Nuclear Technology

Published by American Nuclear Society
Online ISSN: 0029-5450
Publications
Geometry used for the neutron dose determination simulations. Treatment nozzle distances between components were the same for both sets of simulations. Distances between collimators are given from the middle of each one. Note: figure not shown to scale.  
Total neutron fluence as a function of radial distance from the beam's central axis for the three treatment volumes studied: (a) 50 × 50 × 50 mm 3 , (b) 100 × 100 × 100 mm 3 , and (c) 150 × 150 × 150 mm 3 .  
Article
Proton therapy offers low integral dose and good tumor comformality in many deep-seated tumors. However, secondary particles generated during proton therapy, such as neutrons, are a concern, especially for passive scattering systems. In this type of system, the proton beam interacts with several components of the treatment nozzle that lie along the delivery path and can produce secondary neutrons. Neutron production along the beam's central axis in a double scattering passive system was examined using Monte Carlo simulations. Neutron fluence and energy distribution were determined downstream of the nozzle's major components at different radial distances from the central axis. In addition, the neutron absorbed dose per primary proton around the nozzle was investigated. Neutron fluence was highest immediately downstream of the range modulator wheel (RMW) but decreased as distance from the RMW increased. The nozzle's final collimator and snout also contributed to the production of high-energy neutrons. In fact, for the smallest treatment volume simulated, the neutron absorbed dose per proton at isocenter increased by a factor of 20 due to the snout presence when compared with a nozzle without a snout. The presented results can be used to design more effective local shielding components inside the treatment nozzle as well as to better understand the treatment room shielding requirements.
 
Article
Monte Carlo simulations are increasingly used to reconstruct dose distributions in radiotherapy research studies. Many studies have used the MCNPX Monte Carlo code with a mesh tally for dose reconstructions. However, when the number of voxels in the simulated patient anatomy is large, the computation time for a mesh tally can become prohibitively long. The purpose of this work was to test the feasibility of using lattice tally instead of mesh tally for whole-body dose reconstructions. We did this by comparing the dosimetric accuracy and computation time of lattice tallies with those of mesh tallies for craniospinal proton irradiation. The two tally methods generated nearly identical dosimetric results, within 1% in dose and within 1 mm distance-to-agreement for 99% of the voxels. For a typical craniospinal proton treatment field, simulation speed was 4 to 17 times faster using the lattice tally than using the mesh tally, depending on the numbers of proton histories and voxels. We conclude that the lattice tally is an acceptable substitute for the mesh tally in dose reconstruction, making it a suitable potential candidate for clinical treatment planning.
 
Article
The purpose of this study was to evaluate the suitability of the quantity ambient dose equivalent H*(10) as a conservative estimate of effective dose E for estimating stray radiation exposures to patients receiving passively scattered proton radiotherapy for cancer of the prostate. H*(10), which is determined from fluence free-in-air, is potentially useful because it is simpler to measure or calculate because it avoids the complexities associated with phantoms or patient anatomy. However, the suitability of H*(10) as a surrogate for E has not been demonstrated for exposures to high-energy neutrons emanating from radiation treatments with proton beams. The suitability was tested by calculating H*(10) and E for a proton treatment using a Monte Carlo model of a double-scattering treatment machine and a computerized anthropomorphic phantom. The calculated E for the simulated treatment was 5.5 mSv/Gy, while the calculated H*(10) at the isocenter was 10 mSv/Gy. A sensitivity analysis revealed that H*(10) conservatively estimated E for the interval of treatment parameters common in proton therapy for prostate cancer. However, sensitivity analysis of a broader interval of parameters suggested that H*(10) may underestimate E for treatments of other sites, particularly those that require large field sizes. Simulations revealed that while E was predominated by neutrons generated in the nozzle, neutrons produced in the patient contributed up to 40% to dose equivalent in near-field organs.
 
Article
Monte Carlo codes are utilized for accurate dose calculations in proton radiation therapy research. While they are superior in accuracy to commonly used analytical dose calculations, they require significantly longer computation times. The aim of this work is to characterize a Monte Carlo track-repeating algorithm to increase computation speed without compromising dosimetric accuracy. The track-repeating approach reduced the CPU time required for a complete dose calculation in voxelized patient anatomy by more than two orders of magnitude, while on average reproducing the results from the traditional Monte Carlo approach within 4% dose difference and within 1-mm distance to agreement.
 
The geometry of the ocular nozzle and the water phantom used for the simulations in GEANT4 and FLUKA. The components of the nozzle (left to right) are listed in Table I.  
Article
Monte Carlo simulations of an ocular treatment beam-line consisting of a nozzle and a water phantom were carried out using MCNPX, GEANT4, and FLUKA to compare the dosimetric accuracy and the simulation efficiency of the codes. Simulated central axis percent depth-dose profiles and cross-field dose profiles were compared with experimentally measured data for the comparison. Simulation speed was evaluated by comparing the number of proton histories simulated per second using each code. The results indicate that all the Monte Carlo transport codes calculate sufficiently accurate proton dose distributions in the eye and that the FLUKA transport code has the highest simulation efficiency.
 
