Transformation of a system consisting of plane isotropic source and unit sphere detector into a system consisting of point isotropic source and plane detector in Monte Carlo radiation transport calculation

In a nuclear power plant accident, radioactive nuclides may be released which are distributed uniformly on the ground. If estimation of dose rate from such a source by a Monte Carlo calculation is attempted, some difficulty is encountered because the calculation efficiency is very low. To solve this low efficiency problem, we show that a plane isotropic source can be transformed into a point isotropic source by changing the detector shape from a unit sphere to a plane. We verified the validity of this transformation by the numerical comparison of unscattered photon fluence. As an example of this transformation, the ambient dose rate D _i was calculated from the uniform radioactive nuclide distribution on the ground using the EGS5 Monte Carlo code. We also measured the radioactivity and ambient dose rate (M) on the KEK campus within a month after the releases from the Fukushima No. 1 Nuclear Power Plant accident. Using radioactivity data and D _i, we calculated the ambient dose rate (C). The calculated and measured ambient dose rates agreed reasonably well; their ratio (C/M) was 0.62 to 1.28.     

Development of a geometric uncertainty model describing the accuracy of position-sensitive, coincidence neutron detection

A diameter of uncertainty (D_u) was derived from a geometric uncertainty model describing the error that would be introduced into position-sensitive, coincidence neutron detection measurements by charged-particle transport phenomena and experimental setup. The transport of α and Li ions, produced by the ¹⁰B(n,α) ⁷Li reaction, through free-standing boro-phosphosilicate glass (BPSG) films was modeled using the Monte Carlo code SRIM, and the results of these simulations were used as input to determine D_u for position-sensitive, coincidence techniques. The results of these calculations showed that D_u is dependent on encoder separation, the angle of charged particle emission, and film thickness. For certain emission scenarios, the magnitude of D_u is larger than the physical size of the neutron converting media that were being modeled. Spheres of uncertainty were developed that describe the difference in flight path times among the bounding-case emission scenarios that were considered in this work. It was shown the overlapping spheres represent emission angles and particle flight path lengths that would be difficult to resolve in terms of particle time-of-flight measurements. However, based on the timing resolution of current nuclear instrumentation, emission events that yield large D_u can be discriminated by logical arguments during spectral deconvolution.     

Propagation of uncertainties in problems of structural reliability

The problem of controlling a variable Y such that the probability of its exceeding a specified design limit L is very small, is treated. This variable is related to a set of random variables X_i by means of a known function Y = ƒ(X_i). The following approximate methods are considered for estimating the propagation of error in the X_i's through the function ƒ(·): linearization; method of moments; Monte Carlo methods; numerical integration. Response surface and associated design of experiments problems as well as statistical inference problems are discussed.     

Uncertainty analysis in Monte Carlo criticality computations

Uncertainty analysis is imperative for nuclear criticality risk assessments when using Monte Carlo neutron transport methods to predict the effective neutron multiplication factor (k_eff) for fissionable material systems. For the validation of Monte Carlo codes for criticality computations against benchmark experiments, code accuracy and precision are measured by both the computational bias and uncertainty in the bias. The uncertainty in the bias accounts for known or quantified experimental, computational and model uncertainties. For the application of Monte Carlo codes for criticality analysis of fissionable material systems, an administrative margin of subcriticality must be imposed to provide additional assurance of subcriticality for any unknown or unquantified uncertainties. Because of a substantial impact of the administrative margin of subcriticality on economics and safety of nuclear fuel cycle operations, recently increasing interests in reducing the administrative margin of subcriticality make the uncertainty analysis in criticality safety computations more risk-significant. This paper provides an overview of two most popular k_eff uncertainty analysis methods for Monte Carlo criticality computations: (1) sampling-based methods, and (2) analytical methods. Examples are given to demonstrate their usage in the k_eff uncertainty analysis due to uncertainties in both neutronic and non-neutronic parameters of fissionable material systems.     

