Fission Product Behavior Research Articles

Effect analysis of ISLOCA pathways on fission product release at Westinghouse 2-loop PWR using MELCOR

As the amount of fission product released from ISLOCA was overestimated because of conservative assumptions in the past, several studies have been recently conducted to evaluate the actual release amount. Among several pathways for the ISLOCA, most studies were focused on the pathway with the highest possibility. However, different ISLOCA pathways may have different fission product release characteristics. In this study, fission product behavior was analyzed for various pathways at the Westinghouse two-loop plant using MELCOR. Four pathways are considered: the pipes from a cold leg, from a downcomer, from a hot leg to the outlet of RHR heat exchanger, and the pipe from the hot leg to the inlet of RHR pump (Pathway 1–4). According to the analysis results, cladding fails at around 2.5 h in Pathways 1 and 2, and on the other hand, about 3.3 h in Pathways 3 and 4 because the ISLOCA pathways affect the safety injection flow path. While the release amount of cesium and iodine ranges between 20 and 26% in Pathways 1 to 3, Pathway 4 allows only 5% to the environment because the break location is submerged. Also, as more than 90% of cesium released to the environment passes through the personnel door, reinforcing the pressure capacity of the doors would be a significant factor in the accident management of the ISLOCA.

Distribution and behavior of fission product 95Nb in FLiBe salt

Abstract The 235,238UF4 was irradiated by photo-neutrons, distribution and behavior of the fission product 95Nb from irradiated 235,238UF4 in FLiBe salt were investigated by the measurement of its activity in the salt with the γ-ray spectroscopy. The experiments indicated that a part of 95Nb deposited on the surfaces of graphite and Hastelloy, as the moderator and the structural materials of molten salt reactor (MSR), respectively, and the majority of 95Nb maintained in molten salt. Addition of lithium metal made 95Nb in salt to be reduced and settled, leading to the decrease in its activity. Degree of the decrease was found to be correlated with niobium concentration. The experimental results supported the statement proposed early by ORNL, that 95Nb might be used as a redox indicator for MSR. Finally, the problem met with on-site monitoring for redox potential in MSR was pointed, and a possible protocol to resolve the problem was proposed.

Major results of the Research Committee on Fission Product Behaviors under Severe Accident and further developments in the succeeding committee

40年余に及ぶ福島第一原子力発電所(1F)廃炉プロジェクトの推進においては，核分裂生成物(FP)に向き合うことが避けられない。このため，FPの化学的挙動解明に対する新たな取り組みが必須と考え，2017年に本研究専門委員会を発足し，4年間の活動を続けた。本解説記事では，この活動成果を総括すると共に，長期廃炉作業において評価が必要なFPおよび燃料デブリ挙動のうち，特に重点的に検討すべき課題，新たに顕在化した課題を提示した。また，これらの課題に対する次期｢1F廃炉に係るFP挙動｣研究専門委員会での具体的な対応策についてまとめた。特に，実機データを反映した長期的なFP挙動解析手法とそれを支える基盤実験の推進策を紹介し，会員諸氏のご批判を受けつつ，目標達成に向かって注力したい。

Investigation of fuel microstructure at the top of a metallic fuel pin after a reactor overpower transient

The influence of a reactor overpower transient on irradiated metallic fuel performance was investigated in this work. Such transient studies support safety and performance optimization of future sodium fast reactors. A ternary metallic fuel alloy (U-19Pu-10Zr) in HT-9 cladding which was irradiated in the Experimental Breeder Reactor II (EBR-II) to a burnup of 11 at. %, and then subjected to a 30% overpower transient was studied. The sample was selected from the top of a fuel pin where failure is expected to first appear in the EBR-II pin design. This can occur when the pin is subjected to conditions beyond its design envelope. Main elemental distribution, phase characterization, fission product behavior, chemical form and fuel-cladding chemical interaction were analyzed in this study. Simulation of the temperature profile during the transients were performed using BISON and show that the moderate transient did not maintain prolonged high temperature. For the first time, the porous structure commonly found at the top of metallic fuel pin (termed “fluff” structure given its appearance) was examined in detail via postirradiation characterization. The fluff contains the major fuel elements (U, Pu, Zr) and presents a highly porous microstructure (over 40% area fraction). The presence of fission products relevant to source term was investigated. However, semi-volatile and volatile elements (namely Cs, I, Xe, Eu), expected in this region, could not be detected probably due to sample preparation. Europium, also expected in this region, was only detected in precipitates along with the other rare earth elements. Finally, no fluff (fuel) fragments detached from the fuel slug were detected in the plenum region, indicating the fluff stability under the tested moderate ramp. These analyses indicate that the moderate transient experienced by the fuel pin did not significantly influence fuel performance under this transient condition.