Article
The accuracy of proton therapy is partially limited by uncertainties that result from changing pathological conditions in the patient such as tumor motion and shrinkage. These uncertainties can be minimized with the help of a time-resolved range telescope. Monte Carlo methods can help improve the performance of range telescopes by tracking proton interactions on a particle-by-particle basis thus broadening our understanding on the behavior of protons within the patient and the detector. This paper compared the proton multiple coulomb scattering algorithms in the Monte Carlo codes MCNPX and Geant4 to well-established scattering theories. We focus only on beam energies associated with proton imaging. Despite slight discrepancies between scattering algorithms, both codes appear to be capable of providing useful particle-tracking information for applications such as the proton range telescope.
 
Article
The aim of this study was to quantify stray radiation dose from neutrons emanating from a proton treatment unit and to evaluate methods of reducing this dose for a pediatric patient undergoing craniospinal irradiation. The organ equivalent doses and effective dose from stray radiation were estimated for a 30.6-Gy treatment using Monte Carlo simulations of a passive scattering treatment unit and a patient-specific voxelized anatomy. The treatment plan was based on computed tomography images of a 10-yr-old male patient. The contribution to stray radiation was evaluated for the standard nozzle and for the same nozzle but with modest modifications to suppress stray radiation. The modifications included enhancing the local shielding between the patient and the primary external neutron source and increasing the distance between them. The effective dose from stray radiation emanating from the standard nozzle was 322 mSv; enhancements to the nozzle reduced the effective dose by as much as 43%. These results add to the body of evidence that modest enhancements to the treatment unit can reduce substantially the effective dose from stray radiation.
 
Article
A simple dosimeter design is established to monitor the space proton dose to a distributed body organ as a linear combination of ion chambers with varying wall thickness. Even dosimetric quantities, including quality and distribution factors, can be monitored.
 
Conference Paper
When a nuclear power reactor is shut down between successive operation cycles, refueling or reloading is needed. Developing a good refueling or reloading pattern is called "loading pattern optimization". It is a large, combinatorial optimization problem with a nonlinear objective function and nonlinear constraints. An algorithm based on the genetic algorithm was developed to generate optimized boiling water reactor (BWR) reloading patterns. The proposed algorithms are demonstrated in an actual BWR plant. In test calculations, candidates that shuffled fresh and burned fuel assemblies within a reasonable computation time were obtained
 
Conference Paper
Ever since the first power plant had been built in 1977, the electricity by nuclear power plants has been increased in Korea. It has been required that the spent nuclear fuels of nuclear power plants should be managed safely and therefore the technology about this should be also developed. Korea Atomic Energy Research Institute has developed the devices to manage these spent nuclear fuels. Due to high radioactivity, all these devices should be operated in a hot cell, which is a sealed room. Since these devices should be very high reliable, a real-time monitoring is necessary to check that they are working correctly. In this paper, a real-time 3 dimensional graphic simulator is proposed for the monitoring and control of the spent nuclear fuel dismantlement robots through the Internet. In order to reduce the visualization time of the devices, the abstraction of graphics data is performed. Also, simple operation information from large sensor information is extracted and an efficient message format and its communication scheme are defined to reduce the communication time over the Internet
 
Article
A new method of correlating radiation damage data is tested using actual measured data taken from the open literature. This method, the activation fluence method, was found to be as accurate as other contemporary models with which it was compared. The new method also has several ad vantages over the other methods. The method employs a new entity, the activation fluence (time-integrated specific activation rate), as the independent variables in a regression model. Temperature at which the irradiation takes place is also a variable. Although the method was tested for a specific type of damage (change in nilductility transition temperature for A302-B steel) it has no inherent restrictions and is limited only by the imagination of the user.
 
Article
The response of a 300- mu m-thick silicon detector to an incident polyenergetic neutron beam has been evaluated by the use of analytical techniques. The analysis indicates that for neutrons less than 6 MeV the response of a 300- mu m silicon detector to neutrons emanating from a plutonium dioxide (RTG) heat source is basically due to elastic scattering reactions and the contribution from other reactions is less than 2%. The contribution from radiative reactions is even smaller and therefore is ignored. For neutron energies up to 6 MeV, the maximum response for a 300- mu m silicon detector is less than 4 multiplied by 10**-**3 counts/n within the range of bias energies 25 to 250 keV. If the effects of pulse height defect and the true angular distribution of scattered neutrons are included, the response will be reduced to 1. 3 multiplied by 10**-**3 counts/n.
 
Article
The intense burst of neutrons from the d-d reaction in a plasma-focus apparatus is exploited to produce a fissioning uranium plasma. The plasma-focus apparatus consists of a pair of coaxial electrodes and is energized by a 25 kJ capacitor bank. A 15-g rod of 93% enriched U-235 is placed in the end of the center electrode where an intense electron beam impinges during the plasma-focus formation. The resulting uranium plasma is heated to about 5 eV. Fission reactions are induced in the uranium plasma by neutrons from the d-d reaction which were moderated by the polyethylene walls. The fission yield is determined by evaluating the gamma peaks of I-134, Cs-138, and other fission products, and it is found that more than 1,000,000 fissions are induced in the uranium for each focus formation, with at least 1% of these occurring in the uranium plasma.
 
Article
An expansion is derived for the solution to the transport equation in two dimensions subject to boundary conditions given for an arbitrary convex region. Questions of high-energy transport are considered along with the properties of the dose response function. The expansion of the solution of the transport equation is presented in terms of a parameter which measures the lateral dispersion of an unidirectional beam. This parameter is usually small and the expansion is expected to converge rapidly. The dominant term in the expansion is related to fluence-to-dose conversion factors in a semiinfinite slab for normal incidence. A convenient parameterization of the conversion factors is provided along with numerical examples.
 
