Safety study of the coupling of a VHTR with a hydrogen production plant

The present paper deals with specific safety issues resulting from the coupling of a nuclear reactor (very high temperature reactor, VHTR) with a hydrogen production plant (HYPP). The first part is devoted to the safety approach consisting in taking into account the safety standards and rules dedicated to the nuclear facility as well as those dedicated to the process industry. This approach enabled two main families of events to be distinguished: the so-called internal events taking place in the coupling circuit (transients, breaks in pipes and in heat exchangers) and the external events able to threat the integrity of the various equipments (in particular the VHTR containment and emergency cooling system) that could result from accidents in the HYPP. By considering a hydrogen production by means of the iodine/sulfur (IS) process, the consequences of the both families of events aforementioned have been assessed in order to provide an order of magnitude of the effects of the incidents and accidents and also in order to propose safety provisions to mitigate these effects when it is necessary. The study of transients induced by a failure of a part of the HYPP has shown the possibility to keep the part of the HYPP unaffected by the transient under operation by means of an adapted regulation set. Moreover, the time to react in case of transfer of corrosive products in the VHTR containment has been assessed as well as the thermohydraulic loading that would experience the coupling pipes in case of very fast uncoupling of the facilities aiming at avoiding an excessive pressurization of the VHTR containment. Regarding the external events, by applying a method used in the process industries, the bounding representative scenarios have been identified on the basis of their consequences but also on the basis of their occurrence frequency. The consequences of the selected bounding scenarios, calculated taking into the source-term, the atmospheric dispersion and the pressure and toxic effects induced respectively by a hydrogen unconfined vapour cloud explosion (UVCE) and a sulfur dioxide release have been assessed. The resulting safety distance of about 100 m for the UVCE is fairly acceptable in terms of performance (head loss and thermal loss) of the coupling system. However, the longer safety distance (about 1.5 km) calculated for a SO₂ release implies to foresee a long distance to settle the control room of the site or to foresee provisions able to stop very fast the SO₂ leak.     

Safety assessment aspects of type B(U) packages containing wet intermediate level waste

Abstract

In Germany, the mechanical and thermal safety assessment of approved packages for the transport of RAM is carried out by BAM as the competent authority according to the International Atomic Energy Agency regulations. BAM was involved in several approval procedures with ductile cast iron containers containing wet intermediate level waste. These contents, which are not dried, only drained, consist of saturated ion exchange resin and a small amount of free water. Compared to the safety assessment of packages with dry content, attention must be paid to some more specific points. The physical and chemical compatibility of the content itself and of the content with materials of the package must be shown. From the mechanical resistance point of view, the package has to withstand the forces resulting from the freezing liquid. The most interesting point, however, is the pressure build-up inside the package due to vapourisation. This could be caused by radiolysis of the liquid and must be taken into account for the storage period. The paper deals primarily with the pressure build-up inside the package caused by the regulatory thermal test (30 min at 800°C) as part of the cumulative test scenario under accident conditions of transport. To determine the pressure, the temperature distribution in the content must be calculated for the whole period from the beginning of the thermal test until cooling down. In this case, calculating the temperature distribution requires, besides the consideration of conduction and heat radiation, consideration of evaporation and condensation including the associated processes of transport.     

Evaluation study on properties of a macroporous silica-based CMPO extraction resin to be used forminor actinides separation from high level liquid waste

Basic properties of a silica-based octyl(phenyl)-N,N-diisobutylcarbamoyl-methylphosphine oxide (CMPO) extraction resin (CMPO/SiO2-P) was investigated.Adsorption behavior for some rare earth elements (RE) which are constituents of high level liquid waste (HLLW) and the long-term stability of the extraction resin in nitric acid solution were examined.The CMPO extraction resin was significantly stable in 3 mol·L?1 HNO3 solution at 50oC.Furthermore,the RE(III) were efficiently separated from non-adsorptive fission product (FP) elements such as Sr(II) in a column experiment using a highly nitric acid solution.The separation behaviors of the elements are considered to result from the difference in their adsorption and elution selectivity based on the complex formation with CMPO.There was no strong dependency of RE(III) separation efficiency on feed solution flow rate.Only from the perspectives of the acid-resistant behavior of CMPO extraction resin and the elution kinetics for the metal ions with the extraction resin,the CMPO extraction resin can be used in the modified MAREC process for HLLW partitioning.     

Dosimetric characteristic of VVR-M reactor with respect to tritium

Translated from Atomnaya Énergiya, Vol. 67, No. 2, pp. 111–113, August, 1989.     

Experience with piping in German NPPs with respect to ageing-related aspects

Thirteen nuclear power plants (NPPs) with pressurized water reactors (PWRs) and six plants with boiling water reactors (BWRs) are currently in operation in Germany. For almost 25 years, GRS has been systematically evaluating the operating experience of these plants. In this paper, the operating experience relating to piping damage in safety-relevant systems of German plants with light water reactors (LWRs) is evaluated with respect to ageing-related effects. The experience with actions taken against piping degradation is illustrated by examples. The results of the evaluation confirm the conservativeness of the safety concept chosen for the design of German NPPs with LWRs, as well as the effectiveness of ageing management.     

