Dan Ilas

The U.S. Department of Energy's Office of Scientific and Technical Information

This study describes a new approach employing the Dancoff correction method to model the TRISO-based fuel form used by the Advanced High-Temperature Reactor (AHTR) design concept. The Dancoff correction method is used to perform isotope depletion analysis using the TRITON sequence of SCALE and is verified by code-to-code comparisons. The current AHTR fuel design has TRISO particles concentrated along the edges of a slab fuel element. This geometry prevented the use of the DOUBLEHET treatment, previously developed in SCALE to model spherical and cylindrical fuel. The new method permits fuel depletion on complicated geometries that traditionally can be handled only by continuous-energy-based depletion code systems. The method was initially tested on a fuel configuration typical of the Next Generation Nuclear Plant, where DOUBLEHET treatment is possible. A confirmatory study was performed on the AHTR reference core geometry using the VESTA code, which uses the continuous-energy MCNP5 code as a transport solver and ORIGEN2.2 code for depletion calculations. Comparisons of the results indicate good agreement of whole-core characteristics, such as the multiplication factor and the isotopics, including their spatial distribution. Key isotopes analyzed included ²³⁵U, ²³⁹Pu, ²⁴⁰Pu, and ²⁴¹Pu. The results from this study indicate that the Dancoff factor method can generate estimates of core characteristics with reasonable precision for scoping studies of configurations where DOUBLEHET treatment cannot be performed.

Using experimental data published in the International Handbook of Evaluated Reactor Physics Benchmark Experiments for the fresh cold core of the High Temperature Engineering Test Reactor, a comprehensive validation study has been carried out to assess the performance of the SCALE code system for analysis of high-temperature gas-cooled reactor configurations. This paper describes part of the results of that effort. The studies performed included criticality evaluations for the full core and for the annular cores realized during the fuel loading, as well as calculations and comparisons for excess reactivity, shutdown margin, control rod worths, temperature coefficient of reactivity, and reaction rate distributions. Comparisons of the SCALE results with both experimental values and MCNP-calculated values are presented. The comparisons show that the SCALE calculated results, obtained with both multigroup and continuous energy cross sections, are in reasonable agreement with the experimental data and the MCNP...

The Advanced High-Temperature Reactor (AHTR) is a 3400-MW(thermal) fluoride salt-cooled high-temperature reactor that uses coated particle fuel compacted into slabs rather than spherical or cylindrical fuel compacts. Simplified methods are required for parametric design studies to perform burnup analysis on the entire feasible design space. These simplifications include fuel homogenization techniques to increase the speed of neutron transport calculations and equilibrium depletion analysis methods to analyze systems with multibatch fuel management schemes.This paper presents three elements of significant novelty. First, the reactivity-equivalent physical transformation (RPT) methodology usually applied in systems with cylindrical and spherical geometries has been extended to slab geometries. Second, implementing this RPT homogenization, a Monte Carlo-based depletion methodology was developed to search for the maximum discharge burnup in a multibatch system by iteratively estimating the beginning of equili...

The U.S. Department of Energy's Office of Scientific and Technical Information

United States Nuclear Regulatory Commission (USNRC) regulations limit the dose to the skin to 500 mSv per year. This is also the dose limit recommended by the International Commission on Radiological Protection (ICRP). The operational quantity recommended by ICRP for quantifying dose to the skin is the personal dose equivalent, Hp(0.07) and is identical to NRC's shallow dose equivalent, Hs, also measured at a skin depth of 7 mg cm-2. However, whereas ICRP recommends averaging the dose to the skin over an area of 1 cm regardless of the size of the exposed area of skin, USNRC requires the shallow dose equivalent to be averaged over 10 cm. To monitor dose to the skin of the hands of workers handling radioactive materials and particularly in radiopharmaceutical manufacturing facilities, which is the focus of this work, workers are frequently required to wear finger ring dosimeters. The dosimeters monitor the dose at the location of the sensitive element, but this is not the dose required to show compliance (i.e., the dose averaged over the highest exposed contiguous 10 cm of skin). Therefore, it may be necessary to apply a correction factor that enables estimation of the required skin dose from the dosimeter reading. This work explored the effects of finger ring placement and of the geometry of the radioactive materials being handled by the worker on the relationship between the dosimeter reading and the desired average dose. A mathematical model of the hand was developed for this purpose that is capable of positioning the fingers in any desired grasping configuration, thereby realistically modeling manipulation of any object. The model was then used with the radiation transport code MCNP to calculate the dose distribution on the skin of the hand when handling a variety of radioactive vials and syringes, as well as the dose to the dosimeter element. Correction factors were calculated using the results of these calculations and examined for any patterns that may be useful in establishing an appropriate correction factor for this type of work. It was determined that a correction factor of one applied to the dosimeter reading, with the dosimeter placed at the base of the middle finger, provides an adequate estimate of the required average dose during a monitoring period for most commonly encountered geometries. Different correction factors may be required for exceptional or unusual source geometries and must be considered on a case-by-case basis.

This paper describes the characterization of radiation doses to the hands of nuclear medicine technicians resulting from the handling of radiopharmaceuticals. Radiation monitoring using ring dosimeters indicates that finger dosimeters that are used to show compliance with applicable regulations may overestimate or underestimate radiation doses to the skin depending on the nature of the particular procedure and the radionuclide being handled. To better understand the parameters governing the absorbed dose distributions, a detailed model of the hands was created and used in Monte Carlo simulations of selected nuclear medicine procedures. Simulations of realistic configurations typical for workers handling radiopharmaceuticals were performed for a range of energies of the source photons. The lack of charged-particle equilibrium necessitated full photon-electron coupled transport calculations. The results show that the dose to different regions of the fingers can differ substantially from dosimeter re...

A deterministic method is described for performing three-dimensional (3D) photon transport calculations of a LINAC head and phantom/patient geometry to obtain dose distributions for therapy planning. The space, energy, and directional-dependent photon flux density is obtained by numerically solving the Boltzmann equation in general 3D geometry using the method of characteristics. The deterministic transport calculations use similar ray tracing routines as found in Monte Carlo (MC) codes. A special treatment is developed to better represent the impact of scattering from accelerator head components. Equations are presented for computing the water kerma distribution due to the uncollided and collided photon flux density field in the patient region. Kerma results obtained from the deterministic computation are compared to Monte Carlo values for a variety of source spectra and field sizes. The agreement for kerma values in the beam is usually within the MC uncertainties. It is concluded that the deterministic method is a rigorous, first-principles approach that could provide a superior alternative to Monte Carlo calculations for some types of problems. However additional development is needed to provide capability for 3D electron transport calculations.

A typical advanced boiling water reactor (BWR) pin cell configuration is used to develop numerical benchmarks for the effects of moderator and fuel properties on the eigenvalue. Results are developed for pin cell configurations with different moderator densities and temperatures and different fuel exposures and temperatures. These configurations (98) correspond to those typical of cold, hot-operating, and transient conditions in B WRs. All of the calculations were performed using the Monte Carlo code MCNP with the continuous-energy cross sections derived from version VI of the Evaluated Nuclear Data Files ENDF/B.