Search

The performance of heat exchangers in the pressurized water test loop is a critical factor in ensuring the achievement of irradiation parameters for fuel assemblies and the safety of experimental operations. The effect of the heat exchange area margin on the heat exchangers in the pressurized water test loop for the fuel assembly during the steady-state irradiation is analyzed. Additionally, optimization methods for determining the margin of heat exchange area and corresponding design strategies are further investigated. It shows that the effect of the heat exchange area margin on the heat exchange power is less affected by the inlet temperature of the primary water and is primarily influenced by the flow rate of the primary water. A decrease in the flow rate of the primary water reduces the compensatory effect of the cooling section on power and enhances the weakening effect of the regeneration section on power. Meanwhile, the correspondence between the margin of the regeneration section and the cooling section, established based on design conditions, can be applicable when there are changes in the inlet temperature of the primary water, but it is not suitable when there are changes in the flow rate of the primary water. When the flow rate of the primary water decreases, the cooling section margin required to compensate for the decrease in power caused by the regeneration section margin will increase significantly. In addition, short-circuiting the heat exchange tubes in the regeneration section can effectively enhance the heat transfer capability. Furthermore, setting the heat exchange area margins of the regeneration and cooling sections to zero can serve as a termination condition for iterative calculations in the verification of regenerative heat exchangers under off-design conditions. [ABSTRACT FROM AUTHOR]

The performance of heat exchangers in the pressurized water test loop is a critical factor in ensuring the achievement of irradiation parameters for fuel assemblies and the safety of experimental operations. The effect of the heat exchange area margin on the heat exchangers in the pressurized water test loop for the fuel assembly during the steady-state irradiation is analyzed. Additionally, optimization methods for determining the margin of heat exchange area and corresponding design strategies are further investigated. It shows that the effect of the heat exchange area margin on the heat exchange power is less affected by the inlet temperature of the primary water and is primarily influenced by the flow rate of the primary water. A decrease in the flow rate of the primary water reduces the compensatory effect of the cooling section on power and enhances the weakening effect of the regeneration section on power. Meanwhile, the correspondence between the margin of the regeneration section and the cooling section, established based on design conditions, can be applicable when there are changes in the inlet temperature of the primary water, but it is not suitable when there are changes in the flow rate of the primary water. When the flow rate of the primary water decreases, the cooling section margin required to compensate for the decrease in power caused by the regeneration section margin will increase significantly. In addition, short-circuiting the heat exchange tubes in the regeneration section can effectively enhance the heat transfer capability. Furthermore, setting the heat exchange area margins of the regeneration and cooling sections to zero can serve as a termination condition for iterative calculations in the verification of regenerative heat exchangers under off-design conditions.

Nowadays, the WWR-K research reactor is operated with low-enriched uranium fuel and a beryllium neutron reflector. Irradiation positions located in the core make it possible to irradiate objects no larger than 50 mm in size. In the reactor tank, behind the beryllium reflector, there are irradiation positions with a diameter of 200 mm, but in them, in comparison with irradiation positions in the core, the neutron intensity is lower. This motivated us to carry out computational studies on the generation of a large diameter neutron trap in the center of the WWR-K reactor core. The paper presents the results of neutron-physical calculations on the generation of a neutron trap in the center of the WWR-K reactor core. The generation of a large diameter neutron trap in the core of the WWR-K reactor will make it possible to carry out radiation tests of large objects. For this, it is necessary to change the layout of the core. The influence of the transformation of the core on the experimental characteristics of the WWR-K reactor is shown.

