Tower Shielding Facility Research Articles

Dismantlement of the TSF-SNAP Reactor Assembly

This paper describes the dismantlement of the Tower Shielding Facility (TSF)-Systems for Nuclear Auxiliary Power (SNAP) reactor, a SNAP-10A reactor used to validate radiation source terms and shield performance models at Oak Ridge National Laboratory (ORNL) from 1967 through 1973. After shutdown, it was placed in storage at the Y-12 National Security Complex (Y-12), eventually falling under the auspices of the Highly Enriched Uranium (HEU) Disposition Program.To facilitate downblending of the HEU present in the fuel elements, the TSF-SNAP was moved to ORNL on June 24, 2006. The reactor assembly was removed from its packaging, inspected, and the sodium-potassium (NaK) coolant was drained. A superheated steam process was used to chemically react the residual NaK inside the reactor assembly. The heat exchanger assembly was removed from the top of the reactor vessel, and the criticality safety sleeve was exchanged for a new safety sleeve that allowed for the removal of the vessel lid. A chain-mounted tubing cutter was used to separate the lid from the vessel, and the 36 fuel elements were removed and packaged in four U.S. Department of Transportation 2R/6M containers. The fuel elements were returned to Y-12 on July 13, 2006.The return of the fuel elements and disposal of all other reactor materials accomplished the formal objectives of the dismantlement project. In addition, a project model was established for the handling of a fully fueled liquid-metal–cooled reactor assembly. Current criticality safety codes have been benchmarked against experiments performed by Atomics International in the 1950s and 1960s. Execution of this project provides valuable experience applicable to future projects addressing space and liquid-metal–cooled reactors.

大型高速炉遮蔽解析手法の開発 ＪＡＳＰＥＲ（日米共同高速炉遮蔽実験）実験解析の成果

All experiments planned in the JASPER, Japanese-American Shielding Program for Experimental Research, had been conducted between 1986 and 1992 under the cooperation between the U.S. Department of Energy and Power Reactor and Nuclear Fuel Development Corporation of Japan to obtain useful information on FBR shielding analysis accuracy. Tower Shielding Facility (TSF) of Oak Ridge National Laboratory was utilized for all the measurements. Results obtained from the post-experimental analyses are summarized in this report. Both characteristics for bulk shielding attenuation and streaming of neutron through configurations consisting of several materials including boron-carbide (B4C), graphite, stainless steel, sodium and etc. were clarified and the analysis accuracy confirmed. Prediction accuracy in both shielding property for configuration containing fissile material and sodium activation characteristics was confirmed. The shielding analysis system for fast reactors using a 2-dimensional discrete ordinates code as a standard has been improved and verified. Through those activities, useful information to be utilized in designing large FBR is accumulated.

Experimental Verification of Calculated Neutron Streaming in the Bottom Reflector and Core Support Regions of a Large-Scale High-Temperature Gas-Cooled Reactor Design Concept

The analysis of an Oak Ridge National Laboratory Tower Shielding Facility (TSF) experiment in which measurements were made of neutrons streaming through a mockup of a section of the lower core support structure of a large-scale high-temperature gas-cooled reactor (HTGR) design concept is described. The analysis was performed with the same calculational methods used for an analysis of the HTGR design itself, the purpose of the experiment being to provide data against which the validity of the calculational methods could be tested. Also summarized are the HTGR design calculation results; how they affected the design and objectives of the TSF experiment is described. Comparisons of the neutron detector responses observed in the experiment with calculated responses showed satisfactory agreement in most cases, and the implications of these results for the HTGR shield design are highlighted.

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Measurement and Calculation of the Effectiveness of the Gas-Cooled Fast Breeder Reactor Grid-Plate Shield

An analysis of the Oak Ridge National Laboratory Tower Shielding Facility gas-cooled fast breeder reactor (GCFR) grid-plate shield design confirmation experiment was performed and inferences were m...

Measurement and Analysis of the 23Na(n,xγ) Reaction Cross Section for 0.2 ≤ En ≤20 MeV

Differential cross sections for neutron-induced gamma-ray production from sodium have been measured for incident-neutron energies between 0.2 and 20.0 MeV. Gamma rays with energies 0.35 ≤ Eγ ≤ 10.6 MeV were detected with a sodium iodide spectrometer at 125 deg. The data presented are the double-differential cross section, d2σ/dΩdE, for coarse intervals in incident-neutron energy. The measured results are compared with existing data, with calculations based on multistep Hauser-Feshbach theory, and with a benchmark gamma-ray production measurement performed at the Oak Ridge Tower Shielding Facility (TSF). Average agreement between our measured results and model calculations is within 15%. The cross sections measured at the TSF are typically 30% larger than our results, except for gamma-ray energies between 1.1 and 1.5 MeV where the TSF benchmark predicts a yield 20 times greater than we observe. Results of the present measurement have been incorporated for the gamma-ray production in the Evaluated Nuclear Data File.

