The diffusion of fission products from graphite

The emission of fission products embedded in various graphites by fission recoil has been studied by the post-irradiation annealing technique over the temperature range 500° C to 1400° C. Detailed studies of the emission of Xe¹³³ showed the release to be in two parts: an initial rapid release lasting for about 16 hours and a slow diffusion process. The fraction evolved by the initial process increases with the temperature of annealing. Values are given to define its magnitude and some discussion of its origin is given. The latter portion of the release is described by a diffusion coefficient obtained from the application of Fick's law of diffusion to the release. Activation energies for the diffusion of Xe¹³³ have been calculated and the relation of the diffusion coefficients to graphite structure is discussed.     

Numerical method for the containment of fission products at a hypothetical accident in an underground nuclear power plant

As one means to expand the siting of nuclear power plants, construction of underground plants is now under study. An underground nuclear power plant has the feature that ground surrounding the underground cavity can contain the fission products of a hypothetical accident.If it is assumed that in a hypothetical reactor accident the cooling system loses its capacity wholly or partially, and gas containing fission products is emitted into the underground cavity. As a result, temperature, gas concentration and gas pressure in the cavity increase and it can be supposed that the gas leaks up to the surface through the ground, and that ground-water contains and carries fission products. The present paper numerically simulates a course of movement as mentioned above by the finite element method and gives the underground containment effect for fission products from a hypothetical accident.     

Retention factors for fission products from sodium tests to simulate a severe LMFBR accident

In case of an LMFBR whole-core accident, fuel and fission products may be released instantaneously from the HCDA bubble through vessel head leaks or delayed from hot or boiling sodium pools after vessel melt-through. It is necessary to investigate the radiological source terms for both scenarios. In the paper we present and discuss retention factors for simulated fission products and fuel from FAUST under-sodium rupture disk tests on the instantaneous source term and from NALA tests with hot and boiling sodium pools on the delayed source term.     

Investigation of the partitioning of fission products between the metallic and oxide phases of melt in a VVER-1000 vessel during a serious accident

The results of investigations of the interaction of U-Zr-B-C-O melt with steel, performed as part of the OECD MASCA program, are presented. It is established that the presence of Mo, Ru, Sr, Ba, Ce, and La in the melt does not qualitatively affect the interaction with structural steel and the character of the stratification of the melt in the reactor vessel. The partition factors of the fission products between the oxide and metallic phases are determined as a function of the oxidation of the melt, the ratio U/Zr, the composition of the structural steel, and the temperature. __________ Translated from Atomnaya énergiya, Vol. 105, No. 1, pp. 3–7, July, 2008.     

Release of fission products from the reactor fuel of the bilibinsk nuclear power plant under conditions which simulate a severe accident

Power Physics Institute (FÉI). Translated from Atomnaya Énergiya, Vol. 74, No. 5, pp. 416–421, May, 1993.     

Evaluation of plutonium oxide aerosol release from an LMFBR in a hypothetical accident

This work was undertaken to prepare a computer code for the hazard evaluation of plutonium oxide aerosol released to the atmosphere in the event of a hypothetical accident in a 50 MW(th) scale LMFBR. The reactor building structure consists of semi-double containments as follows: the primary containment has a large volume in comparison with the secondary annular containment in which a part is connected to the atmosphere through an emergency filter system. Sodium oxide aerosol containing PuO₂---UO₂ fuel, fission products and structural steel agglomerates quickly by coagulation due to its high concentration. Simultaneously, the aerosol concentration decreases due to settling, plating and thermophoresis. Using the present code, the amount of PuO₂ aerosol leakage to the atmosphere was evaluated.     

The influence of the chemical state of the fission products on the filter properties of the graphite components

Results from the experimental and theoretical studies on the transport and sorption behaviour of Cs in graphite with and without oxygen impurities and in the case of HTR-relevant accident conditions are presented and compared. The transport mechanisms of Cs in graphite are discussed on the basis of the calculations carried out with the PATRAS-CORE computer program and the experimentally determined effects dealt with in detail theoretically. Finally, the consequences arising for the safety of an HTR plant are analyzed.     