• Analyzed of neutron characteristics and material damage on neutron absorber by the irradiation test using HANARO research reactor. • Clarified the model specification of the HANARO reactor core and irradiation system. • Described the characteristics of boron carbide neutron absorber and its nuclear characteristics. • Explained the correlation between the nuclear absorption characteristics and material damage of neutron absorption materials. This study aims to verify the nuclear characteristics and material integrity of neutron absorbers through neutron irradiation using the HANARO research reactor. The nuclear reactions and material damage of boron carbide in the neutron irradiation environment of the external region of the reactor core were evaluated based on the Monte Carlo method. The irradiation conditions of the reactor core were assumed based on equilibrium core operation with a reactor power of 30 MW. In order to evaluate material damage over a one-year period, it was assumed that the concentrated boron carbide undergoes a 2 % boron depletion every 100 days. Reactivity test was assessed to verify the appropriateness of the assumptions established for the irradiation test. As a result, the reactivity level showed approximately 1.06 mk, which sufficiently proves its lower level compared to the criteria of 12.5 mk. In order to analyze the physical characteristics of neutron absorbers influencing material damage, the group fluxes and energy-dependent neutron spectra of the target material were evaluated. The neutron flux in Group I appears low, while the epithermal group flux is observed relatively high. The specimen material closer to the core generally showed higher flux, indicating a maximum difference of about 3 × 1013 #/cm2·sec compared to the minimum one. The 3rd specimen showed the highest in flux, while 7th specimen is the lowest. The shape of all the dpa values showed similar trends to the pattern of neutron absorption reaction rates. The neutron absorption was significant up to 100 days, and after this point, they showed gradual reductions in the damage increase rate. The total dpa values linearly increased along with the irradiation time, and showed about 12–13 dpa over one year of EFPD. The results of this study can be utilized as fundamental data for confirming changes in material damage due to high neutron irradiation on neutron absorbers in connection to nuclear characteristics. [ABSTRACT FROM AUTHOR]

When performing remote tasks using robots in nuclear power plants, a 3D shape measurement system is advantageous in improving the efficiency of remote operations by easily identifying the current state of the target object for example, size, shape, and distance information. Nuclear power plants have high-radiation and underwater environments therefore the electronic parts that comprise 3D shape measurement systems are prone to degradation and thus cannot be used for a long period of time. Also, given the refraction caused by a medium change in the underwater environment, optical design constraints and calibration methods for them are required. The present study proposed a method for developing an underwater 3D shape measurement system with improved radiation tolerance, which is composed of commercial electric parts and a stereo camera while being capable of easily and readily correcting underwater refraction. In an effort to improve its radiation tolerance, the number of parts that are exposed to a radiation environment was minimized to include only necessary components, such as a line beam laser, a motor to rotate the line beam laser, and a stereo camera. Given that a signal processing circuit and control circuit of the camera is susceptible to radiation, an image sensor and lens of the camera were separated from its main body to improve radiation tolerance. The prototype developed in the present study was made of commercial electric parts, and thus it was possible to improve the overall radiation tolerance at a relatively low cost. Also, it was easy to manufacture because there are few constraints for optical design.

Copyright of Large Electric Machine & Hydraulic Turbine is the property of Large Electric Machine & Hydraulic Turbine Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

A neutron irradiation test of static random access memories (SRAMs) was performed using a spallation neutron source at Materials and Life Science Experimental Facility (MLF) in the Japan Proton Accelerator Research Complex (J-PARC). The probability of neutron-induced single-event upsets (SEUs) was measured for 65-nm bulk and silicon on thin buried oxide (SOTB) SRAMs under neutron irradiation at the BL10 experimental facility. The measured SEU data were compared with the previous data of the same SRAMs which were measured at other irradiation facilities having different neutron spectra. The differences in the operating voltage dependence of the measured SEU probabilities are discussed with particular attention to the impact of irradiation side on SEUs. The particle and heavy ion transport code system (PHITS) simulation based on the simple sensitive volume model qualitatively reproduced the operating voltage dependence seen in the measured ratio of SEUs for the Bulk SRAM between the resin side and board side irradiations under different neutron fields. [ABSTRACT FROM AUTHOR]

Negative and positive muon irradiation tests of static random access memories (SRAMs) were performed at Muon Science Innovation Channel (MuSIC) of Research Center for Nuclear Physics (RCNP), Osaka University. The muon-induced single event upset (SEU) cross sections for 65-nm bulk SRAMs were measured. The SRAM device was irradiated by the muon beams with average momenta ranging from 37.8 to 41.0 MeV/c at the beam exit. The incident muon fluence was measured using a one-by-one muon detection system with a plastic scintillator by taking advantage of the direct current (dc) muon beam and reliable absolute values of SEU cross sections were obtained. In addition to the measurement of the SEU cross sections, we measured muonic X-rays from the irradiated device to investigate the relation between the measured SEU cross section and the emission rate of muonic X-rays associated with the stopping position of the muons as a function of incident muon momentum. The result suggested that negative muons have a significant effect on SEUs by the emission of secondary light ions via negative muon capture reaction when negative muons stop near sensitive volumes (SVs) in SRAMs. [ABSTRACT FROM AUTHOR]

