Material Properties of a Natural UF6 Transport Cylinder Vessel at High Temperature

Abstract

The IAEA Regulations for the Safe Transport of Radioactive Material are to be revised in 1996 and the fire test (800°C for 30 min) could become a requirement for the natural UF₆ transport cylinder. ASME SA 516 carbon steel is used as the structural material for this type of cylinder. It is very important to obtain high temperature data for SA 516 steel to be able to evaluate the integrity of the UF₆ transport cylinder vessel in the fire test. CRIEPI has therefore conducted material tests on SA 516 at high temperatures. The AC₁ and AC₃ transformation points of actual SA 516 steels have been measured. Tensile tests up to 900°C were conducted using USA, French and Japanese manufactured materials and the influence of phase transformation assessed. Preliminary creep tests show that assessment by creep strength can give a more conservative estimation than using the tensile strength. Creep deformation equations have been obtained using uniaxial creep tests and internal pressure creep tests. In addition, by the use of internal pressure creep rupture tests, the relation between the circumferential stress, the test temperature and the rupture time has been obtained.     

Modelling of polyurethane foam thermal degradation within annular region subject to fire conditions

Abstract

Nuclear materials are placed in shielded, stainless steel packaging for storage or transport. These drum type packages often employ a layer of foam, honeycomb, wood or cement that is sandwiched between thin metal shells to provide impact and thermal protection during hypothetical accidents, as those prescribed in the Code of Federal Regulations (10 CFR 71·73). The present work discusses the modelling of the thermal degradation of polyurethane (PU) foam within an annular region during an 800°C fire. Measurements and analysis by Hobbs and Lemmon [M. L. Hobbs and G. H. Lemmon: ‘Polyurethane foam response to fire in practical geometries’, Polym. Degrad. Stab., 2004, 84, 183–197.] indicate that at elevated temperatures, PU foam exhibits a two-stage, endothermic degradation. The first stage produces a degraded solid and a combustible gas; the second stage reaction consumes the degraded solid and produces another combustible gas. As a result, during a prolonged fire, a gas filled void develops beside the outer metal shell and grows inward toward the inner shell and the containment vessel. As a result of the radial symmetry in the drum geometry, a one-dimensional finite difference model is constructed for the annular foam region. Heat flux is applied to the inner surface to model the decay heat of the containment vessel contents. Thermal radiation and convection boundary conditions with a specified environmental temperature are applied to the outer surface. The material and reaction rate properties determined by Hobbs and Lemmon are applied to the foam. The annular region temperature and composition are determined as functions of radius and time after the environmental conditions are changed from room temperature to those of an 800°C fire. The effects of surface to surface radiation between the package’s outer skin and the undegraded foam and the reaction rate reduction due to material damage during the reaction are evaluated for fires lasting 20 h. The peak package liner temperature caused by a 30 min fire is predicted to be 129°C, well below the short term limit for containment vessel seal (377°C).     

Behaviour of package for transport of spent fuel assemblies exposed to beyond regulation fires

Abstract

Institut de Radioprotection et de Sûreté Nucléaire (IRSN) performed a study relative to the thermal behaviour of a new TN International package design for transport of spent fuel assemblies called TN®112. The aim of this study is to evaluate the behaviour of the package exposed to fires, with durations and temperatures different from those required in the IAEA regulation TS-R-1 (respectively 30 min and 800°C). Its main objective is to provide quantitative data available for safety assessment in emergency situations involving fires. Moreover it can also be used for a cross comparison with the analysis of the thermal behaviour of the package during the IAEA regulatory fire test presented by the applicant in the package design safety analysis report. This study is based on numerical calculations performed with the code THERMX-PROTEE. The three-dimensional model used represents a quarter of the upper half of the package, where the closure system is located. The thermal behaviour of the neutron-shielding resin located in the cavity plug, the trunnions and the packaging body was modelled to allow simulation of endothermic reactions of vaporisation. During the heating phase of the fire test, the water vapour produced in the heated resin components is transferred and condensed in the nearby colder elements; the associated thermal transfers can rapidly increase the temperature of the colder elements. The part of the vapour which cannot be condensed when most of the nearby resin elements reach a temperature above 100°C is evacuated through the holes that are distributed throughout the external envelope of the packaging and closed by fusible plugs under normal conditions. A specific calculation module has been developed to take into account the corresponding energy transfers. This module was qualified by comparison with the results of experimental fire tests. The calculations performed in the framework of this study cover fire temperatures between 400 and 1000°C. One of the results of those calculations is the time necessary to reach the maximum allowable temperature of the elastomer gaskets.     

