Computer Code System Research Articles

Project SPHINX spent hot fuel incinerator by neutron flux (the development of a new reactor concept with liquid fuel based on molten fluorides)

The SPHINX project is dealing with a solution of some principle problems of a very promising way of nuclear waste treatment, high level wastes from spent nuclear fuel in particular, by means of transmutation of radionuclides by use of a nuclear reactor with liquid fuel based on molten fluorides, which might be a subcritical system driven by a suitable neutron source. Its superiority lies also in the fact that it makes possible to utilize actinides contained, by others, in spent nuclear fuel and so to reach a positive energy effect. The SPHINX project has been proposed by the consortium TRANSMUTATION being established by four leading nuclear research bodies in the Czech Republic (Nuclear Research Institute Rez plc, SKODA Nuclear Machinery plc in Pilsen, Nuclear Physics Institute of Academy of Sciences in Rez and Technical University in Praha) at the end of 1996 to which Technical University in Brno (specialized for a secondary circuit problems) has associated in the year 2000. The project has been supported by the Ministry of Industry and Trade of the Czech Republic, CEZ, a.s. (Czech Electricity Generating Company) and RAWRA (Radwaste Repository Authority). The R&D program of the SPHINX project contains an experimental part, which serves for a verification of design inputs for designing a demonstration unit of a transmuter with liquid fuel based on molten fluorides. The current status of the experimental program performance has been focused upon the irradiation of samples of molten-salt systems as well as structural materials proposed for the blanket of the SPHINX transmuter in the field of high neutron flux of research reactors. The main aims of this program called Irradiated Probes BLANKA are the following: (1) Experimental verification of long time behavior of transmuter blanket which contains molten fluoride salts as a fuel and coolant, (2) Validation of computational code system being developed for the computation of actinides concentration in long- term operation of the transmuter, and (3) Material research on behavior of materials in neutron and gamma fields, and materials interactions on high temperature conditions. At present, two agreements on multinational cooperation in this field have been signed: One with European Commission and one with Russian Kurchatov Institute (joint experimental programs AMPULA containing fluorides of transuranium elements like Np, Pu, Am and Cm in irradiated samples and a joint development of the ISTAR code).

A Study on Induced Activity in the Low-activationized Concrete for J-PARC

In high power proton accelerator facilities, concrete shield can be highly activated, which makes maintenance work of accelerators quite difficult. So, a low-activationized concrete is planned to be partially adopted as a concrete shield at the accelerator facilities of Japan Proton Accelerator Research Complex (J-PARC) aiming at reducing y-ray exposure dose during maintenance period. The low-activationized concrete is the one using limestone as an aggregate. A new quantity, 24Na-equivalent, was introduced as a criterion to assure effectiveness of the low-activationized concrete. In order to verify its effectiveness, test samples of both the low-activationized concrete and the ordinary concrete were irradiated using FNS at JAERI. The measurements were analyzed by using a shielding design calculation code system being used for J-PARC, showing that the measurements and the calculations of induced activity are in good agreement within a factor of two. Furthermore, by using the same code system, y-ray exposure dose was calculated for modeled configuration of J-PARC to find out that y-ray exposure dose due to the low-activationized concrete was about 10 times lower than that due to the ordinary concrete in a period of less than a few days after operation.

Development of a Real-Time Environmental Radiation Dose Evaluation System for the Vicinity of a Nuclear Facility

A computer code system has been developed to evaluate the environmental radiological impact of accidental releases of radionuclides from a nuclear facility in real time. The system is based on a three-dimensional mass-conserving wind field model and a particle dispersion model with a nesting function in order to calculate the atmospheric dispersion of radionuclides with high spatial resolution in the vicinity of the facility. Input includes on-line local meteorological and effluent data at 10-min intervals, as observed and monitored at the site of the facility, and hourly local meteorological data as predicted by an atmospheric dynamic model. Contour maps and tables of internal and external doses are the output. The performance of the system was validated using environmental monitoring data for the air absorbed dose rate as observed at monitoring stations and monitoring posts around the Tokai Reprocessing Plant when 85Kr was discharged during its operation, based on the safety regulations. The observed increase of the dose rate was simulated well. The agreement of calculated and observed dose rate was within a factor of 2 for 42% of observations, and within a factor of 5 for 74%.