Behaviour of fission products under severe PWR accident conditions. The VERCORS experimental programme—Part 3: Release of low-volatile fission products and actinides

The VERCORS analytical programme consisted of a series of tests carried out on irradiated PWR fuel samples. The tests - funded jointly by EDF and IRSN - were carried out by the Commissariat à l’Energie Atomique (CEA) at their Grenoble site. They were performed in a hot cell belonging to the Active Materials Analysis Laboratory (LAMA). The general outline of the programme was set out in a first article (of a series of 3), which described the different levels of fission products (FP) volatility and their characteristics. This led to a classification into five main categories of volatility and/or behaviour: (1) Volatile FP including fission gases, iodine, caesium, antimony, tellurium, cadmium, rubidium and silver; (2) Semi-volatile FP, a category made up of molybdenum, rhodium, barium, palladium and technetium; (3) Low-volatile FP comprising ruthenium, cerium, strontium, yttrium, europium, niobium and lanthanum with generally low but significant release; (4) Non-volatile FP including zirconium, neodymium and praseodymium; and lastly (5) Actinides which group together uranium, plutonium, neptunium, americium and curium. The specific behaviour of fission gases and volatile FP is dealt with in the second article, which also includes the specific characteristics of volatile FP regarding transport. The main variables (i.e. temperature, which is the main variable at least until loss of sample geometry, oxidising-reducing conditions, burn-up, interactions with the cladding and/or the structural components, the nature of the fuel, and finally the state of the fuel) affecting the kinetics and/or the released fraction of these same FP could also be identified.This final article represents the Third Part of the series. It concerns the release of actinides and less volatile FP, in keeping with the classification by categories previously identified, which are as follows: (1) semi-volatile FP, comprising of Mo, Ba, Rh, Pd, Tc, (2) low-volatile FP, comprising of Sr, Y, Nb, Ru, La, Ce, Eu, (3) non-volatile FP, comprising of Zr, Nd and Pr and lastly (4) actinides. The main parameters favouring their release are also highlighted.     

The release of unstable fission products during variable reactor operating histories

A solution to the diffusion equation describing the release of unstable fission products during irradiation has been obtained which may be employed for consecutive periods at different temperatures and ratings. The equations have been used to evaluate the release of species of different half-lives for three idealized irradiation histories which retain the same time-averaged irradiation conditions. The results illustrate the wide spectrum of release characteristics displayed by short-, medium- and long-lived species for these cases. It is concluded that the employment of the present analysis is essential for an accurate assessment of radiological hazard.     

Numerical interpretation of the 8 experiment simulating a hypothetical core disruptive accident

In the case of a hypothetical core disruptive accident (HCDA) in a liquid metal fast breeder reactor (LMFBR), it is assumed that the core of the nuclear reactor has melted partially and that the chemical interaction between the molten fuel and the liquid sodium has created a high-pressure gas bubble in the core. The violent expansion of this bubble loads and deforms the reactor vessel, thus endangering the safety of the nuclear plant. The experimental test 8 simulates the explosive phenomenon in a mock-up included in a flexible vessel with a flexible roof. This paper presents a numerical simulation of the test and a comparison of the computed results with the experimental results and previous numerical ones.     

Experimental determination of the leakage of solid fission products at a temperature simulating an accident at a nuclear power plant

Institute of Nuclear Reactors, Russian Scientific Center "Kurchatovskii Institut." Translated from Atomnaya Énergiya, Vol. 75, No. 5, pp. 363-367, November, 1993.     

Three-dimensional behaviors of the hydrogen and steam in the APR1400 containment during a hypothetical loss of feed water accident

During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by an active reaction of the fuel-cladding and the steam in the reactor pressure vessel and released with the steam into the containment. In order to mitigate hydrogen hazards which could possibly occur in the NPP containment, a hydrogen mitigation system (HMS) is usually adopted. The design of the next generation NPP (APR1400) developed in Korea specifies that 26 passive autocatalytic recombiners and 10 igniters should be installed in the containment for a hydrogen mitigation. In this study, an analysis of the hydrogen and steam behavior during a total loss of feed water (LOFW) accident in the APR1400 containment has been conducted by using the computational fluid dynamics (CFD) code GASFLOW. During the accident, a huge amount of hot water, steam, and hydrogen is released into the in-containment refueling water storage tank (IRWST). The current design of the APR1400 includes flap-type openings at the IRWST vents which operate depending on the pressure difference between the inside and outside of the IRWST. It was found from this study that the flaps strongly affect the flow structure of the steam and hydrogen in the containment. The possibilities of a flame acceleration and a transition from deflagration to detonation (DDT) were evaluated by using the Sigma–Lambda criteria. Numerical results indicate that the DDT possibility was heavily reduced in the IRWST compartment by the effects of the flaps during the LOFW accident.     

