Doppler and Self-shielding Effects of Resonances in Unresolved Energy Region of Niobium

Sorption of strontium on a number of minerals and rocks has been studied by varying aqueous concentrations of strontium. Distribution coefficients (K _d) of strontium determined by a batch method remained constant at a low concentration level and decreased with increasing concentration at a high concentration level. Values of K _d,tr at trace concentrations were considerably different depending on types of geological samples. A linear relationship was observed between the values of K _d,tr and the cation exchange capacity (CEC) for cesium as: log K _d,tr=log CEC+constant. The values of K _d,tr of strontium changed almost linearly to those of cesium for individual samples suggesting a similarity in physicochemical interactions between geological samples and these ions.     

UK Transport Events and International Cooperation

Abstract

In the UK, an accident and incident database has been established for transport events from 1964 onwards. The CEC is promoting a Community transport event databank and the IAEA has established an international system EVTRAM. The objectives of these systems are different and each has merit provided an adequate level of response is established and maintained. Such systems provide information on the effectiveness of regulatory requirements and enable lessons learnt to be made widely available. The reporting procedures in the UK and the results of an analysis of transport events are presented.     

Distribution Coefficient of Cesium and Cation Exchange Capacity of Minerals and Rocks

The sorption behavior of cesium on a variety of minerals and rocks has been investigated by batch equilibration using ¹³⁷Cs. The distribution coefficient (K_d) of cesium increased with decreasing concentration of cesium and approached a constant value in dilute regions. The distribution coefficients determined at a trace level (K_dt,tr) differed more than four orders of magnitude strongly depending on the kinds of samples, for example, about 30 on quartz and 10⁵ on tuffs. The values of K_d,tr are related to the cation exchange capacity (CEC; meq/g) determined from the saturated sorption of cesium for individual samples as : log K_{d, tr} =log CEC + constant. Most K_{d, tr} values were found to approximately follow this linear relationship.     

Solution of group diffusion equation for X-Y geometry by finite Fourier transformation

A new difference equation to the two dimensional diffusion equation for x-y geometry is derived by using the finite Fourier transformation. This difference equation has a form of a coupled equation of the 3 point difference equations for each coordinate, and can be easily solved by the iterative method of the alternative direction implicit method. Group diffusion equations are solved using this difference equation and sample calculations show that accurate results can be obtained with less mesh points than the usual 5 points difference equation.     

An integrated structure and scaling methodology for severe accident technical issue resolution: Development of methodology

Scaling has been identified as a particularly important element of the Severe Accident Research Program because of its relevance not only to experimentation, but also to analyses based on code calculations or special models. Recognizing the central importance of severe accident scaling issues, the United States Regulatory Commission implemented a Severe Accident Scaling Methodology (SASM) development program involving a lead laboratory contractor and a Technical Program Group to guide the development and to demonstrate its practicality via a challenging application. The Technical Program Group recognized that the Severe Accident Scaling Methodology was an integral part of a larger structure for technical issue resolution and, therefore, found the need to define and document this larger structure, the Integrated Structure for Technical Issue Resolution (ISTIR). The larger part of the efforts have been devoted to the development and demonstration of the Severe Accident Scaling Methodology, which is Component II of the ISTIR. The ISTIR and the SASM have been tested and demonstrated, by their application to a postulated direct containment heating scenario. The ISTIR objectives and process are summarized in this paper, as is its demonstration associated directly with the SASM. The objectives, processes and demonstration for the SASM are also summarized in the paper. The full body of work is referenced.     

Development of a 1200 fine group nuclear data library for advanced nuclear systems

Accurate and reliable nuclear data libraries are essential for calculation and design of advanced nuclear systems.A 1200 fine group nuclear data library Hybrid Evaluated Nuclear Data Library/Fine Group (HENDL/FG) with neutrons of up to 150 MeV has been developed to improve the accuracy of neutronics calculations and analysis.Corrections of Doppler,resonance self-shielding,and thermal upscatter effects were done for HENDL/FG.Shielding and critical safety benchmarks were performed to test the accuracy and reliability of the library.The discrepancy between calculated and measured nuclear parameters fell into a reasonable range.     

