Design study on reactor structure of Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR)

Pb–Bi-cooled direct contact boiling water small fast reactor (PBWFR) can produce steam by direct contact of feedwater with primary Pb–Bi coolant above the core, and circulate Pb–Bi coolant by means of buoyancy of steam bubbles. The PBWFR is capable of eliminating components of the cooling system such as primary pumps and steam generators, and thereby making the reactor system simple and compact. The specifications of the PBWFR are as follows: the fuel is Pu–U nitride; the core height is 75 cm; the core diameter is 278 cm; the average burnup is 80 GWd/t; the refueling interval is 10 years; the rated electric power is 150 MWe; the rated thermal power is 450 MWt; the core outlet/inlet temperatures are 460 °C/310 °C, respectively; and the operating steam pressure is 7 MPa. The reactor structure design has been formulated, where reactor vessel sizes are 4200 mm (ID) × 19,750 mm (H), the guard vessel is a closed type, the upper structure is made of chimneys, and the core support structure is hung up. An ultrasonic flow meter is installed inside the vessel. The seismic evaluation, design of refueling procedure and cost evaluation have been performed.     

Medium temperature carbon dioxide gas turbine reactor

A carbon dioxide (CO₂) gas turbine reactor with a partial pre-cooling cycle attains comparable cycle efficiencies of 45.8% at medium temperature of 650 °C and pressure of 7 MPa with a typical helium (He) gas turbine reactor of GT-MHR (47.7%) at high temperature of 850 °C. This higher efficiency is ascribed to: reduced compression work around the critical point of CO₂; and consideration of variation in CO₂ specific heat at constant pressure, C_p, with pressure and temperature into cycle configuration. Lowering temperature to 650 °C provides flexibility in choosing materials and eases maintenance through the lower diffusion leak rate of fission products from coated particle fuel by about two orders of magnitude. At medium temperature of 650 °C, less expensive corrosion resistant materials such as type 316 stainless steel are applicable and their performance in CO₂ have been proven during extensive operation in AGRs. In the previous study, the CO₂ cycle gas turbomachinery weight was estimated to be about one-fifth compared with He cycles. The proposed medium temperature CO₂ gas turbine reactor is expected to be an alternative solution to current high-temperature He gas turbine reactors.     

Influence of ageing heat treatment on alloy A-286 microstructure and stress corrosion cracking behaviour in PWR primary water

The influence of ageing heat treatment on alloy A-286 microstructure and stress corrosion cracking behaviour in simulated Pressurized Water Reactor (PWR) primary water has been investigated. A-286 microstructure was characterized by transmission electron microscopy for ageing heat treatments at 670 °C and 720 °C for durations ranging from 5 h to 100 h. Spherical γ′ phase with mean diameters ranging from 4.6 to 9.6 nm and densities ranging from 8.5 × 10²² m⁻³ to 2 × 10²³ m⁻³ were measured. Results suggest that both the γ′ phase mean diameter and density quickly saturate with time for ageing heat treatment at 720 °C while the γ′ mean diameter increases significantly up to 100 h for ageing heat treatment at 670 °C. Grain boundary η phase precipitates were systematically observed for ageing heat treatment at 720 °C even for short ageing periods. In contrast, no grain boundary η phase precipitates were observed for ageing heat treatments at 670 °C except after 100 h. Hardening by γ′ precipitation was well described by the dispersed barrier hardening model with a γ′ barrier strength of 0.23. Stress corrosion cracking behaviour of A-286 was investigated by means of constant elongation rate tensile tests at 1.5 × 10⁻⁷ s⁻¹ in simulated PWR primary water at 320 °C and 360 °C. In all cases, initiation was transgranular while propagation was intergranular. Grain boundary η phase precipitates were found to have no significant effect on stress corrosion cracking. In contrast, yield strength and to a lesser extent temperature were found to have significant influences on A-286 susceptibility to stress corrosion cracking.     

