During the process of beam extraction in positive ion source under high voltage region, a large number of electrons are produced in the gaps of grids. After back-streaming acceleration, these electrons go back to arc chamber or impinge grids and heat electron dump or grids, which are harmful for the safety of ion source. Under the situation of poor beam extraction optics, a large part of the primary beam ions bombard the surface of suppressor grid. And this process produces a large number of electrons. Due to the huge extracted voltage, the secondary electron emission coefficient of the suppressor grid surface is also great, when beam ions bombard on it. As a result, the grids’ current grows. The curvature of ion emission surface and equipotential surface nearby are mainly connected to the perveance and plasma grid geometry. In order to optimize the beam performance of high current ion source and increase the mean arc efficiency, the plasma grid of accelerator is already replaced from circular cross section grid to diamond cross section grid. As a result, the shape of ion emission surface is only connected to the perveance. According the measurement of the current of suppressor grid and the calculation of the perveance of the corresponding shoot, we can analyze the effect of beam divergence angle on back-streaming electron. When the beam divergence angle increases, the number of back-streaming electrons increases rapidly, and grids current changes significantly, especially the current of gradient grid and suppressor grid. The results can guide the parameters operating on the ion source for EAST-NBI and find the reasonable operation interval of perveance and the best one to ensure the safety and stable running of the ion source, which has great significance on the development of long pulse, high power ion source.

The Experimental Advanced Superconducting Tokamak (EAST) is one of the fully superconducting tokamak, its aim at the long-pulse operation (1000s) to study the physics of steady-state operation for nuclear fusion sciences. In order to support the steady-state operation and physical research, the high power neutral beam injection (NBI) system need to be employed on the EAST for the plasma heating and current driving. According to the scientific study schedule of the EAST, the designed NBI system includes two beam lines which will be constructed in two phases. Each beam line will deliver a deuterium neutral beam with beam energy of 50–80 keV with beam duration of 10–100 s. Each beam line has the maximum beam power of 4MW. The high current ion source is one of the most important parts in the high power NBI. A hot cathode positive ion based source was developed for EAST-NBI, which shown in Fig. 1. The ion source contains a bucket hot cathode arc chamber with 650 mm long, 260 mm width and 300 mm depth. There are 32 hairpin filaments with diamond of 1.5 mm and 160 mm long to supply primary electrons. A tetrode type accelerator with slit type used to extract the ions from the plasma and accelerated to the desired energy. The beam extraction area is 120 mm × 480 mm (can be changed) with beam transmittance of 60 %. The designed beam species is deuterium with beam power of 2–4 MW and beam energy of 50–80 keV and beam pulse length of 10–100 s. The ion source needs to be conditioning before operation on the EAST-NBI. An ion source test bed was designed and developed for the ion source performance tests and the optimization verification. The characteristics of ion source were tested with hydrogen beam, and each ion source should achieve 4MW ion beam with beam energy of 80 keV. The optimum beam perveance and arc efficiency were analyzed too. The optimum beam perveance was 2.8 μp with beam energy of 50 keV and the arc efficiency was 0.55A/kw. Long pulse operation was one of the requests of EAST ion source. The real-time feedback control method was employed and can got stable plasma and ion beam. The beam extraction was tested to achieve 100 s on the test bed. Consider the high power deposited on the calorimeter, the beam was modulated with suitable frequency and duty ratio. When the conditioning finished, the ion sources were moved to the EAST-NBI. The deuterium beam was extracted and injected into the EAST plasma. Details of the performance of positive ion source on the test bed and EAST-NBI will be presented.