Soft error originating from cosmic ray is a serious concern for reliability demanding applications, such as autonomous driving, supercomputer, and public transportation system. Also, as the number of electronic devices increases, consumer electronics may require higher reliability than ever. On the other hand, device miniaturization and lower voltage operation degrade the immunity of SRAM and flip-flops in VLSI, and then soft error countermeasures will be demanded in more and more products. This paper characterizes and discusses soft error rates of SRAM and flip-flops in the terrestrial environment. Measured soft error rates of bulk and FDSOI SRAMs are presented. Also, the multiplicity and mechanism of multiple cell upsets (MCUs), which can spoil error correction code (ECC), are discussed. Flip-flops are also sensitive to radiation, but its protection is not well established since ECC cannot be applied. This paper reviews redundant FFs that are developed aiming at higher radiation immunity and demonstrates the resilience improvement with irradiation test results. Also, the importance of layout design is pointed out with a comparative study. Simulation, on the other hand, is a key technology to understand the soft error mechanism and guide radiation aware design for higher reliability. This paper outlines a physics-based multi-scale Monte Carlo simulation framework tailored for soft error simulation. The simulation flow and the model construction are explained, and some important implications are derived from the simulation results that assume 65-nm to 25-nm SRAMs. Finally, this article touches on future trends regarding device structure and overlooked secondary cosmic rays. The advantage and unique features of FinFET recently reported in literature are introduced. Also, muon-induced soft error is discussed focusing on the difference between positive and negative muons. [ABSTRACT FROM AUTHOR]

We investigated the effects of irradiation beam conditions on the performance degradation of silicon and triple-junction solar cells for use in space. The fluence rates of electron and proton beams were varied. Degradation did not depend on the fluence rate of protons for both cells. A higher fluence rate of electrons caused greater degradation of the Si cell, but the dependence was due to the temperature increase during irradiation. Two beam-area expansion methods, defocusing and scanning, were examined for proton irradiation of various energies (50 keV–10 MeV). In comparing the output degradation from irradiation with defocused and scanned proton beams, no significant difference in degradation was found for any proton energy. We plan to reflect these findings into ISO standard of irradiation test method of space solar cells.

Highlights • Tritium production using high temperature gas-cooled reactor (HTGR) is considered. • An irradiation test on High Temperature engineering Test Reactor (HTTR) is assumed. • The Li rod suited to the HTTR and produces 30 g of tritium in a year was designed. • The scenario of the irradiation test to confirm the Li rod performance was presented. • Li rod irradiation test to demonstrate the tritium production in the HTGR was proposed. Abstract Large quantities of tritium are required to start up fusion reactors and conducts engineering tests using tritium for a fusion blanket system. However, tritium is very rare and kg orders of tritium must be produced artificially. Tritium production, by 6Li(n,α)T reaction using a high temperature gas-cooled reactor (HTGR) has been proposed. This method considers the loading of Li rods into burnable poison holes in the HTGR. In this paper, the Li rod suited for use in the High Temperature engineering Test Reactor (HTTR) was designed, and tritium production and leakage from Li-rod capsules were evaluated by adjusting the thicknesses of LiAlO 2 , alumina, and Zr layers. An irradiation test scenario to be conducted in the HTTR for demonstration of the Li rod's tritium production and containment performance was presented. [ABSTRACT FROM AUTHOR]

Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U–Zr fuel is a driver for the initial core of the PGSFR, and U–transuranics (TRU)–Zr fuel will gradually replace U–Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U–Zr fuel, work on U–Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U–TRU–Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic–martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea) in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4–5 g-U/cm3 were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr), additional protective coatings (silicide or nitride), and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

