Prospects for the development of neutron-activation analytical facilities based on powerful antimony-beryllium sources

Conclusions When using antimony-beryllium neutron sources for activation analysis of the composition of a substance, the problem of the analytical monitoring of many elements can be solved with a limit of determination of 10^–3–10^–5%. Based on the data obtained by the authors concerning the spatial distribution of the neutrons from a¹²⁴Sb–Be-source in different moderators, a beryllium-graphite assembly with a powerful¹²⁴Sb source has been designed, manufactured, and introduced into operation, for neutron-activation analysis.As applicable to the¹²⁴Sb–Be-graphite assembly, a procedure has been developed for the neutron-activation determination of gold, with a limit of determination of 2·10^–5%. The possible limits of the neutron-activation determination of certain other elements have been estimated. In order to ensure operation of the facility with recharging of the source once in 6 months, it is advisable to carry out the preparation of a source in a neutron flux with a density of 3·10^–13 neutrons/ (cm²·sec) during 30–50 days (for a mass of metallic antimony of 500 g) with subsequent two-week cooling in order to reduce the¹²²Sb activity. Initial data have been obtained for the design of a transportation container for powerful¹²⁴Sb sources.Translated from Atomnaya Énergiya, Vol. 53, No. 4, pp. 255–260, October, 1982.     

Implementation and assessment of the renormalization group (RNG), quadratic and cubic non-linear eddy viscosity k– models in GOTHIC

In GOTHIC, the standard k– model is used to model turbulence. However, for practical reasons, one usually employs relatively large meshes near physical boundaries (walls). In an attempt to enhance the turbulence modelling in the code for simulation of mixing driven by highly buoyant discharges, in the framework of this simplified approach appropriate for containment analysis codes, we have implemented three additional models which are modifications/extensions of the standard k– model: the renormalization group k– model, and the non-linear (quadratic and cubic) eddy viscosity k– models. These models which for the time being, are only implemented in the ‘gas’ phase, were tested with different simple test-problems and their predictions were compared to the corresponding ones of the standard k– model. Furthermore, a simple study was performed to assess the sensitivity of the predictions to the mesh size.     

Averaging methods of the gap heat transfer coefficients and the loss form coefficients of nuclear reactor cores loaded with different fuel bundles

When performing transient analysis in heterogeneous nuclear reactors loaded with different types of fuel bundles is necessary to model the reactor core by a few representative fuel elements with average properties of a region containing a large number of fuel elements. The properties of these representative fuel bundles are obtained by averaging the thermal–hydraulic properties of the fuel elements contained in each region. In this paper we study the different ways to perform the averaging of the thermal–hydraulic properties that can have an influence on the transient results for licence purposes. Also we study the influence of the different averaging methods on the peak clad temperature (PCT) evolution for a LOCA, and on the critical power ratio (CPR) in the hot channels for a turbine trip transient without bypass credit.     

Thermal compatibility of centrifugally atomized U–Mo powders with aluminium in a dispersion fuel

The thermal compatibility of centrifugally atomized U–Mo alloys with aluminum has been studied. Samples of extruded dispersions of 24 vol.% spherical U–2 wt.% Mo and U–10 wt.% Mo powders in an aluminum matrix were annealed for over 2000 h at 400°C. No significant dimensional changes occurred in the U–10 wt.% Mo/aluminum dispersions. The U–2 wt.% Mo/aluminum dispersion, however, increased in volume by 26% after 2000 h at 400°C. This large volume change is mainly due to the formation of voids and cracks resulting from nearly complete interdiffusion of U–Mo and aluminum. Interdiffusion between U–10 wt.% Mo and aluminum was found to be minimal. The different diffusion behavior is primarily due to the fact that U–2 wt.% Mo decomposes from an as-atomized metastable γ-phase (bcc) solid solution into the equilibrium α-U and U₂Mo two-phase structure during the experiment, whereas U–10 wt.% Mo retains the metastable γ-phase structure throughout the 2000 h anneal and thereby displays superior thermal compatibility with aluminum compared to U–2 wt.% Mo.     