Measurement of the individual dose of external irradiation

The relation between the dose recorded with an individual tissue-equivalent luminescence dosimeter and the effective dose obtained by the human body is calculated. The scattering of γ rays in the human body and the surrounding space is taken into account. It is shown that for the conditions under which the dose load is formed, which exist under the normal operating conditions of a nuclear power plant, the effective dose obtained by the worker is related with the dose measured with an individual tissue-equivalent dosimeter by the relation D_eff=0.56±0.125D_equiv where D_eff is the effective dose, in Sv; D_equiv is the measured dose. in Gy. 4 tables, 2 references. All-Russia Scientific-Research Institute of Nuclear Power Plants. Translated from Atomnaya énergiya, Vol. 87, No. 3, pp. 222–226, September, 1999.     

Monte Carlo uncertainty quantification of the effective delayed neutron fraction

The applicability of Monte Carlo techniques, namely the Monte Carlo sensitivity method and the random-sampling method, for uncertainty quantification of the effective delayed neutron fraction β_eff is investigated using the continuous-energy Monte Carlo transport code, MCNP, from the perspective of statistical convergence issues. This study focuses on the nuclear data as one of the major sources of β_eff uncertainty. For validation of the calculated β_eff, a critical configuration of the VENUS-F zero-power reactor was used. It is demonstrated that Chiba's modified k-ratio method is superior to Bretscher's prompt k-ratio method in terms of reducing the statistical uncertainty in calculating not only β_eff but also its sensitivities and the uncertainty due to nuclear data. From this result and a comparison of uncertainties obtained by the Monte Carlo sensitivity method and the random-sampling method, it is shown that the Monte Carlo sensitivity method using Chiba's modified k-ratio method is the most practical for uncertainty quantification of β_eff. Finally, total β_eff uncertainty due to nuclear data for the VENUS-F critical configuration is determined to be approximately 2.7% with JENDL-4.0u, which is dominated by the delayed neutron yield of ²³⁵U.     

Direct measurements of the<Superscript>90</Superscript>Sr activity in water using Cherenkov radiation

A computational-experimental investigation of Cherenkov radiation due to⁹⁰Sr−⁹⁰Y in water samples was performed. The Monte Carlo method was used to simulate the generation of photons from the decay of⁹⁰Sr−⁹⁰Y and other isotopes in water in the range 220–600 nm. The Cherenkov radiation was measured using a low-noise photomultiplier with a tellurium-rubidium photocathode on a MgF₂ entrance window. Experiments on measuring the amplitude distributions and counting rates due to Cherenkov radiation from the radioactive solutions of⁹⁰Sr−⁹⁰Y,¹³⁷Cs−¹³⁷Ba, and⁴⁰K were performed. The sensitivity and lowest measurable activity for water samples of⁹⁰Sr−⁹⁰Y was estimated on the basis of the results obtained, 4 figures, 3 tables, 11 references. Russian Science Center “Kurchatov Institute.” Translated from Atomnaya énergiya, Vol. 88, No. 4, pp. 282–286, April, 2000.     

Measurement of Anisotropic Diffusion Coefficients in H2O-Al Slab

Diffusion coefficients were measured for an H₂)O-Al slab lattice (H₂O 8 mm, Al 8 mm) in parallel (D_∥) and perpendicular (D_⊥) directions, applying the pulsed neutron technique. The measured results were compared with those of theoretical calculations based on the collision probability method, taking into account anisotropic scattering, and assuming a source neutron distribution of Maxwellian configuration.

Fairly good agreement between theory and experiment was obtained for D _⊥ but a considerable discrepancy, of the order of 19%, resulted for D _∥. The relation between measured and theoretical quantities in a heterogeneous system are also discussed.     

Monte-Carlo modeling of isotopic kinetics

An algorithm for direct Monte Carlo modeling of the isotopic kinetic is described. The nuclear reactions, including those that change the isotopic composition (n, α), (n, 2n), and others, are recorded for each neutron or group of neutrons when performing a Monte Carlo calculation of neutron transport. At the same time as the neutron transport calculations, the change of the concentration of the nuclides and the irradiation time are calculated on the basis of the number of recorded interactions and energy released. The computational results presented show that it is in principle possible to apply this approach to isotopic kinetic problems.     