• This article provides results of comprehensive study on irradiated MOX fuel properties. • MIMAS MOX fuel was studied with references of SBR MOX and UO 2 fuels. • MOX fuels with a Pu content up to 14% and a burnup up to 74 GWd/tU were studied. • This study includes thermal conductivity, FP gas distribution, and volume change. • SIMS analysis confirmed that Pu agglomerates retain more FP gas bubbles than matrix. This paper presents the major results and outcome from a study on the properties of uranium and plutonium mixed oxide (MOX) fuel, aiming to expand the technical bases concerning MOX fuels and to review the previous reports on MOX fuel properties issued by the former Japanese nuclear safety commission. An irradiation test was conducted on MOX fuels fabricated through different processes, up to a local burnup of about 74 GWd/t, which is beyond the current Japanese licensing limit of 62 GWd/t in pellet burnup. Another irradiation test was conducted on MOX fuels with an initial Pu content of 14.3%, which also exceeds the Japanese licensing limit of 13%. After the irradiation tests, MOX fuel properties were studied using different measuring methods in the post-irradiation examination. The results showed no obvious deviation from the MOX fuel properties in the previous reports, while various new useful results were obtained. Elemental analysis via secondary ion mass spectrometry on the MOX fuels revealed that Pu rich agglomerates retained more fission gas bubbles than the matrix. Fuel centerline temperature measurements during irradiation and thermal diffusivity measurements using the laser flash method revealed that the thermal conductivity of MOX fuel was lower than that of UO 2 fuel prior to irradiation, whereas it became comparable with that of the UO 2 fuel by irradiation. Volume change measurements using both a non-destructive method and destructive methods indicated that the swelling rates of the MOX fuel with Pu rich agglomerates were slightly higher than the UO 2 swelling rate in a reference. [Display omitted] [ABSTRACT FROM AUTHOR]

To measure the heat generation rate of nuclear fuel rods during an irradiation test, the temperature deviation of the coolant that passes by the fuel rod and its flow velocity need to be measured. Although the temperature of the coolant can be measured by thermocouples instrumented in the nuclear fuel test rig, the flowrate of the coolant is difficult to measure owing to the narrow space of the test rig. Therefore, a noise analysis technique using fluctuation signals from thermocouples installed at both end parts of the fuel rod has been developed by several advanced groups. Although the noise analysis technique was also developed at the Korea Atomic Energy Research Institute, out of pile test results showed that the accuracy of the coolant flow measurement was more than 40%. In this study, the ground pattern of the control board and earth connection are enhanced to eliminate external noise. In addition, a heater unit is installed in the coolant flow simulator to obtain a strong fluctuation signal from the coolant. Thus, the coolant flow simulator is improved, and the accuracy of the coolant flow measurement shows an error of less than 5%. [ABSTRACT FROM AUTHOR]

The heat generation rate is one of the most important characteristics in evaluating the performance of nuclear fuel. The heat generation rate of a nuclear fuel rod can be measured indirectly by calculating the heat flux that coolant absorbs while passing by the nuclear fuel rod. To calculate the heat flux from the fuel rod, it is necessary to measure the temperature deviation and flow velocity at both end parts of the fuel rod. Thus, measuring the flow velocity in the test rig during the irradiation test of nuclear fuel is important to measure the heat generation rate. In this study, a cross correlation technique that uses signals obtained from two thermocouples instrumented at both end parts of the fuel rod has been developed to measure the flow velocity of coolant in the test rig. [ABSTRACT FROM AUTHOR]

Neutronic and thermal hydraulic characteristics of U–Mo fuel mini plates irradiated in the HANARO reactor were analyzed for the safety assessment of these plates. A total of eight fuel plates were double-stacked; each stack contained three 8.0 gU/cc U–7Mo fuel plates and one 6.5 gU/cc U–7Mo fuel plate. The neutronic and thermal hydraulic analyses were carried out using the MCNP code and TMAP code, respectively. The core status used in the study was the equilibrium core, and four Control Absorber Rod (CAR) locations were considered: 350 mm, 450 mm, 550 mm, and 650 mm away from the bottom of the core. For the fuels in the lower stack, the maximum heat flux was found at the CAR located at 450 mm. For the fuels in the upper stack, the maximum heat flux was found at the CAR located at 650 mm. The axial power distributions for the upper and lower stacks were selected on the basis of thermal margin analyses. A mock-up irradiation target assembly was designed and tested at the out-of-pile test facility to measure the flow rate through the irradiation site, given that the maximum flow rate through the irradiation site at the HANARO reactor is limited to 12.7 kg/s. For conservative analyses, measurement and correlation uncertainties and engineering hot channel factors were considered. During normal operation, the minimum ONB temperature margins for the lower and upper stacks are 41.6 °C and 31.8 °C, respectively. This means that boiling does not occur. However, boiling occurs during the limiting accidents. Nevertheless, the fuel integrity is maintained since the minimum DNBR are 1.96 for the Reactivity Insertion Accident (RIA) and 2.10 for the Locked Rotor Accident (LRA). From the neutronic and thermal hydraulic analyses of the U–Mo fuel mini plates irradiated in the HANARO reactor, it is concluded that boiling does not occur during normal operation and that the fuel integrity is maintained during limiting accidents. [ABSTRACT FROM AUTHOR]