Dynamic analysis of a nuclear reactor with fluid–structure interaction: Part I: Seismic loading, fluid added mass and added stiffness effects

The present paper is related to the dynamic (seismic) analysis of a naval propulsion ground prototype (land-based) nuclear reactor with fluid–structure interaction modelling. Many numerical methods have been proposed over the past years to take fluid–structure phenomenon into account in various engineering domains, among which nuclear engineering in seismic analysis. The purpose of the present paper is to make a comparative study of these methods on an industrial case, namely the pressure vessel and internals of a nuclear reactor. A simplified model of the pressure vessel and the internal structure is presented; fluid–structure interaction is characterised by added mass, added stiffness and coupling effects. The basic principles of the mathematical techniques for fluid–structure modelling and dynamic methods used in the analysis are first presented and then applied to compute the eigenmodes and the dynamic response of the fluid–structure coupled system with various numerical procedures (quasi-static, spectral and temporal approaches). Numerical results are presented and discussed; fluid–structure interaction effects are highlighted. As a main conclusion, added mass effects are proved to have a significant influence on the dynamic response of the nuclear reactor.     

Review of liquid metal corrosion issues for potential containment materials for liquid lead and lead–bismuth eutectic spallation targets as a neutron source

Lead (Pb) and lead–bismuth eutectic (44Pb–56Bi) have been the two primary candidate liquid metal target materials for the production of spallation neutrons. Selection of a container material for the liquid metal target will greatly affect the lifetime and safety of the target subsystem. For the liquid lead target, niobium–1 wt% zirconium (Nb–1Zr) is a candidate containment material for liquid lead, but its poor oxidation resistance has been a major concern. In this paper, the oxidation rate of Nb–1Zr was studied based on the calculations of thickness loss resulting from oxidation. According to these calculations, it appeared that uncoated Nb–1Zr may be used for a 1-year operation at 900°C at P_O2=1×10^–6 Torr, but the same material may not be used in argon with 5-ppm oxygen. Coating technologies to reduce the oxidation of Nb–1Zr are reviewed, as are other candidate refractory metals such as molybdenum, tantalum, and tungsten. For the liquid lead–bismuth eutectic target, three candidate containment materials are suggested, based on a literature survey of the materials’ compatibility and proton irradiation tests: Croloy 2-1/4, modified 9Cr–1Mo, and 12Cr–1Mo (HT-9) steel. These materials seem to be used only if the lead–bismuth is thoroughly deoxidized and treated with zirconium and magnesium.     

Experimental investigation of heat transfer from a circular pipe to lead coolant with controlled oxygen content

The results of experimental investigations of heat transfer from a circular pipe to lead coolant with the oxygen content being controlled and monitored are presented. The heat-transfer investigations are conducted for Peclet numbers 800–3550, Prandtl numbers 0.0123–0.0211, and Reynolds numbers 40,000–190,000 with specific heat flux ~40 kW/m ² and thermodynamically active oxygen content in lead 10^-7 –10⁰ . The experimental dependences of the Nusselt numers on the Prandtl numbers with different oxygen content in the lead coolant are obtained.Translated from Atomnaya Énergiya, Vol. 97, No. 5, pp. 345–349, November, 2004.     