XPS and optical studies of Xe-implanted and annealed YSZ single crystals

Xe⁺ ion implantation with 200 keV was completed at room temperature up to a fluence of 1 × 10¹⁷ ion/cm² in yttria-stabilized zirconia (YSZ) single crystals. Optical absorption and X-ray photoelectron spectroscopy (XPS) were used to characterize the changes of optical properties and charge state in the as-implanted and annealed crystals. A broad absorption band centered at 522 or 497 nm was observed in the optical absorption spectra of samples implanted with fluences of 1 × 10¹⁶ ion/cm² and 1 × 10¹⁷ ion/cm², respectively. These two absorption bands both disappeared due to recombination of color centers after annealing at 250 °C. XPS measurements showed two Gaussian components of O_1s spectrum assigned to Zr–O and Y–O, respectively, in YSZ single crystals. After ion implantation, these two peaks merged into a single peak with the increasing etching depth. However, this single peak split into two Gaussian components again after annealing at 250 °C. The concentration of Xe decreased drastically after annealing at 900 °C. And the XPS measurement barely detected the Xe. There was no change in the photoluminescence of YSZ single crystals with a fluence of 1 × 10¹⁷ ion/cm² after annealing up to 900 °C.     

Three pass core design proposal for a high performance light water reactor

The paper describes a novel core concept for a nuclear reactor cooled with supercritical water, in which the coolant is heated up from 280 °C at the reactor inlet to 500 °C at the outlet in four steps: a first heat-up step is provided by heat transfer from fuel assemblies to the moderator water in gaps and moderator boxes, a second step is foreseen in a central “evaporator” and two further steps in a first and a second superheater surrounding it. The coolant flow scheme includes upward and downward flow through the core with intermediate mixing in chambers above and below the core to eliminate hot streaks. A preliminary single channel analysis, concentrating on an average flow channel and on the hottest one only, indicates that such core design can match the limits of cladding materials available today. Even though the resultant pressure drop of the coolant will be higher than usual, it is expected that the assembly boxes can be designed with acceptable deformations.     

Carbon ion-implanted optical waveguides in Nd:YLiF4 crystal: Refractive index profiles and thermal stability

We report on the optical planar waveguides in Nd:YLiF₄ laser crystals fabricated by 6.0 MeV C³⁺ ion implantation at doses of 1 × 10¹⁵ or 2.5 × 10¹⁵ ions/cm², respectively. The refractive index profiles, which are reconstructed according to the measured dark mode spectroscopy, show that the ordinary index had a positive change in the surface region, forming non-leaky waveguide structures. The extraordinary index is with a typical barrier-shaped distribution, which may be mainly due to the nuclear energy deposition of the incident ions into the substrate. In order to investigate the thermal stability of the waveguides, the samples are annealed at temperature of 200–300 °C in air. The results show that waveguide produced by higher-dose carbon implantation remains relatively stable with post-irradiation annealing treatment at 200 °C in air.     

Core performance of new concept passive-safety reactor “kamado” - safety, burn-up and uranium resource problem -

New concept of a passive-safety reactor “KAMADO” has a negligible possibility of core melting and flexibility of total reactor power. The reactor core of KAMADO consists of fuel elements of graphite blocks, which have UO₂ fuel rods and cooling water holes. These fuel elements are located in a reactor water pool of atmospheric pressure (1 atm) and low temperature (< 60°C). In case of LOCA, decay heat from fuel rods is removed by conduction heat transfer to the reactor water pool. Since the cooling water does not contact a fuel rod directly, core design has much flexibility without considering dry-out limitation and Minimum Critical Power Ratio (MCPR). Additionally an effective use of spent fuel is expected.     

Reactor core design optimization of the 200 MWt Pb–Bi cooled fast reactor for hydrogen production

In this study reactor core geometrical optimization of 200 MWt Pb–Bi cooled long life fast reactor for hydrogen production has been conducted. The reactor life time is 20 years and the fuel type is UN-PuN. Geometrical core configurations considered in this study are balance, pancake and tall cylindrical cores. For the hydrogen production unit we adopt steam membrane reforming hydrogen gas production. The optimum operating temperature for the catalytic reaction is 540 °C. Fast reactor design optimization calculation was run by using FI-ITB-CHI software package. The design criteria were restricted by the multiplication factor that should be less than 1.002, the average outlet coolant temperature 550 °C and the maximum coolant outlet temperature less than 700 °C. By taking into account of the hydrogen production as well as corrosion resulting from Pb–Bi, the balance cylindrical geometrical core design with diameter and height of the active core of 157 cm each, the inlet coolant temperature of 350 °C and the coolant flow rate of 7000 kg/s were preferred as the best design parameters.     

