The utilization of thorium in Generation IV reactors

The main goal of this paper is to show how thorium, as an alternative nuclear fuel, could be applied as fuel in a Generation IV reactor. The paper focuses on the multiplication factor, the produced ²³³U and delayed neutron fraction in infinite lattice models. For the investigations, simplified models of a fuel assembly of five design types of the six reactor concepts were elaborated. The MSR reactor type is out of scope of this paper due to the fact that it is designed for the utilization of thorium. Although the fissile isotope content was not increased to compensate the thorium caused multiplication factor decrease, the burnup calculations suggest that the designs of ESFR (European Sodium-Cooled Fast Reactor) and ELSY (European Lead-cooled System) are the most promising types according to the trend of the multiplication factor changes and the amount of produced fissionable ²³³U.     

Neutronics assessment of the use of thorium fuels in current pressurized water reactors

The use of thorium fuel in current PWRs in a once-through fuel cycle is an attractive option due to potential advantages such as high conversion ratio and low minor actinide generation. The current neutronics assessments indicate that the thorium fuel cycle could supplement the current uranium–plutonium fuel cycle to improve operational performance and spent fuel consideration in current PWRs without core and subassembly modifications. Neutronics safety parameters in the PWR cores with the thorium fuels are within the range of current PWRs.The PWR cores with thorium fuels have significantly higher conversion ratios which could enable efficient fuel utilization. Further, it is shown that the use of thorium as a fertile material can reduce minor actinide generation and the radio-toxicity of spent fuels. In considerations related to proliferation resistance, the results of the current analyses show no significant difference between the studied thorium fuels and the standard oxide fuel for the assumed characteristics and burnup levels.     

超临界水冷混合堆快谱区组件物理——热工分析

提出了超临界水冷混合堆快谱区多层燃料组件设计方案.应用MCNP程序为该组件建立计算模型,并进行了相应的物理计算;同时运用子通道分析程序STAFAS对多层燃料组件子通道进行了初步的稳态热工分析.计算结果表明:超临界水冷混合堆快谱区多层燃料组件燃料转换比超过1.0,并且获得负的冷却剂空泡反应性系数;燃料包壳表面最高温度约为595℃,低于设计准则规定的上限值,同时组件各子通道出口冷却剂温度均匀性较好.通过对燃料棒径敏感性分析可知,较大棒径组件燃料转换比较大,但也会导致热通道包壳表面温度峰值升高.     

Plutonium management with thorium-based fuels and inert matrix fuels in thermal reactor systems

The plutonium that is produced by light water reactors worldwide is currently re-used to a limited extent. In the last century, the expected introduction of fast reactors and the associated need for large amounts of plutonium did not take place. The result is that worldwide a stockpile of excess plutonium has formed, which is the dominant contributor to the radiotoxicity of spent nuclear fuel for storage times from 10² to 10⁵ years. One option to reduce or stabilize the plutonium stockpile is to utilize this plutonium in advanced fuel types, such as thorium-based and inert matrix fuels. Because these fuels do not contain uranium, the plutonium consumption rate is very high. In this paper, the status of the fuel research and some recent developments are given.     

The utilization of thorium for long-life small thermal reactors without on-site refueling

Thorium cycle has many advantages over uranium cycle in thermal and intermediate spectrum nuclear reactors. In addition to large amount of resources in the world which up to now still not utilized optimally, thorium based thermal reactors may have high internal conversion ratio so that they are very potential to be designed as long-life reactors without on-site refueling based on thermal spectrum cores. In this study preliminary study for application of thorium cycle in some of thermal reactors has been performed.

We applied thorium cycle for small long-life high temperature gas reactors without on-site refueling. Calculation results using SRAC code show that 10 years lifetime without on-site refueling can be achieved with excess reactivity of about 10% dk/k.

