VERCORS tests, and the comparisons was as favorable or better than the original CORSOR-M MELCOR default release model. These modified ORNL-Booth parameters, input to MELCOR 1.8.5 as 'sensitivity coefficients' (i.e. user input that over-rides the code defaults) are recommended for the interim period until improved release models can be implemented into MELCOR. For the case of ruthenium release in air-oxidizing conditions, some additional modifications to the Ru class vapor pressure are recommended based on estimates of the RuO vapor pressure over mildly hyperstoichiometric UO. The increased vapor pressure for this class significantly increases the net transport of Ru from the fuel to the gas stream. A formal model is needed. Deposition patterns in the Phebus FPT-1 circuit were also significantly improved by using the modified ORNL-Booth parameters, where retention of lower volatile CsMoO is now predicted in the heated exit regions of the FPT-1 test, bringing down depositions in the FPT-1 steam generator tube to be in closer alignment with the experimental data. This improvement in 'RCS' deposition behavior preserves the overall correct release of cesium to the containment that was observed even with the default CORSOR-M model. Not correctly treated however is the release and transport of Ag to the FPT-1 containment. A model for Ag release from control rods is presently not available in MELCOR. Lack of this model is thought to be responsible for the underprediction by a factor of two of the total aerosol mass to the FPT-1 containment. It is suggested that this underprediction of airborne mass led to an underprediction of the aerosol agglomeration rate. Underprediction of the agglomeration rate leads to low predictions of the aerosol particle size in comparison to experimentally measured ones. Small particle size leads low predictions of the gravitational settling rate relative to the experimental data. This error, however, is a conservative one in that too-low settling rate would result in a larger source term to the environment. Implementation of an interim Ag release model is currently under study. In the course of this assessment, a review of MELCOR release models was performed and led to the identification of several areas for future improvements to MELCOR. These include upgrading the Booth release model to account for changes in local oxidizing/reducing conditions and including a fuel oxidation model to accommodate effects of fuel stoichiometry. Models such as implemented in the French ELSA code and described by Lewis are considered appropriate for MELCOR. A model for ruthenium release under air oxidizing conditions is also needed and should be included as part of a fuel oxidation model since fuel stoichiometry is a fundamental parameter in determining the vapor pressure of ruthenium oxides over the fuel. There is also a need to expand the MELCOR architecture for tracking fission product classes to allow for more speciation of fission products. An example is the formation of CsI and CsMoO and potentially CsOH if all Mo is combined with Cs such that excess Cs exists in the fuel. Presently, MELCOR can track only one class combination (CsI) accurately, where excess Cs is assumed to be CsOH. Our recommended interim modifications map the CsOH (MELCOR Radionuclide Class 2) and Mo (Class 7) vapor pressure properties to CsMoO, which approximates the desired formal class combination of Cs and Mo. Other extensions to handle properly iodine speciation from pool/gas chemistry are also needed.
In a nuclear power plant (NPP), a severe accident is a low probability sequence that can lead to core fusion and fission product (FP) release to the environment (source term). For instance during a loss-of-coolant accident, water vaporization and core uncovery can occur due to decay heat. These phenomena enhance core degradation and, subsequently, molten materials can relocate to the lower head of the vessel. Heat exchange between the debris and the vessel may cause its rupture and air ingress. After lower head failure, steam and air entering in the vessel can lead to degradation and oxidation of materials that are still intact in the core. Indeed, Zircaloy-4 cladding oxidation is very exothermic and fuel interaction with the cladding material can decrease its melting temperature by several hundred of Kelvin. FP release can thus be increased, noticeably that of ruthenium under oxidizing conditions. Ruthenium is of particular interest because of its high radio-toxicity due to 103Ru and 106Ru isotopes and its ability to form highly volatile compounds, even at room temperature, such as gaseous ruthenium tetra-oxide (RuO4). It is consequently of great need to understand phenomena governing steam and air oxidation of the fuel and ruthenium release as prerequisites for the source term issues.
VERCORS is an analytical experimental programme focusing on the release of fission products (FP) and actinides from an irradiated fuel rod, under conditions representative of those encountered during a severe PWR accident. The 17 tests – financed jointly by EDF and IRSN – were conducted by the CEA on its Grenoble site in a specific high-activity cell at the Laboratory for Active Materials (LAMA) over a 14-year period (1989–2002), in accordance with three test phases. A first series of six tests (VERCORS 1–VERCORS 6) was conducted between 1989 and 1994 on UO2 fuel close to the relocation. Next, two different test series – VERCORS HT (three tests) and RT (eight tests) – were alternately performed between 1996 and 2002 at higher temperature up to the fuel sample collapse. These tests focused on UO2 and MOX fuels with a variety of initial configurations (intact or debris beds). This programme made it possible to precisely quantify fission product releases in all the situations explored, as well as to identify similar behavioural patterns between some of these fission products, thus making it possible to classify them schematically into four groups with decreasing volatility: (1) volatile FP including fission gases, iodine, caesium, antimony, tellurium, cadmium, rubidium and silver with very high releases (practically total release) at temperatures of around 2350°C; (2) semi-volatile FP, a category composed of molybdenum, rhodium, barium, palladium and technetium with releases of 50–100%, but very sensitive to oxygen potential and with marked redeposits nearby the emission point; (3) FP that are low volatile, such as ruthenium, cerium, strontium, yttrium, europium, niobium and lanthanum, with significant releases of around 3–10% on average, but capable (for some elements and under particular conditions) of reaching 20–40%; (4) non-volatile FP composed of zirconium, neodymium and praseodymium, for which no release can be measured by gamma spectrometry for the envelope conditions of the VERCORS test grids. Actinides each have their own type of behaviour. They can nevertheless be subdivided into two categories, the first including U and Np, with releases of up to 10% and behaviour similar to that of the low volatile FP, and the second (Pu) with very low releases, typically less than 1%.