Two-group,three-dimensional model based study of reactivity induced transients in upgraded LEU material test reactors

A two-group, three-dimensional diffusion theory based methodology coupled with one-dimensional single-phase heat transfer calculations has been developed for the transient analysis of typical material test reactors (MTRs). This methodology has been implemented in a FORTRAN based computer program MTRAP3. It uses the CITATION computer program as a subroutine for static neutronic calculations while the group constant generation is performed by employing the WIMS-D/4 code. The MTRAP3 program uses Cranck–Nicolson (CN) based numerical scheme for solution of time dependant neutron diffusion calculations while time-implicit strategy is employed for detailed heat-transfer calculations. The CN-scheme has been found to remain stable for much larger time steps (Δt ∼ 10⁻⁵ s) as compared with the time-explicit scheme which is limited to very small time steps only (Δt ∼ 10⁻¹⁰ s). For step as well as for ramp reactivity induced transients, the predicted values of core integrated reactor power and core average temperatures has been found to agree well with the corresponding values found by using the PARET computer program. The assembly-wise power profile as found by the MTRAP3 program has been found consistent with the corresponding experimental measurements.     

Development of Worldwide Version of System for Prediction of Environmental Emergency Dose Information: WSPEEDI, (III)

The numerical models of WSPEEDI (Worldwide version of System for Prediction of Environmental Emergency Dose Information) were revised. The new models are formulated with a terrain following vertical coordinates with the compressibility of the atmosphere being considered. The models have been integrated into a computer code system together with system control software, worldwide geographic database, meteorological data pre-processor and graphics. The performance of the code system has been evaluated against the Chernobyl case with reliable source terms, well established meteorological input and a comprehensive monitoring database, which were provided in ATMES (Atmospheric Transport Model Evaluation Study) coordinated by IAEA, WMO and CEC.     

Design,Analysis and Optimization of the First Wall Cooling System of the CFETR Blanket

China Fusion Engineering Test Reactor (CFETR) is a superconducting tokamak which is designed by China National Integration design Group for Magnetic Confinement Fusion. CFETR Blanket, as a plasma-facing component withstand very high heat load, is very critical for fusion reactor operation. The first wall (FW) is one of the most significant components of the blanket. The cooling system of the FW has been designed. Meanwhile, thermal–dynamic calculations are performed to obtain the coolant feature and temperature distribution of the FW using ANSYS CFX code. Besides, thermo-mechanical coupling analysis is carried out using the temperature distribution from thermal–dynamic calculation as boundary condition. In addition, cooling channel optimization is proposed according to the analysis results. Analysis results of the optimization cooling channel indicate that the maximum temperature and thermal stress satisfy the design requirements of the FW.     

Helium Production in Solid Target and Metallic Window Materials Irradiated with Intermediate and High Energy Protons

The helium isotope formation cross-section has been obtained for iron, tantalum and tungsten irradiated with protons at energies from the reaction threshold up to several GeV. The cross-section evaluation has been performed using the results of model calculations and by the analysis of available experimental data.

The numerical calculations were carried out using the modified ALICE code and the CASCADE/INPE code.     

处理后的粘土矿物的阳离子交换容量

研究了高岭土类、蒙脱石类、海泡石和沸石类矿物经强酸、强碱和80℃处理后的阳离子交换吸附容量。结果表明：pH值和80℃处理温度对粘土矿物的阳离子交换容量均有影响，粘土矿物的阳离子交换容量主要受其晶体结构的影响；沸石类、高岭土类及海泡石粘土矿物的耐酸、碱及热稳定性良好；蒙脱石类粘土矿物耐酸、碱及热稳定性较差。     

The ‘FRENCH’ (flux reconstructed explicitly by neutronic calculated harmonics) method in multidimensional and multigroup theory

The explicit high-order perturbation formulation which allows reconstruction of reactivity variations as well as the neutronic flux for very strong perturbations has been implemented for multidimensional calculations. A new method, the ‘FRENCH’ method, has been developed to overcome the difficulties found by the ‘Standard’ method for multigroup calculations. Practical applications on cases of interest have shown a significant improvement in the agreement with the results obtained by direct calculations.     