Article
Until recently, the majority of work in geothermal energy development has been devoted to technical considerations of resource identification and extraction technologies. The increasing interest in exploiting the variety of geothermal resources has prompted an examination of the institutional barriers to their introduction for commercial use. A significant effort was undertaken by the Jet Propulsion Laboratory as a part of a national study to identify existing constraints to geothermal development and possible remedial actions. These aspects included legislative and legal parameters plus environmental, social, and economic considerations.
 
Chapter
A numerical analysis met hodformelting/solidcation phenomena has been developed to evaluate feasibility of the several candidate techniques in the nuclearfuel cycle. Our method is based on the extended finite element method, which has been usedfor moving boundary problems. The basic idea of the extendedflnite element method is to incorporate the signed distance function into the standard finite element interpolation to represent a discontinuous gradient of the temperature at a moving solid- liquid interface. This technique makes it possible to simulate movement of the solid-liquid interface without the use of a moving mesh. Construction of the finite element equation from the energy equation in the case of melting/solidification problems has been discussed and is reported here. The technique of quadrature and the method to solve the governing equations for the problem involving liquid flows have also been constructed in the present work. The numerical solutions of the basic problems-a one-dimensional Stefan problem, solidification in a two-dimensional square corner, and melting of pure gallium-were compared to the exact solutions or to the experimental data. Through these verifications, validity of the newly developed numerical analysis method has been demonstrated.
 
Article
A study is made of utilizing electron trapping in dielectrics as a means of reducing bremsstrahlung in spacecrafts at synchronous altitude. Traps retain electrons, and large internal electric fields are induced within the dielectric. Electrons penetrating the insulating material can lose most of their kinetic energy to the electric field with a subsequent decrease in energy loss to bremsstrahlung. This acts to reduce bremsstrahlung production. It also lowers the average radiation energy of that which is produced, with consequent increase in probability of absorption by the wall. Breakdown phenomenon causes the shielding effect of the trapped electrons to be cyclic. A thin layer of dielectric material on the external surface of a spacecraft should provide an effective, light, and inexpensive shield against bremsstrahlung while not interfering with any of the system functions. Electron-trap shielding is applicable not only to space, but wherever a dielectric-charge layer is allowed to accumulate.
 
Article
The assumption that spherical-shell ion chamber rssponse is equal to the dose in the center of the cavity is found to be a poor approximation for ion chambers used for area monitoring in the space program. The dose response is calculated using the appropriate areal density distribution function. Effects of nuclear reaction are evaluated using proton buildup factors. Errors of up to 100% are found for some components of the space radiation environments.
 
Article
The main problem for fluid mechanics analysis in the rocket engine is that of predicting the contained fuel mass for various propellant-to-fuel flow ratios. The analysis described here predicts a dimensionless measure of fuel mass called the fuel volume fraction. This analysis uses a coaxial free-jet computer code, and eddy viscosity equations developed for this code. The calculated variation of volume fraction with flow ratios, fuel radius, and fluid density is shown to be in general agreement with previous data. The analysis and the data predict that the required fuel volume fraction of 0.20 at the flow ratio of 50 can be obtained at a density ratio of 1.0 and a radius ratio of 0.7.
 
Article
The emission coefficient for uranium plasmas (temperature: 8000 K) was measured for the wavelength range from 1200 to 6000 A. The results were compared to theoretical calculations and other measurements. Reasonable agreement between theoretical predictions and our measurements was found in the region from 1200 to 2000 A. Although it was difficult to make absolute comparisons among the different reported measurements, considerable disagreement was found for the higher wavelength region. A short discussion regarding the overall comparisons is given, and final suggestions are made as to the most appropriate emission coefficient values to be used in future design calculations. The absorption coefficient for the same wavelength interval is also reported.
 
Article
A nuclear analysis using transport theory was made of an open cycle gas-core reactor for assumed operating conditions. Calculations were made for cavity diameters from 2.44 to 4.88 m, for hydrogen (cavity) bypass variation from 0 to 99%, for reflector thickness from 0.61 to 1.07m, and for both isotopes ²³⁵U and ²³³U as fuel. The results for these configurations indicated that ²³³U and some bypass hydrogen may be necessary to keep critical mass levels low enough to give system pressures of <1000 atm.
 
Article
The major sources of neutrons from plutonium dioxide nuclear fuel are considered in detail. These sources include spontaneous fission of several of the plutonium isotopes, (α,n) reactions with low Z impurities in the fuel, and (α,n) reactions with ¹⁸0. For spontaneous fission neutrons a value of (1.95 ± 0.07) × 10³ n/sec/g PuO2 is used. The neutron yield from (α,n) reactions with oxygen is calculated by integrating the reaction rate equation over all alpha-particle energies and all centerofmass angles. The results indicate a neutron emission rate of (1.14 ± 0.26) × 10⁴ n/sec/g PuO2. The neutron yield from (α,n) reactions with low Z impurities in the fuel is presented in tabular form for 1 ppm of each impurity. The total neutron yield due to the combined effects of all the impurities depends on the fractional weight concentration of each impurity. The total neutron flux emitted from a particular fuel geometry is estimated by adding the neutron yield due to the induced fission to the other neutron sources.
 