The management of gaseous wastes from reprocessing containing volatile fission products

The aim of the paper is to present the many factors important in the management of volatile fission products from reprocessing of radioactive fuels. Emissions of ¹³¹I (half-life—8 days) and Xe radioisotopes (half-lives up to 12 days) are controlled by cooling time to negligible levels, leaving ³H, ¹⁴C, ⁸⁵Kr and ¹²⁹I which are very diverse species. These isotopes when discharged to the atmosphere cause exposure at very low levels to very large populations due to combination of half life and dispersibility. Methods for isolation from reprocessing effluents are described, also proposed immobilization and disposal modes are examined in terms of relative merits of discharge/storage/disposal of the species. Comparisons of natural and nuclear cycle productions and cost-benefit factors in choice of control options as indicated by the many contemporary studies are reviewed.     

低温冷阱法分离裂变产物中138Cs

为将短寿命核素138Cs从裂变产物中分离出来,考虑从其母核放射性惰性气体核素(Xe)开始分离.利用不同气体的饱和蒸气压的不同,设计加工了一套低温冷阱分离装置来实现惰性气体核素的分离.结合气体反冲传输技术,该装置成功地应用于从复杂裂变产物中将138Cs与Kr、Rb等干扰核素分离,大部分Υ核素的去污系数达到104以上.     

The release of fission products from nuclear fuel during irradiation by both lattice and grain boundary diffusion

An analysis has been made to describe the steady state release of radioactive fission products from nuclear fuel during irradiation, by both lattice and grain-boundary diffusion. The analysis is appropriate for calculating the release of volatile fission products from regions of fuel pins prior to the development of gross interlinked porosity. In general terms, the dependence of fractional release on decay constant, temperature and solute atom-grain boundary interaction energy has been investigated. The analysis is applied to a specific irradiation experiment where the release of the rare gases, ¹³³I, ¹³¹I, and ¹³²Te were measured from both small and very large grain UO₂ samples.     

Effect analysis of the intentional depressurization on fission product behavior during TMLB' severe accident

It has been found that the pressure in the reactor coolant system (RCS) remains high in some severe accident sequences at the time of reactor vessel failure, with the risk of causing direct containment heating (DCH).Intentional depressurization is an effective accident management strategy to prevent DCH or to mitigate its consequences. Fission product behavior is affected by intentional depressurization, especially for inert gas and volatile fission product. Because the pressurizer power-operated relief valves (PORVs) are latched open, fission product will transport into the containment directly. This may cause larger radiological consequences in containment before reactor vessel failure. Four cases are selected, including the TMLB' base case and the opening one, two and three pressurizer PORVs. The results show that inert gas transports into containment more quickly when opening one and two PORVs,but more slowly when opening three PORVs; more volatile fission product deposit in containment and less in reactor coolant system (RCS) for intentional depressurization cases. When opening one PORV, the phenomenon of revaporization is strong in the RCS.     

The restraining effect of austenitic cladding on the extension of a three-dimensional crack into the wall of a water-cooled nuclear reactor pressure vessel during a hypothetical overcooling accident

A very simple analysis gives an upper bound to the restraining effect of the austenitic cladding on the stress intensity for the extension, at the deepest point, of a three-dimensional under-clad crack into the wall of a water-cooled nuclear reactor pressure vessel during a hypothetical overcooling accident. By comparing the upper bound results with results for a two-dimensional crack, it is concluded that cladding restraint is unlikely to provide a significant reduction in the stress intensity at the deepest point. This viewpoint is confirmed by additional results for a special case, namely that of a semi-circular under-clad crack.     

The measurement of the radioactivity of short-lived gases in a mixture of fission products

The article describes a procedure for measuring the activity of short-lived isotopes of inert gases in a mixture of fission products of U²³⁵. The daughter products of the gases are selectively accumulated in filters set up in a flow-through system and are measured by a radiometric unit. A sensitivity of the order of 10^–15 grams of inert gas is obtained. The procedure is intended for the investigation of the discharge of gaseous fission fragments from fuel elements.The author takes this opportunity to express his gratitude to V. M. Makurin and S. N. Stepanov for their participation in the conduction of the experiments.