An ENDF-6 compatible evaluation for neutron induced reactions of Th in the unresolved resonance region

An evaluation for neutron induced reactions of ²³²Th has been performed in the unresolved resonance region from 4 to 100 keV. A generalized single-level representation compatible with the energy-dependent options of the ENDF-6 format has been used. The average partial cross sections have been expressed in terms of transmission coefficients by applying the Hauser–Feschbach statistical reaction theory including width-fluctuations. The evaluation is based on a combined analysis of experimental average cross section data (including the most recent capture cross section data obtained at the time-of-flight facilities GELINA and n_TOF), and results of dispersive coupled-channel optical model calculations. Experimental data on transmission and self-indication measurements as well as integral benchmark calculations have been used for validation. The evaluation also includes covariance information.     

Pin power reconstruction for CANDU reactors using a neuro-fuzzyinference system

A neuro-fuzzy inference system has been developed for reconstructing fuel pin powers from Canada deuterium uranium (CANDU) core calculations performed with a coarse-mesh finite difference diffusion approximation and single-assembly lattice calculations. The neuro-fuzzy inference system is trained by a genetic algorithm and a least-squares method using the partial core calculation results of two 6×6 fuel bundle models. Verification tests have been performed for two partial core benchmark problems composed of other 6×6 and 3×3 fuel bundles. The reconstructed pin powers are compared with the reference solutions obtained with the detailed collision probability calculations using the HELIOS lattice analysis code. The results indicate that the proposed reconstruction algorithm is accurate, yielding the error due to the reconstruction scheme of less than 0.5%     

Dosimetry study for a new in vivo X-ray fluorescence (XRF) bone lead measurement system

A new ¹⁰⁹Cd γ-ray induced bone lead measurement system has been developed to reduce the minimum detectable limit (MDL) of the system. The system consists of four 16 mm diameter detectors. It requires a stronger source compared to the “conventional” system. A dosimetry study has been performed to estimate the dose delivered by this system. The study was carried out by using human-equivalent phantoms. Three sets of phantoms were made to estimate the dose delivered to three age groups: 5-year old, 10-year old and adults. Three approaches have been applied to evaluate the dose: calculations, Monte Carlo (MC) simulations, and experiments. Experimental results and analytical calculations were used to validate MC simulation. The experiments were performed by placing Panasonic UD-803AS TLDs at different places in phantoms that representing different organs. Due to the difficulty of obtaining the organ dose and the whole body dose solely by experiments and traditional calculations, the equivalent dose and effective dose were calculated by MC simulations. The result showed that the doses delivered to the organs other than the targeted lower leg are negligibly small. The total effective doses to the three age groups are 8.45/9.37 μSv (female/male), 4.20 μSv, and 0.26 μSv for 5-year old, 10-year old and adult, respectively. An approval to conduct human measurements on this system has been received from the Research Ethics Board based on this research.     

Parallelization of Monte Carlo Code MCACE for Shielding Analysis and Estimation of Its Efficiency by Simulator

An improved coarse-mesh discrete ordinates method has been developed for three-dimensional hexagonal transport calculations of high-conversion light water reactors and fast reactors. This method employs a new weighted diamond difference approximation which is obtained by using the neutron balance equations in divided submeshes. The weight is a function of neutron direction and scaler flux, and this method can be easily incorporated into conventional discrete ordinates transport codes.