Structural integrity evaluation method for overheating rupture of FBR steam generator tube

This paper describes a structural integrity evaluation method for a SG tube of FBR in case of sodium–water reaction and creep rupture tests to obtain the strength of the tube material. In the SG of FBR, if intermediate size of water/steam leak (1–2 kg s⁻¹) would occur from a tube, it could cause overheating rupture of the multiple tubes surrounding the initially failed tube due to generated sodium–water reaction heat. In the ultra-high temperature condition, the creep strength of the material is one of the dominant factors for failure behavior. Accordingly, we tried to apply the creep failure criterion for the overheating rupture of the SG tube. The creep rupture tests have been performed at ultra-high temperature conditions ranging from 1223.2 to 1323.2 K. The test material is ‘Mod .9Cr–1Mo steel’ which is one of the candidate materials for the tubes of the future SG of FBR. The test results have shown that tube rupture depends on the creep strength of the material; hence, instantaneous rupture does not occur even if the stress exceeds the design value of ultimate tensile strength. The test data have been suitably expressed using the Larson–Miller Parameter, and a structural integrity evaluation method based on the sum of the use-fraction associated with the creep damage has been proposed. Based on this method, the structural integrity of the tube in the sodium–water reaction flame has been evaluated. The results show that it is important to detect the initial leak of the tube within a short period and to reduce the steam pressure more rapidly by SG blowdown.     

Fluid-to-fluid modeling of critical heat flux in 4 × 4 rod bundles

Fluid-to-fluid modeling of critical heat flux (CHF) is to simulate the CHF behaviors for water by employing low cost modeling fluid, and the flow scaling factor is the key to apply the technique to fuel bundles. The CHF experiments in 4×4 rod bundles have been carried out in Freon-12 loop in equivalent nuclear reactor water conditions (P=10.0–16.0 MPa, G=488.0–2100.0 kg/m² s, X_cr=−0.20–0.30). The models in fluid-to-fluid modeling of CHF is verified by the CHF data for Freon-12 obtained in the experiment and the CHF correlation for water obtained by Nuclear Power Institute of China (NPIC) in the same 4×4 rod bundles. It has been found that the S.Y. Ahmad Compensation Distortion model, the Lu Zhongqi model, the Groeneveld model and Stevens–Kirby model overpredict the bundles CHF values for water. Then an empirical correlation of flow scaling factor is proposed. Comparison of the CHF data in two kinds of test sections for Freon-12, in which the distance of the last grid away the end of heated length is different, shows that the spacer grid, which is located at 20 mm away from the end of the heated length, has evidently influenced on the CHF value in the 4×4 rod bundles for Freon-12. This is different from that for water, and the need for further work is required.     

Thermal–hydraulic feasibility analysis on uprating the HTR–PM

In order to meet energy demand in China, the high temperature gas-cooled reactor–pebble-bed module (HTR–PM) is being developed. It adopts a two-zone core, in which graphite balls are loaded in the central zone and the outer part is fuel ball zone, and couple with a steam cycle. Outer diameter of the reactor core is 4.0 m and height of the core is 9.43 m. The helium inlet and outlet temperature are 250 and 750 °C, respectively. The reactor thermal power is 380 MW. Preliminary studies show that the HTR–PM is feasible technologically and economically. In order to increase the reactor thermal power of the HTR–PM, some efforts have been made. These include increasing the height of reactor core, optimizing the thickness of fuel zone and better selection of the scheme of central graphite zone, etc. Basic design concepts and thermal–hydraulic parameters of the HTR–PM are given. Measures to increase the thermal power are introduced. Thermal–hydraulic analysis results are presented. The results show that, from the viewpoint of thermal–hydraulics, it is possible to increase the reactor power.     

Radiation-Induced Change in the Properties of Isotropic Structural Graphite

Three types of samples of isotropic graphite with different grain density and size were irradiated in a BOR-60 reactor up to neutron fluence (1.7–2.8)·10²⁶ m^–2 (E > 0.18 MeV) at 360–400°C. After irradiation, the change in the dimensions, resistivity, linear thermal expansion coefficient and dynamic elastic modulus were investigated. It was determined that the density in the range 1.67–1.76 g/cm³ results in an increase of the maximum weight and depth of volume shrinkage of isotropic fine-grain graphite. An equation was proposed for fitting the temperature dependence of the critical neutron fluence in the range 380–780°C for the experimental graphite samples.     