Study on small long-life LBE cooled fast reactor with CANDLE burn-up – Part I: Steady state research

Small long-life reactor is required for some local areas. CANDLE small long-life fast reactor which does not require control rods, mining, enrichment and reprocessing plants can satisfy this demand. In a CANDLE reactor, the shapes of neutron flux, nuclide number densities and power density distributions remain constant and only shift in axial direction.

The core with 1.0 m radius, 2.0 m length can realize CANDLE burn-up with nitride (enriched N-15) natural uranium as fresh fuel. Lead–Bismuth is used as coolant.

From steady state analysis, we obtained the burn-up velocity, output power distribution, core temperature distribution, etc. The burn-up velocity is less than 1.0 cm/year that enables a long-life design easily. The core averaged discharged fuel burn-up is about 40%.     

ENHS reactor power level enhancement possibilities

The ENHS thermal hydraulic optimization code was modified and applied to search for the maximum attainable power from a wide range of ENHS design options subjected to the following constraints: maximum permissible hot channel coolant outlet temperature of 600 °C, clad inner temperature of 650 °C and primary coolant temperature rise of either 150 °C or 90% of the theoretical limit for accelerated corrosion rate. The T–H optimization variables include the intermediate heat exchanger number of channels, channel width and elevation; diameter of the riser and diameter of a flow-splitting shroud in the riser. It was found possible to increase the attainable power from the nominal 125 MW_th up to 311 MW_th for the reference core, 400 MW_th for a reference-like core having equilibrium composition fuel and 372 MW_th for a flattened power core with 9 plutonium concentration zones. A power level exceeding 400 MW_th may be achieved by flattening the power distribution of the equilibrium core or using nitride fuel with enriched nitrogen rather than metallic fuel. With forced circulation it is possible to operate the flattened power core at up to 532 MW_th corresponding to 223 MW_e.     

Low temperature neutron irradiation effects on microstructure and tensile properties of molybdenum

Polycrystalline molybdenum was irradiated in the hydraulic tube facility at the High Flux Isotope Reactor to doses ranging from 7.2 × 10⁻⁵ to 0.28 dpa at 80 °C. As-irradiated microstructure was characterized by room-temperature electrical resistivity measurements, transmission electron microscopy (TEM) and positron annihilation spectroscopy (PAS). Tensile tests were carried out between −50 and 100 °C over the strain rate range 1 × 10⁻⁵ to 1 × 10⁻² s⁻¹. Fractography was performed by scanning electron microscopy (SEM), and the deformation microstructure was examined by TEM after tensile testing. Irradiation-induced defects became visible by TEM at 0.001 dpa. Both their density and mean size increased with increasing dose. Submicroscopic three-dimensional cavities were detected by PAS even at 0.0001 dpa. The cavity density increased with increasing dose, while their mean size and size distribution was relatively insensitive to neutron dose. It is suggested that the formation of visible dislocation loops was predominantly a nucleation and growth process, while in-cascade vacancy clustering may be significant in Mo. Neutron irradiation reduced the temperature and strain rate dependence of the yield stress, leading to radiation softening in Mo at lower doses. Irradiation had practically no influence on the magnitude and the temperature and strain rate dependence of the plastic instability stress.     

Electronic excitation effects in CeO2 under irradiations with high-energy ions of typical fission products

In order to understand the formation mechanism of a crystallographic re-structuring in the periphery region of high-burnup nuclear fuel pellets, named as “rim structure”, information on the accumulation process of radiation damage and fission products (FPs), as well as high-density electronic excitation effects by FPs, are needed. In order to separate each of these processes and understand the high-density electronic excitation effects, 70–210 MeV FP ion (Xe^10–14+, I⁷⁺ and Zr⁹⁺) irradiation studies on CeO₂, as a simulation of fluorite ceramics of UO₂, have been done at a tandem accelerator of JAEA-Tokai and the microstructure changes were determined by transmission electron microscope (TEM). Measurements of the diameter of ion tracks, which are caused by high-density electronic excitation, have clarified that the effective area of electronic excitation by high-energy fission products is around 5–7 nm  and the square of the track diameter tends to follow linear function of the electronic stopping power (S_e). Prominent changes are hardly observed in the microstructure up to 400 °C. After overlapping of ion tracks, the elliptical deformation of diffraction spots is observed, but the diffraction spots are maintained at higher fluence. These results indicate that the structure of CeO₂ is still crystalline and not amorphous. Under ion tracks overlapping heavily (>1 × 10¹⁵ ions/cm²), surface roughness, with characteristic size of the roughness around 1 μm, is observed and similar surface roughness has also been observed in light-water reactor (LWR) fuels.     