Fluorination of Uranium Metal by Fluorine Gas

A study has been made on the fluorination of uranium metal chips to UF₆ with fluorine gas in the temperature range of 100° to 400°C. The formation of UF₆ was influenced by the concentration and the flow rate of fluorine gas and by the reaction temperature. The reaction occurred apparently above 200°C. Uranium metal was first converted to low fluorine content compound such as UF₃ or UF_4-x, then to UF₄ and finally to UF₆. The intermediate compounds were confirmed by X-ray analysis and by their color.     

Evaluation of UF6 Vapor Release in Postulated Accident

The rupture of UF₆ gas line connected to hot UF₆ cylinder, being one of various accidents in UF₆ vapor leak-out, is considered as a postulated accident for uranium enrichment plants. For this type of rupture, we will estimate the amount of UF₆ vapor release based on a simplified calculation model and then make an evaluation of UF₆ vapor release through a ventilation system of feed vaporization facility. Assuming an instantaneous steady state for the change of UF₆ states, an unsteady state thermodynamics process is solved. Numerical examples show that about 52% of the initial UF₆ quantity are vaporized at 80°C (the temperature of the liquid UF₆ in the cylinder). Furthermore, by using the amount of released UF₆ vapor and the collection capacity of HEPA filter for IiF gas, the amount of gaseous UO₂F₂, HF which may be dissipated to the environment are conservatively estimated.     

Preparation of Uranium Phosphide from Uranium Tetrafluoride

A study was made on UP preparation by two-step reactions starting with UF₄, Si and red phosphorus.

The first step was to produce an intermediate uranium phosphide U₃P₄ from the reaction of UF₄ with Si under phosphorus vapor at temperatures above 900°C.

In the second step, the intermediate phosphide was heat treated under vacuum. To obtain single phase UP from the intermediate phosphide, the treatment required a temperature above 1,050°C, at which temperature the minimum holding time was 60 min.     

Gas Chromatography of Uranium Hexafluoride at Low Temperatures

Gas chromatography of UF₆ at low temperature was studied with the use of columns of polytrifluoromonochloroethylene oils as liquid phase. The dependence of retention time and HETP of UF₆ on (1) degree of polymerization of oils, (2) liquid phase loading, and (3) kinds of solid support were studied in the temperature range between ?10° and 40°C, and the most favorable conditions for quantitative analysis of UF₆ are discussed.

The relation between the gas chromatographic characteristics of the columns and the B.E.T. surface area of the solid support has been briefly examined.     

Kinetic Studies of Fluorination of Uranium Carbides by Fluorine, (I)

The rate of the fluorination reaction of UC with fluorine to form UF₆ has been studied in the temperature range between 240° and 300°C by following with a thermobalance the change in weight of the solid phase. The overall reaction takes place in two steps: In the first step, UC is rapidly converted to UF₄, which, in the second step, fluorinates to UF₆. During the first step, polymeric fluorocarbon is also produced and is found dispersed throughout the intermediate UF₄ particles. At reaction temperatures below 270°C, the second step reaction roughly follows the linear law. At temperatures above 270°C, it comes to follow the parabolic law, presumably due to the formation of a compact layer of polymeric fluorocarbon on the surface of the particles. Analysis of the gaseous products using mass and infrared spectrophotometers revealed the formation of several kinds of lower fluorocarbons.     