Article
The design of a real-time rem-rad dosimeter with sufficient generality for inclusion of dose distribution factors for space applications is discussed. This generalized dosimetric system is only slightly more complex than dosimeters in current use.
 
Article
It is anticipated that many future manned space operations will be radiation limited and that laminated wall structures and the use of new materials will be required to reduce radiation exposure. Methods for electron shield analysis are reviewed in light of anticipated needs in the space program. The most general method is still the Monte Carlo method, which is of limited usefulness for shield analysis due to excessive computer requirements. Methods based on energy deposition coefficients or energy transmission and reflection factors are quite accurate, but are currently limited to aluminum shield material. Analytical methods based on Mar's approximation for the electron transmission factor are relatively general and computer efficient but seriously underestimate shield requirements. A correction to methods using Mar's approximate transmission factor is derived herein and results in a slightly conservative estimate of shield requirements. Techniques for laminated shield design are still lacking.
 
Article
Fissioning uranium plasmas are the gaseous fuel in high-temperature cavity reactors, originally conceived for nuclear rocket propulsion in space. A predominantly pragmatic research effort, sponsored by the National Aeronautics and Space Administration, has led to the determination of the most important characteristics of the uranium nuclear fireball in gaseous core reactors. For achieving thrust at a specific impulse up to 5000 sec, the nuclear fuel must bum at a temperature in excess of 10 000 K. For criticality the uranium particle density must be not less than the molecular density of gases at standard conditions, which, in combination with the high temperature, results in a uranium plasma pressure of several hundred atmospheres. The plasma is confined by a peripherally injected propellant flow, which simultaneously intercepts the thermal radiation from the nuclear fireball and provides for an effective mechanism for heat transfer. Results of extensive research indicate that the plasma core reactor scheme is feasible. In these investigations it was assumed that because of the high pressure the fissioning plasma is optically thick. It is now believed that in gases, the energy release of fissions can lead to distributions of ionized and excited states that deviate from Maxwell-Boltzmann distributions. In that case, the fissioning plasma, or gas, exists in a nonequilibrium state and is optically thin. This condition can be exploited for the direct conversion of fission fragment energy into coherent light, that is, for the nuclear-pumped lasers. In current research, the nonequilibrium conditions of fissioning plasmas and gases are emphasized, culminating in the first successful demonstrations of experimental nuclear-pumped lasers, and in a program of gaseous fuel reactor experiments with enriched uranium hexafluoride. A variety of applications of plasma core reactors and nuclear-pumped lasers is now envisioned for benefits in space and on earth. Such benefits include advanced propulsion in space, terrestrial power generation approaching 70% efficiency, the possibility of nuclear bumup of transuranium actinides wastes, and the breeding of ²³³U from thorium. The research on gaseous fuel reactors and nuclear-pumped lasers predominantly requires expertise in nuclear engineering, plasma, atomic, and molecular physics, and fluid mechanics and chemistry. A multidisciplinary effort is seen as a logical approach.
 
Article
A conceptual two-stage microthruster employing sublimed molecules as the propellant of the first stage and alpha particles as the propellant of the second stage is analyzed. Transient thermal analysis is developed as a tool to provide design information on material compositions. The numerical result suggests that the radioisotope heating of the nozzle wall could help flatten the subliming surface temperature. The modified Monte Carlo technique for analyzing the vapor flow through the nozzle is shown to be very accurate and flexible. The result clearly points the direction to follow to optimize the performance. Applications of this device are particularly suited in the spin stabilization and the precession damping of small scientific probes.
 
Article
An Adaptive Neural Fuzzy Inference System (ANFIS) modeling technique is introduced for sensor and associated instrument channel calibration validation. This method uses an inferential modeling technique after a genetic algorithm search is used to empirically determine the appropriate combinations of input variables to optimally model each signal to be monitored. These variables are used as input to a fuzzy inference system which is trained to estimate the monitored signals. The estimates are compared to the actual signals and a statistical decision technique known as the Sequential Probability Ratio Test (SPRT) is used to detect sensor anomalies. The sensor fault detection system is demonstrated using data supplied from Florida Power Corporation's Crystal River #3 Nuclear Power generating station. 4 I. INTRODUCTION In large power generating systems and process control systems, outputs from many different channels are used in control systems, safety critical systems, and for plant s...
 
Article
This work presents an empirical modeling approach combining a bilinear modeling technique, Partial Least Squares, with the universal function approximation abilities of single hidden layer non-linear artificial neural networks. This approach, referred to as Neural Network Partial Least Squares, is compared to the common Autoassociative Artificial Neural Network. The Neural Network Partial Least Squares model has been embedded into a graphical user interface and implemented at the Electrical Power Research Institute's Instrumentation and Control Center located at Tennessee Valley Authority's Kingston fossil power plant. Results are presented for 51 process signals with an average absolute estimation error of ~1.7% of the mean value, and sample drift detection performances are shown.
 
Article
The application of a symbiosis between light water reactors (LWRs) and ²³⁵U-Pu advanced pressurized water reactors (APWRs) has been found to have certain positive features as a strategy interim to the introduction of fast breeders and Pu-Udepl APWRs. On the basis of a particular model for the two-component system, it has been quantitatively shown how, as a result of the lower Pufiss inventory of the ²³⁵U-Pu APWR as well as its self-sufficiency in plutonium, the installed APWR capacity can grow faster than is the case for Pu-Udepl APWRs. The benefits, however, are to be realized at the expense of an increased absolute uranium ore consumption, since the ²³⁵U-Pu APWR does require a finite enriched-uranium feed. While, from the point of view of global energy policy, the fast breeder clearly holds the key to a nuclear generating capacity in the terawatt(electric) range, the present delays in its large-scale commercialization render it important to evaluate the pros and cons of alternative interim strategies. It is seen that such evaluations need to be made from the twin viewpoints of (a) improved uranium utilization, relative to standard L WRs, and (b) the quantities of effectively “stored” fissile plutonium.
 