Calorimetric and thermodynamic analysis of palladium-deuterium electrochemical cells

An analysis is made of the chemical heats liberated from a palladium-deuterium electrochemical cell operating inside a calorimeter. It is important, in such an analysis, to carefully identify the chemical and electrochemical sources of heat, before any “excess heats” can be ascribed to non-chemical reactions.⁽¹⁾ The calorimeter measures the enthalpy (ΔH _r ) of the reaction; whereas, the electrochemical voltage of the cell reflects the free energy (ΔG _r ) of the reaction within the Pd-D electrolysis cell. The heat energy from the calorimeter cell therefore doesnot equal the electrical energy supplied to the cell, as might initially be expected. The magnitudes of the differing calorimetric and electrochemical energies were found to be related through the “thermoneutral potential” (ξ_H) of the electrochemical reaction. The chemical heat theoretically expected from the calorimeter is given by (1) I(ξ_L-ξ_H), the cell current (I), multiplied by the difference between the operating cell voltage (ξ_L) and the thermoneutral potential (ξ_H), rather than (2) Iξ_L, the electrical input power. This was verified empirically using a freon vaporization calorimeter, which operates on the principle of accurate measurement of the vaporization rate of liquid freon which completely surrounds the electrochemical cell. The calorimetrically-measured heats observed from a Pt/D₂O, 0.1M LiOD/Pd electrochemical cell were within 2% of the thermoneutral potential predicted value, I(ξ_L-ξ_H); but were found to be 15–30% less than the electrical work supplied to the cell, Iξ_L. Measurements of D₂O consumed by the cell reactions also verified that essentially no significant recombination of D₂ and O₂ gases occurred within the cell. No “excess heats” were observed from this Pd cell during the 36 days of its electrolytic operation. Likewise, no increase in the neutron flux around the cell was found, using three³He radiation detectors.     

Improvement of Correlated Sampling Monte Carlo Methods for Reactivity Calculations

Two correlated Monte Carlo methods, the similar flight path and the identical flight path methods, have been improved to evaluate up to the second order change of the reactivity perturbation. Secondary fission neutrons produced by neutrons having passed through perturbed regions in both unperturbed and perturbed systems are followed in a way to have a strong correlation between secondary neutrons in both the systems. These techniques are incorporated into the general purpose Monte Carlo code MORSE, so as to be able to estimate also the statistical error of the calculated reactivity change.

The control rod worths measured in the FCA V-3 assembly are analyzed with the present techniques, which are shown to predict the measured values within the standard deviations. The identical flight path method has revealed itself more useful than the similar flight path method for the analysis of the control rod worth.     

Requirements for a radiation monitoring system in connection with risk and dose

Conclusions In summary, we have arrived at the seemingly paradoxical conclusion that the requirements for the radiation-monitoring system in region 2, where by definition the average individual dose is less than in region 1, are higher. Conversely, it would seem that if the dose load is smaller, then there is no need to complicate the monitoring system; for example, it is sufficient to establish a monitoring level that is the same as in region 1. The paradox is solved if one takes account of the fact that the lower dose loads in region 2 are associated with the lower content of radionuclides in objects in the environment, i.e., the relatively small useful signals, detected by the given monitoring system. This requires better monitoring systems, since weak signals must be detected and discriminated under conditions of random noise, due to background sources of inoizing radiation and other types of noise. In conclusion, we note that the materials presented in this paper make it possible to distinguish the radioecological environment in two regions quantitatively. Specifically, it must be assumed that regions 1 and 2 have the same radioecological state ifK ₁ ^* <K ₂ ^* ;D ₂<D ₁;N ₁≪N ₂ and ln(K ₁ ^* /K ₂ ^* )<K _thr(K ₂ ^* −K ₁ ^* )/(K ₁ ^* K ₂ ^* ). The latter approximate relation follows from Eq. (3) and the conditiong ₁(K _thr)<g ₂(K _thr), whereK _thr is the threshold value or the control level of the yearly effective collective dose andg(K) is the distribution density of the effective collective dose. Under these conditions, the requirements imposed on the radiation monitoring system in region 2 are more stringent than for an analogous system in region 1. Moscow Scientific and Industrial Association “Radon.” Translated from Atomnaya énergiya, Vol. 88, No. 6, pp. 476–480, June, 2000.     