Development of selective adsorbents is very important subject for the effective multi-nuclide decontamination related to the severe accident of Fukushima Daiichi Nuclear power Station (Fukushima NPS). In this study, the adsorption properties for nine kinds of zeolites (Zeolite A, Zeolite X, Zeolite Y, Zeolite L, Modified Chabazite, Phillipsite, Erionite, Synthetic Mordenite, Natural Mordenite and Clinoptilolite) are evaluated in the presence of sodium salts, boric acid and seawater. The present study deals with (1) selective adsorption properties for single nuclide ions (Cs + , Sr 2+ , Eu 3+ , I − , UO 2 2+ , Am 3+ and NpO 2 + ), and (2) multi-nuclide adsorption properties of 26 elements (typical elements in Advanced Liquid Processing System (ALPS) in Fukushima NPP-1) for the above zeolites. The distribution coefficient ( K d , ml/g) and uptake ( R , %) were estimated by batch method using NaI (Tl) scintillation counter, ICP-AES and AAS. Zeolites with different crystal structures have the diversity of the adsorption selectivity for various radioactive nuclides. Chabazite, mordenite and clinoptilolite with lantern or tunnel structure were very effective for the adsorption of monovalent Cs + ions even in real seawater. Zeolite A and X with three-dimensional cage structures were effective for the adsorption of divalent Sr 2+ and Co 2+ ions under the practical condition (30% diluted seawater). Zeolite L was effective for the adsorption of Eu 3+ ions under the practical condition. As for I − adsorption, Ag-zeolites are found to be effective, and the uptake (%) of I − (NaI in pure water) for Ag-zeolites was estimated to be above 98% in pure water. As for actinoid adsorption, the distribution profile, K d vs pH, had a maximum depending on the hydrolysis pH. Zeolite A, Zeolite L and Zeolite X showed an excellent adsorption property for UO 2 2+ , Am 3+ and NpO 2 + , respectively. Selective adsorption tendencies of different zeolites were evaluated for 26 elements referred to ALPS. Comparing the uptake results for different zeolites, the following tendency of adsorbability was observed. Mordenite had adsorption selectivity for monovalent alkali metal ions of Rb + and Cs + . Zeolite A and X exhibited relatively high adsorption selectivity for divalent ions of Sr 2+ and Co 2+ . Zeolite L had adsorption selectivity for trivalent lanthanide ions such as Ce 3+ and Eu 3+ . These tendencies were the same as those without boric acid. Thus, the zeolites with diverse adsorption selectivity are effective for the multi-nuclide decontamination of radioactive contaminated water.

A silica-based ammonium molybdophosphate (AMP/SiO 2 ) was successfully applied to remove Cs(I) from radioactive waste water in our previous study. The structure characterization of irradiated AMP/SiO 2 was carried out by X-ray diffraction (XRD). The effect of the initial concentration of metal ions and acidity concentration for removal Cs(I) by irradiated AMP/SiO 2 were investigated. No obvious difference was observed for AMP/SiO 2 with 300 kGy absorbed dose from the XRD patterns, except the slightly shift of (110) peak toward a small angle. Relatively large distribution coefficient (8×10 3 cm 3 /g) of Cs(I) was obtained in 3M HNO 3 by irradiated AMP/SiO 2 . The maximum adsorption capacity of AMP/SiO 2 irradiated at 300 kGy was estimated as 22.4 mg/g. The loss of molybdenum from AMP/SiO 2 in 3 M HNO 3 with 200 kGy absorbed dose is less than 5%.

Hydration is a main process following mixing of cement with water, producing a chemical reaction between cement particles and water molecules known as Calcium-Silica-Hydroxide (C-S-H). The C-S-H acts as a binder between coarse and fine mix aggregates. During mixing, capillary voids are formed where water can easily become trapped. The van der Waals force being the principal source of strength between such compounds, voids become detrimental to long-term strength of concrete because they decrease the amount of surface area aggregate and cement particles have to come into contact with each other. An overall decrease in both pores size and their distribution in concrete increase density, therefore increasing the van der Waals force between these particles. In order to reduce capillary pores size and their distribution, in this paper we present a result from preliminary investigation on how gamma ray exposure during the early stages of the curing process affect the concrete mixture strength values. The gamma ray enhanced curing is expected to break the water molecules trapped in capillary pores into H + and OH − , allowing these ions to escape and therefore result in densifying the concrete mixture. Concrete cubes of 5cm 3 in volumes are exposed to isotropic 137 Cs gamma emitting source. The absorbed dose in cubes is measured as well as calculated using MCNP6. Each concrete cube is exposed to a 630 MBq 137 Cs source for seven continuous days. The absorbed dose is experimentally measured using a Landauer nanoDot system. The average measured dose is 1.12 Gy at the front and 0.33 at the back of a concrete, totaling 0.79 Gy. The average measured dose is 0.98 Gy at the front and 0.29 Gy at the back, totaling 0.69 Gy of absorbed dose, thus showing a good agreement between the numerical predictions and experimental measurements. Compressive strength of concrete mixes exposed to gamma rays during the curing time and those cured without exposure to gamma radiation are tested. Preliminary data shows that the average compressive strength for gamma ray enhanced curing of concrete mixes is in order of 9,000 psi while the average com-pressive strength for the cubes cured in dry air is on average about 5,000 psi. A number of new experiments is planed until the time of the Conference in providing more exact answer to the question: does gamma ray enhanced curing of concrete increase its strength and reduces the curing time?