Horizontal sectional views of the ~a! subcritical and ~b! critical cores: fuel subassembly; reflector; control rod.
Vertical sectional views of the ~a! subcritical and ~b! critical systems: 1 reactor core, 2 fuel zone, 3 diagrid, 4 riser channel, 5 inner vessel, 6 reactor vessel, 7 safety vessel, 8 reactor roof, 9 vessel support, 10 rotating plug, 11012 above core structures, 13 target unit, 14 intermediate heat exchangers, and 15 transfer machine.
Isometric view of two secondary loops: 1 intermediate heat exchangers ~2 per loop!, 2 air cooler ~3 per loop!, 3 pumps ~1 per loop!, and 4 natural circulation bypass ~1 per loop!.
Nodalization diagram of the primary circuit ~subcritical design!.
Nodalization diagram of the secondary circuit.
Article
A consistent analytical comparison has been made of the transient behavior of critical and subcritical fast-spectrum reactor systems, the basic core design assumed in each case being that of the 80-MW(thermal) mixed-oxide-fueled, Pb-Bi-cooled, experimental accelerator driven system (XADS). The transient calculations were performed using the FAST code system developed at the Paul Scherrer Institute. The present study demonstrates a high level of self-protection of both the critical and subcritical systems over a wide range of postulated events, including transient overpower due to reactivity insertion, loss of flow, station blackout, loss of coolant, and core overcooling accidents. The relative advantages and shortcomings of the two system types, from the viewpoint of transient behavior, are discussed on the basis of the corresponding simulation results obtained.
 
The Intersecting Storage Rings (ISR), the world's first hadron collider (photo CERN)
Attenuation of radiation by the RF chicane in LEP measured at six beam energies (80.5 to 102 GeV). The chicane separated the straight section (SS) housing the superconducting RF cavities (left) from the arc (right). The chicane was built to prevent synchrotron radiation streaming from the arc into the SS. The figure shows that an even stronger radiation source, produced by the operation of the cavities, was actually present in the SS. Dose rates are normalised to 1 Ah of total circulating beam current (Gy per Ah) [10].
Article
From the early days, accelerator shielding and radiation protection in general have been influenced by the increasing knowledge of the health effects of ionizing radiation, which has progressively decreased the dose rates allowed in occupied areas. At the same time the tools available for estimating shielding have benefited from increasing computing power. Nonetheless, the simplified models of the early days are often still useful for a first estimate before going into complex and detailed Monte Carlo simulations. This paper provides a brief historical overview of accelerator shielding. As the subject is vast, it is restricted to a review of the various phases of accelerator shielding studies at CERN. CERN is a good example as its accelerators and the related shielding problems span 50 yr of history and cover all major aspects that can be encountered in accelerator radiation protection: various types of accelerators with a wide range of beam intensities producing many varieties of accelerated and secondary particles with energies from MeV to TeV.
 
Article
Some of the most frequently presented scenarios for the initial power excursion of the Chernobyl accident are evaluated based on computer simulations. The applied transient model uses one-dimensional descriptions of the reactor core and the main flow circuit. According to the simulations, a slow flow decrease caused by gradual slowing down of the four main circulation pumps could have initiated the accident only if the void reactivity coefficient had been considerably larger than the original Soviet figure. On the other hand, a faster flow reduction, such as pump cavitation or deliberate stopping of even some of the pumps, would have produced enough void for prompt criticality. However, this scenario is sensitive to the size of the void coefficient and to the amount of flow reduction. The most probable initiator was considered to be the positive scram caused by the graphite followers of the manual control rods. Such a mechanism would naturally have brought the additional reactivity to the bottom half of the reactor, and the timing of the power surge would have been the reported one. The simulations indicated that the positive scram was possible only because of the double-humped axial power profile that probably prevailed in the reactor before the accident. The simulations also demonstrated the inability of the shutdown system in this sequence.
 
Article
Hygroscopic aerosols were studied at 40° C at different relative humidity levels in a flow-type reactor chamber. The main interest was in growth of hygroscopic aerosols under higher humidity conditions. The time development of the aerosol in the 0.01- to 17-µm size range was determined using electrical aerosol analyzer and optical particle counter aerosol analyses. Low velocity and laminar flow were used to facilitate the comparison with theoretical considerations. Cesium hydroxide (CsOH) and sodium hydroxide (NaOH) were used as hygroscopic materials. Cesium is one of the most abundant species in core melt release, and NaOH is well known for its hygroscopic properties. The primary particles were produced by a constant output atomizer. The dry particle size, as volume median diameter (VMD), for CsOH was 1.8 µm. The observed airborne particle size after 2 min of travel was 6.3 µm, after 5 min 5.3 µm, and after 10 min 3.4 µm at saturated conditions. For dry NaOH aerosol, the measured initial VMD was 2.7 µm. At saturated conditions, the observed VMD was 6.7 µm after a 10-min travel. Theoretical calculations with the modified NAUA code showed that during travel through the chamber, the particle size change can be attributed to hygroscopic growth and sedimentation.
 