A parametric study on pressure–temperature limit curve using 3-D finite element analyses

In order to operate a reactor pressure vessel (RPV) safely, it is necessary to keep the pressure–temperature (P–T) limit during the heatup and cooldown process. While the ASME Code provides the P–T limit curve for safe operation, this limit curve has been prepared under conservative assumptions. In this paper, the effects of conservative assumptions involved in the P–T limit curve specified in the ASME Code Sec. XI were investigated. Three different parameters, the crack depth, the cladding thickness and the cooling rate, were reviewed based on 3-D finite element analyses. Also, the constraint effect on P–T limit curve generation was investigated based on J–T approach. It was shown that the crack depth and constraint effect change the safety region in P–T limit curve dramatically, and thus it is recommended to prepare a more precise P–T limit curve based on finite element analysis to obtain P–T limit for safe operation of a RPV.     

Variants of Aperiodic Pulsed Reactors with Boosted Pulse Parameters

Single (one-section) reactors, containing a neptunium–gallium or uranium–molybdenum alloy core, and coupled two-section systems, consisting of a single pulsed reactor and a driven subcritical assembly, consisting of a uranium–molybdenum alloy or dispersed uranium–graphite material, are studied. Calculations of the neutron and dynamical characteristics of these systems are performed. The information obtained, together with data from previous calculations, made it possible to draw a conclusion about the structure of reactor systems characterized by a short radiation pulse and large-volume cores and cavities for holding specimens. It is shown that a reactor with a single neptunium–gallium alloy core and a coupled cascade-type system consisting of this single driven reactor together with a driven subcritical uranium–graphite assembly have the best pulse parameters.     

The effect of neutron irradiation on the structure and mechanical properties of alloy steels

We have studied the effect of neutron irradiation at temperatures of 200–500°C with various integral doses (1.5·10²⁰·10²¹ neutrons/cm²) on the properties and microstructure of some steels with different chemical compositions and initial structures. We have shown the effect of alloying by various elements on the sensitivity of the steel to irradiation and the temperature of annealing of radiation defects of hardening.Translated from Atomnaya Énergiya, Vol. 15, No. 1, pp. 30–37, July, 1963     

Gyrotrons for ECH applications

Gyrotron oscillators have served as effective sources for electron cyclotron heating (ECH) applications in the area of magnetic confinement fusion. Successful development programs at frequencies at 28 GHz, 60 GHz, and 140 GHZ, have led to the availability of wide-range gyrotron sources with high-average-power capabilities. Since 1975, over 100 pulsed and CW gyrotrons with typical power levels of 200 kW at frequencies ranging from 28–106 GHz have been used by various fusion laboratories. Present development activity is aimed at providing sources that will generate power levels up to 1 MW CW at frequencies in the range 100–140 GHz for the ECH experiments that are currently being planned. Initial experimental efforts in this area have verified many of the concepts to be employed in forthcoming 1-MW CW test vehicles. Source requirements, that are even more formidable, are foreseen for the next generation magnetic fusion facilities. Frequencies ranging from 200–300 GHz with power generation capabilities of 1–2 MW CW per tube are being considered for these future applications. To this end, various gyrotron designs have been conceived that address these demanding specifications.     

Thyroid-Gland Irradiation Dose in the Russian Population Due to the Fallout of Radioactive Iodine After the Chernobyl Accident

The average irradiation dose to the thyroid gland is estimated for the people living in 4105 populated points in the Bryanskaya, Tul'skaya, Orlovskaya, and Kaluzhskaya oblasts. The basic principles of the method used to reconstruct the dose are presented. The people living in Bryanskaya oblast have the highest irradiation dose to the thyroid gland: in children less than 3 yr old the individual dose reached 10 Gy; the average dose exceeded 2.5 Gy in 12 populated points. In children living in Bryanskaya oblast, for populated points with soil contamination density above 37 kBq/m² the irradiation dose exceeded 0.05 Gy. The highest average irradiation dose to the thyroid gland in children living in Tul'skaya, Orlovskaya, and Kaluzhskaya oblasts is 0.3–1 Gy. The collective irradiation dose for the four most strongly contaminated oblasts is estimated to be as follows: Bryanskaya – 60, Tul'skaya – 20, Orlovskaya – 13, Kaluzhskaya – 3.5 thousand·people·Gy.     