Photoluminescence of Au formed in 12CaO · 7Al2O3 single crystal by Au-implantation

Au⁺ ion implantation with fluences from 1 × 10¹⁴ to 3 × 10¹⁶ cm⁻² into 12CaO · 7Al₂O₃ (C12A7) single crystals was carried out at a sample temperature of 600 °C. The implanted sample with the fluence of 1 × 10¹⁵ cm⁻² exhibited photoluminescence (PL) bands peaking at 3.1 and 2.3 eV at 150 K when excited by He–Cd laser (325 nm). This was the first observation of PL from C12A7. These two PL bands are possibly due to intra-ionic transitions of an Au⁻ ion having the electronic configuration of 6s², judged from their similarities to those reported on Au⁻ ions in alkali halides. However, when the concentration of the implanted Au ions exceeded the theoretical maximum value of anions encaged in C12A7 (2.3 × 10²¹ cm⁻³), surface plasmon absorption appeared in the optical absorption spectrum, suggesting Au colloids were formed at such high fluences. These observations indicate that negative gold ions are formed in the cages of C12A7 by the Au⁺ implantation if an appropriate fluence is chosen.     

Experimental studies on steel corrosion in Pb-Bi with steam injection

Experimental studies on steel corrosion were performed in simulated PBWFR (Pb-Bi cooled direct contact boiling water fast reactor) coolant environment. Some candidate steels of high Cr contents were immersed in steam-injected liquid Pb-Bi pool to investigate how their Cr contents and oxygen potential in Pb-Bi or (P_H2/P_H2O) in the steam influence their corrosion behaviors at temperature range of the reactor operation. Test specimens were made from eight types of steel with Cr contents ranged from 8 to 18%. The experiments were conducted by exposing these specimens to Pb-Bi pool where steam was injected. (P_H2/P_H2O) ratios of the steam were employed as experimental parameter, ranged from < 3×10⁻⁷ to 1×10⁻⁵ to control oxygen potential of Pb-Bi. Exposure temperatures were studied of 400, 450 and 500°C. It was found that 12Cr steel (HCM12/HCM12A) was the most resistant to corrosion and therefore a candidate reactor material.     

Mechanism of dynamic strain aging and characterization of its effect on the low-cycle fatigue behavior in type 316L stainless steel

Low-cycle fatigue tests were carried out in air in a wide temperature range from 20 to 650 °C with strain rates of 3.2 × 10⁻⁵–1 × 10⁻² s⁻¹ for type 316L stainless steel to investigate dynamic strain aging (DSA) effect on the fatigue resistance. The regime of DSA was evaluated using the anomalies associated with DSA and was in the temperature range of 250–550 °C at a strain rate of 1 × 10⁻⁴ s⁻¹, in 250–600 °C at 1 × 10⁻³ s⁻¹, and in 250–650 °C at 1 × 10⁻² s⁻¹. The activation energies for each type of serration were about 0.57–0.74 times those for lattice diffusion indicating that a mechanism other than lattice diffusion is involved. It seems to be reasonable to infer that DSA is caused by the pipe diffusion of solute atoms through the dislocation core. Dynamic strain aging reduced the crack initiation and propagation life by way of multiple crack initiation, which comes from the DSA-induced inhomogeneity of deformation, and rapid crack propagation due to the DSA-induced hardening, respectively.     