Pyrohydrolysis Reactions of UF4 and UO2F2

This study was made for clarification of the pyrohydrolysis behaviors of UF₄ and UO₂F₂. The progress of the reaction was measured by titrating the amount of produced hydrogen fluoride. When nitrogen was used as carrier gas of water, the pyrohydrolysis of UF₄ proceeded at the temperatures of 350–400°C, and UO_2+x (x?0.3) was formed; the value of x decreased with the decrease of oxygen dissolved in the water. In the case of oxygen carrier gas, the pyrohydrolysis of UF₄ formed U₃O₈-UO₂F₂ mixture as the reaction product in the above temperature region and the reaction was markedly retarded in the course because of lower rate in the pyrohydrolysis of UO₂F₂. The pyrohydrolysis of UO₂F₂ proceeded at a little [higher temperature of 450–500°C in both cases of nitrogen and oxygen carrier gas and α-UO₃ was formed.     

Kinetic Studies of Fluorination of Uranium Carbides by Fluorine, (II)

The fluorination reaction of uranium dicarbide with fluorine to form UF₆ has been studied in the temperature range between 220° and 300°C using a thermobalance. The overall mechanism of reaction is similar to the case of uranium monocarbide, i.e. in the first step, UC₂ is rapidly converted to UF₄, polymeric fluorocarbon and gaseous fluorocarbon, while in the second step, UF₄ and polymeric fluorocarbon are converted rather slowly to UF6 and gaseous fluorocarbon. The amount of polymer is much larger than the case of uranium monocarbide, its weight ratio to UF₄ being 0.2 to 0.25 in the early stages of the reaction. The fluorination of ₄ to UF₆ always follows the linear law derived from the diminishing sphere model. The apparent activation energy was determined to be 19.5 kcal/mole. Differences in the fluorination between UC and UC₂ are discussed, with the effect of polymer taken into consideration.     

Compatibility between several heat resistant alloys and sintered Li2O in static helium gas environment

The reaction of sintered Li₂O discs with several commercial heat resistant alloys has been investigated under the conditions of 3.3 × 10⁴Pa ($\text{1}\text{3}\text{atm}$) static He gas atmosphere in the temperature range of 500 and 750° C. Reaction products were identified by X-ray diffraction analysis to be two phases of Li₅FeO₄ and LiCrO₂. The former was dominant below 650° C and the latter was dominant above 650° C. The activation energies of the reaction were determined by the Arrhenius plots for weight changes and sub-scale thickness measurements. The reactivity of the four Fe-Ni-Cr alloys decreased according to the order of Incoloy 800, 316 SS, Hastelloy X-R and Inconel 600. Grain boundary penetration was observed above 500° C for Incoloy 800, 550° C for 316 SS and 600° C for Inconel 600. There was no grain boundary penetration in Hastelloy X-R.     

Preparation of Uranium Nitrides from Uranium Tetrachloride

A study was made of UN preparation by two-step reactions with UCl₄, Al and N₂ gas. In the first step, an intermediate uranium nitride with an N/U ratio of 1.68 resulted from the reaction of UCl₄ with A1 under constant flow of N₂ at reaction temperatures higher than 800°C. The X-ray pattern of the intermediate nitride did not correspond to any previously known uranium nitride. A maximum conversion efficiency of about 70% was obtained at temperatures between 800° and 1,000°G for the first reaction. In the second reaction, the intermediate nitride was heat treated under vacuum. To obtain single phase UN from the intermediate nitride, the heat treatment required a temperature of at least 1,100°C, at which the minimum holding time was 60 min.     

Influence of fission products on ruthenium oxidation and transport in air ingress nuclear accidents

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 RuO_x 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.     