The author recently found that there should exist a “radiation-induced electrolytic (RIE)” mechanism in the reactor water inducing severe interaction between structural materials and their environments in aged LWRs. This mechanism was identified while trying to theoretically reconstruct the potential differences observed in two in-pile test loops; NRI-Rez in Czech Republic and INCA Loop in Sweden. These results are indicating that the in-core potential is approximately 0.1/0.4volt higher, in BWR(NWC)/PWR water chemistry respectively, when compared to the out-core regions. Through modeling studies, it was found that the concentrations of (DH)/(DO) for PWR/BWR(NWC) are higher/lower respectively, in the in-core region compared with the out-of-core region. These solute species in high concentrations should spontaneously decompose at the out-of-core region, enabling control of their water chemistry. This mode of corrosion cell has been dismissed in the nuclear community considering that the transport of ions with flow is insignificant due to high purity of reactor water. Part 1 of this paper focuses on how the RIE phenomena are prompted although the reactor water is kept in high purity. The stable molecular species in the reactor water flow transport the valence electrons. They are released at the cathodic in-core region and are recovered at the anodic out-of-core region. Thus estimated potential differences have been benchmarked with the published in-pile test results for both PWR- and BWR water chemistry environments as explained in Part 2 of this series (1).

This paper deals with near past, ongoing and planned R&D works on fission products (FPs) behaviour in Reactor Cooling System (RCS), containment building and in Filtered Containment Venting Systems (FCVS) for severe accident (SA) conditions. For the last topic, in link with the Fukushima post-accident management and possible improvement of mitigation actions for such SA, the FCVS topic is again on the agenda (see Status Report on Filtered Containment Venting, OECD/NEA/CSNI, Report NEA/CSNI/R(2014)7, 2014.) with a large interest at the international scale. All the researches are collaborative works; the overall objective is to develop confident models to be implemented in ASTEC SA simulation software. After being initiated in the International Source Term Program (ISTP), researches devoted to the understanding of iodine transport through the RCS are still ongoing in the frame of a bilateral agreement between IRSN and EDF with promising results. In 2017, a synthesis report of the last 10 years of researches, which have combined experimental and fundamental works based on the use of theoretical chemistry tools, will be issued. For containment, the last advances are linked to the Source Term Evaluation and Mitigation (STEM) OECD/NEA project operated by IRSN. The objective of the STEM project was to improve the evaluation of Source Term (ST) for a SA on a nuclear power plant and to reduce uncertainties on specific phenomena dealing with the chemistry of two major fission products: iodine and ruthenium. More precisely, the STEM project provided additional knowledge and improvements for calculation tools in order to allow a more robust diagnosis and prognosis of radioactive releases in a SA. STEM data will be completed by a follow-up, called STEM2, to further the knowledge concerning some remaining issues and be closer to reactor conditions. Two additional programmes deal with FCVS issues: the MItigation of outside Releases in the Environment (MIRE) (2013–2019) French National Research Agency (NRA) programme and the Passive and Active Systems on Severe Accident source term Mitigation (PASSAM) (2013–2016) European project. For FCVS works, the efficiencies for trapping iodine with various FCVS, covering scrubbers and dry filters, are examined to get a clear view of their abilities in SA conditions. Another part, performed in collaboration with French universities (Lorraine and Lille 1), is focused on the enhancement of the performance of these filters with specific porous materials able to trap volatile iodine. For that, influence of zeolites materials parameters (nature of the counter-ions, structure, Si/Al ratio …) will be tested. New kind of porous materials constituted by Metal organic Frameworks (MOF) will also be looked at because they can constitute a promising way of trapping.

Around near-surface underground, an unsaturated zone is formed in such porous layer composed of mainly silicate and silica sand. In the unsaturated zone in which pores are not completely filled with groundwater but also air, the migration of RNs is influenced by the presence of air phase. So far, though many studies have been reported about the unsaturated zone, little studies on the relation of the unsaturated zone and the migration of RNs have been carried out. In this study, the change of the migration of tracer ions in the unsaturated zone was examined by using a bed column, and analyzed by applying advection dispersion equation to the experimental results. As tracer ions, cesium (Cs) and strontium (Sr) were chosen in this study. These elements are the most important elements for the decontamination of Eastern Japan area caused by Fukushima the Daiichi Nuclear Power Plant accident. The experimental results showed the facilitation of the migration of tracer ions in the unsaturated zone. This causes the decrease in the sorption of tracer ions due to the inhibition of air phase. That is, the increase in the retardation effect by the clogging with air phase in flow paths was not remarkable. In the assessment of RNs migration under the unsaturated condition, these competing effects should be reflected adequately.