Modelling of fission products release in VERDON-1 experiment with cGEMS: Coupling of severe accident code MELCOR with GEMS thermodynamic modelling package

The treatment of chemistry and thermodynamics in the integral severe accident codes is typically limited. A more accurate treatment of the chemistry during the severe accident modelling is, therefore, of great interest. For this purpose, the work is focused on the development of coupling of the MELCOR code with chemical thermodynamic calculations using GEMS codes and HERACLES database. Developed coupling between the two codes, called cGEMS, allowed for the improved thermodynamic description of the fission product release from the nuclear fuel under severe accident conditions.VERDON-1 test was selected as an experimental reference for the simulations. Experimental release behaviour of Mo, Cs and Ba observed in VERDON-1 experiment was reproduced by the developed coupled code. Performed simulations provided detailed information about the fission product speciation at different redox conditions.The obtained information and the developed code provides a more accurate description of the fission product behaviour and release during severe accidents.

Overview and outcomes of the OECD/NEA benchmark study of the accident at the Fukushima Daiichi NPS (BSAF), phase 2 – Results of severe accident analyses for unit 2

This is the second paper in a series of 3 in which results of severe accident analyses for Unit 2 of Fukushima Daiichi are presented, gained in phase 2 of the OECD/NEA project “Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF)”. Nine organizations of six countries (CNL Canada; CRIEPI, IAE, JAEA and NRA Japan; IRSN France; KAERI Korea; NRC/DOE/SNL U.S. and VTT Finland) submitted results of their calculated severe accident scenarios for Unit 2 of Fukushima Daiichi using different severe accident codes (ASTEC, MAAP, MELCOR, SAMPSON, THALES-KICHE).The present paper describes the findings of the comparison of the participants’ results for Unit 2 against each other and against plant data, the evaluation of the accident progression and the final status inside the reactors. Special focus is on reactor pressure vessel status, melt release and fission product behavior and release. Unit 2 specific aspects will be highlighted and points of consensus as well as remaining uncertainties and data needs will be summarized.

Overview and outcomes of the OECD/NEA benchmark study of the accident at the Fukushima Daiichi NPS (BSAF) Phase 2 – Results of severe accident analyses for Unit 1

Phase 2 of the OECD/NEA Project “Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF)” was established in mid-2015. The objectives have been similar to Phase 1 of the project but with an extended analysis period of 3 weeks from the occurrence of the earthquake, a major focus on fission product (FP) behaviour and releases to the environment and the comparison to various data including radiological data and results of backwards calculations of the source term. Nine organizations of six countries (Ciemat Spain; IAE, JAEA and NRA Japan; CEA, IRSN France; IBRAE Russia; KAERI Korea; NRC/DOE/SNL U.S.A.; VTT Finland) submitted results of their calculated severe accident scenarios for Unit 1 at the Fukushima Daiichi site using different severe accident codes (ASTEC, MAAP, MELCOR, SAMPSON, SOCRAT, THALES-KICHE). This paper describes the findings of the comparison of the participants results for Unit 1 against each other and against plant data, the evaluation of the accident progression and the final status inside the reactors. Special focus is on RPV status, melt release and FP behaviour and release. Unit specific aspects will be highlighted and points of consensus as well as remaining uncertainties and data needs will be summarised. The results for Units 2 and 3 are presented next in separated papers.