Helium bubbles and trace of lithium in B4C control rod pellets used in JOYO experimental fast reactor

B₄C pellets used in the control rod of the experimental fast reactor ‘JOYO’ with different ¹⁰B burnups from lower than 10 × 10²⁰ captures/cm³ to 80 × 10²⁰ captures/cm³ and irradiated at less than 800 °C were examined by transmission electron microscopy (TEM). In a B₄C pellet irradiated in an irradiation capsule of ‘JOYO’ at 800 °C up to 30 × 10²⁰ captures/cm³, intragranular helium bubbles appeared in flat plate-shapes with the plane of the plate parallel to the (111) rhombohedral plane. However, in the other specimens that were taken from an actual control rod, the helium gas formed very tiny spherical intergranular bubbles with a diameter of a few nanometers . These tiny bubbles make wavy arrays roughly parallel to the (111) plane. The B₄C specimens were heated on a TEM in situ heating holder up to 1040 °C for 10 min. Clustering of tiny bubbles was observed, but did not extend to the plate-shaped bubbles. In high burnup specimens, large bubbles/cracks were rarely found along the {100} planes, which may correspond to the amorphous bands caused by the slip. While heating the specimens in TEM over 800 °C, liquid phases of lithium-bearing compoundsappeared on the surface of specimen.     

Fluorination of Uranium Compounds by Bromine Trifluoride Vapor, (I)

The fluorination of U₃O₈ powder by BrF₃ vapor was attempted. The reaction proceeded even at 67°C and under 10 mmHg BrF₃ partial pressure, producing UF₆. The reaction rate increases with temperature up to about 220°C, and its activation energy is 0.9 kcal/mol. The reaction rate, however, decreases at temperatures slightly above 220°C and rises again after passing through a minimum at 225°–230°C.     

Thermal analysis of NAC-LWT cask under normal and fire accident conditions

Abstract

Two- and three-dimensional thermal models of a Nuclear Assurance Corporation Legal Weight Truck (NAC-LWT) cask were constructed using the PATRAN commercial finite element package. The two-dimensional model included the effect of radial stiffeners in the package’s external neutron shield but the three-dimensional model did not. A normal conditions of transport (NCT) simulation using both models predicted the peak cladding temperature was roughly 210°C. The NCT package temperatures were used as initial conditions for transient fire/post-fire simulations. Different assumptions were used to determine when the neutron shield liquid drained from the tank and was replaced by air. When the liquid was assumed to remain within the tank during and after the fire, the peak cladding temperature was predicted to exhibit a temporal maximum of roughly 300°C, ～6 h after the end of the fire. If the liquid drained from the tank during the fire, the cladding temperature did not exhibit a temporal peak. Rather, it eventually reached a maximum temperature of roughly 280°C, which is the steady state NCT peak temperature when air is in the neutron shield tank. This undergraduate project will be used to lay down a foundation for further research on NAC-LWT casks. Two and three dimension package of the cask will be constructed using ANSYS, and simulations will be run for NCT and fire/post-fire conditions. The models will also be linked to Container Analysis Fire Environment (CAFE) to predict response of the package in fire.     

Wetting by sodium at high temperatures in pure vapour atmosphere

The wetting angle of sodium on different pure metals (Ni, Ta, Mb, Mo, W), alloys (stainless steel 304L, Inconel 600, RTG 36, TZM) and oxides (Feldmühle E37, Degussa AL23, sapphire, ZrO₂, Y₂O₃) was measured in the temperature range from 520 to 720°C. A new measurement method was applied consisting of a combination of the classical sessile drop with a gas controlled heat pipe. The method is especially suited for high temperatures where the previous methods are impracticable because of excessive sodium evaporation. It has furthermore the advantage that the measurements are made under pure sodium vapour. In the temperature range mentioned all investigated materials are well wetted. The maximum measured wetting angle is 7.5°. The wetting angles are nearly independent from temperature, but they show a time-dependency consisting in general of a rapid initial decrease and a reaching of the equilibrium value after about 1 hour.     

