Survey of dust production in pebble bed reactor cores

Graphite dust produced via mechanical wear from the pebbles in a pebble bed reactor is an area of concern for licensing. Both the German pebble bed reactors produced graphite dust that contained activated elements. These activation products constitute an additional source term of radiation and must be taken under consideration during the conduct of accident analysis of the design. This paper discusses the available literature on graphite dust production and measurements in pebble bed reactors. Limited data is available on the graphite dust produced from the AVR and THTR-300 pebble bed reactors. Experiments that have been performed on wear of graphite in pebble-bed-like conditions are reviewed. The calculation of contact forces, which are a key driving mechanism for dust in the reactor, are also included. In addition, prior graphite dust predictions are examined, and future areas of research are identified.     

An investigation for usage of graphite powder-helium suspension as reactor coolant in a HTGR

This study's aim is to perform a numerical experiment to examine the enhancement of heat transfer in high-temperature gas-cooled reactors (HTGR) by using gas-particulate (graphite powder-helium) suspension as reactor coolant. At very high temperatures, the radiative heat can be profoundly improved compared to gas alone because of the large absorptivity of the cloud of fine particles. For every high-temperature flow situation where radiation is a significant mode of heat transfer, there exists an optimum loading ratio (mass of graphite to that of He) and particle size for which maximum heat-transfer rates are achieved. The study uses typical HTGR data to examine the heat-transfer phenomena and determine the mass of graphite, optimum particle size, and the influence of other parameters which will result in maximum heat transfer.     

Application of radiation chemistry research to the operation of nuclear reactors

The theme of this review is the application of radiation chemistry research to improve the operating efficiency of nuclear reactors. The intense radiation fields in reactor cores produce a hostile environment for incore materials; this report describes how recent research helped overcome the chemistry problems caused by the radiation.Examples discussed are the inhibition of graphite moderator corrosion and prevention of carbon deposition in gas-cooled reactors, suppression of radiolysis of the cooling water in concrete pressure vessels, hydrogen formation following a loss of coolant accident in a PWR and improving the stability of decontamination reagents for water reactors.     

Examination of the radiation state of the graphite masonry of commercial reactors

The results of an examination of the radiation state of the graphite masonry in three commercial uranium-graphite reactors at the Mayak Industrial Association are presented. On the basis of these results, conclusions are drawn about the nuclear safety of the masonry and radiation licenses are composed. The comprehensive radiation examination made it possible to determine the level, composition, and distribution of radioactive contamination of the masonry and the level and distribution of neutron and γ radiation and to predict the variation of the radionuclide activity in the graphite as a function of the holding time. These data are required to make design decisions about further reactor decommissioning stages. __________ Translated from Atomnaya énergiya, Vol. 102, No. 2, pp. 113–116, February, 2007.     

Computational Estimates of the Radiation Characteristics of Irradiated Graphite after Final Shutdown of a Nuclear Power Plant with an RBMK Reactor

A method of calculating the radiation characteristics of irradiated graphite masonry of an RBMK reactor is described. The MCNP computer code is used to determine the spatial distribution of the neutron flux density in the interior volume of the graphite, the CHAIN code is used to determine the isotopic composition and the radition characteristics of the irradiated graphite on the basis of the MCNP fluxes.The results of the calculation of the radiation characteristics of graphite from the reactors in the Nos. 2 and 3 units of the Leningrad nuclear power plant and the No. 1 unit of the Chernobyl nuclear power plant are presented and the contribution made by the accident to the flow of fuel mass into the masonry is estimated.     

