The Direct Internal Recycling concept to simplify the fuel cycle of a fusion power plant

A new concept, the Direct Internal Recycling (DIR) concept, is proposed, which minimizes fuel cycle inventory by adding an additional short-cut between the pumped torus exhaust gas and the fuelling systems. The paper highlights quantitative modelling results derived from a simple fuel cycle spreadsheet which underline the potential benefits that can be achieved by implementation of the DIR concept into a fusion power plant.DIR requires a novel set-up of the torus exhaust pumping system, which replaces the batch-wise and cyclic operated cryogenic pumps by a continuous pumping solution and which offers at the same time an additional integral gas separation function. By that, hydrogen can be removed close to the divertor from all other gases and the main load to the fuel clean-up systems is a smaller, helium-rich gas stream. Candidate DIR relevant pump technology based on liquid metals (vapour diffusion and liquid ring pumps) and metal foils is discussed.     

A neutron poison tritium breeding controller applied to a water cooled fusion reactor model

The generation of tritium in sufficient quantities is an absolute requirement for a next step fusion device such as DEMO due to the scarcity of tritium sources. Although the production of sufficient quantities of tritium will be one of the main challenges for DEMO, within an energy economy featuring several fusion power plants the active control of tritium production may be required in order to manage surplus tritium inventories at power plant sites. The primary reason for controlling the tritium inventory in such an economy would therefore be to minimise the risk and storage costs associated with large quantities of surplus tritium. In order to ensure that enough tritium will be produced in a reactor which contains a solid tritium breeder, over the reactor's lifetime, the tritium breeding rate at the beginning of its lifetime is relatively high and reduces over time. This causes a large surplus tritium inventory to build up until approximately halfway through the lifetime of the blanket, when the inventory begins to decrease. This surplus tritium inventory could exceed several tens of kilograms of tritium, impacting on possible safety and licensing conditions that may exist.This paper describes a possible solution to the surplus tritium inventory problem that involves neutron poison injection into the coolant, which is managed with a tritium breeding controller. A simple PID controller and is used to manage the injection of the neutron absorbing compounds into the water coolant of a stratified blanket model, depending on the difference between the required tritium excess inventory and the measured tritium excess inventory. The compounds effectively reduce the amount of low energy neutrons available to react with lithium compounds, thus reducing the tritium breeding ratio. This controller reduces the amount of tritium being produced at the start of the reactor's lifetime and increases the rate of tritium production towards the end of its lifetime. Thus, a relatively stable tritium production level may be maintained, allowing the control system to minimize the stored tritium with obvious safety benefits. The FATI code (Fusion Activation and Transport Interface) will be used to perform the tritium breeding and controller calculations.     

Tritium dose criteria and radiological impact of a tritium plant

Radiation safetry criteria adopted in Russia (in the former USSR) distinguish five classes of tritium compounds. The lowest permissible tritium concentration in the air is set for insoluble tritium compounds (3.10⁵ times lower than that for HT). Russia's criteria for tritiated radioactive waste are outlined. It is explained why the tritium weighting factor of two is used as a basis for the tritium dose criteria development in this country. The ecological situation nearby a large tritium processing plant is considered. Amounts of tritiated waste produced at the plant, sources of tritium effluents, tritium content in the air, water, snow, soil and vegetation as well as HTO sorption parameters of various food products are reported. On the basis of HTO near-surface concentrations in the air and public doses measured 3 km away from the plant stack, the tritium dose factor was calculated.     

三氧化二铝PLOT柱气相色谱法分离氢同位素自旋异构体

氢同位素核自旋异构体正-仲态比例影响氢同位素的低温物性,有必要对其比例进行测定。本文利用活性三氧化二铝多孔层开管(Porous Layer Open Tubular,PLOT)柱实现了正-仲氢同位素(氕、氘)的基线分离,发展了一种可在液氮温度下测定同核分子正-仲态比例的色谱分析技术。研究结果表明,与传统三氧化二铝填充柱相比,高效PLOT柱实现了正、仲氕(o-H_2、p-H_2)以及正、仲氘(o-D2、p-D2)的基线分离(分离度R_s大于1.5),当流量为5 m L·min-1时,分离度R_s(p-H_2,o-H_2)=6.9,R_s(o-D2,p-D2)=1.8。正仲态分离度与进样量、流量均有关系。根据峰面积的积分结果,常温(298 K)下正、仲氕比例为2.77:1,正、仲氘比例为1.78:1,与理论测算值基本符合。HD与o-H_2实现了部分分离,R_s(o-H_2,HD)=0.5,根据理论预测,实现HD与o-H_2的基线分离(R_s达到1.5),理论塔板数需要达到3.9×105。     

Hypothetical integration of a fast reactor to the fuel cycle in Slovakia

The inspiration for dealing with the topic of fuel cycle back-end was attendance at a European project called RED-IMPACT – Impact of Partitioning Transmutation and Waste Reduction Technologies. This paper includes an image how to re-use energetic potential of stored spent fuel and at the same time how to effectively reduce spent fuel and radioactive waste volumes aimed for deep repositories. The first part is based on the analysis of Pu and minor actinides (MA) content in actual VVER-440 spent fuel stored in Slovakia. The next parts present the hypothetical possibilities of reprocessing and Pu re-use in a fast reactor under Slovak conditions. For the hypothetical transmutation of heavy nuclides (Pu and MA) contained in Slovak spent fuel a SUPERPHENIX (SPX) fast reactor with increased power was chosen because a fast nuclear reactor cooled by sodium belongs to the group of Generation IV reactor systems. This article deals with the analysis of power production and fuel cycle indicators. The indicators of the SPX calculation model were compared with the results of the VVER-440 spent fuel with the initial fuel enrichment of 4.25% U-235 + 3.35% Gd₂O₃. The created SPX model in the spectral computer code HELIOS 1.10 consists of a fissile (fuel) and a fertile part (blanket). All kinds of calculations were performed by the computer code HELIOS 1.10. This study also exposes the HELIOS modelling and simulating borders.     