Article
Methods to calculate bounding values for the generation of fuel vapor during disassembly and during expansion of the fuel after disassembly in a fast-reactor core-disruptive accident were developed. Isentropic expansion of the fuel following disassembly with no fuel mixing before expansion was assumed. It was necessary to develop consistent thermodynamic fuel properties for the analysis. The method was applied to a liquid-metal fast breeder reactor disassembly, first with sodium in the core and then with sodium removed. Bounding values were also compared to lower values obtained by assuming mixing and thermal equilibrium of the fuel prior to expansion. For the bounding calculation with sodium removed, 4.6% of the fuel vaporized when the expanded fuel occupied all of the available volume. This value was reduced to 0.9% when mixing and thermal equilibrium prior to expansion was assumed.
 
Thermodynamic stability of RuO 4 (g), RuO 3 (g) and RuO 2 (s) as a function of temperature for respectively 1 mole of Ru, 100 moles of H 2 O, and 200 moles of O 2 (15% molar).
Air radiolysis compounds concentration profiles as a function of time ~DR 10 kG0h at 373 K, steam mass fraction 30%!.  
Article
During an hypothetical severe accident on a Pressurized Water Reactor (PWR), Fission-Products (FPs) are released from the nuclear fuel and may reach the reactor containment building. Among the FPs, ruthenium is of particular interest due to its ability to form volatile oxide compounds in highly oxidizing conditions. In addition, Ru is a very hazardous compound because it is chemically toxic and also because of its radiotoxicity. The topic of ruthenium is examined in terms of nuclear safety issues. A review of the literature regarding ruthenium oxides properties, gaseous and aqueous chemistry is compiled. The study focuses on the ruthenium tetroxide (RuO4) which is highly reactive and volatile, and is the most likely gaseous chemical form under the conditions prevailing in the containment. The interactions between ruthenium oxides and containment surfaces, which could be of main importance in the overall Ru behaviour, are also discussed. Finally, an evaluation of the possible revolatilisation phenomena of ruthenium adsorbed on PWR containment surfaces, or dissolved in the sump, under super-oxidizing conditions (radiolysis), is also presented. In this case, ruthenium dioxide (RuO2) must also be considered. Knowledge of all these phenomena is required to accurately predict ruthenium behaviour, and to make best estimate assessment of the potential ruthenium source-term.
 
Article
Severe water ingress accidents in the 200-MW HTR-module were assessed to determine the safety margins of modular pebble-bed high-temperature gas-cooled reactors (HTR-module). The 200-MW HTR-module was designed by Siemens under the criteria that no active safety protection systems were necessary because of its inherent safe nature. For simulating the behavior of the HTR-module during severe water ingress accidents, a water, steam, and helium multiphase cavity model was developed and implemented in the dynamic simulator for nuclear power plants (DSNP) simulation system. Comparisons of the DSNP simulations incorporating these models with experiments and with calculations using the time-dependent neutronics and temperature dynamics code were made to validate the simulation. The analysis of the primary circuit showed that the maximum water concentration increase in the reactor core was
 
Article
The objective of this paper is validation of a combined MCNP&ORIGENS 3D model for reactivity predictions of the entire BR2 core during reactor operation. MCNP is used for evaluation of the effective multiplication factor keff and 3D space dependent specific power distribution. The 1D code ORIGENS is used for calculation of isotopic fuel depletion versus burnup and preparation of a database (DB) with depleted fuel compositions. The approach taken is to evaluate the 3D power distribution at each time step and along with DB to predict the 3D isotopic fuel depletion at the next step and to deduce the corresponding shim rods positions of the reactor operation. The reactor has a complex operation, with important shutdowns between cycles, and its reactivity is strongly influenced by poisons 3He and 6Li from the beryllium reflector, and burnable absorbers 149Sm and 10B in the fresh UAlx fuel. Our computational predictions for the shim rods position at various restarts are within +/-30 mm. The computational time for a Monte Carlo simulation by MCNP of one depletion time step for the 3D full scale heterogeneous geometry reactor model, containing ~ 4000 cells with varied fuel depletion, is 10000 histories/minute on a single PC PENTIUM-4/2GHz.
 
Article
Thesis (M.S.E.(Nuclear))--University of Washington, 1981. Includes bibliographical references (leaves [114]-116).
 
Article
To develop an advanced partitioning process by extraction chromatography using a minimal organic solvent and compact equipment to separate minor actinides such as Am and Cm from nitrate acidic high-level waste (HLW) solution, several novel silica-based extraction resins have been prepared by impregnating organic extractants into the styrene-divinylbenzene copolymer, which is immobilized in porous silica particles (SiO2-P). The extractants include octyl(phenyl)-N, N-diisobutylcarbamoylmethylphosphine oxide (CMPO), di(2-ethylhexyl)-phosphoric acid (HDEHP), and bis(2,4,4-trimethylpentyl)dithiophosphinic acid (Cyanex 301). Compared to conventional polymer-matrix resins, these new types of extraction resin are characterized by rapid kinetics and significantly low pressure loss in a packed column. The results of separation experiments revealed that trivalent actinides and lanthanides can be separated from other fission products, such as Cs, Sr, and Ru in simulated HLW solution containing concentrated nitric acid by extraction chromatography using a CMPO/SiO2-P resin-packed column. Satisfactory separation between Am(III) and a macro amount of lanthanides from simulated HLW solution with pH 4 was achieved by using a newly purified Cyanex 301/SiO2-P resin. However, the Am(III) separation was very sensitive to the purity of Cyanex 301, and the improvement of its stability is an important task for practical utilization.
 