Experimental strategy of fuel performance testing with respect to PCI

Basic metallurgical investigations have revealed that stress corrosion cracking of Zircaloy tubes requires critical stress and iodine concentration for a minimum of time. Based on this observation KWU has structured its experimental strategy on the RSST approach. This means the evaluation of a defect-free power Range below a PCI defect threshold and a defect-free LHGR Step (naturally beyond the threshold), a limited Speed of power increase if both limits are exceeded, and a minimum Time for any mechanism to become effective.KWU has initiated a large ramp test program at HFR, Petten, the results of which are backed by the participation in international ramp test programs at Studsvik. The first target was the determination of the failure threshold as a function of burnup for the different fuel rod designs. Then, the allowable safe speed for passing the defect threshold was investigated. The defect-free range was confirmed by power reactor experiments on a broad statistical basis with rods of original length. In forthcoming experiments in Petten the verification of safe steps shall be a point of main priority. The permissible time above the thresholds may be controlled purely by the crack nucleation. In the Studsvik Demo Ramp II Program this is a point of special consideration.Detailed PIE results show that these performance limits can well be interpreted by observable phenomena like grain growth, fission product redistribution, fission gas release etc.     

中低放固体放射性废物处置中的α废物在线检测及位置判定方法初探

根据在核废料回取分类、压缩减容的处置工艺及部分专用设备开发等方面的研究和初步实践,提出了利用现代测试技术和自动控制原理对a型废物进行在线探测、位置判定、目标成像、定性分析、比活度粗测的方法。     

Air exposure and sample storage time influence on hydrogen release from tungsten

In investigations of hydrogen retention in first wall components the influence of the conditions of the implanted target storage prior to analysis and the storage time is often neglected. Therefore we have performed a dedicated set of experiments. The release of hydrogen from samples exposed to ambient air after irradiation was compared to samples kept in vacuum. For air exposed samples significant amounts of HDO and D₂O are detected during TDS. Additional experiments have shown that heavy water is formed by recombination of releasing D and H atoms with O on the W surface. This water formation can alter hydrogen retention results significantly, in particular - for low retention cases. In addition to the influence of ambient air exposure also the influence of storage time in vacuum was investigated. After implantation at 300 K the samples were stored in vacuum for up to 1 week during which the retained amount decreased significantly. The subsequently measured TDS spectra showed that D was lost from both the high and low energy peaks during storage at ambient temperature of ∼300 K. An attempt to simulate this release from both peaks during room temperature storage by TMAP 7 calculations showed that this effect cannot be explained by conventional diffusion/trapping models.     

碱矿渣水泥固化核废物研究现状

着重从碱矿渣的结构特征、碱矿渣水泥水化产物及碱矿渣水泥固化体的性能等方面论述了碱矿渣水泥固化核废物的研究现状,并初步分析了固化机理,展望了其应用前景。     

Simulation study of carbon impurity dynamics on tungsten surfaces exposed to hydrogen ions

The depth profile of C impurity deposited on a W target exposed to H⁺ and C⁺ impurities at a concentration of C: 0.8% has been calculated in terms of segregation, diffusion and chemical erosion. For the segregation, the Gibbsian model has been used. For the diffusion, a concentration dependent diffusion model (C in WC and/or C) has been utilized. For the chemical erosion, the chemical erosion yield much lower than that for the H-C system has been applied. The calculated depth profiles at 653 K and 913 K are in good agreement with the XPS data. The agreement indicates that there is a significant contribution of segregation, which shifts the maximum C concentration to the top surface in the depth profiles. On the other hand, there are little contributions from diffusion and chemical erosion, which are related closely to formation of WC in the target.     

Design study of lead-bismuth cooled ADS dedicated to nuclear waste transmutation

Research and development on nuclear waste transmutation are being carried out with a special emphasis placed on dedicated accelerator-driven systems at the Japan Atomic Energy Research Institute under the Japanese OMEGA Program. The reference accelerator-driven system design employs eutectic lead-bismuth as spallation target material and coolant. The fuel for the subcritical core is minor-actinide mononitride. The system consists of a 1.5GeV, 14mA proton accelerator and an 800MWt subcritical core with an effective neutron multiplication factor of 0.95. The transmutation rate of minor actinides is approximately 250 kg/y at 80% load factor. The design has salient features that the coolant inventory is large due to the tank-type configuration, the temperature rise through the core is relatively low, and the power conversion is operated on a saturated steam turbine cycle. These features make the plant response to a beam trip slow and much less demanding.     

聚变高温制氢反应堆概念设计研究

在深入分析相关领域研究发展状况的基础上,提出了具有较好技术可行性的聚变高温制氢反应堆概念(称之为FDS-Ⅲ),包括具有先进等离子体物理和技术水平的聚变堆芯、先进高温锂铅包层(HTL)、可减少热流分布密度的"垂直靶板"偏滤器以及相应的功率转换系统。尤其是提出了HTL包层新概念,其特点是选用技术基础相对成熟的低活化铁素体/马氏体钢作结构材料,在锂铅流道中使用可耐高温的多层流道插件,实现约1000℃的出口温度,可应用于制氢。初步性能分析表明FDS-Ⅲ制氢堆及其包层概念具有较好的技术可行性。     

Some aspects on the optimum repository plug design with respect to the escape of gas

BfS is in the progress of developing a closure concept for the repository for low and intermediate level radioactive waste in Morsleben (ERAM). In the course of this work, the optimal design of the plug is currently being evaluated with respect to gas escape and the exchange of potentially contaminated brine through the plug. For the sealing to behave well in the long term, it is important that the gas formation processes do not disrupt the plug or enhance the radionuclide release, e.g. by means of excessive pressure build-up. The object has been to study different scenarios for gas and brine transport for two alternative plug concepts, by using the multi-phase flow model TOUGH. Rock convergence due to creep has been included in the modelling. The results of the calculations indicate that the closure concepts restrict he exchange of brine and allow escape of gas; an excavation-damaged zone around tunnels is a potential pathway for gas and brine, and the effect of the rock convergence is small. The results also indicate that a very dense plug results in excessive pressurisation of the repository, whereas a permeable plug results in an increased exchange of brine.