Shielding and Activation Analyses in Support of the Spallation Neutron Source (SNS) ES and H Requirements

Shielding and activation analyses play an important part in determining how to meet the Environmental, Safety and Health (ES and H) requirements of an intense high-energy accelerator facility like the proposed Spallation Neutron Source (SNS). The shielding and activation analyses described in this paper were performed primarily using the CALOR code system coupled with MCNP for radiation transport, the ORIHET95 isotope generation and depletion code for activation analysis, and the DOORS multi-dimensional discrete ordinates transport code system for shielding analyses. Additionally, a portion of the shielding calculations were performed with the semi-empirical code - CASL. This paper gives an overview of relevant ES and H policies and requirements, and provides detailed discussions of the shielding and activation analyses completed in support of those policies and requirements

The Spallation Neutron Source (SNS) conceptual design shielding analysis

The shielding design is important for the construction of an intense high-energy accelerator facility like the proposed Spallation Neutron Source (SNS) due to its impact on conventional facility design, maintenance operations, and since the cost for the radiation shielding shares a considerable part of the total facility costs. A calculational strategy utilizing coupled high energy Monte Carlo calculations and multi-dimensional discrete ordinates calculations, along with semi-empirical calculations, was implemented to perform the conceptual design shielding assessment of the proposed SNS. Biological shields have been designed and assessed for the proton beam transport system and associated beam dumps, the target station, and the target service cell and general remote maintenance cell. Shielding requirements have been assessed with respect to weight, space, and dose-rate constraints for operating, shutdown, and accident conditions. A discussion of the proposed facility design, conceptual design shielding requirements calculational strategy, source terms, preliminary results and conclusions, and recommendations for additional analyses are presented

Development of the activation analysis calculational methodology for the Spallation Neutron Source (SNS)

For the design of the proposed Spallation Neutron Source (SNS), activation analyses are required to determine the radioactive waste streams, on-line material processing requirements remote handling/maintenance requirements, potential site contamination and background radiation levels. For the conceptual design of the SNS, the activation analyses were carried out using the high-energy transport code HETC96 coupled with MCNP to generate the required nuclide production rates for the ORIHET95 isotope generation code. ORIHET95 utilizes a matrix-exponential method to study the buildup and decay of activities for any system for which the nuclide production rates are known. In this paper, details of the developed methodology adopted for the activation analyses in the conceptual design of the SNS are presented along with some typical results of the analyses

Preliminary Shielding Analysis and Design of the Remote Maintenance Cells for the Proposed National Spallation Neutron Source (NSNS)

Radiation shielding analysis and design calculations were performed for remote maintenance cells of the proposed National Spallation Neutron Source (NSNS) facility. In the analysis, a calculational strategy utilizing coupled high energy Monte Carlo calculations and multi-dimensional discrete ordinates calculations was implemented to perform an activation analysis and shielding assessment of the NSNS remote handling cells. A general description of the remote maintenance cells, the methodology employed, and preliminary results of the shielding analysis and recommendations are presented

Conversion factor and uncertainty estimation for quantification of towed gamma-ray detector measurements in Tohoku coastal waters

Factors to convert the count rate of a NaI(Tl) scintillation detector to the concentration of radioactive cesium in marine sediments are estimated for a towed gamma-ray detector system. The response of the detector against a unit concentration of radioactive cesium is calculated by Monte Carlo radiation transport simulation considering the vertical profile of radioactive material measured in core samples. The conversion factors are acquired by integrating the contribution of each layer and are normalized by the concentration in the surface sediment layer. At the same time, the uncertainty of the conversion factors are formulated and estimated. The combined standard uncertainty of the radioactive cesium concentration by the towed gamma-ray detector is around 25 percent. The values of uncertainty, often referred to as relative root mean squat errors in other works, between sediment core sampling measurements and towed detector measurements were 16 percent in the investigation made near the Abukuma River mouth and 5.2 percent in Sendai Bay, respectively.Most of the uncertainty is due to interpolation of the conversion factors between core samples and uncertainty of the detector?s burial depth. The results of the towed measurements agree well with laboratory analysed sediment samples. Also, the concentrations of radioactive cesium at the intersection of each survey line are consistent. The consistency with sampling results and between different lines? transects demonstrate the availability and reproducibility of towed gamma-ray detector system