Article
This work is related to the design of the core of the eXperimental demonstration of the technological feasibility of Transmutation in an Accelerator-Driven System (XT-ADS) facility in the framework of the EUROpean Research Programme for the TRANSmutation of High Level Nuclear Waste in an Accelerator Driven System (EUROTRANS) project. The design specifications for the proton accelerator of the XT-ADS are 600 MeV and up to 3.5 mA for the beam energy and current, respectively. The proton beam impinges on a liquid target consisting of a lead-bismuth-eutectic mixture. The state-of-the-art Monte Carlo code MCNPX was used to assess the neutronics performance and shielding properties of the system. The nuclear data-processing system NJOY 99 was also used. The work consisted of the optimization of the core configuration (geometry, number, and location of the fuel and absorber assemblies) and the appropriate fuel composition in order to reduce radiation damage (namely, the displacement per atom values) on the core barrel and top grid plate, while maintaining the high neutron fluxes (10¹⁵ n·cm–2·s–1) and the keff of the system of ~0.95. The assessment of the core configuration and fuel composition was performed, resulting from the interplay among parameters such as the desired high neutron fluxes, the keff value wanted for safety and core performance reasons, the as-low-as-possible radiation damage of the core barrel and top grid plate, and the fuel composition, among others.
 
Article
Three-dimensional continuous-energy coupled neutron-gamma Monte Carlo models of the Advanced Neutron Source (ANS) final preconceptual and conceptual reference core designs have been developed using the Monte Carlo Neutron and Photon transport code (MCNP) Version 3b. These models contain the reactor core with control rods, the heavy water reflector tank with shutdown rods and some beam tubes, and the outer light water pool. Eighty homogenized fuel zones per fuel element are used to represent the radial and axial ²³⁵U fuel distribution. These models are the most sophisticated, physically accurate reactor physics models of the ANS currently available. The use of MCNP methods and applications to the ANS are demonstrated. Beam tube studies, coolant voiding studies, and many criticality studies have already been performed, as have studies with variance reduction techniques. In comparison with deterministic methods, MCNP proves superior in calculating the core multiplication factor and neutron fluxes in the reflector. The MCNP code offers the ANS project the capability of performing complicated reactor physics calculations not currently possible with most deterministic methods.
 
Geology of the Mol site.
Reduction of the 3D real world problem to a 2D model Table III gives the values of the parameters in the clay layer used for the transport calculations.  
Evolution of the dose via the river pathway per TWh(e) for the considered fuel cycle scenarios
Article
We made evaluations of the impact of advanced fuel cycles on the dimensions and the radiological consequences of a geological repository in a clay layer for the disposal of high-level radioactive waste and spent fuel. The thermal output of the high-level radioactive waste arising from advanced fuel cycles is significantly lower than the one of spent fuel. This allows to reduce the dimensions of the geological repository. The impact of advanced fuel cycles on the radiological consequences in the case of the expected evolution scenario is rather limited. The maximum dose, which is expected to occur a few tens of thousands year after the disposal of the waste, is essentially due to fission products and their amount is about proportional to the heat generated by nuclear fission. By considering the evolution of the radiotoxicity of the waste, it can be expected that the consequences of human intrusions into the repository will be significantly lower in case of waste arising from advanced fuel cycles.
 
Article
An exact closed-form solution for the stresses and strains in an idealized nuclear reactor fuel pin under operational conditions is presented. The fuel is considered as a single region, either solid or annular, which may or may not interact with the surrounding cladding, depending on initial fuel-cladding gap and subsequent reactor operating parameters. Temperature-dependent thermal conductivity and irradiation swelling and temperature-independent creep in both fuel and cladding are allowed. Although the model is considerably simplified from those used in the more detailed numerical simulations, design parameters of interest can be easily and readily studied, and the important mechanisms contributing to cladding deformation can be identified. More importantly, however, the exact solutions can be used as a benchmark to check the accuracy of the more detailed but necessarily approximate numerical techniques. Example calculations are presented for a fuel pin operating under typical liquid-metal fast breeder reactor conditions for cases with and without fuel-cladding interaction occurring over the lifetime of the pin.
 
Article
Thesis (M.S.)--Pennsylvania State University, 2006. Library holds archival microfiches negative and service copy.
 
Article
We have studied a new aqueous reprocessing system which consists of anion exchange as main separation method, electrolytic reduction for reducing U(VI) to U(IV) and extraction chromatography for MA partitioning. In this work, hot tests were carried out on the main flow sheet (U and Pu recovery) using a nitric acid solution of a spent commercial BWR-fuel with burnup of 55,000 MWd/tHM. Firstly, a separation experiment was conducted using a column packed with AR-01 anion exchanger, and the separation behavior of about 20 elements was examined. Then electrolytic reduction was performed for the U(VI)-eluate from the 1st column using a flow type electrolysis cell. Subsequently, the reduced U-solution was applied to the 2nd AR-01 column to separate the U(IV) from contaminated FPs. Most amounts of Pu(IV)-Np(IV) were successfully separated and recovered in the 1st column. In the electrolysis, U(VI), Np(V,VI) and a trace amount of Pu(VI) were reduced to U(IV), Np(IV) and Pu(IV), respectively. In the 2nd column, the U(IV) with small amounts of Np(IV) and Pu(IV) was completely separated from FPs. These results demonstrated that the proposed U and Pu recovery process is essentially feasible, though more effective elution methods for Pd and Tc need to be investigated further.
 