Investigations about the release, transportation and deposit of fibrous insulation material (FIM), corrosion products as well as resulting compounds and debris mixtures become more important to reactor safety research, when considering long-term behavior of emergency core cooling systems (ECCS) during loss of coolant accidents (LOCA). Debris released by a leakage jet leads to head loss buildups at the sump strainers, the debris filters and the spacers of fuel assemblies. However, these complex processes may influence the decay heat transfer out of the reactor core. A similar but newer scenario implies that the boric acid coolant in pressurized water reactors (PWR) can support corrosion processes at hot-dip galvanized installations, leading to a significant release of ionic zinc into the coolant during the sump recirculation phase. A long-term change of chemical properties of the coolant (e.g. pH, zinc ion concentration) has to be considered in safety analyses. Chemical analyses showed that the solubility of zinc in boric acid coolant is inversely proportional to the coolant temperature. Consequently, zinc ions can be dissolved at lower temperatures in the containment sump. Precipitations of zinc borate (ZBP) are possible at hot spots in the reactor core. The ZBP can be formed as a flocculent disperse phase in the coolant or as solidified layers at hot fuel rod surfaces. Layer spalling could lead to the release of further solid particles into coolant flow inside the reactor core. In several joint research projects between the Zittau/Görlitz University of Applied Sciences (HSZG) and the Helmholtz-Zentrum Dresden-Rossendorf (HZDR) investigations on the impact of these processes to the head loss buildup and the heat transfer from core were done at laboratory and semi-technical scaled test facilities. Generic experiments showed the formation of ZBP in heating rod configurations. The ZBP may remain in the core structures or can be transported on debris filter cakes in upstream and downstream components of ECCS and influence the head loss. After this, research addressed the systematical clarification of physico-chemical mechanisms and their influence on thermal-hydraulic-dynamic processes occurring as a consequence of flow induced corrosion effects during LOCA. This paper includes a description of the most important involved test facilities, applied measuring techniques, an overview of boundary conditions considered experimentally and selected results.

Generic investigations regarding the influence of corrosion processes of hot-dip galvanized PWR containment installations on strainer clogging as well as on the coolant chemistry and possible resulting in-core effects are carried out within joint research projects of the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), TU Dresden (TUD) and Zittau-Görlitz University of Applied Sciences (HSZG). Lab-scale experiments at HZDR and TUD are focused on elucidation of physico-chemical corrosion and precipitation processes as well as resulting clogging effects. Results of generic experiments in a lab-scale corrosion test facility suggest that there is a multi-stage corrosion process. The first stage comprises dissolution of the zinc layer in the coolant forming zinc ions and in turn affecting the coolant chemistry. During the second stage, the base material (steel) corrodes forming insoluble corrosion particles, which can subsequently lead to accelerated clogging of fiber-laden strainers within a few hours. The main influences on corrosion were identified as impact of the coolant jet onto the corroding surface, water chemistry and zinc surface / coolant volume ratio. Furthermore, retrograde solubility of zinc corrosion products in boric acid containing coolants with increasing temperature was observed. Thus, formation and deposition of solid corrosion products cannot be ruled out if zinc containing coolant is heated up during its recirculation into hot downstream components (e.g. hot-spots in core). Corrosion experiments, which included formation of corrosion products at a heated cladding tube, proved that zinc, dissolved in the coolant at low sump temperatures, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces. Due to alternating heating and cooling of the coolant during sump recirculation operation, a cycle of zinc corrosion and zinc borate precipitation may be initiated, which may eventually influence the thermal hydraulics in downstream components during the post-LOCA stage. The results obtained at lab-scale were confirmed by corresponding experiments in semi-technical test facilities of the project partner HSZG. Based on the experimental results, water chemical measures were tested to reduce corrosion and/or zinc borate precipitation effects.