The next application of thorium cycle has been employed in long-life small and medium PWR cores without on-site refueling. Relatively high fuel volume fraction is also applied to get relatively hard spectrum, small size, and high internal conversion ratio. In the current study we have been able to reach more than 10 years lifetime without on-site refueling for 20–300 MWth PWR with maximum excess reactivity of a few %dk/k.

The last application of thorium cycle has been employed in long-life BWR cores without on-site refueling. Relatively high fuel volume fraction is applied to get relatively hard spectrum, small size, and high internal conversion ratio. In the current study we have been able to reach more than 10 years lifetime without on-site refueling for 100–600 MWth BWR with maximum excess reactivity of a few %dk/k.     

Reactor physics ideas for large scale utilization of thorium in gas cooled reactors

The physics principles for maximizing the fertile to fissile conversion were used in developing reactor concepts for large scale utilization of thorium in thermal and fast reactors (Jagannathan & Pal, 2006; Jagannathan et al., 2008). It is recognized that these principles are very well suited for ‘He’ gas cooled reactors with graphite moderator since both helium gas coolant and the graphite moderator have low neutron absorption characteristics and thus gives better neutron economy. In this paper, these ideas are applied to the High Temperature Test Reactor (HTTR) core of Japan to assess its advantage over the present day gas cooled reactors. HTTR is helium cooled and graphite moderated system. Significant amount of thorium has been loaded in the HTTR core with some minimal changes in the existing core design. The modified design is called HTTR-M core.In the HTTR-M core, the fuel is changed from enriched UO₂ fuel to Pu in ThO₂ fuel. The locations of boron type burnable poison rods within each fuel assembly of HTTR are replaced by one cycle irradiated thoria rods. Also, the B₄C type control assembly around the HTTR core is replaced by fresh seedless thorium assembly. The fertile thoria assembly are scattered uniformly in the HTTR-M core. The equilibrium core of HTTR-M shows very small burnup reactivity swing. The core excess reactivity is ∼18 mk at BOC and reduces to 1 mk at 660 days. It is interesting to note that this small reactivity change is intrinsically achieved by the choice of seed and fertile dimensions and their contents without the use of burnable poison rods or mechanical control rods which are used in HTTR core. The burnup reactivity swing in the latter after using burnable poison is ∼100 mk. The fissile seed inventory ratio (FIR) in a fuel cycle is 0.90 as compared with 0.717 of HTTR core. Since ²³³U is a better fissile nuclide with highest ‘η’ value in thermal range, the above conversion ratio can be regarded as quite good.     

A numerical neutronics benchmark study for inert matrix plutonium fuels in uranium dioxide and mixed plutonium-uranium dioxide environments

A numerical benchmark exercise has been under way for comparing the results of different calculational methods/data sets used for the analysis of light water reactor (LWR) configurations employing Pu inert matrix fuels (IMFs). The first phase of the exercise was devoted to infinite arrays of identical IMF cells. The main feature investigated in the second phase has been the influence of the neutron spectra in UO₂ and MOX cores on individual IMF cells. Phase 3 is concerned with the more realistic situation of an IMF assembly surrounded by UO₂ assemblies. Significant discrepancies have been observed for power peaking effects and delayed neutron parameters in Phase 2. In Phase 3, neutron balance differences for the IMF, particularly at EOL, are found to be significantly larger than were observed in Phase 1.     

Burning of MOX fuels in LWRs; fuel history effects on thermal properties of hull and end piece wastes and the repository performance

The thermal impacts of hull and end piece wastes from the reprocessing of MOX spent fuels burned in LWRs on repository performance were investigated. The heat generation rates in MOX spent fuels and the resulting heat generation rates in hull and end piece wastes change depending on the history of MOX fuels. This history includes the burn-up of UO₂ spent fuels from which the Pu is obtained, the cooling period before reprocessing, the storage period of fresh MOX fuels before being loaded into an LWR, as well as the burn-up of the MOX fuels. The heat generation rates in hull and end piece wastes from the reprocessing of MOX spent fuels with any of those histories are significantly larger than those from UO₂ spent fuels with burn-ups of 45 GWd/THM. If a temperature below 80°C is specified for cement-based materials used in waste packages after disposal, the allowable number of canisters containing compacted hull and end pieces in a package for 45 and 70 GWd-MOX needs to be limited to a value of 0.4–1.6, which is significantly lower than 4.0 for 45 GWd-UO₂.     