The present method was applied to hexagonal fuel assembly calculations of high-conversion reactor and fast reactor core calculations, and the results were compared with those of Monte- Carlo calculations. The values of kefi and power distributions agreed with each other within 0.5 and 3%, respectively, verifying accuracy of the present improved coarse-mesh discrete ordinates transport calculation method.     

Separation of the alpha-emitting radioisotopes actinium-225 and bismuth-213 from thorium-229 using alpha recoil methods

An innovative method has been demonstrated for separating alpha-emitting isotopes for medical radiotherapy applications. The method relies on recoil-ion separation rather than on conventional wet chemistry techniques to separate medical isotopes from their precursor sources. The isotopes ²²⁵Ac and ²¹³Bi have been separated from electro-deposited sources of ²²⁹ThO₂. Separations of ²²⁵Ac were carried out by placing nickel recoil collector foils in firm contact with the ²²⁹ThO₂ sources. One-stage recoil-ion separations of ²²⁵Ac from ²²⁹Th have been performed as well as two-stage separations of ²¹³Bi from previously recoil separated ²²⁵Ac. In addition, a direct recoil separation of ²¹³Bi from ²²⁹Th has been demonstrated. The ²¹³Bi from the one-stage direct separation has a high isotopic purity, but contains small amounts of long-lived ²²⁵Ac alpha activity. The two-stage separations of ²¹³Bi produce high isotopic purity material (>99.9999%), but result in lower isotopic yields. Range-energy calculations have been carried out to determine the yields of recoil ions as a function of alpha-particle energy and ThO₂ thickness. The results of the calculations have been benchmarked with recoil separation measurements carried out using ThO₂ electro-deposits over a range of thickness. A computer code based on the generalized Bateman equations has been developed to allow calculations of the amounts of any isotope in the ²²⁹Th decay chain as a function of recoil separation exposure time and elapsed time after the separation. An excellent match has been obtained between the predictions of the Bateman calculations and the results of recoil separation measurements. The recoil separation method has proven to be a simple and effective way of separating medically useful isotopes such as ²¹³Bi. In addition, the method has been shown to produce no chemical or radioactive wastes, in contrast to radiochemical separation methods, which generate mixed (chemical and radioactive) waste.     

Five-detector method for core-barrel motion analysis

For surveillance of core-barrel motion and vibration of PWRs, a simple approach has been proposed in the framework of VIBREAL by Åkerhielm et al. The new method can separate noise-source spectra by simple arithmetic calculations on the CPSDs of five ex-vessel neutron detectors, and also can determine the direction of shell-mode core-barrel motion. A practical application to Borssele reactor noise has been shown in comparison with the results by SPEC-DEC as well as the original VIBREAL.     

Time-dependent behaviour of the neutron field in two interacting cylindrical disks

The influence 6f a void on the neutron flux in a moderating system has been studied, mainly by the Monte Carlo method. The calculations simulate the decay of the neutron field in a pulsed neutron-source measurement. The neutron flux was studied as a function of space, energy and time for a system of two flat cylindrical polyethylene disks. The slab thickness was varied between 1.1 and 4.4 cm and the radius was 9.0 cm. The gap between the slabs was varied from 0 to 18 cm. Some calculations have also been made for absorbers in the gap. The purpose of these absorbers was to eliminate the time-delay effect for the low-velocity neutrons accumulating in the gap. From these results the interaction parameter for the two slabs in the corresponding stationary cases has been calculated. The agreement with measurements made by Grosshög is good. In the one-velocity case the variational method has been used to predict the decay rates. The agreement with the Monte Carlo results is good for small gaps.     

Organic iodine chemistry

A shared-cost action on Organic Iodine Chemistry has been completed as part of the CEC 4th Framework programme on Nuclear Fission Safety. Organisations from four EC countries are involved in an integrated programme of experiments and analysis to help clarify the phenomenology, and to increase confidence in the modelling of iodine behaviour in containment. The project is focused on identifying the main routes for organic iodine formation, and providing new experimental kinetic data which will be used to improve existing models and to stimulate code development.