Analyses of THAI 1 hydrogen deflagration using MELCOR code version 2.1 and 2.2

The Fukushima Daiichi NPP (Nuclear Power Plant) accident pointed out the hydrogen explosion issue as one of the main problems that can affect the NPP containment integrity. During a severe accident scenario, the hydrogen combustion can occur and lead to containment integrity failure, since it generates local and global pressure and heat spikes. Such topic was analysed in several research programs addressed all around the world. An important series of test campaigns was done in OECD/NEA WGAMA (Organisation for Economic Co-operation and Development/Nuclear Energy Agency Working Group on Analysis and Management of Accidents) program called THAI (Thermal-hydraulics, Hydrogen, Aerosol and Iodine). The THAI goal is to simulate several phenomena related on hydrogen and Fission Product behaviour in the containment to obtain data relevant for the code benchmarking and validation. Therefore, theoretical analyses are needed, in order to obtain a reliable prediction of the accidental scenario. The facility allows to investigate safety relevant effects under thermal-hydraulics conditions of severe accidents. The experiments performed cover from hydrogen deflagration to iodine and aerosol behaviour under different thermal-hydraulics conditions. Three representative experiments were chosen from the THAI campaign to be modeled and simulated using the MELCOR code with versions 2.1 and 2.2 and compare the results with the experimental ones.This work aims to assess the MELCOR code capability pointing out on the limitation in simulating the hydrogen deflagration and underling possible method to reduce their effect on the simulate results. The benchmarks were addressed with old version of MELCOR however the new version presented slightly different results due to the modification in the parametric model and the default sensitivity coefficients.

A Review of Circulating Voids and Gaseous Fission Product Behavior in Molten Salt Reactors

Abstract Molten salt reactors (MSRs) are a family of advanced, generation-IV, nuclear reactors that use a circulating molten alkali fluoride fuel salt as both a primary working fluid and a fuel matrix. Since the fuel salt is liquid, gases become entrained in the fuel salt flow, and these entrained gases are called circulating voids. This paper reviews some aspects of circulating voids on reactor behavior. A formal definition of the circulating void fraction is given. A time-line of Oak Ridge reports on circulating void behavior is presented. Methods to measure the circulating void fraction are described. Foam induced by gas entrained in the fuel salt is discussed. The discovery of circulating voids in the molten salt reactor experiment (MSRE) is detailed. Aspects related to bubble size and interfacial area are reviewed. The Laplace pressure is examined. The void and pressure coefficients of reactivity are examined. Some requirements for a bubble model for xenon behavior analysis are proposed. The effects of bubble surface mobility are examined. Finally, sources of gas production in the fuel salt are investigated.

Impact of fission product inclusion on phase development in U3Si2 fuel

Due to its high thermal conductivity and uranium density, U3Si2 has been considered as a candidate for use as an accident tolerant fuel (ATF). In order to fully assess its suitability and performance as a fuel, the impact of fission products (FPs) on the stability and performance of U3Si2 must be investigated. The interactions of FPs and U3Si2 have had relatively little study until now and require experimental and computational examination. U3Si2 was doped with individual FPs to explore U-Si-FP interactions and phase equilibria that may impact the performance of the ATF during irradiation. Elemental Ce, Mo, Y, or Zr were used to individually dope U3Si2 at a concentration of 5 wt% FP. A diffusion couple of a 1:1 Mo:Zr alloy and U3Si2 was heated to 1200 °C in order to consider the impacts of multiple FPs on the stability and structure of the fuel. Samples were characterized for FP solubility and secondary phase formation using electron microscopy, energy dispersive spectroscopy, and x-ray diffraction. First principles density functional theory calculations complemented the experimental effort to understand FP behavior. Experimental and computational findings were used in the development of a thermodynamic database containing 8 major FPs and their associated silicide phases. Fuel compositions generated from depletion calculations were used to thermodynamically model the equilibrium phases of the fuel undergoing burnup.