Structural investigation of thorium in molten lithium-calcium fluoride mixtures for salt treatment process in molten salt reactor

X-ray absorption fine structure (XAFS) measurements on thorium fluoride in molten lithium-calcium fluoride mixtures and molecular dynamics (MD) simulation of zirconium and yttrium fluoride in molten lithium-calcium fluoride mixtures have been carried out. In the molten state, coordination number of thorium (N_i) and inter ionic distances between thorium and fluorine in the first neighbor (r_i) are nearly constant in all mixtures. However the fluctuation factors (Debye-Waller factor (σ²) and C₃ cumulant) increase until xCaF₂ = 0.17 and decrease by addition of excess CaF₂. It means that the local structure around Th⁴⁺ is disordered until xCaF₂ = 0.17 and stabilized over xCaF₂ = 0.17. The variation of fluctuation factors is related to the number density of F⁻ in ThF₄ mixtures and the stability of local structure around Th⁴⁺ increases with decreasing the number density of F⁻ in ThF₄ mixtures. This tendency is common to those in the ZrF₄ and YF₃ mixtures. However, in the case of YF₃ mixtures, the local structure around Y³⁺ becomes disordered until xCaF₂ = 0.40 and it becomes stabilized by addition of excess CaF₂. The difference between ThF₄ mixtures and YF₃ mixtures is related to the difference of Coulumbic interaction between Th⁴⁺-F⁻ and Y³⁺-F⁻. Therefore, the variation of local structure around cation is related to not only number density of F⁻ in molten salts but also the Coulumbic interaction between cation and anion.     

First-principles energy and vibration spectrum simulations of Cr/V interacting with H in W-based alloy in a fusion reactor

We perform first-principles total energy and vibration spectrum calculations to study the effect of Cr/V on the H formation and diffusion at temperature range 300–2100 K in dilute W–Cr/W–V alloy. Temperature and H chemical potential are two important factors to affect the H formation energy and migrating energy. The H formation energy referring to the static/temperature-dependent H chemical potential decreases/increases with the temperature. At each given temperature, the presence of Cr in W reduces the H formation energy, while the existence of V in W has little effect on the H formation energy. The diffusion energy barrier and pre-exponential factor strongly depend on the temperature and increase with the temperature from 300 to 2100 K. Both Cr and V additions in W has a large consequences on H migrating energy. The energy barriers at any given temperature can be reduced by ～0.05 and 0.10 eV in W–Cr and W–V alloys, respectively. The current study reveals that vibration free energy plays a decisive role in the formation energy and migrating energy of H with the temperature, while the thermal expansion energy has little influence on the H formation energy and migrating energy with the temperature.     

Non destructive determination of PuO2 content in MOX fuel pins for fast reactors using Passive Gamma Scanning

Use of Passive Gamma Scanning for non destructive evaluation of PuO₂ content in mixed oxide (MOX) fuels for fast reactors is demonstrated. Experiments have been carried out on MOX fuel pins for the hybrid core of Fast Breeder Test Reactor having nominal PuO₂ content of 44% and MOX pins having nominal PuO₂ content of 21% for the Prototype Fast Breeder Reactor. A comparison of results obtained using a conventional NaI(Tl) detector and that using a through well shaped detector is also presented.     