Cement is a practical material for constructing the geological disposal system of radioactive wastes. However, such materials alter groundwater up to 13 in pH around the repository, changing the permeability of natural barrier. So far, the authors have examined the relation of permeability change with dissolution process by flowing a high pH solution (NaOH, 0.1 mM) into a bed packed with amorphous silica particles. Here, the particle diameters were adjusted to a size fraction of 74 to 149 μm by sieving. Its specific surface area was estimated as 350 m 2 /g by the BET method using nitrogen gas. The experimental results showed that the permeability did not immediately change although the soluble silicic acid continuously flowed out of the packed bed. This study proposes a new mathematical model considering the diffusion and dissolution processes in the inner pore of the particle. This model assumed that each packed particle (74 to 149μm in diameter) consists of the sphere-shaped aggregation of smaller particles (20 nm in diameter). OH − ions diffuse into the pore between such small particles, and simultaneously consumed by the reaction with small particles. The radius of the each packed particle (sphere-shaped aggregation of small particles) was defined by the length from the center of the aggregation to the region where the small particles still remains. Since the outer small particles more easily dissolve than inner small particles because of diffusion process of OH − ions, each packed particle gradually shrinks. The fundamental equations consist of a simple diffusion equation of spherical coordinates of OH − ions considering the reaction term, which is linked by the equation to describe the size change of small particles with time. Here, this model also considered a change (time and space) of the diffusion oefficient caused by the change of the porosity between small particles. Besides, the change of over-all permeability of the packed bed was evaluated by Kozeny-Carman equation and the calculated radii of packed particles. The dissolution rate constant already reported was used. The calculated result was able to well describe the experimental result, though there was no fitting parameter in the comparison with the experiment results. While the flow paths of underground cannot be simply simulated by a packed bed, this approach suggested that the dynamic behavior of permeability in a natural barrier depends also on non-uniformity of dissolution processes in inner pores (secondary pores) of minerals.

The development of selective adsorbents for radioactive Sr ions is one of the most important subjects for the safety decontamination in Fukushima NPP-1 [1] . In this study, the selective adsorption properties of Sr, characterization and stable solidification were clarified by using the novel adsorbent of potassium titanates (KT-1). The adsorption properties of Sr 2+ ions for original and calcined specimens were examined by batch method under the following conditions; V/m = 100 cm 3 /g, Mixed solution: 10,000 ppm Na + , 10 ppm Cs + , 10 ppm Ca 2+ , 1 ppm Mg 2+ and 1 ppm Sr 2+ , 85 Sr tracer: 5,000 cpm/cm 3 , centrifugation: 2,500 rpm, 25°C, shaking time: 1∼24 h, calcination temp.: 300∼900°C. Relatively large uptake percentage above 90% was obtained for the original and calcined specimens below 800°C, while the Sr uptake for calcined specimens above 900°C was lowered due to the thermal decomposition of K 2 Ti 2 O 5 · x H 2 O. The Sr distribution in the column was examined by flowing the mixed solution through the columns packed with KT-1. The Sr distribution profiles were obtained by the measurement of γ-activity in the column at 5 mm intervals. In either case, no breakthrough of Sr was observed. The distribution profile tended to smooth with increasing flow rate; Sr adsorption band and flow rate have a linear relationship. The leachability of Sr for the solid forms was further examined under the following leaching conditions; leachant: pure water and 1 M HCl; leachant temp.: 25°C and 90°C, leaching period: 4 weeks; calcining temp.: 500∼1,100°C. The leached percentage of Sr in pure water was less than the detection limit of ICP-AES, and that in 0.1 M HCl tended to markedly decrease with calcining temperature; the formation of SrTiO 3 phase above 800°C was effective for the lowering of leachability. The novel adsorbent of KT-1 is thus effective for the selective decontamination and stable solidification of Sr in Fukushima NPP-1.