Investigation of loss factors of aerosol particles in the sampling systems of a nuclear power plant

The loss factor of particles with different sizes in a sampling system of a nuclear power plant has been experimentally determined. As shown by optical particle size measurements, the particle size distribution of the test aerosol titanium dioxide (aerodynamic diameter between 0.6 and 8 μm) agrees well with that of the ambient aerosol in the nuclear power plant. The total loss factor for titanium dioxide lies between 1.3 and 1.6. This means a loss of particles of about 23–38% and shows that the transfer properties of the sampling system are good. The experimental results for small particles (aerodynamic diameter 0.6–8 μm) are in good agreement with those of theoretical estimations. The transfer properties of the sampling system for larger particles (aerodynamic diameters up to 3 mm) have been investigated, too. The results of the measured loss factors show that the sampling system is suitable even in this case.     

Core loss during a severe accident (COLOSS)

The COLOSS project was a 3-year shared-cost action, which started in February 2000. The work-programme performed by 19 partners was shaped around complementary activities aimed at improving severe accident codes. Unresolved risk-relevant issues regarding H₂ production, melt generation and the source term were studied through a large number of experiments such as (a) dissolution of fresh and high burn-up UO₂ and MOX by molten Zircaloy, (b) simultaneous dissolution of UO₂ and ZrO₂, (c) oxidation of U–O–Zr mixtures, (d) degradation–oxidation of B₄C control rods.Corresponding models were developed and implemented in severe accident computer codes. Upgraded codes were then used to apply results in plant calculations and evaluate their consequences on key severe accident sequences in different plants involving B₄C control rods and in the TMI-2 accident.Significant results have been produced from separate-effects, semi-global and large-scale tests on COLOSS topics enabling the development and validation of models and the improvement of some severe accident codes. Break-throughs were achieved on some issues for which more data are needed for consolidation of the modelling in particular on burn-up effects on UO₂ and MOX dissolution and oxidation of U–O–Zr and B₄C–metal mixtures. There was experimental evidence that the oxidation of these mixtures can contribute significantly to the large H₂ production observed during the reflooding of degraded cores under severe accident conditions.The plant calculation activity enabled (a) the assessment of codes to calculate core degradation with the identification of main uncertainties and needs for short-term developments and (b) the identification of safety implications of new results.Main results and recommendations for future R&D activities are summarized in this paper.     

Neutron irradiation effects on optical absorption of KU1 and KS-4V quartz glasses and Infrasil 301

In this work we have compared the effects of neutron (10²¹–10²² n/m² fluences) and gamma irradiation (23.8 MGy dose) on the IR–vis–UV optical absorption spectra of high purity silica with different OH content: KU1 (800 ppm), KS-4V (<0.2 ppm), and commercial silica Infrasil 301 (<8 ppm). The results show that the UV–vis optical degradation of the silica, after neutron irradiation at the highest fluence is similar for the three grades studied, while gamma-induced optical absorption depends on the material grade (KS-4V shows the lowest optical absorption). The effects of both types of radiation on the IR band related with the hydroxyl group (3650 cm⁻¹) depend on the silica grade. For KU1, the shape of this band changes with neutron fluence. For Infrasil 301 gamma and neutron irradiated, this band height increases, possibly due to free molecular or hydrogen atoms. The shift to lower energies observed for the 2260 cm⁻¹ band in the three neutron irradiated silica grades, reflects the changes induced by neutrons in the lattice bonding angle distribution.