基于热堆的钍铀转换过程中232U生成的模拟计算

在钍铀燃料循环过程中生成的232U的衰变子体具有强放射性,对燃料循环具有重要影响。本工作采用ORIGEN2、SCALE5程序,以及基于Bateman方法编写的程序,分析了在不同条件下,热堆中钍反应生成232U的规律。一般情况下,232U主要由232Th的(n,2n)反应链生成,而在中子能谱更软情况下,230Th对232U生成贡献增大;CANDU型重水堆和压水堆的含钍燃料组件的燃耗计算结果表明,铀中232U含量随燃耗深度增加而变大,同时初始230Th/Thtotal大小直接线性影响卸料燃耗时232U/Utotal或232U/233U。     

A Boundary Element-Response Matrix method for criticality diffusion problems in xyz geometry

The Boundary Element-Response Matrix (BERM) method shown in the paper aims to represent an alternative to the Finite Element method in order to solve 3D multigroup diffusion (criticality) problems in xyz geometry. The theory extends the previous work on the diffusion equations in two dimensions and new techniques for the evaluation of the integrals involved in the boundary integral equations, as well as new procedures for solving the resulting linear system, have greatly enhanced the performances of the method. Results show that BERM can achieve an excellent accuracy, still keeping a good computational efficiency.     

The evaluation of the pyrochemistry for the treatment of Gen IV nuclear fuels - Inert matrix chlorination studies in the gas phase or molten chloride salts

The structure of the fuels for the future Gen IV nuclear reactors will be totally different from those of PWR, especially for the GFR concept including a closed cycle. In these reactors, fissile materials (carbides or nitrides of actinides) should be surrounded by an inert matrix. In order to build a reprocessing process scheme, the behavior of the potential inert matrices (silicon carbide, titanium nitride, and zirconium carbide and nitride) was studied by hydro- and pyrometallurgy. This paper deals with the chlorination results at high temperature by pyrometallurgy. For the first time, the reactivity of the matrix towards chlorine gas was assessed in the gas phase. TiN, ZrN and ZrC are very reactive from 400 °C whereas it is necessary to be over 900 °C for SiC to be as fast. In molten chloride melts, the bubbling of chlorine gas is less efficient than in gas phase but it is possible to attack the matrices. Electrochemical methods were also used to dissolve the refractory materials, leading to promising results with TiN, ZrN and ZrC. The massive SiC samples used were not conductive enough to be studied and in this case specific SiC-coated carbon electrodes were used. The key point of these studies was to find a method to separate the matrix compounds from the fissile material in order to link the head to the core of the process (electrochemical separation or liquid-liquid reductive extraction in the case of a pyrochemical reprocessing).     

Method based on isotope ratio mass spectrometry for evaluation of carbon activation in the reactor graphite

The general idea of this work is to introduce an evaluation method to restore the irradiation parameters of graphite or other carbonaceous materials using experimental and modelling results of ¹³C generation in the irradiated material. The method is based on coupling of stable isotope ratio mass spectrometry and computer modelling of the reactor core to evaluate the realistic characteristics of the reactor core such as the neutron fluence in any position of the reactor graphite stack or other graphite constructions.The generation of carbon isotopes ¹³C and ¹⁴C in the irradiated graphite of the RBMK-1500 reactor has been estimated by modelling of the reactor core with computer codes MCNPX and CINDER90. Good agreement of simulated and measured Δ¹³C/¹²C values in graphite of the central part of the reactor core indicates that the neutron flux (1.40 × 10¹⁴ n/cm² s) is modelled accurately in the graphite sleeve of the fuel channel. The simulated activity of ¹⁴C is compared with the one measured by the β spectrometry technique. Results indicate that production of ¹⁴C from ¹⁴N in the RBMK-1500 reactor is considerable and has to be taken into account in order to make proper evaluation of ¹⁴C activity. Measured ¹⁴C specific activity values correspond to 15 ± 4 ppm impurity of ¹⁴N in graphite samples from the RBMK-1500 reactor core.     

一种基于PC/104的便携式建材放射性检测仪

介绍了一种新型的便携式建材放射性检测仪。给出了仪器总提框图,讨论了其工作原理、具体的电路实现和仪器下位机软件框图。仪器以PC／104为核心,结合A／D采集卡和PKD-01等集成芯片．具有电路结构简单．可靠性和稳定性高的特点。