• Two different types of piezoelectric actuators were developed for ITER diagnostics to operate in vacuum under high thermal and neutron fluxes. • Results of irradiation test of the piezoelectric actuators are analyzed. • Four types of the most common piezoelectric ceramic materials were tested. • Effects of radiation on piezoelectric ceramics are discussed as a challenge for piezoelectric actuators at ITER. • It is shown that depolarization of the piezoelectric ceramics related to irradiation and multiple baking in ITER can be reversed remotely. As a part of ITER diagnostics design and development, ultrasonic and stick-slip piezoelectric motors and several types of piezoelectric ceramics were tested under severe neutron and gamma ray irradiation (∼ 1019 per cm2 at E > 0.1 MeV) in a pool-type fission reactor. The fission neutron and gamma ray spectra were corrected to comply with conditions expected under ITER divertor cassettes. Taking into account possible neutron-induced amorphization of crystal structure, four types of piezoelectric ceramics were pre-selected. To restore the electronic structure of piezoelectric ceramic damaged by gamma rays, recovery polarization (i.e. heating while applying reverse bias) was suggested. The motors and piezoceramic samples were assessed for performance before and after irradiation as well as after recovery polarization. Recovery polarization was performed at 330 °C as it is expected close to the divertor cassettes during baking. Two most promising types of piezoelectric ceramics survived multiple cycles of depolarization/polarization at both 330 °C and 200 °C as appropriate for baking ITER divertor and vacuum vessel, respectively. All the experiments and procedures are described in detail. [ABSTRACT FROM AUTHOR]

This paper presents some of the key technologies in the area of fuel performance that Korea Atomic Energy Research Institute (KAERI) has developed for a dual-cooled annular fuel, which should be available before the annular fuel can be considered to be used in a commercial nuclear power plant. First, considering the characteristics of the annular fuel--that it has two coolant channels, outer and inner, and also two gaps between the pellet and cladding--KAERI has developed a computer code DUOS that calculates temperature, swelling, densification, and stress and strain in the annular fuel. The DUOS code was verified by comparing it with either ABAQUS or analytical solutions. The first irradiation test of sintered annular fuel pellets with different initial densities was performed in the HANARO reactor up to a pellet burnup of 10.9 MWd/kg U and then subjected to postirradiation examination. Gamma scanning along the axial direction of the irradiated fuel rods showed the geometrical integrity of the annular fuel pellets, ruling out the possibility that fragmented annular pellet cracks could move down along the axial direction of the fuel rod and hence the pellet stack length could be reduced. Macroscopy of the annular fuel pellets revealed many radial and circumferential cracks that could lead to different outer and inner gap sizes along the axial direction of the annular fuel rod, which would suggest that heat transfer to both the outer and inner coolant channels during the irradiation of annular fuel rods would depend on the axial profile of the two gaps along the axial direction. The swelling rate derived from density measurement of the annular fuel pellets with 98.0% theoretical density was 0.25 to O. 60 vol % per 10 MWd/kg U, corresponding to the one observed for solid fuel pellets irradiated at low temperature. [ABSTRACT FROM AUTHOR]