Effect of Neutron Irradiation on Mechanical Properties of Al-Li Alloys

An increase in yield stress at room temperature was observed in Al-0.6W/0 Li alloy irradiated to thermal neutron doses of 2.9 × 10¹⁹ to 7.2 × 10¹⁹ cm^?2. The hardening of as-irradiated specimens is accompanied with yield point followed by jerky yield-elongation in the stress-strain curve. The radiation hardening could not be annealed out by heating for 30 min at temperatures up to 350°C, whereas the yield-elongation disappeared gradually with increasing heating temperature in the l mm diam. specimens; with the 2 mm diam. specimens the yield-elongation still remained even after post-irradiation heating for 30 min at 350°C. Strengthening accompanied by jerky yield-elongation is considered to be due to He atom clusters precipitated along the dislocation. The hardening observed in the specimens heat-treated after irradiation at temperatures above 250°C is caused by randomly distributed gas bubbles.

In heavily cold-worked Al-0.6%W/o Li specimens, recovery of work hardening occurred during neutron irradiation to 4.2 × 10¹⁹ cm^?2. Hardening due to gas bubbles was also observed in the cold-worked specimens. In Al-2.7W/0 Li alloy, an increase in yield stress took place in the specimens irradiated to 4.2 × 10¹⁹ cm^?2 and heated for 30 min at temperatures of 155° to 260°C. The hardening is thought to be due to re-precipitation of β-phase resolved during the neutron irradiation.     

Oxidation and Gas Release Behavior of UC,U(C,N) and UN at High Temperature

A study was made of the oxidizing behavior at high temperature (800°–1,800°C) in vacuum of UC, UN and U(C,N) samples containing added oxygen in excess amounts, through observations of gas release, X-ray diffraction analysis, and microphotography.

The oxidation in vacuum of UC and U(C,N) was found to proceed above 1,200°C by stepwise reactions from one temperature interval to the next, the process differing however according to the chemical state of the oxygen present in the samples. In the temperature range below 1,200° C, the UC and U(C,N) samples reacted violently with the free oxygen present in dissolved state, to form UO₂. Between 1,200° and 1,400° C the UO₂ thus produced reacted with the UC or U(C,N), forming solid solutions of U(C,O) and U(C,N,O) respectively: Above 1,600°C, these solid solutions gradually decomposed back into UC and U(C,N), and U. In all stages of oxidation, large amounts of CO—and N₂ in the case of U(C,N)—evolved from the samples as reaction products. In the case of UN, no reaction was observed below 1,200°C, and only oxidized above that temperature to form UO₂ and N₂ by the action of the dissolved oxygen present.

These results indicate that in the case of UC and U(C,N), the quantity of gases evolving from the oxidation is dictated by the total amount of oxygen contained in the samples, while that from UN is dependent on the amount of molecular oxygen alone.     

Effect of Barriers on Sodium Mist Deposition on LMFBR Components

Tests were made to evaluate the effect of two barriers (a convection barrier and a drip receiver) against sodium mist deposition on LMFBR conponents in the cover gas space. Two models of LMFBR rotating shield plugs (1,800 mm in height, 680 and 850 mm in diameter) and a sodium test tank were manufactured for this purpose, and the mist deposition rate on the walls of these models was measured both in the cases with and without a barrier. The sodium pool temperature during the test was 580°C. In parallel with these tests, the relation between the form of deposit and the wall temperature was examined to determine the critical wall temperature above which deposits do not accumulate. This was conducted by exposing a vertical stainless steel cylinder to cover gas entrained with mist over 500°C sodium for 30 to 1,000 h.

The test results revealed that the barriers functioned effectively, and they reduced the local deposition rate near top of the annulus by three orders of magnitude relative to the cases without a barrier. The critical wall temperature to avoid deposit accumulation was found to be 100 to 150°C.