In separate effect tests at 1000–1200 °C Ru oxidation rate and content of Ru in escaping air flow have been studied with special emphasis on effects of other fission product elements on the Ru oxidation and transport. The results showed that in the decreasing temperature section (1100–600 °C) most of the RuO3 and RuO4 (≈95%) decomposed and formed RuO2 crystals; while the partial pressure of RuO4 in the escaping air was in the range of 10−6 bar. The re-evaporation of deposited RuO2 resulted in about 10−6 bar partial pressure in the outlet gas as well. Measurements demonstrated the importance of surface quality in the decreasing temperature area on the heterogeneous phase decomposition of ruthenium oxides to RuO2. On the other hand water or molybdenum oxide vapour in air appears to decrease the surface catalyzed decomposition of RuOx to RuO2 and increases RuO4 concentration in the escaping air. High temperature reaction with caesium changed the form of the released ruthenium and caused a time delay in appearance of maximum concentration of ruthenium oxides in the ambient temperature escaping gas, while reaction with barium and rare earth oxides extended Ru escape from the high temperature area.
In the case of a severe accident in a nuclear Light Water Reactor (LWR), the high radiation fields reached in the reactor containment building due to the release of fission products from the reactor core would induce air radiolysis. The air radiolysis products (ARP) could, in turn, oxidise gaseous molecular iodine (I2) into aerosol-borne iodine–oxygen–nitrogen compounds, abbreviated as iodine oxides (IOx). These reactions involve the conversion of a gaseous iodine compound resulting in a change of the iodine depletion rate from the containment atmosphere. Kinetic data were produced within the first part of PARIS project on the air radiolysis products formation and destruction. The second part of the PARIS project as presented in this paper deals with the impact of the ARP on the conversion of I2 into IOx. The objective was to provide a database to develop new or to validate existing kinetic models of formation and destruction of iodine oxides.
The iodine tests of the PARIS project, performed at very low, realistic iodine concentrations, constitute an important database to further develop or validate empirical and mechanistic models on radiolytic I2 oxidation. In the presence of painted surface areas or silver aerosol surface areas, radiolytic I2 oxidation is negligible compared to I2 adsorption on these surfaces for the conditions examined. However, radiolytic I2 oxidation remains very efficient if surface areas are small or if they are made of the relatively non-reactive stainless steel.
NUREG-1465 was a major change to the preceding figures defining in-containment source term. Since then, the most important research venture in the arena of severe accidents has been the PHEBUS-FP project. Experimental data and interpretations brought in along the course of the project have highlighted similarities and discrepancies with those insights given in the NUREG-1465.This paper sets comparisons between NUREG-1465 and PHEBUS-FP in three key aspects: the release of radionuclides into containment, the in-containment aerosol behaviour and, finally, iodine chemical behaviour in the containment. The experimental basis of discussions is the FPT0, FPT1, FPT2 and FPT3 series, although the latter cannot be openly addressed yet. Similarities have been found regarding qualitative gap and early in-vessel releases, quantitative net release of noble gases and iodine releases, dominance of sedimentation as natural removal mechanism for in-containment aerosols, etc. Nonetheless, PHEBUS-FP data have also drawn attention to major discrepancies with respect to NUREG-1465. Examples are cesium and tellurium releases and possible massive iodine release under specific conditions. But, in addition, PHEBUS-FP has brought new insights, such as potential formation of in-sump iodine precipitates or the need of revisiting NUREG-1465 element grouping.
An overview of experimental programs that have been conducted to better understand core melt progression phenomena and fission product behaviour during severe reactor accidents in light water reactors is presented. This discussion principally focuses on the melting and liquefaction of core materials at different temperatures, materials oxidation and relocation, hydrogen generation behaviour, and the release and transport of fission products and aerosols. A comparison of fission product release results from annealing and in-reactor experiments is also presented.