The fundamental heat conduction solution for a thermal shock on a finite orthotropic cylindrical thin shell

Because graphites have excellent thermal, nuclear, and thermal shock resistance, and actually get stronger as temperature increases to about 4000°F, they are frequently used as components of reactors. The crystal structure of graphite has orthotropic symmetry, and this property must be included in the analysis or design of components using this material. Besides, ribbed steel containment vessels in nuclear reactors can be analyzed as shell structures with orthotropic properties. In this paper, the problem of a thermal shock on an orthotropic thin cylindrical shell is investigated. Its fundamental solution for a thermal shock on a thin orthotropic shell of finite length is obtained for each of various boundary conditions, by methods of transform calculus and by the method of reflections. The materials assumed for use in the numerical examples are grades ATJ and ZTA graphite, which can be considered as orthotropic media. Curves for heat storage and temperature distribution in a thin cylindrical shell of finite length are plotted as functions of time. They exhibit the effects of thermally orthotropic property of the material upon the heat dissipation in the shell.     

Air ingress benchmarking with computational fluid dynamics analysis

The air ingress accident is a complicated accident scenario that may limit the deployment of high-temperature gas reactors. The complexity of this accident scenario is compounded by multiple physical phenomena that are involved in the air ingress event. These include diffusion, natural circulation, and complex chemical reactions with graphite and oxygen. In an attempt to better understand the phenomenon, the FLUENT-6 computational fluid dynamics code was used to assess two air ingress experiments. The first was the Japanese series of tests performed in the early 1990s by Takeda and Hishida. These separate effects tests were conducted to understand and model a multi-component experiment in which all three processes were included with the introduction of air in a heated graphite column. MIT used the FLUENT code to benchmark these series of tests with quite good results. These tests are generically applicable to prismatic reactors and the lower reflector regions of pebble-bed reactors. The second series of tests were performed at the NACOK facility for pebble bed reactors as reported by Kuhlmann [Kuhlmann, M.B., 1999. Experiments to investigate flow transfer and graphite corrosion in case of air ingress accidents in a high-temperature reactor]. These tests were aimed at understanding natural circulation of pebble bed reactors by simulating hot and cold legs of these reactors. The FLUENT code was also successfully used to simulate these tests. The results of these benchmarks and the findings will be presented.     

The development of a stress analysis code for nuclear graphite components in gas-cooled reactors

Most of the UK nuclear power reactors are gas-cooled and graphite moderated. As well as acting as a moderator the graphite also acts as a structural component providing channels for the coolant gas and control rods. For this reason the structural integrity assessments of nuclear graphite components is an essential element of reactor design. In order to perform graphite component stress analysis, the definition of the constitutive equation relating stress and strain for irradiated graphite is required. Apart from the usual elastic and thermal strains, irradiated graphite components are subject to additional strains due to fast neutron irradiation and radiolytic oxidation. In this paper a material model for nuclear graphite is presented along with an example of a stress analysis of a nuclear graphite moderator brick subject to both fast neutron irradiation and radiolytic oxidation.     

Above-ground HTGR design for passive decay heat removal

The design of a small high-temperature gas-cooled reactor (HTGR) for passive decay heat removal which could be located deeply underground was proposed previously. In the present work, analogue design analyses of passive decay heat removal for an above-ground long-life small prismatic HTGR was carried out to obtain the conditions for successful decay heat removal by radiation and conduction inside the reactor building, and by radiation and natural cooling by air at the outer surface of the reactor building. Sensitivity analysis of the peak temperatures of both the core and the reactor building after reactor shutdown was performed by changing the physical characteristics of the reactor regions. Enlarging the reactor building was found to be an effective way to reduce the peak reactor building temperature to within its design limit. By using the obtained condition for design parameters, the appropriate sizes of reactor core and reactor building were evaluated for some reactors. Consequently, criticality and burnup analyses for the proposed reactors were performed to confirm the possibility of designing a long-life core for the core size and reactor power which meet the condition of removing decay heat successfully. Using our design, all the reactors with 20 wt% uranium enrichment could be critical for over nine years.     