Fission product releases from thoria and urania fuels under reactor accident conditions: Hot cell experiment #6 (HCE6)

The accident-conditions fission-product release experiment HCE6 was performed on samples of thoria (ThO2) fuel and urania (UO2) fuel irradiated at water reactor operating conditions in the pressurized water loop in the NRU reactor. The tests in the experiment were intended to compare fission-product releases from ThO2 fuel and UO2 fuel under similar circumstances, and to explore the behaviour of fission products in ThO2 fuel under oxidizing conditions. Under inert conditions, or temperatures of ~1720 K, the releases from ThO2 were significantly lower than from UO2. A 1920 K test on a sheathed sample of ThO2 exhibited a delay of ~3000 s in noble gas releases relative to an inert-environment ThO2 fragment test, possibly due to a decrease in oxygen potential by contact with Zircaloy at high temperature. In a steam environment, the releases of Kr, Xe, I, and Cs, were quite similar for the ThO2 and UO2 samples, and the release timing illustrated similar effects of Zircaloy oxidation on the release of these fission products. In an air-steam mixed environment, lower releases of noble gases and volatile fission products from the ThO2 sample were observed, probably due to the shorter exposure to oxidizing environment and greater amount of Zircaloy present. None of the tests showed significant releases of 95Nb, 103Ru, 140Ba, or 233Pa.

MOX fuel microstructural evolution during the VERDON-3 and 4 tests

The VERDON-3 and -4 tests were part of the VERDON-ISTP programme that aimed at studying the fuel and fission products (FP) behaviour in severe accident conditions. The main objective of these two complementary tests was the study of MOX fuel behaviour and FP release under oxidising (VERDON-3) and reducing (VERDON-4) conditions at very high temperature (>2300 °C). Complementary to the on-line gamma spectrometry measurements performed during the two tests, post-test characterisations were carried out in order to tackle these tasks. The two samples recovered after the VERDON-3 and -4 tests were compared to a third one extracted from the same father rod and left as irradiated. This comparison enabled to highlight the effect of temperature and atmosphere on the fuel behaviour. These three samples were characterised by several techniques available at the LECA-STAR facility of the CEA Cadarache. Experimental observations showed that an interaction between the fuel and the cladding occurred in both types of conditions by interdiffusion mainly between U and Zr. This phenomenon led to the formation of a UyZr1-yO2±x cubic phase at the periphery of the fuel pellet which melted in the VERDON-4 test conditions, penetrating through the cracks of the sample and dissolving the fuel matrix. No liquid was formed during the VERDON-3 test despite the formation of a large fuel-cladding interaction zone.

Fission product speciation in the VERDON-3 and VERDON-4 MOX fuels samples

Within the framework of the International Source Term Programme, the VERDON-3 and VERDON-4 complementary tests were devoted to studying MOX fuel and fission product (FP) behaviour at very high temperature (>2300 °C) under oxidising and reducing atmospheres. Post-test qualitative and quantitative characterisations of the VERDON-3 and VERDON-4 samples were carried out in the LECA-STAR facility at the CEA Cadarache centre. This article focuses on the impact of the atmosphere on fission product speciation and more precisely on the behaviour of Cs, Mo and Ba. The well-known white inclusions composed of metallic Mo, Ru, Rh, Pd and Tc underwent transformations during the two tests implying melting and/or composition changes. The semi-volatile behaviour of Mo was once again confirmed. Almost no Cs was seen to remain in the samples after the VERDON-3 and VERDON-4 tests, which is consistent with its volatile behaviour. The behaviour of Ba, which is known to be released in high quantities in reducing atmospheres, provided complex to assess as significant releases seems to have taken place in both VERDON-3 and VERDON-4 tests.

Analytical bond-order potential for silver, palladium, ruthenium and iodine bulk diffusion in silicon carbide

The analytical bond-order potential has been developed for simulating fission product (Ag, Pd, Ru, and I) behavior in SiC, especially for their diffusion. We have proposed adding experimentally available elastic constants and physical properties of the elements as well as important defect formation energies calculated from density functional theory simulation to the list of typical properties as the extensive fitting database. The results from molecular dynamics simulations are in a reasonable agreement with defect properties and energy barriers of their experimental/computational counterparts. The successful validation of the new potential has established a good reliability and transferability of the potentials, which enables the ability of simulation in extended scale. The kinetic behavior such as diffusion of different interstitials is then realized by applying the new interatomic potentials. The bulk diffusion is less likely to dominate the transport of the four fission products under pure thermal condition, when we refer to their extremely small values of the effective diffusion coefficients. The interstitial mechanism is hard for Pd, Ru, and I to access due to the high formation energy and high migration energy. However, it is found that the migration energy of silver is relatively low, which indicates Ag diffusion via an interstitial mechanism being feasible, especially under irradiation condition, where massive interstitials can be formed in high-temperature nuclear reactors.