Article
The NEPTUN test facility at Würenlingen, Switzerland, has been modified to enable light water high conversion reactor (LWHCR) representative reflooding and boiloff experiments to be carried out. Results from a first series of forced feed reflooding tests, simulating cold-leg injection, are presented for a range of values of the flooding rate, rod power, and initial rod temperature parameters. Rewetting of the LWHCR fuel bundle simulator was found to be possible in each case. Analysis of the NEPTUN-III reflooding experiments with RELAP5/MOD2 yield discrepant results, and it has been shown, in the context of calculations of the boiloff experiments, that some LWHCR specific models and correlations need to be developed.
 
Article
The subject is estimation of the maximum values of the heat flux at steady-state nominal operating conditions in the reactor BR2. A strong variation of the fuel depletion and the heat flux with the azimuthal direction and dependence on the orientation of the fuel element in the core are obtained. The full-scale 3-Dimensional MCNP&ORIGEN-S heterogeneous geometry model of BR2 with detailed 3-D isotopic fuel depletion profile, including a detailed azimuth fuel modeling, is developed. The relative azimuth power distribution is calculated with MCNP and introduced into ORIGEN-S to evaluate the azimuth isotopic fuel profile. The applied detailed azimuth fuel modeling is compared with homogeneous one in the hot plane. An increase of the maximum heat flux value of 5% for low burnt fuel and 20% for highly burnt fuel due to the azimuth fuel modeling is obtained. A strong variation of the heat flux with the orientation of the fuel element in the core, modeled with azimuthal fuel profile is observed. Perturbation effects in the maximum heat flux values of 10% for a low burnt fuel and ~ 40% for highly burnt fuel, correlated to different orientations of the fuel element in the core are obtained.
 
Article
The minimum overall size of a reflected pebble-bed reactor is, in general, considerably smaller than that of the corresponding bare-critical assembly. It becomes meaningful, therefore, to refer to reflector savings in terms of overall size and not just in terms of core dimensions or fuel requirements, as is usually the practice for other types of thermal reactors. This paper considers, for the purpose of illustration, numerical results for a spherical reactor fueled with low-enriched uranium (LEU) pebble-bed fuel elements at average burnup. The factors that contribute to the significant size-savings from the reflector are discussed, as are certain practical implications such as the possible dual criticality of a system of fixed outer dimensions.
 
Article
Investigation of the initial core poisoning of the pebble bed high temperature reactor has been made by experiments and by checking computations. In following the example of the thorium high-temperature reactor THTR-300, THTR absorber elements poisoned with hafnium-boron were added to the THTR fuel and graphite elements of the KAHTER core. Three different hafnium-boron poisoned core loadings, corresponding to 2.7, 5.3, and 8% reactivity compensation, were used in the experiments. For purposes of comparison, two cores poisoned exclusively with boron were also studied. The poisoning of these cores corresponds to 2.7 and 8% reactivity compensation, respectively. The experiments and checking computations should serve to test the accuracy of the theoretical models and data sets in modeling the reactivity effects of absorber poisoned elements in the THTR. In particular, the applicability of the nuclear data of hafnium and the treatment of resonance calculations should be verified. In addition to determining critical masses and keff, special emphasis was placed in the experiments on the exact determination of all reactivity effects. In some cases, repeated loading of a configuration also provided a measure of the reproducibility of keff. The experiments were checked computationally using the GAMTEREX code package and the program system RSYST. These two computation packages contain different data bases, although the hafnium data are identical, and the computing models differ in certain phases of the calculations. Both code systems compute keff values to within the present accuracy requirements, whereas the program system RSYST gives better agreement with experimental measurements.
 
Article
Finnish spent nuclear fuel final disposal is planned to be based on the Kärnbränslesäkerhet 3-Vertical concept, which was originally planned for fractured crystalline bedrock. Within this concept, the role of the bentonite buffer is considered central. The aim of the study was to model the evolution of the final repository during the thermal phase (heat-generating period of spent fuel) when the bentonite is initially only partially saturated. There is an essential need to determine how temperature influences saturation and how both of these factors affect the chemistry of bentonite. In this study the Long-Term Test of Buffer Materials A2 parcel test at the Äspö hard rock laboratory in Sweden was modeled using TOUGHREACT code. The results focused on the following phenomena occurring in the bentonite: cation exchange, changes of bentonite pore water, mineral alterations, saturation, and pressure changes in bentonite buffer. The results show similarity with experimental data. However, the results are open to questions, and further study is needed to confirm the validity of the results. Differences between modeled and experimental results can be explained, for example, so that the experimental results are not from the fracture position as our one-dimensional model assumes.
 
Top-cited authors
Pavel Hejzlar
  • Massachusetts Institute of Technology
Michael Lorne Fensin
  • Los Alamos National Laboratory
R. A. Forster
  • Los Alamos National Laboratory
Thomas Edward Booth
  • Los Alamos National Laboratory
Roger Martz
  • Los Alamos National Security