Molecular-dynamics simulation of threshold displacement energies in zircon

Molecular-dynamics simulations were used to examine the displacement threshold energy (E_d) surface for Zr, Si and O in zircon using two different interatomic potentials. For each sublattice, the simulation was repeated from different initial conditions to estimate the uncertainty in the calculated value of E_d. The displacement threshold energies vary considerably with crystallographic direction and sublattice. Based on the present simulations and previous experimental studies, this work recommends E_d values of 75, 75 and 60 eV for Zr, Si and O, respectively, to be used in Monte Carlo simulations of irradiation damage profile in zircon.     

Metadata on the radioactive contamination of environmental objects on the territories of Russia and the USSR

A unique data archive, accumulated at the Taifun Scientific and Industrial Association in 1954–2005, on the radioactive contamination of the environment on the territory of the USSR and Russia is presented. The archive contains data on the yearly total β activity of atmospheric fallout on the underlying surface, the total volume β activity in the atmosphere at the ground, the results of measurements of the ⁹⁰Sr and ¹³⁷Cs content in samples combined over one month or quarter, atmospheric aerosols and fallout on individual points, the volume activity of tritium and ⁹⁰Sr in water, rivers, lakes and seas, and the radionuclide contamination density of the territories of populated points as a result of the Chernobyl accident. __________ Translated from Atomnaya énergiya, Vol. 101, No. 2, pp. 149–152, August, 2006.     

Measurement of Acoustic Velocity and Attenuation in an Air-Water Two-Phase Medium

Actinides, mainly responsible for the long term risk of spent fuel, are the principal candidates to transmutation due to their large absorption cross sections.

Systems driven by particle accelerators have been investigated in the past to produce fissile material. Recently these systems have been reconsidered to destroy minor actinides (MA) and long-lived fission products (LLFP), reducing the need for the traditional final confinement of radioactive waste.

Two Monte Carlo calculation models have been developped to determine the criticality safety conditions and the burning capability of MAS and of Pu.

A Pu burner, whose core is poisoned with Th to compensate by producing ²³³U the burnup reactivity due to the even Pu isotopes, can operate at a low proton current using perhaps a cyclotron, incinerating 70% of the charged Pu; its burning capability would be the production of about 1.5 PWRs.

Liquid fuel accelerator driven systems can be used in the future (due to the accelerator dimensions) for MA burning using D₂0 as carrier in a homogeneous core; such a system can burn the production of more than 15PWRs.

In the future, also the problem of LLFP burning could be solved definitively using a system with D₂0 as carrier.     

Gamma-ray dose analysis for ITER JA WCCB-TBM

To evaluate the nuclear properties of the International Thermonuclear Experimental Reactor (ITER) JA Water-Cooled Ceramic Breeder Test Blanket Module (WCCB-TBM) and to ensure its design conforms to nuclear licensing regulations, nuclear analyses have been performed for the WCCB-TBM's components, including its frame, shield, flange, port extension, pipe forest, bio-shield and Ancillary Equipment Unit (AEU). Utilising Monte Carlo code MCNP5.14, activation code ACT-4 and the Fusion Evaluated Nuclear Data Library FENDL-2.1, this paper focusses on the shutdown dose rate calculation for the WCCB-TBM. Monte Carlo N-Particle Transport Code (MCNP) geometry input data for the TBM are created from computer-aided design (CAD) data using the CAD/MCNP automatic conversion code GEOMIT, and other geometry input data are created manually. The ‘Direct 1-Step Monte Carlo’ method is adopted for the decay gamma-ray dose rate calculation. Behind the bio-shield, the effective dose rates 1 day after shutdown are about 0.2 μSv h⁻¹, which are much lower than 10 μSv h⁻¹, the upper limit for human access. Behind the flange, the effective dose rates 10⁶ s after shutdown are 50–80 μSv h⁻¹, which are lower than 100 μSv h⁻¹, the upper limit for human hands-on access for workers performing maintenance.     