HANARO (High flux Advanced Neutron Application ReactOr) is an open-tank-in-pool-type, multipurpose research reactor. Since the commencement of HANARO operations in 1995, a significant number of experimental facilities have been developed and installed, and continued efforts to develop more facilities are in progress. Owing to a stable operation of the reactor and its rapid utilization, more experimental facilities are being continuously added to satisfy various fields of study and diverse applications. The equipment is classified into a capsule and a fuel test loop (FTL) for the irradiation tests of the nuclear fuels and materials in HANARO. Capsules for the irradiation tests of nuclear fuels in HANARO, under the conditions of its coolant and materials in dry conditions, have been developed. The FTL can conduct an irradiation test of a nuclear fuel under the operating conditions of commercial nuclear power plants. Three test fuel rods can be irradiated in HANARO by using the FTL. Some instrumentation such as thermo-couples, linear variable differential transformers, small heaters, fluence monitors, and self-powered neutron detectors are used to perform the irradiation tests by using these irradiation devices. In this paper, the status of HANARO and its irradiation devices including the instrumentations and some results from these instrumentations during irradiation tests are described. [ABSTRACT FROM AUTHOR]

Attrition-milling technology for fabricating mixed oxide (MOX) fuel was developed to mix the plutonium in UO2 fuels as homogeneously as possible. The fabricated MOX fuels were instrumented with temperature and pressure gauges that enabled one to measure the fuel temperature and rod internal pressure online. An irradiation test in the Halden reactor was performed to investigate the in-pile behavior of the fabricated MOX fuel. The irradiation of 1020 effective full-power days was successfully accomplished with good integrity of the test fuel rods. The rod average burnup reached -50 MWd/ kg HM, and the measured fuel centerline temperature was ∼1000°C for the MOX fuels. A significant fission gas release was observed due to the high power level. The online measured in-pile performance data of the two attrition-milled MOX fuel rods were analyzed and compared with the fuel performance code COSMOS. COSMOS simulated the fuel centerline temperature and rod internal pressure for both MOX fuel rods. The analysis by COSMOS showed good agreement with the online measured in-pile behavior of MOX fuel. [ABSTRACT FROM AUTHOR]

Irradiation test results demonstrate the validity of a scan FF technology for separately evaluating SET and SEU soft error rates (SERs) in logic VLSI systems. The SET and SEU soft errors mean the upset caused by latching an SET pulse that originates in combinational logic cell blocks and the upset caused by a direct ion hit to the FF, respectively. A test chip is fabricated using a 0.2-μm fully-depleted silicon-on-insulator standard cell library and irradiated under an LET of 40 MeV-cm2 /mg. The SET and SEU soft error rates are successfully measured by the scan FFs on the test chip. A theoretical SET SER estimation from measured SET-pulse widths is also experimentally validated. [ABSTRACT FROM AUTHOR]

A scan flip-flop (FF) is designed to observe both single event transient (SET) and single event upset (SEU) soft errors in logic VLSI systems. The SET and SEU soft errors mean the upset caused by latching an SET pulse that originates in combinational logic blocks and the upset caused by a direct ion hit to the FF, respectively. An irradiation test method using the scan FF is proposed to obtain SET and SEU soft-error rates at each FF distributed in logic VLSI systems. A test chip is designed using a 0.2-µm fully-depleted silicon-on-insulator standard cell library. The basic concepts have been validated with Verilog timing simulations. The cell-level implementation costs of the proposed scan FF are estimated to be reasonable. [ABSTRACT FROM AUTHOR]

This paper describes an irradiation test of highpower-density internally and externally cooled annular fuel samples in the 5-MW Massachusetts Institute of Technology (MIT) research reactor MITR-II. The design of the irradiation facility is briefly reviewed, with an emphasis on the thermal-hydraulic behavior of the irradiation capsules. The irradiation test is described, including the thermal history of the two irradiated samples. A discussion of the observed asymmetrical temperature profiles is provided. Results of preliminary postirradiation examination consisting of collimated gamma scans of the irradiation capsules to confirm burnup estimates and estimate fission gas release (FGR) are also presented. It is concluded that the vibropacked fuel samples' FGR is below 1%, and that is within the predictable range by a specially equipped FRAPCON model. [ABSTRACT FROM AUTHOR]