Radioactive waste treatment is planned in LWTF (Low-level radioactive Waste Treatment Facility, JAEA) for LLW generated from the Tokai-reprocessing facility. The target LLW consists of highly concentrated sodium nitrate (5 M NaNO 3 ) containing low-level 90 Sr. In this study, selective adsorption properties of Sr 2+ for highly functional A type zeolites (A51-JHP, A51-J (Union Showa) and A-4, X type zeolite (F-9) and Titanic acid-PAN (polyacrylamide) were clarified by batch and column adsorption methods. The irradiation stabilities of these adsorbents were also evaluated. The distribution properties of Sr 2+ on different adsorbents were compared in simulated waste solution (5 M NaNO 3 , 0.1 ppm Sr 2+ , 85 Sr as tracer). The order of distribution coefficients ( K d,Sr ) was Titanic acid-PAN > A51-JHP > A51-J > A-4 > F-9. The largest value of K d,Sr for titanic acid-PAN was estimated to be 218 cm 3 /g, while the saturated capacity ( Q max ) was very small. Titanic acid-PAN had also the largest uptake rate of Sr 2+ ions and the uptake attained equilibrium within 8 h. On the other hand, A51-JHP had a relatively large K d,Sr value above 100 cm 3 /g and a Q max value of 0.65 mmol/g. The breakthrough properties of Sr 2+ were examined by varying cations present (single and mixed solutions) and flow rate (0.08 and 0.17 cm 3 /min). The components for the single solution were 400 g/L NaNO 3 , 100 ppm Sr 2+ , 85 Sr as tracer, and the mixed solution contains 200 ppm Cs + , 100 ppm Ca 2+ , 50 ppm Mg 2+ , 50 ppm RuNO 3+ in addition to the single solution components. The breakthrough curve for Titanic acid-PAN column using single solution had an S -shaped profile, while the “concentration phenomenon” exceeding C/C 0 (breakthrough ratio) = 1 was observed in the case of mixed solution. As for the A51-JHP column, the breakthrough curve for single solution was similar to that for mixed solution and the 5% breakpoint was enhanced by decreasing the flow rate. The A51-JHP was stable under 60 Co-irradiation up to 2.54 MGy; K d,Sr and Q max values were almost constant. In contrast, Titanic acid-PAN was affected above 0.28 MGy, due to the radiolysis of PAN matrix, and this surface alteration led to the release of active component of titanic acid. The novel A type zeolite (A51-JHP) is thus expected for the selective removal of Sr 2+ in LWTF. The optimization of particle size and flow rate should be examined before practical use.

The development of selective adsorbents has become very important for the effective multi-nuclide decontamination. In this study, the selective adsorption properties of 26 nuclides for different types of zeolites (A, L, natural mordenite (NM), Ag-NM) were examined in the presence of boric acid. The batch adsorption experiments were carried out using four kinds of test solutions containing boric acid and calcium hydroxide; (1)DW (distilled water) + H 3 BO 4 : 3,000 ppm + LiOH: 10 ppb, (2)DW + Ca(OH) 2 : 500 ppm + H 3 BO 4 : 3,000 ppm + LiOH: 10 ppb, (3)Seawater (30% diluted) + H 3 BO 4 : 3,000 ppm, (4)Seawater + H 3 BO 4 : 3,000ppm. The uptake (%) of Sr 2+ for zeolite A (A-51J), Cs + for natural mordenite (NM, 2460#, Ayashi, Sendai), and I − for Ag-NM was determined under the following conditions; Concentration of Sr 2+ , Cs + and I − ions: 10 ppm, V/m = 100 cm 3 /g, 25°C, 24 h. The uptake (%) of Sr 2+ , Cs + and I − ions was estimated to be above 90%, while tended to decrease in the presence of seawater. Especially, the uptake (%) of I − ions for Ag-NM markedly decreased in the presence of seawater. As for the zeolites A and L, the uptake (%) of 26 elements was determined by using two kinds of test solutions; (1)DW (distilled water) + H 3 BO 4 : 3,000 ppm + LiOH: 10 ppb + 26 nuclides: 10 ppm, (2)Seawater (30% diluted) + H 3 BO 4 : 3,000 ppm + 26 nuclides: 10 ppm. Zeolite A has relatively large uptake percentage for Sr, Co, Ni and Zn, and zeolite L has high adsorbability to lanthanoid group of Eu, Ce and Pr. The increase in pH led to the enhancement of uptake (%), while the hydrolysis of metal ions should be also considered. The multi-nuclides separation is thus expected by considering the difference in uptake properties of zeolite A, L and natural mordenite.