Elemental dissolution study of Pu-bearing borosilicate glasses

Single-pass flow-through tests were conducted to study the effects of self-radiation damage from alpha decay on dissolution kinetics of three radiation-aged Pu-bearing (1 mass% PuO₂) borosilicate glasses over a pH interval of 9–12 at 80–88 °C. The chemical compositions of the glasses were identical except the ²³⁹Pu/²³⁸Pu isotopic ratio, which was varied to yield accumulated doses of 1.3 × 10¹⁶, 2.9 × 10¹⁷ and 2.6 × 10¹⁸ -decays/g at the time of testing. Release of Al, B, Cs, Na, Si and U to solution increased with increasing pH, whereas Ca, Pu and Sr were invariant over the pH interval. Average dissolution rates, based on B release, were identical within experimental uncertainty for all three glass compositions and increased from 0.17 ± 0.07 at pH(23 °C) 9 to 10.6 ± 2.7 (g/(m² d¹)) at pH(23 °C) 12. Release rates of Pu were 10²- to 10⁵-fold slower compared to all other elements and were not affected by isotopic composition, self-radiation damage sustained by the glass, or pH. These data demonstrate that self-radiation damage does not affect glass dissolution rates, despite exposure to internal radiation doses for >20 years.     

Study of steam, helium and supercritical CO2 turbine power generations in prototype fusion power reactor

Power generation systems such as steam turbine cycle, helium turbine cycle and supercritical CO₂ (S-CO₂) turbine cycle are examined for the prototype nuclear fusion reactor. Their achievable cycle thermal efficiencies are revealed to be 40%, 34% and 42% levels for the heat source outlet coolant temperature of 480 °C, respectively, if no other restriction is imposed. In the current technology, however, low temperature divertor heat source is included. In this actual case, the steam turbine system and the S-CO₂ turbine system were compared in the light of cycle efficiency and plant cost. The values of cycle efficiency were 37.7% and 36.4% for the steam cycle and S-CO₂ cycle, respectively. The construction cost was estimated by means of component volume. The volume became 16,590 m³ and 7240 m³ for the steam turbine system and S-CO₂ turbine system, respectively. In addition, separation of permeated tritium from the coolant is much easier in S-CO₂ than in H₂O. Therefore, the S-CO₂ turbine system is recommended to the fusion reactor system than the steam turbine system.     

Surface morphology of the Ar ion-irradiated Ag/Si(1 1 1) system

In the present study, a 500 Å thin Ag film was deposited by thermal evaporation on 5% HF etched Si(1 1 1) substrate at a chamber pressure of 8×10⁻⁶ mbar. The films were irradiated with 100 keV Ar⁺ ions at room temperature (RT) and at elevated temperatures to a fluence of 1×10¹⁶ cm⁻² at a flux of 5.55×10¹² ions/cm²/s. Surface morphology of the Ar ion-irradiated Ag/Si(1 1 1) system was investigated using scanning electron microscopy (SEM). A percolation network pattern was observed when the film was irradiated at 200°C and 400°C. The fractal dimension of the percolated pattern was higher in the sample irradiated at 400°C compared to the one irradiated at 200°C. The percolation network is still observed in the film thermally annealed at 600°C with and without prior ion irradiation. The fractal dimension of the percolated pattern in the sample annealed at 600°C was lower than in the sample post-annealed (irradiated and then annealed) at 600°C. All these observations are explained in terms of self-diffusion of Ag atoms on the Si(1 1 1) substrate, inter-diffusion of Ag and Si and phase formations in Ag and Si due to Ar ion irradiation.     

The excitation power density effect on the Si nanocrystals photoluminescence

In the present work we have studied the photoluminescence (PL) behavior from Si nanocrystals (NCs) as a function of the excitation power density and annealing time. The NCs were produced in a SiO₂ matrix by Si implantations from room temperature (RT) up to 700 °C, followed by post-annealing in N₂ atmosphere at high temperature. With this aim we have changed the excitation power density (from 2 × 10⁻³ W/cm² up to 15 W/cm²) and the annealing time (from 10 min up to 15 h). The strong PL signal, which at 15 W/cm² is composed by a single-peak structure (650–1000 nm) centered at around 780 nm, expands up to 1200 nm showing a two-peak structure when measured at 20 × 10⁻³ W/cm². The peak structure located at the short wavelength side is kept at 780 nm, while the second peak, starting at around 900 nm, redshifts and increases its intensity with the implantation temperature and annealing time. The effect of the annealing time on the PL spectra behavior measured at low excitation power agrees by the first time with the Si NC growth according to quantum confinement effects.