Analysis of Mixed Oxide Fuel Critical Experiments with Neutronics ANalysis Codes for Boiling Water Reactors

Critical experiments of UO₂ and full mixed oxide (MOX) fuel cores conducted at the Tank-type Critical Assembly (TCA) were aNalyzed using BWR design-purpose codes HINES and CERES with ENDF/B files and Monte Carlo fine analysis codes VMONT and MVP with the JENDL-3.2 library.

The averaged values of the multiplication factors calculated with HINES/CERES, VMONT and MVP agreed with those of experiments within 0.3%ΔAk. The values by the design-purpose codes showed a small difference of 0.1%Δk between UO₂ and MOX cores. Monte Carlo code results showed that the JENDL-3.2 library had a tendency to overestimate the multiplication factors of UO₂ cores by about 0.3%Δk compared with those values of MOX cores. The root mean square errors of calculated power distributions were less than 1% for HINES/CERES and VMONT.

These results showed that (1) the accuracy of these codes when applied to full MOX cores was almost the same as their accuracy for UO₂ cores, which confirmed the accuracy of present core design codes for full MOX cores; and (2) the accuracy of the 190-energy-group Monte Carlo calculation code VMONT was almost the same as that of the continuous-energy Monte Carlo calculation code MVP.     

Dose–time dependence of steel critical brittleness temperature for VVER-1000 vessel service life prediction

The high conservativeness of the evaluation of the service life of VVER-1000 vessels is due to the operational reliability margin and the high error in determining the initial computational data. A structural analysis of the error in critical brittleness temperature investigations is presented and the factors which have a large effect on the error in constructing the dose–time dependences of the critical brittleness temperature are determined. In constructing dose–time curves, the use of the critical temperature of a calibration metal and heat-treatment regimes of calibration samples makes available and permits checking a wide range of structures that is characteristic for large blanks, makes it possible to take account of radiation embrittlement effects and radiation-stimulated reversible brittleness, decrease the error in determining the guaranteed values to T _cr ± 5°C, and eliminate the large effect of structural nonuniformity on objective predictions made for a reactor vessel.     

Consideration on Effective Diffusivity of Strontium in Granite

Abstract

Diffusive behavior of strontium and certain kinds of divalent cations in Inada granite were studied by a through-diffusion method. In order to examine the effect of sorption onto overall diffusive behavior, two kinds of solutions were used: 0.1M KCl solution and deionized water. The effective diffusion coefficient (D_e ) and rock capacity factor (a) were (2.0–3.6) xl0–¹³ m²/s and less than 0.022 in 0.1M KCl solution and (0.32–1.7) x 10 ^?13m²/s and 1.5–2.4 in deionized water respectively. The D_e , and α in deionized water were much larger than those in 0.1M KCl solution. These results are well explained by taking into account the diffusion of sorbed ion or the surface diffusion. In support of this mechanism, most D_e , values of Sr reported for various rocks are found proportional to the sorptivity ( _ρR_d )-In the case that the sorptivity is low, D_e of Sr depends on porosity like that of nonsorbed iodide. The effective diffusion coefficient of Sr in rocks was well explained by taking into account pore and surface diffusion and was expressed as D_e=2.1 xl0^?10 _? ^1.3+3.5xl0″^?12 _ρR_d . The effective diffusion coefficient of divalent cations in the granite was found proportional to their diffusion coefficients in bulk solution.     

Light and heavy water dynamic structure factor for neutron transport codes

In this study, we report on recent neutron inelastic scattering experiments performed at the Institut Laue-Langevin (ILL) for H₂O and D₂O. The measured dynamic structure factors S(q, ω) have been reduced, normalised and transformed into the S(α, β) formalism, where α and β stand for the unit-less momentum and energy transfers, respectively. The measurements were complemented with molecular dynamics simulations. After processing with NJOY, new water neutron scattering cross-sections have been generated for use with e.g. the Monte Carlo N-Particle (MCNP) software in view to improve the accuracy of the nuclear facility models. As an example, we present improved accuracy calculations for the safety rod insertion impact on the criticality factor k_eff for the ILL high flux research reactor.