State of the art neutron detectors lack capabilities required by the fields of homeland security, health physics, and even for direct in-core nuclear power monitoring. A new system being developed at Purdue’s Metastable Fluid and Advanced Research Laboratory in conjunction with S/A Labs, LLC provides capabilities the state of the art lacks, and simultaneously with beta ( β ) and gamma ( γ ) blindness, high (> 90% intrinsic) efficiency for neutron/alpha spectroscopy and directionality, simple detection mechanism, and lowered electronic component dependence. This system, the Tensioned Metastable Fluid Detector (TMFD) [3], provides these capabilities despite its vastly reduced cost and complexity compared with equivalent present day systems. Fluids may be placed at pressures lower than perfect vacuum (i.e. negative) [4, 5], resulting in tensioned metastable states. These states may be induced by tensioning fluids just as one would tension solids. The TMFD works by cavitation nucleation of bubbles resulting from energy deposited by charged ions or laser photon pileup heating of fluid molecules which are placed under sufficiently tensioned (negative) pressure states of metastability. The charged ions may be created from neutron scattering, or from energetic charged particles such as alphas, alpha recoils, fission fragments, etc. A methodology has been created to profile the pressures in these chambers by lasing, called Laser Induced Cavitation (LIC), for verification of a multiphysics simulation of the chambers. The methodology and simulation together have lead to large efficiency gains in the current Acoustically Tensioned Metastable Fluid Detector (ATMFD) system. This paper describes in detail the LIC methodology and provides background on the simulation it validates.

Background of the experimental and methodical work is a loss of coolant accident (LOCA) in a Pressurized Water Reactor (PWR). During LOCA, both the leakage water jet as well as the water in the containment can lead to corrosion of galvanized installations (e.g. grates). That means a change of consistency and chemical properties of the coolant water during the sump recirculation operation, in particular an increase of the zinc concentration (Zn 2+ ions in dissolution) and the pH-value. Furthermore, it leads to separation of layer-forming zinc compounds (e.g. zinc borate) at hotspots (fouling at fuel assemblies) and/or the crystallization of zinc corrosion products out of the coolant in hot areas [HOF10]. This in turn causes a decrease of pH-value and, consequently, an increase of the corrosiveness of the coolant flow. This can come along with a chemical transformation of zinc borate to nearly indissoluble zinc oxide by thermo hydrolysis. Spalling parts of the coating on the surface of the fuel rods could lead to the release of particles into coolant water flow. In combination with the release of fibrous insulation material, which can be transported into the reactor containment, containment building sump and connected systems, this process causes pressure build-ups at the strainers and the spacers of the fuel assemblies and a reduction of the pump performance. In this case, efficient heat dissipation cannot be guaranteed. In a joint research project between the Institute of Process Technology, Process Automation and Measuring Technology (IPM) of University Zittau/Goerlitz and the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), the influence of different boundary conditions like the flow structures in the reactor pressure vessel, temperature distributions and the temperature gradient in the central core regions will be investigated. Therefore, in a final step a test facility containing a 3×3 heated fuel rod dummy configuration was designed and constructed. The design and development of the test facility was accompanied by pre-test calculations and CFD simulations to get information about local surface and fluid temperatures under different flow conditions.

Ferritic/martensitic steels are candidates for fast reactors because of their sodium compatibility, superior resistance to corrosion and radiation damage, including swelling, and excellent thermal conductivity and thermal expansion coefficient. One significant limitation of any cladding material is its susceptibility to swelling at high doses. While HT9 has neutron irradiation performance data up to ∼200 dpa, dose requirements for the Traveling Wave Reactor (TWR) may be much higher. Obtaining higher-dose data will take many years, but in the interim, heavy ion irradiation could provide a useful tool toward predicting the swelling trends beyond 200 dpa. In this study, HT9 was irradiated from 440–480°C using 5 MeV Fe ++ ions. The samples are compared to a portion of HT9 fuel assembly duct from FFTF, which was characterized after neutron irradiation at 440°C with an accumulated dose of 155 dpa. Comparisons are made of the void size and density using transmission electron microscopy (TEM). The increase in dose from 280 dpa to 375 dpa increased void size, number density and swelling at 440°C, while swelling was generally lower at 480°C for the same helium pre-implantation conditions. Helium generally enhanced the nucleation of voids, as measured by the void density.