Monitored Drift Tube chambers (MDT) are the precision tracking detectors in the muon spectrometer of the ATLAS detector at the Large Hadron Collider (LHC). From simulations of the LHC operating conditions, the count rate in the inner barrel chamber of the muon spectrometer is expected to reach 500 Hz/cm² and can be ten times higher in the luminosity upgrade under discussion (SLHC). A degradation of the performances of drift tube chambers after extended operation in high-rate environment is usually caused by the formation of deposits on the anode wire. To study MDTs aging, three bundles of short tubes equipped like a final MDT chamber were irradiated in different periods with neutrons and photons using the TAPIRO nuclear reactor and the CALLIOPE gamma facility at the ENEA-Casaccia Research Center (Italy). The MDT performances with respect to accumulated dose were studied with cosmic rays and results on drift and tracking properties are presented. [ABSTRACT FROM AUTHOR]

Nuclear robots must be hardened against radiation for effective operation in high-radiation environments. These robots require special components that usually have few functions, high costs and long delivery times. In the present study, we propose a modulated and hierarchical electrical circuit system for robot components such as actuators. The hierarchical system is set into a design that can withstand radiation without reducing or limiting the functionality of the robot. The hierarchical system is achieved by minimizing the variety of integrated circuit (IC) components to be used and selecting specific ICs that require radiation testing. These ICs are defined by spotted ICs with regard to the integration density and operational speed. In some cases, it is possible to select off-the-shelf IC components, but certain key IC components are needed for on-line irradiation tests, depending on robot operation reliability and the radiation environment. Since the serial numbers of off-the-shelf IC components are random, it is important that a sufficiently large number of IC components are tested under the operating conditions for statistical evaluation. If the radiation lifetime is random, then the radiation lifetime of the sorted IC components can be estimated statistically. Here, the robot user decides the operation based on this lifetime. By doing so, the experimental cost can be minimized in accordance with the proposed method at the user's operational risk. Of course, if on-line irradiation tests are performed for various types of IC components, the radiation lifetime of the robot can be predicted with a relatively high accuracy. As a result, the design of a robot that operates under certain radiation conditions, while having a certain lifetime and cost, is possible. This design technique was applied to a nuclear disaster prevention robot and the radiation lifetime of the spotted IC components was shown to be distributed randomly, indicating the validity of the proposed method. [ABSTRACT FROM AUTHOR]

This study describes the mechanical compatibility of the direct use of spent pressurized water reactor fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel and fuel handling system in the reactor core by both experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design, which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high-power and high-burnup conditions even though some material properties, such as the thermal conductivity, are a little lower compared to the uranium fuel. However, it is required that the current DUPIC fuel design be changed slightly to accommodate the high internal pressure of the fuel element. It is also strongly recommended that more irradiation tests of the DUPIC fuel be performed to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor. [ABSTRACT FROM AUTHOR]

We present the results of a global irradiation test of full size triple-GEM detectors operated with a CF4-based gas mixture. This study has been performed in the framework of an R&D activity on detectors for the innermost region of the first muon station of the LHCb experiment. The prototypes have been irradiated at the Calliope facility of the ENEA-Casaccia with a high intensity 1.25 MeV gamma from a 60Co source. After the irradiation test the detectors' performances have been measured with X-rays and with a 3 GeV pion beam at CERN. A SEM analysis on several samples of the detectors has been performed to complete the understanding of the physical processes occurring in the GEM detector during the strong irradiation [ABSTRACT FROM PUBLISHER]

Rotating disk centrifugal atomization technology has been utilized to fabricate uranium silicide (U3Si, U3Si2) and U-Mo nuclear fuel powders having high uranium content per unit volume for high performance research reactor fuels. Atomized nuclear fuel powders have characteristics of a spherical shape with narrow size distribution, small specific surface area and high purity. The heat treatment time for the formation of U3Si by a peritectoid reaction was reduced from about 72 hours for comminuted powders to about 6 hours for atomized powders due to more rapid solidification of atomized powder with a finer microstructure. The homogeneity of fuel particles in fuel meats was improved by mixing atomized fuel powders with Al powders using a V-shaped tumbler mixer. The atomized powders with a spherical shape and smooth surface were extruded under lower force than the comminuted powders with angular rough surfaces. The dispersed fuel meat with atomized powders resulted in an increased fuel powder loading density and higher thermal conductivity in the heat flow direction. The thermal swelling of dispersed fuel meat decreased due to the reduced specific surface area of spherical atomized nuclear powder. The atomized U-10wt.%Mo dispersion fuel showed less bubble formation than the comminuted fuel after an irradiation test with 40% burnup. [ABSTRACT FROM AUTHOR]