Segregation behavior and embrittling effect of lanthanide La, Ce, Pr, and Nd at Σ3(111) tilt symmetric grain boundary in α-Fe**Project supported by the National Natural Science Foundation of China (Grant No. U1867217), the National Science and Technology Major Project of the Ministry of Science and Technology of China (Grant No. 2019ZX06004009), and the China National Nuclear Corporation Centralized Research and

The migration of lanthanide fission products to cladding materials is recognized as one of the key causes of fuel–cladding chemical interaction (FCCI) in metallic fuels during operation. We have performed first-principles density functional theory calculations to investigate the segregation behavior of lanthanide fission products (La, Ce, Pr, and Nd) and their effects on the intergranular embrittlement at Σ3(111) tilt symmetric grain boundary (GB) in α-Fe. It is found that La and Ce atoms tend to reside at the first layer near the GB with segregation energies of −2.55 eV and −1.60 eV, respectively, while Pr and Nd atoms prefer to the core mirror plane of the GB with respective segregation energies of −1.41 eV and −1.50 eV. Our calculations also show that La, Ce, Pr, and Nd atoms all act as strong embrittlers with positive strengthening energies of 2.05 eV, 1.52 eV, 1.50 eV, and 1.64 eV, respectively, when located at their most stable sites. The embrittlement capability of four lanthanide elements can be determined by the atomic size and their magnetism characters. The present calculations are helpful for understanding the behavior of fission products La, Ce, Pr, and Nd in α-Fe.

TEM characterization of high burnup fast-reactor MOX fuel

This work examines the radial evolution of a high burnup fast reactor mixed-oxide (MOX) fuel pellet irradiated at the Fast Flux Test Facility (FFTF) using scanning (SEM) and transmission electron microscopy (TEM). At 21.3% fissions per initial metal atoms (FIMA), the sample analyzed in this work is likely among the highest burnup oxide fuel samples ever evaluated using TEM. Initial SEM examination focused on the characterization of the fuel microstructure and solid fission product behavior as a function of radial position. The fuel pellet underwent extensive restructuring as demonstrated by columnar grain formation near the central void. Oxide fission product phases were confined to near the fuel centerline and decomposed from perovskite BaZrO3 to the fluorite ZrO2 type crystal structure. Metallic five metal precipitates (FMPs) were observed at all radial positions and the Pd-rich metallic phase increased in area fraction moving toward the pellet surface. TEM examination focused on the determination of dislocation density ranging from the fuel pellet center to the pellet surface. Dislocation density was highest in the cooler regions of the fuel pellet, with the outer edge experiencing grain refinement and high defect concentrations consistent with the rim structure. The results of the paper highlight the variable dislocation behavior as a function of radial position in high burnup MOX fuels, as well as describe the largely unexplored change in crystal structure of the oxide solid fission product phase.

Investigation of differences of phenomena of diffusiophoresis and thermophoresis from aerosol deposition in modified ART Mod 2 code

Fission product behavior assessment in containment vessel (CV) is one of the important areas in safety analyses because the CV is the final release protection. Modified ART Mod 2 code is used to study aerosol deposition phenomena in the CV. From Phébus FPT experiments, the phenomena of diffusiophoresis and thermophoresis on wet condensers were occurred in the CV. The two phenomena are recognized to be the same thing. However, this is not a universal view on these phenomena, which are evaluated separately in the modified ART Mod 2 code. In previous study on the modified ART Mod 2 calculation of Economic Simplified Boiling Water Reactor (ESBWR) and Phébus FPT experiments, there was more diffusiophoresis than thermophoresis in the case of simulating single volume of the CV while there was more thermophoresis for multiple volume case. This research aim is to conduct a ceteris paribus sensitivity analysis of representative parameters affecting diffusiophoresis and thermophoresis namely aerosol size, pressure, and gas temperature, under different nodalizations in the modified ART Mod 2 code using the Phébus FPT experiments to investigate the differences. As a result, differences of gas temperature between atmosphere and wall in single volume most affect diffusiophoresis while differences of temperature between atmosphere and wall of multiple volumes most affect thermophoresis.