Particle behaviour depends strongly on classic characteristics, e.g., size, and less macroscopic ones involving structure and composition these being especially important in situations of strong differential forces on a particle, i.e., surface impact or intensely-shearing flows. The former situation may lead to particle deposition or break-up and re-entrainment (with potential accident-management implications). This paper reviews information on aerosols from prototypical experiments identifying common features and typical variations. It emerges that a particle comprising one-third metal, one-third metal oxide and one-third a mixture of fission-product species would not be out of place in any potential reactor-accident sequence. Particle shapes appear relatively compact without branching chain-like structures. On size and structure, aerosols in the upstream part of the primary circuit would comprise a near-lognormal population with AMMD no more than 2 μm and geometric standard deviation around 2, particles comprising agglomerates of highly-coordinated clusters as small as 0.1 μm. In the containment, aerosols can typically be represented by primary-circuit particles and their agglomerates though particular circumstances (core–concrete interaction, hot-leg accident sequence) can alter this simple picture.
A particular concern in the event of a hypothetical severe accident is the potential release of highly radiotoxic fission product (FP) isotopes of ruthenium. The highest risk for a large quantity of these isotopes to reach the containment arises from air ingress following vessel melt-through. One work package (WP) of the source term topic of the EU 6th Framework Network of Excellence project SARNET is producing and synthesizing information on ruthenium release and transport with the aim of validating or improving the corresponding modelling in the European ASTEC severe accident analysis code. The WP includes reactor scenario studies that can be used to define conditions for new experiments.The experimental database currently being reviewed includes the following programmes:•AECL experiments conducted on fission product release in air; results are relevant to CANDU loss of end-fitting accidents;•VERCORS tests on FP release and transport conducted by CEA in collaboration with IRSN and EDF; additional tests may potentially be conducted in more oxidizing conditions in the VERDON facility;•RUSET tests by AEKI investigating ruthenium transport with and without other FP simulants;•Experiments by VTT on ruthenium transport and speciation in highly oxidizing conditions.In addition to the above, at IRSN and at ENEA modelling of fission product release and of fuel oxidation is being pursued, the latter being an essential boundary condition influencing ruthenium release.Reactor scenario studies have been carried out at INR, EDF and IRSN: calculations of air ingress scenarios with respectively ICARE/CATHARE V2; SATURNE-MAAP; and ASTEC codes provided first insights of thermal-hydraulic conditions that the fuel may experience after lower head vessel failure.This paper summarizes the status of this work and plans for the future.
The Phebus fission product and source term international programs
- B Clément
- Zeyen
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product and source term international
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Overview of the VERDON-ISTP program and the main insights from the VERDON-2 air ingress test
- A Gallais-During
A. Gallais-During et al., Overview of the
VERDON-ISTP program and the main
insights from the VERDON-2 air ingress
test, Proc. 7 th European Review Meeting
on Severe Accident Research (ERMSAR-2015), Marseille, France, 24-26 March
2015, http://www.sar-net.eu/
?q=node/56.
A study of the revaporisation behaviour of deposits on the metallic vertical line of Phébus FPT3
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revaporisation behaviour of deposits on
the metallic vertical line of Phébus FPT3,
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2012, http://www.sar-net.eu/
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Primary system fission product release and transport: A state of the art report, NUREG/CR-6193, NEA/ CSNI/R(94)2, 1994
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product release and transport: A state of
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Experimental study on the reactions between cadmium and caesium iodide on the surface of primary circuit piping during a severe nuclear accident
- M Gouëllo
M. Gouëllo et al., Experimental study on the
reactions between cadmium and caesium iodide
on the surface of primary circuit piping during a
severe nuclear accident, Proc. 7 th European
Review Meeting on Severe Accident Research
(ERMSAR-2015), Marseille, France, 24-26 March
2015, http://www.sar-net.eu/?q=node/56.
Study of the ruthenium transport under prevailing conditions in the primary circuit in the case of a PWR severe accident
- O Leroy
O. Leroy et al., Study of the ruthenium transport under prevailing conditions in the primary
circuit in the case of a PWR severe accident,
Proc. 7 th European Review Meeting on Severe
Accident Research (ERMSAR-2015), Marseille,
France, 24-26 March 2015,
http://www.sar-net.eu/?q=node/56.
Experimental study of the boron and air effects on iodine transport in the primary circuit during severe nuclear accident
- M Gouëllo
M. Gouëllo et al., Experimental study of the
boron and air effects on iodine transport in the
primary circuit during severe nuclear accident,
To be published in the proceedings of the International Congress on Advances in Nuclear
Power Plants (ICAPP2016), San Francisco, CA,
USA, 17 -20 April 2016.
In-Containment, Presented at the 1 st Source Term Workshop of NUGENIA
- S Dickinson
S. Dickinson, In-Containment, Presented at
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Marseille, France, 1-2 April 2015, unpublished
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Progress in iodine chemistry in SARNET2
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T. Haste et al., Progress in iodine chemistry
in SARNET2, Proc. 21 st Int. Conf. on Nuclear
Energy for New Europe, Ljubljana, Slovenia,
5-7 September 2012.