The ORNL Benchmark Experiment for Neutron Transport in Thick Sodium

An experiment concerning deep neutron penetration in sodium was performed, and experimental results were obtained which provide a basis for verification of the accuracy of sodium cross sections to be used in transport calculations. The experiment was conducted at the Tower Shielding Facility of Oak Ridge National Laboratory and included measurements of both the neutron fluence and the neutron spectra through a large diameter sample of sodium up to 15 ft thick. Calculated results for the experiment were also compared with the experimental measurements. These results were obtained using the multigroup Monte Carlo code, MORSE, and a two-dimensional discrete ordinates code, DOT-III. One-hundred group data sets were developed from both a preliminary and the final version of the ENDF/III set (MAT-1156) for sodium for use in the calculations. Comparisons of the calculations with experiment indicate that (a) the preliminary version is slightly superior to the final version and (b) using the preliminary set, the total neutron leakage above thermal energies penetrating through 15 ft of sodium agrees to within ∼15%; and the absolute spectra penetrating through 12.5 ft of sodium, when integrated over the energy range of the measurement, agrees to within 20%. Using the final set, the corresponding comparisons are 30% and 60%.

Moments Method Calculations of Neutron Distributions in Concrete

The moments method is applied to the problem of calculating nuclear distributions in an infinite medium. Several comparisons are given of these results with similar data calculated by the discrete ordinates method. New calculations are presented on the distribution of doses from neutrons, originating in a plane-slant fission source and incident, at various angles, on concrete utilized in radiation measurements at the Tower Shielding Facility of the Oak Ridge Nationai Laboratory (TSF concrete). For a given set of neutron cross sections, these results give reliable estimates of the dose distribution at deep penetrations, i.e., attenuation of six orders of magnitude or more. Functional representations of the distributions are included in order to facilitate the use of the data in shield design calculations. (auth)

A comparison of folding and unfolding techniques for determining the gamma spectrum from thermal neutron capture in aluminum

The gamma-ray spectrum above 1 MeV resulting from thermal-neutron capture in aluminum is determined by both a folding and an unfolding analysis of data taken at the Tower Shielding Facility. This results in a comparison of the relative merits of the two analysis techniques and demonstrates the accuracy that can be obtained in unfolding a complex spectrum. It is concluded that the accuracy of the final photon spectrum is not governed by the technique employed to unfold the measured counts.

Analytical radiation shielding calculations for concrete — formulas and parameters

Formulas for analytical radiation shielding calculations are discussed in view of their applicability to concrete shielding. Based on neutron removal-diffusion theory and gamma-ray buildup factors, they provide a reasonable tool to compare shielding properties of different concretes. Therefore, parameters for these formulas were calculated from recent cross sections. A sample calculation is given for the concrete tested at the Oak Ridge National laboratory (USA) Tower Shielding Facility.

Neutron Activation of Sodium in Anthropomorphous Phantoms

The production of Na24 in a body by neutron irradiation is an important part of a system of monitoring nuclear accidents. The feasibility of using phantoms filled with a homogeneous aqueous solution of NaCl in the determination of the relationship between neutron dose and Na24 production was investigated using phantoms, burros and the LASL Godiva II and ORNL Tower Shielding Facility II reactors. The Na24 production per unit neutron dose was measured in man-shaped phantoms, cylinders and burros. It was found that: (1) the ratio of the concentrations of Na24 and Na23, (Na24/Na23), in circulated blood is equal to the ratio in the whole body; (2) the inhomogeneity of the distribution of Na in living animals is not significant in the production of Na24; (3) a phantom filled with NaCl solution can be used to approximate a living animal in experiments involving the production of Na24 by neutron irradiation; (4) the size of a body and its orientation with respect to the source of neutrons has an effect on the magnitude of the Na24 production; and (5) if analyses of blood samples are used to determine Na24 production in the burro, the exact time of sampling, within the range of 2–24 hrs post-irradiation, is not important.