Box modelling of gas-phase atmospheric iodine chemical reactivity in case of a nuclear accident

A new atmospheric iodine gas-phase mechanism containing 248 reactions combined to the RACM air quality mechanism was implemented into a 0D box model in order to evaluate the behaviour of iodine fission products released in the atmosphere. Different simulation scenarios were performed unravelling the influence of several parameters on the iodine speciation (season, day/night, low/medium/high iodine concentration release, nature of the injected iodine species). The main outcomes of our work are the following: (i) the I2 reactivity is faster than that of CH3I, (ii) iodine nitroxides INOx are mainly formed during night with low and medium releases in winter, (iii) iodine oxides are promoted with a high release of I2 during summer, (iv) the formation of organic iodine compounds can be highlighted only in the case of a low I2 release. In this work, I2O, I2O5, and HOIO2 are the most abundant iodine oxides because their atmospheric removal pathways in the gas-phase are not yet known. The main iodinated organic species is CH3OOI arising from the reaction between I atom and the methyl peroxy radical. In case of a CH3I release, before the I atom formation, the reaction mechanism leads to either the formation or the destruction of several other organic iodinated compounds such as CH2IOH and CH2IOO.

Leaching behavior of gamma-emitting fission products, calcium, and uranium from simulated MCCI debris in water

ABSTRACT Gamma-ray radionuclides (Cs-137, Ba-140, I-131, Ru-103, and Zr-95) were produced by neutron irradiation of simulated molten core–concrete interaction (MCCI) debris, which was synthesized by the heat treatment of a mixture of UO2 with concrete components at a relatively low temperature of 1473 K under reducing and oxidizing conditions. The major uranium solid phases were unreacted UO2 and calcium uranate. The leaching ratio of the radionuclides in the powdered sample of the simulated MCCI debris was investigated under atmospheric conditions at 298 K in 0.1 mol/dm3 NaClO4 after filtration of the leachate through a 0.45-µm pore membrane. The uranium molar concentration in the filtrate was affected by the oxidation state in the solids. In the present study, however, the effect of various solid phase conditions on the leaching ratio normalized to that of uranium matrix could not be clarified. It was found that the leaching ratio of various fission products (R M) was proportional to that of Cs (R Cs), and this trend did not depend on the oxidation state of uranium, the type of uranium complex (including a colloidal species), or the presence of Ca, Si, cement, or Zr.

Characterization of solid fission products in 13.7% FIMA MOX fuel using electron microscopy techniques

This work utilizes electron microscopy-based techniques to examine the radial behavior of solid fission products in plutonium (Pu) bearing mixed oxide (MOX) fuel irradiated to a burnup of 13.7% fissions per initial metal atom (FIMA). Metallic precipitates primarily consist of five fission products: ruthenium (Ru), rhodium (Rh), technetium (Tc), molybdenum (Mo), and palladium (Pd). The five metal precipitates (FMPs) examined in this work have low concentrations of Pd and Mo, with no major compositional differences along the fuel radius. A secondary Pd–Te metallic phase forms in cooler regions of the pellet, likely due to the diffusion of gaseous species away from the central void. X-ray chemical maps indicate that the Pd–Te phase can nucleate on the surface of FMPs before precipitating into separate particles. These particles were also found to alloy with iron (Fe) near the surface of the fuel pellet due to interdiffusion with the stainless-steel cladding. The insoluble perovskite oxide phase was found to form near the central void and at intermediate radial positions, but not at the fuel edge. These findings suggest that solid fission product phases form at varying counts and compositions along the fuel pellet radius, and thus should be considered when describing the thermal behavior of the fuel.