In Fukushima NPP-1, large amounts of high-activity-level water (HALW) accumulated in the reactor, turbine building and the trench in the facility is treated by circulating injection cooling system. The development of highly functional adsorbents and stable solidification method contributes to the advancement of the decontamination system and environmental remediation. The present study deals with (1) preparation of insoluble ferrocyanide loaded zeolites, (2) selective uptake of Cs + in seawater, (3) estimation of Cs immobilization ratio and stable solidification. Various kinds of Cs-selective composites loaded with insoluble ferrocyanides (CoFC, NiFC) into the zeolites (zeolite A (A51, A-51J), zeolite X (LSX), chabazite (modified chabazite) and natural mordenite (SA-5)) matrices have been prepared using successive impregnation/precipitation methods by Tohoku University. As for Cs + adsorption, these composites had relatively large uptake (%) above 95%, distribution coefficients ( K d ) above 10 3 cm 3 /g and excellent adsorption kinetics even in seawater. The immobilization ratio (%) of Cs for the CoFC saturated with Cs + was estimated at different calcination temperatures up to 1,200°C in advance. The immobilization ratio was less than 0.1% above 1,000°C, indicating that the adsorbed Cs + ions are completely volatilized and insoluble ferrocyanides had no immobilization ability for Cs. In contrast, the insoluble ferrocyanide-loaded zeolites had excellent Cs immobilization ability; in the case of insoluble ferrocyanide-loaded natural zeolites (NiFC-SA-5, CoFC-modified chabazite), the immobilization ratio was above 99% and 96% even after calcination at 1,000°C and 1,100°C, respectively, indicating that nearly all Cs ions are immobilized in the sintered solid form. On the other hand, the immobilization ratio for the insoluble ferrocyanide-loaded A and X zeolites (NiFC-A (A51, A51J), NiFC-X) tended to decrease with calcining temperature; for example, the immobilization ratio for NiFC-X at 1,000°C and 1,100°C was estimated to be 74.9% and 55.4%, respectively, and many spots concentrating Cs were observed on the surface. The difference in immobilization behavior between natural zeolites and synthetic ones is probably due to the phase transformation and surface morphology at higher temperature above 1,000°C. The stable solidification of insoluble ferrocyanides was thus accomplished by using the excellent Cs immobilization abilities of zeolite matrices (Cs trapping and self-sintering abilities).

Calcium silicate hydrate (CSH) is a main component of cement-based material required for constructing the geological repository. As in many countries, since the repository in Japan is constructed below water table, we must consider the interaction of radionuclide with cement materials altered around the repository after the backfill. Using fluorescence emission spectra, so far, the authors have investigated the interaction of Eu(III) (as a chemical analog of Am(III)) with CSH gels formed with no dried process, considering a condition saturated with groundwater. However, in such fluorescence emission behaviors, a deexcitation process of OH vibrators of light water and a quenching effect caused by Eu-Eu energy transfer between Eu atoms incorporated in the CSH gel must be considered. This study examined the fluorescence emission behavior of Eu(III) sorbed on CSH gels formed with no dried process, by using La(III) (non-fluorescent ions) as a diluent of Eu(III). Furthermore, the CSH samples were synthesized with CaO, SiO 2 , and heavy water (D 2 O) as a solvent in order to avoid the obvious deexcitation process of OH vibrators of light water. This study prepared CSH samples with the Ca/Si ratio set to 1.6, 1.0, and 0.5. A 1 mM solution of a given combination of Eu(III) and La(III) (Eu(III) content: 100%, 67%, 50% or 33%) was added into CSH gel sample. The contact time-period of the CSH gel with the Eu(III)/La(III) solution was set to 60 days. In the results, the peak around 618 nm was split into two peaks of 613 nm and 622 nm in the cases of Ca/Si = 1.0 and 1.6. Then, the peak of 613 nm decreased with increment of Eu(III)/La(III) ratio. This means that the relative intensity of 613 nm is useful to quantify the amount of Eu(III) incorporated in CSH gel. Besides, the intensity peak of 584 nm decayed with increment of Eu/La ratio, suggesting a quenching effect due to Eu-Eu energy transfer. However, the decay behavior of the fluorescence emission did not depend on the Eu/La concentration ratio. That is, such a quenching effect is neglectable. Additionally, the low Ca/Si ratio samples underwent slow attenuation of fluorescence and showed profiles similar to those of high Ca/Si ratio samples. Therefore, low Ca/Si ratio samples also include the reaction forming a complex on the surface of CSH gel with Eu(III). In other words, even if Ca/Si ratio is lower than 1.0, CSH gels would retard the migration of radionuclides released from the repository.

A linked facility between an ion accelerator and STM enabled to elucidate cascade damage process experimentally with an atomistic scale. By adjusting experimental conditions, we have successfully attained very clear images of Si (111) surface irradiated by very low-flux and stable ion beams. We can obtain the information of displacement damage produced by one injected ion. Much higher density of defects is observed than the number of injected ions throughout the area of the image, indicating that there are processes for defect formation other than direct collision with injected particles.