State-of-the-art computer code RELAP5 validation with RBMK-related separate phenomena data

Validation of the RBMK model, developed by employing best estimate system computer code RELAP5 is performed by employing the data from NPPs operation or from integral and separate effects facilities. Validation of the models on the basis of separate phenomena is necessary to perform due to the fact that RELAP5 code has been developed for PWRs, which operate at different conditions (pressure, temperature, coolant void fraction, etc.) from RBMKs. In addition to that, there is a number of phenomena specific for RBMK type reactors (oscillatory flow rate behaviour in parallel channels, flow stagnation in channels, stratification in long horizontal piping, etc.), which have not been studied during RELAP5 validation for PWRs. In the paper, RELAP5 models for separate effects related to RBMK-1500 are presented and modelling of transients is performed. Obtained results are compared with experimental data.

Measurements of decay heat and validation of the European activation code system for fusion power plant applications

A very efficient detector able to measure both electron and photon heats simultaneously and separately has been developed at ENEA Frascati. The detector consists of a well-type 22.8×22.8 cm 2 CsI scintillator (for photon detection) and a small plastic electron detector inserted in the well. This system has been used to measure the decay heat induced in 15 natural elements, irradiated with 14 MeV neutrons. The experimental results have been used to validate the calculation code system EASY-2001, especially developed for the prediction of induced radioactivity in fusion devices. The Calculation versus experiment (C/E) results are reported. For those results where deviations from the C/E=1 line were found, outside the experimental uncertainty band, new effective (integral) cross-sections are proposed, able to bring the results inside the uncertainty estimations. The new cross-section values are discussed in the text.

Theoretical Approach and Computer Code System for Nuclear Data Evaluation of 20-1000 MeV Neutron Induced Reactions on Heavy Nuclei

New approach for fission cross-section description in the energy region from 20 to 1000 MeV has been recently worked out. The developed code system is based on a detailed description of all stages of nucleon- induced reactions on heavy nuclei for E > 20 MeV including: description of entrance channel in the framework of coupled channel method with new optical model potentials for neutron- and proton-induced reactions; simulation of direct processes with the intranuclear cascade model; simulation of pre-equilibrium particle emission with hybrid Monte-Carlo model; description of fission/evaporation competitions on the base of detailed statistical model. The approach developed is adopted for evaluation of nuclear data for energies up to 1 GeV as follows: yields of evaporation residuals; yields of fission products; spectra of secondary gammas, protons and neutrons, including emission from fission fragments; total, elastic, fission and (n, xnyp) reaction cross-sections.

Deployment of MAGS, a comprehensive tool for magnet design and safety analysis, for quench and arc simulation of an ITER-FEAT coil

The coil system of the ITER-FEAT research tokamak stores an energy of about 40 GJ. This amount of energy has to be removed in a safe and controlled manner from the system in case of an accident. As experimental possibilities are limited, numerical models are in great demand. MAGS (MAGnet System) represents a computer code system to simulate accident scenarios of magnet systems of fusion plants. It was used in the past to investigate accident scenarios of the ITER assembly which are required for licensing questions. In this work, the MAGS code is used to perform the safety analysis for a toroidal field (TF) coil of the new ITER-FEAT design becoming necessary due to changes in the magnet system compared to the previous ITER assembly. The analysis includes quench and arc scenarios.

Analysis of divertor cooling loop failures for the ITER-FEAT plant

A Tokamak type fusion reactor is characterized by very high heat loads to plasma facing surfaces of components surrounding the plasma. In order to maintain structural integrity of in-vessel components, it is essential to provide sufficient cooling of those components during normal operation as well as during accident conditions. In relation to the ITER-FEAT design efforts, component thermal responses have been evaluated for a number of specified transient events. In this paper results are presented from thermal-hydraulic analyses of two postulated transients in the divertor cooling loop system, (i) trip of the main coolant pump in the divertor primary heat transport system, and (ii) loss of heat sink in the divertor cooling system; trip of the secondary side pump. The analyses have been made using the ATHENA thermal-hydraulic system computer code.

Estimates of Collective Doses from a Hypothetical Accident of a Nuclear Submarine

The collective dose to the Japanese population has been estimated from a hypothetical accident of a nuclear submarine if it sinks in an offshore region around Japan. A computer code system DSOCEAN has been used for assessing the collective dose due to radionuclides released to the ocean from a sunken nuclear submarine. The estimated collective effective dose commitment from all of the radionuclides released after the break of the fuel pellets is estimated to be 2.5 x104 man-Sv. The contribution of 241Am to the total collective effective dose commitment is the highest, followed by 137Cs, 238Pu, 240Pu, 239Pu and 241 Pu. The maximum of the estimated collective effective dose by the annual intake of marine products after radionuclide releases for one year is approximately 0.3% of the annual average dose by the natural radiation that is reported by UNSCEAR.