Recombination Effect as Component of Residual Defect in Silicon Surface Barrier Detector

The thermal conductivity of graphite components used as in-core components in high-temperature gascooled reactors (HTGRs) is reduced by neutron irradiation during reactor operation. The reduction in thermal conductivity is expected to be reversed by thermal annealing when the irradiated graphite component is heated above its original irradiation temperature. In this study, to develop an evaluation model for the thermal annealing effect on the thermal conductivity of IG-110 graphite for the HTGRs, the thermal annealing effect evaluated quantitatively at irradiation temperatures of up to 1,200°C and neutron fluences of up to 1.5 dpa. Moreover, the thermal conductivity of IG-110 graphite was calculated by using a modified thermal resistance model considering the thermal annealing effect. The following results were obtained. (1) The thermal annealing effect on the thermal conductivity of IG-110 graphite could be evaluated quantitatively and a thermal annealing model was developed based on the experimental results at irradiation temperatures of up to 1,200°C and neutron fluences of up to 1.5 dpa. (2) The thermal conductivities of IG-110 graphite calculated by using the modified thermal resistance model considering the thermal annealing effect showed good agreement with experimental measurements. This study has shown that it is possible to evaluate the annealed thermal conductivity of IG-110 graphite by using the modified thermal resistance model at irradiation temperatures of 550–1150°C and irradiation fluences of up to 1.5 dpa.     

反应堆气态流出物中放射性微粒物的监测和数据处理

反应堆等其他重要核设施所产生的气载放射性微粒物(P)、放射性碘(I)和放射性气体(G),通常简称为PIG.这类气载放射性物质,不仅存在于场所空气中,是场所空气的重要放射性污染源;也存在于核设施各类排放空气流中,是烟囱排放的重要监测对象.对于PIG的监测,通常是智能化的"在线"式自动监测,用采样与测量同时进行的累积式监测方法.对这种监测方法给以简单介绍后,主要对这类监测中放射性微粒物的数据处理方法,也即对如何正确报告放射性微粒物的监测结果的方法给予了研究和讨论.弄清这些问题,不仅是PIG监测仪器研制中正确设计运作模式和数据处理软件的基础,也是检测或验证PIG监测系统是否正确运行的基础.     

HTR-PM两根一回路连接管断裂的进气事故分析

进气事故是模块式高温气冷堆关注的超设计基准事故之一,石墨氧化腐蚀反应可能导致反射层结构强度减弱、燃料元件完整性和包容裂变产物能力被破坏,以及产生可燃气体等较严重后果。进气事故的分析研究对进一步掌握高温气冷堆的事故特性以及提高反应堆的安全设计具有重要意义。本文基于200MWe球床模块式高温气冷堆示范工程（HTR-PM）的初步设计,假设与一回路压力边界上、下相连的燃料元件进料管和卸料管同时发生断裂,从而形成烟囱效应并导致空气进入堆芯,利用高温气冷堆专用系统分析软件TINTE对自然循环建立及后续的进气腐蚀过程进行了研究,分析了自然循环流量、堆内石墨腐蚀速率、舱室氧气消耗量、燃料元件温度等关键参数的变化。结果表明,即使考虑腐蚀反应的不均匀性,事故后约60h时才会出现首个燃料包覆颗粒裸露现象,燃料元件最高温度峰值低于1620℃的设计限值,保持完好的燃料包覆颗粒仍具有包容放射性裂变产物的能力。同时,如果在相应的时间内采取措施切断进气源,使石墨腐蚀反应不能继续发展,将不会对反应堆的安全造成严重的影响。     

Prediction of fuel channel—graphite gas-gap behaviour in RBMK reactors

To maintain thermal contact between the fuel assembly and the graphite moderator, RBMK design reactors employ graphite split rings, which are alternatively tight on the pressure tube or tight on the graphite brick central bore. The split in the graphite rings allows a helium/nitrogen gas mixture to flow up the fuel channel. This prevents oxidation of the graphite and can be sampled to detect pressure tube leaks. The initial clearance between the rings and pressure tube or graphite brick is approximately 2.7 mm (1.35 mm each side). Due to material property changes of the pressure tubes and graphite during operation of the reactor, the size of the clearance between the rings and the pressure tube/brick, called the “gas-gap”, varies. Closure of these gaps has been identified as a possible safety case issue by reactor designers and by independent reviews carried out as part of TACIS reviews and as part of the Ignalina Safety Analysis Report. The reasons for this are that gas-gap closure would cause the pressure tube to be tightly gripped by the graphite bricks via the split rings, which could lead to:

• Extra loading on the upper pressure tube zirconium/steel transition joint, particularly during shut down and emergency transients.