The Laser Ion Source Trap (LIST) coupled to a gas cell catcher

The proof of principle of the Laser Ion Source Trap (LIST) coupled to a gas cell catcher system has been demonstrated at the Leuven Isotope Separator On Line (LISOL). The experiments were carried out by using the modified gas cell-based laser ion source and the SextuPole Ion Guide (SPIG). Element-selective resonance laser ionization of neutral atoms was taking place inside the cold jet expanding out of the gas cell catcher. The laser path was oriented in longitudinal as well as transverse geometries with respect to the atoms flow. The enhancement of beam purity and the feasibility for in-source laser spectroscopy were investigated in off-line and on-line conditions.     

Neutron poison distribution in the central reflector to reduce the DLOFC temperature of a Th-LEU fueled OTTO PBMR DPP-400 core

In a previous study using a mixture of thorium and 20 a/o% LEU at 16 gram per fuel sphere heavy metal loading and adjusting the effective fuel enrichment to produce the same amount of cumulative energy per fuel sphere as with the 10 a/o% Low Enriched Uranium (LEU), the maximum Depressurized Loss Of Forced Cooling (DLOFC) temperature was reduced from 2273 to 1925 °C and 1811 °C for a symmetric and asymmetric core, respectively using an once-through-then-out (OTTO) fuelling scheme. This article presents an additional strategy for reducing the maximum DLOFC temperature by placing an optimized distribution of neutron poisons in the central reflector. This strategy produced maximum DLOFC temperatures of 1509 and 1448 °C for the symmetric and the asymmetric cores, respectively. These results are impressive as it means that the less complicated OTTO cycle with its lower capital cost achieved the same cumulative energy produced per fuel sphere than the standard six-pass refuelling scheme and that at substantially lower maximum DLOFC temperatures. Both the addition of the neutron poisons to the central reflector and the creation of a radially asymmetric core resulted in lower burn-ups that had to be reversed by increasing the enrichment of the fuel.     

Using thorium to reduce the maximum fuel temperatures during depressurized loss of coolant accidents in a once-through-then-out (OTTO) PBMR DPP-400 core

This article presents the results for the PBMR-DPP-400, but for a once-through-then-out (OTTO) refueling scheme. An optimization attempt of the axial and radial power profiles is reported. The main aim was to reduce the maximum depressurized loss of forced coolant (DLOFC) temperature by adding thorium to the fuel and making the fuel layout radially asymmetric by placing lower enriched fuel in the inner and higher enriched fuel in the outer fuel flow regions. These measures (1) flattened the peaks in the axial power profiles and thus suppressed the hotspots in the axial DLOFC temperature profiles and (2) ‘pushed’ the power radially outwards, so as to reduce the distance that the decay heat must be evacuated towards the outside of the fuel core. This resulted in a huge reduction in the maximum DLOFC temperature for the OTTO cycle from 2273 to 1811 °C, which is still above the 1600 °C limit but represents a remarkable result. Maximum DLOFC temperature below the 1600 °C limit was obtained by reducing the power output. The results obtained and the proposed strategies for further improvement are applicable to the Chinese HTR-PM and could produce even better results in Prismatic Block Reactors such as the Japanese HTTR.     

A thermal model to characterize the flattening effect of a nuclear fuel element in an annular channel using simple analytical approach

In the present work, axial power flattening effect due to insertion of adjuster rods is modeled and a heat generating fuel pin in an annular channel is characterized. The effect of insertion of single adjustor rod at the peak heat generation region is modeled as two heat sources and one heat sink. The results obtained from the developed model indicate that insertion of adjuster rods beyond a critical number (3–4); flattening effect does not improve much. The peak surface and centerline temperatures of fuel pin are observed to show a cyclic variation and shift towards end as compared to unadjusted profile. Results are in accordance with the best estimate RELAP code.     

Thermal-hydraulic calculations using MARS code applied to low power and shutdown probabilistic safety assessment in a PWR

The methods developed for full-power probabilistic safety assessment, including thermal-hydraulic methods, have been widely applied to low power and shutdown conditions. Experience from current low power and shutdown probabilistic safety assessments, however, indicates that the thermal-hydraulic methods developed for full-power probabilistic safety assessments are not always reliable when applied to low power and shutdown conditions and consequently may yield misleading and inaccurate risk insights. To increase the usefulness of the low power and shutdown risk insights, the current methods and tools used for thermal-hydraulic calculations should be examined to ascertain whether they function effectively for low power and shutdown conditions. In this study, a platform for relatively detailed thermal-hydraulic calculations applied to low power and shutdown conditions in a pressurized water reactor was developed based on the best estimate thermal-hydraulic analysis code, MARS2.1. To confirm the applicability of the MARS platform to low power and shutdown conditions, many thermal-hydraulic analyses were performed for the selected topic, i.e. the loss of shutdown cooling events for various plant operating states at the Korean standard nuclear power plant. The platform developed in this study can deal effectively with low power and shutdown conditions, as well as assist the accident sequence analysis in low power and shutdown probabilistic safety assessments by providing fundamental data. Consequently, the resulting analyses may yield more realistic and accurate low power and shutdown risk insights.