Nuclear Simulation and Radiation Physics Investigations of the Target Station of the European Spallation Neutron Source

The European Spallation Neutron Source (ESS) delivers high-intensity pulsed particle beams with 5-MW average beam power at 1.3-GeV incident proton energy. This causes sophisticated demands on material and geometry choices and a very careful optimization of the whole target system. Therefore, complex and detailed particle transport models and computer code systems have been developed and used to study the nuclear assessment of the ESS target system. The purpose here is to describe the methods of calculation mainly based on the Monte Carlo code to show the performance of the ESS target station. The interesting results of the simulations of the mercury target system are as follows: time-dependent neutron flux densities, energy deposition and heating, radioactivity and afterheat, materials damage by radiation, and high-energy source shielding. The results are discussed in great detail. The validity of codes and models, further requirements to improve the methods of calculation, and the status of running and planned experiments are given also.

MORSE: present capabilities and future directions

The MORSE (Multigroup Oak Ridge Stochastic Experiment) Monte Carlo code is a three-dimensional (3D) multipurpose neutron and gamma-ray transport code that has been in use for nearly 30 years and has undergone several evolutions during that time. This article will describe the current capabilities and the future directions of the two versions that are distributed by Oak Ridge National Laboratory: MORSE-SGC and MORSE-CGA. MORSE-SGC is part of the SCALE computer code system, both as a stand-alone program and as part of two of the shielding analysis sequences. MORSE-CGA differs from MORSE-SGC in its cross-section handling and in its requirement for user-supplied analysis routines.

The IRAC Code System to Calculate Activation and Transmutation in the TIARA Facility

A computer code system IRAC has been updated to calculate nuclide transmutation and induced radioactivity for incident particles of neutron, proton, deuteron, alpha up to 150 MeV, and l2C, l4N, l6O,20Ne, 40Ar up to 500 MeV in arbitrary multi-layer target system of three-dimensional geometry. The system provides four libraries that consist of ACSELA for isotope production cross-sections, DECAYLIB for decay data, GAMMALIB for photon emission data, and MASWPS for atomic mass data. Features and the system performance have been presented. To validate accuracy of the code system, radioactivities of residual isotopes for incident 260-MeV 20Ne ions on cobalt target were measured by using gamma-spectrometry. Comparison of radioactivity has been shown between calculated results and experimental data for each cooling time.

Two-Dimensional Discrete Ordinates Photon Transport Calculations for Brachytherapy Dosimetry Applications

The DANTSYS discrete ordinates computer code system is applied to quantitative estimation of water kerma rate distributions in the vicinity of discrete photon sources with energies in the 20- to 800-keV range in two-dimensional cylindrical r-z geometry. Unencapsulated sources immersed in cylindrical water phantoms of 40-cm diameter and 40-cm height are modeled in either homogeneous phantoms or shielded by Ti, Fe, and Pb filters with thicknesses of 1 and 2 mean free paths. The obtained dose results are compared with corresponding photon Monte Carlo simulations. A 210-group photon cross-section library for applications in this energy range is developed and applied, together with a general-purpose 42-group library developed at Los Alamos National Laboratory, for DANTSYS calculations. The accuracy of DANTSYS with the 42-group library relative to Monte Carlo exhibits large pointwise fluctuations from –42 to +84%. The major cause for the observed discrepancies is determined to be the inadequacy of the weighting function used for the 42-group library derivation. DANTSYS simulations with a finer 210-group library show excellent accuracy on and off the source transverse plane relative to Monte Carlo kerma calculations, varying from –4.9 to 3.7%. The P3 Legendre polynomial expansion of the angular scattering function is shown to be sufficient for accurate calculations. The results demonstrate that DANTSYS is capable of calculating photon doses in very good agreement with Monte Carlo and that the multigroup cross-section library and efficient techniques for mitigation of ray effects are critical for accurate discrete ordinates implementation.

Neutron response calculation on the basis of variable track etch rates along the secondary particle trajectories in CR-39

The calculation of the response of CR-39 detectors exposed to neutrons is of high importance for their dosimetric application. A computer code system has been developed for this purpose. Whereas the generation of secondary charged particles is carried out using non-analogue Monte-Carlo techniques with variance reduction the simulation of the track formation process is treated without any free parameter starting from the etch rate ratio V(REL) only. Results are given for the contribution of recoil protons to the response as a function of the neutron energy and angle of incidence. Furthermore, the influence of an external radiator has been studied. The comparison of the calculated values with experimental data confirm the reliability of the track etch model applied.

Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel—I: Methodology Overview

A conservative methodology is presented that would allow taking credit for burnup in the criticality safety analysis of spent nuclear fuel (SNF) packages. The method is based on the assumption that the isotopic concentration in the SNF and cross sections of each isotope for which credit is taken must be supported by validation experiments. The method allows credit for the changes in the 234U, 235U, 236U, 238U, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, and 241Am concentration with burnup. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps:1. Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. Chemical assay benchmarks are used for this purpose, in conjunction with a method for assessing the calculational bias and uncertainty for each isotope.2. Validate a computer code system to predict the subcritical multiplication factor keff of an SNF package by use of UO2 and UO2/PuO2 critical experiments. The method uses an upper safety limit on keff (which can be a function of trending parameters) to ensure that the calculated keff when increased for the bias and uncertainty is <0.95.3. Establish conditions for the SNF (depletion analysis) and package (criticality analysis) that bound keff. Bounding axial and horizontal profiles must be established to ensure that the “end effect” and “horizontal effect” are accounted for conservatively.4. Use the validated codes and bounding conditions to generate package-loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment.5. Verify by measurement that SNF assemblies meet the package-loading criteria, and confirm proper assembly selection prior to loading.

A Comparison of Measured and Predicted XV-15 Tiltrotor Surface Acoustic Pressures

Predicted XV-15 exterior surface acoustic pressures are compared with previously published experimental data. Surface acoustic pressure transducers were concentrated near the tip-path-plane of the rotor in airplane mode. The comparison emphasized cruise conditions which are of interest for tiltrotor interior noise - level flight for speeds ranging from 72 m/s to 113 m/s. The predictions were produced by components of the NASA Langley Tiltrotor Aeroacoustic Code (TRAC) system of computer codes. Comparisons between measurements and predictions were made in both the time and frequency domains, as well as overall sound pressure levels. In general, the predictions replicated the measured data well. Discrepancies between measurements and predictions were noted. Some of the discrepancies were due to poor correlation of the measured data with the rotor tach signal. In other cases limitations of the predictive methodology have been indicated.

Time-Spatial Peculiarities of Dose Formation after Inhalation of Alpha Emitters

A detailed study of the time dynamics of radiation dose formation after inhalation of alpha-emitting aerosols has been carried out, using the Internal Dosimetry Support System (IDSS) computer code. The deformation of the primary aerosol size distribution has been calculated for all respiratory tract regions. The role of endogenous short-range radiation sources (body fluids) in the irradiation of lung tissues had been studied. The ICRP 66 Respiratory Tract model takes no account of energy self-absorption within aerosol particles so non-standard calculations of alpha particle energy transport were performed to investigate this effect. For UO2 particles of 20 and 100 μm AMAD containing uniformly distributed 239 Pu the committed equivalent dose to the extrathoracic tissues is reduced by factors of 1.1 and 3.9, respectively. The corresponding factors for committed effective dose are 1.1 and 3.2, respectively.

Monte Carlo simulation of physicochemical processes of liquid water radiolysis. The effects of dissolved oxygen and OH scavenger.

The paper describes developments of the physicochemical part of a computer code system that estimates DNA strand break induction on plasmid pBR322 DNA. In order to test the reliability of the model, we evaluated the dielectric function and the time-dependent yield of chemical species in the presence of OH radical scavenger or dissolved oxygen. Results agree with measurements on the radiolysis of liquid water. When a hybrid model of a liquid inelastic cross-section and a vapour elastic cross-section is used, energy deposition by vibrational excitations is estimated to be approximately 11% of total energy deposition.

Generation and validation of ORIGEN-S libraries for depletion and transmutation calculations based on JEF2.2 and EAF3 basic data

The data libraries for light elements, actinides and fission products of the ORIGEN-S code for depletion and transmutation calculations in the SCALE4.1 computer code system have been updated with respect to cross-section data, radioactive-decay data and fission-product yield data using JEF2.2 as the basic data source and EAF3 as an additional source. This required the fission-product library to be extended with 201 new fission-product nuclides or isomeric states. The effect of the update of different quantities involved is evaluated with a burn-up benchmark. When ORIGEN-S is used as a stand-alone code, i.e. without regular update of cross-sections of the major nuclides due to changes in the neutron spectrum during burn-up, the results show appreciable differences in actinide and fission-product densities due to the cross-section update. The effects of updates of decay data and fission-product yields are generally small, but with noticeable exceptions. The update of fission and capture reaction energies gives a small but systematic change in actinide and fission-product concentration. The new ORIGEN-S libraries have also been converted for use with the SCALE4.2 package.

The Application of Probabilistic Safety Techniques to the Safe Transport of Radioactive Material

Information is presented about the various parts which make up the computer code system for the assessment of the radiological consequences and risks involved in the transport of radioactive material and which is known as the INTERTRAN 2 package. The INTERTRAN 2 package has been developed over the past seven years under a Coordinated Research Programme of the International Atomic Energy Agency (IAEA).