• Splitting of the graphite brick, leading to loss of thermal contact between the pressure tube and graphite. As approximately 5.6% of the heat in graphite-moderated reactor is generated within the moderator through neutron and gamma-heating, loss of thermal contact would result in higher graphite temperatures, accelerating the rate of graphite expansion and hence increasing the loading of the core radial restraint.

• Graphite debris may become lodged in inter-brick gaps, leading to increased axial pressure tube loading during shut down and emergency transients.

The authors have carried out deterministic assessments based on the Ignalina RBMK-1500 reactors in Lithuania, modelling the behaviour of the graphite under irradiation and have predicted graphite bore diameter changes that are in good agreement with the measurements of graphite bore diameters taken at Ignalina Nuclear Power Plant (NPP). A probabilistic model has been developed using the actual results of the deterministic calculations with non-linear graphite behaviour. Statistical analysis of the measurements of tube and graphite diameters taken from Units 1 and 2 at Ignalina NPP has been carried out. Further work has been carried out to try to determine the uncertainty inherent in the predictions of the gas-gap closure from the calculations. The overall objective of the studies is to aid prediction of the gas-gap closure process, and help to identify a suitable monitoring strategy for gas-gap closure that could be used for any RBMK reactor.     

Analyzing the effect of displacement rate on radiation-induced segregation in 304 and 316 stainless steels by examining irradiated EBR-II components and samples irradiated with protons

Recent studies have indicated that, at temperatures relevant to fast reactors and light water reactors, void swelling in austenitic alloys progresses more rapidly when the radiation dose rate is lower. A similar dependency between radiation-induced segregation (RIS) and dose rate is theoretically predicted for pure materials and might also be true in complex engineering alloys. Radiation-induced segregation was measured on 304 and 316 stainless steel, irradiated in the EBR-II reactor at temperatures near 375 °C, to determine if the segregation is a strong function of damage rate. The data taken from samples irradiated in EBR-II is also compared to RIS data generated using proton radiation. Although the operational histories of the reactor irradiated samples are complex, making definitive conclusions difficult, the preponderance of the evidence indicates that radiation-induced segregation in 304 and 316 stainless steels is greater at lower displacement rate.     

The relationship between irradiation induced dimensional change and the coefficient of thermal expansion: a modified Simmons relationship

In the 1960s, a theoretical relationship between the dimensional changes and the coefficient of thermal expansion of irradiated graphite was derived by J.H.W. Simmons. The theory was shown to be comparable with experimental observations at low irradiation doses, but shown to diverge at higher irradiation doses. However, various modified versions of this theory have been used as the foundation of design and life prediction calculations for graphite-moderated reactors.This paper re-examines the Simmons relationship, summarising its derivation and assumptions. The relationship was then modified to incorporate the high dose, high strain changes that were assumed to be represented in the changes in Young’s modulus with irradiation dose. By scrutinising the behaviour of finite element analyses, it was possible to use a modified Simmons relationship to predict the dimensional changes of an isotropic and anisotropic graphite to high irradiation doses.These issues are important to present high-temperature reactors (HTRs) as the life of HTR graphite components is dependent upon their dimensional change behaviour. A greater understanding of this behaviour will help in the selection and development of graphite materials.     

核级石墨氧化腐蚀模型研究

高温气冷堆内应用到大量核级石墨材料,对其长期氧化腐蚀行为进行研究至关重要。文章建立了综合考虑石墨内部孔隙率变化及失重率影响的石墨氧化模型,对气体在石墨内部的瞬时氧化腐蚀情况进行了模拟计算。提出氧化深度的概念,研究发现反应温度越高,反应气体在石墨内部的氧化深度越小;并与实验结果及其他模型的计算结果进行了对比,验证了模型的有效性。     

Modelling of the spatial distribution of the induced activities in the RBMK-1500 reactor graphite blocks and rings/sleeves

There is one nuclear power plant (NPP) in Lithuania – the Ignalina NPP – which is under decommissioning now. The Ignalina NPP has two units with RBMK-1500 reactors, which are the most powerful and the most advanced versions of RBMK-type reactor design. Unit 1 of the Ignalina NPP was shut down at the end of 2004 and Unit 2 was shut down at the end of 2009. RBMK is a water-cooled graphite-moderated channel-type power reactor and the decommissioning of these reactors faces specific challenges for proper characterisation and disposal of irradiated reactor graphite.Apart from radiological inventory, the spatial distribution of radionuclides in the reactor graphite is also very important because it could indicate the possibilities for decontamination/treatment of the irradiated graphite. This is important for consideration of the near surface disposal option for irradiated graphite, as without treatment it usually does not meet the waste acceptance criteria.Based on that, the work presented in this paper is focused on the modelling of the induced activity spatial distribution in the Ignalina NPP RBMK-1500 reactor graphite components: blocks and rings/sleeves. The modelling was performed with MCNP and SCALE computer codes and consisted of two mains stages: modelling of the neutron flux in the reactor graphite components, and then modelling of the neutron activation in them using the already modelled neutron flux. In such a way, the spatial induced activity distribution in the analysed reactor components was obtained. Modelling results show that the thermal neutron flux is more intensive in the outer radial regions of the graphite components and this, in general, results in higher induced activities there.     

Nuclear graphite friction properties and the influence of friction properties on the pebble bed

Recently, the role of friction and wear in the safety of pebble bed reactors was reconsidered. The friction properties of graphite were analyzed in this paper. In the first part of this paper, we investigated the past research of graphite and analyzed the influence of environment gas, temperature and radiation on friction coefficient. The variation laws of friction coefficient with the effect factors were given, especially in helium environment (including pure helium and HTGR helium). In the second part, the mechanical behavior of pebble bed is discussed while considering the influence of the friction properties of graphite. The results indicated that the fuel elements cannot be crushed failure even though the fuel elements have a large friction coefficient.     

Variants of Aperiodic Pulsed Reactors with Boosted Pulse Parameters

Single (one-section) reactors, containing a neptunium–gallium or uranium–molybdenum alloy core, and coupled two-section systems, consisting of a single pulsed reactor and a driven subcritical assembly, consisting of a uranium–molybdenum alloy or dispersed uranium–graphite material, are studied. Calculations of the neutron and dynamical characteristics of these systems are performed. The information obtained, together with data from previous calculations, made it possible to draw a conclusion about the structure of reactor systems characterized by a short radiation pulse and large-volume cores and cavities for holding specimens. It is shown that a reactor with a single neptunium–gallium alloy core and a coupled cascade-type system consisting of this single driven reactor together with a driven subcritical uranium–graphite assembly have the best pulse parameters.     

Water-ingress analysis for the 200 MWe pebble-bed modular high temperature gas-cooled reactor

Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor adopting steam-turbine cycle, which will cause a positive reactivity introduction, as well as the chemical corrosion of graphite fuel elements and reflector structure material. Besides, increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes and flammable water gas. The analysis of such a kind of important and particular accident is significant to verify the inherent safety characteristics of the modular HTR plants.Based on the preliminary design of the 200 MWe high temperature gas-cooled reactor pebble-bed modular (HTR-PM), the design basis accident of a double-ended guillotine break of one heating tube and the beyond design basis accident of a large break of the main steam collection plate have been analyzed by using TINTE code, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature, the primary loop pressure, the graphite corrosion, the water gas releasing amount, as well as the natural convection influence on the condition of failing to close the blower flaps, have been studied in detail. The calculation results indicate that even under some severe hypothetical postulates, the HTR-PM is able to keep the inherent safeties of the modular high temperature gas-cooled reactor and has a relatively good natural plant response, which will not result in environmental radiation hazard.