A Two-Dimensional Point-Kernel Model for Dose Calculations in a Glovebox Array D. E. Kornreich and D. E. Dooley ~- Los Alamos National Laboratory I. Background An associated paper details a model of a room containing gloveboxes using the industry standard dose equivalent (dose) estimation tool MCNP.2 Such tools provide an excellent means for obtaining relatively reliable estimates of radiation transport in a complicated geometric structure. However, creating the input deck that models the complicated geometry is equally complicated. Therefore, an alternative tool is desirable that provides reasonably accurate dose estimates in complicated geometries for use in engineering-scale dose analyses. In the past, several tools that use the point-kernel model for estimating doses equivalent have been constructed (those referenced are only a small sample of similar This new tool, the Photon And Neutron Dose Equivalent Model Of Nuclear materials Integrated with an Uncomplicated geometry Model (PANDEMONIUM), combines point-kernel and diffusion theory calculation routines with a simple geometry construction tool. PANDEMONIUM uses VisioTM5 to draw a glovebox array in the room, including hydrogenous shields, sources and detectors. This simplification in geometric rendering limits the tool to two-dimensional geometries (and one-dimensional particle transport calculations).
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II. TheModels The geometry module currently contains a library of four types of geometric structures: 1) gloveboxes; 2) hydrogenous shields (e.g., polyethylene shields or persons); 3) dose measurement points (either single or grid-structured); and 4)sources. Attributes of the gloveboxes include dimensions, position, orientation, and shielding makeup; attributes of the hydrogenous shield include dimensions, position, and orientation; and, attributes of the source include isotopic composition and number densities, chemical form, bulk volume, source shielding, and position. The primary purpose of the geometry model is to calculate the source-to-detector distances and thicknesses for all shielding materials. The geometric model is exported from VisioTMvia a Visual BasicTMmacro, which writes an input file that can be read by a FORTRAN code. This information is sent to the dose calculation module, which uses the standard point-kernel model to obtain photon doses equivalent and diffusion theory solutions in spherical geometry to obtain neutron doses equivalent. The neutron current leaving the surface of a spherical source is radially attenuated to give the flux at a given distance from the source. Neutron dose rates are given by 0,=hi R2 J(R) ( R + a)2 9 where hi = fluence-to-dose factor [mrem-cm2-s/hr-n],6 and R = the radius of the source [cm], a = the surface-to-detector distance [cm], and 2
J(R) = the current at the surface of the source [n/cm2-s]. A model to account for neutron thermalization by hydrogenous material is also included in the calculation of h i. The transport code ONEDANT~was used to obtain a numerical fit for the attenuation of fast neutrons through increasing thicknesses of water shields. Using this fit takes into account the decrease in the dose equivalent from neutron thermalization and absorption in water. Neutron sources from spontaneous fission and (a,n) reactions are also calculated. Spontaneous fission sources are calculated from decay data and (a,n) reaction sources are calculated according to data included in the SOURCES-3A code.8 Photon dose calculations are slightly more complicated as a result of a required multi-group treatment (only in-group interactions are considered). The scalar flux is calculated and accounts for radial and shielding attenuation. Attenuation coefficients for all shielding materials are obtained from ANSI? The photon dose rates are given by where hl (Ei ) = fluence-to-dose factor at energy Ei (mrem-cm2-s/hr-y), @, ( E j )= scalar uncollided neutron flux at energy Ei (y/cm2-s), ' ( E, ) = buildup factor for photons of energy Ei. 3
III. Problems and Results A sample problem considers a very important component of plutonium processing. Figure 1 contains a floorplan of the gloveboxes in a fictitious pyrochemical processing room. Three important processes occur in this room: electrorefining (ER - purifying plutonium of impurities), molten salt extraction (MSE - removing americium-241 from plutonium), and multicycle direct oxide reduction (MC-DOR - converting plutonium oxide to metal). In addition to the processing gloveboxes, a control area behind a polyethylene shield is also included in the floorplan. In each of these processes, plutonium metal is present at some point. 4 kg of high-grade plutonium (93% Pu-239) metal is assumed to be the source for each process. The ER and MC-DOR sources contain 200 ppm of Am-241, and the source in the MSE glovebox contains 1000 ppm of Am241. All gloveboxes provide 0.25 in of shielding except the MSE glovebox, which provides an addition 0.125 in of lead shielding. The EDE matrices for plutonium and americium metal are provided in Table 1. Clearly, the largest contributor to a worker s dose is the immediate process on which he is working; however, this matrix allows for an estimation of the dose he might receive from other processes in the room. A complete analysis of the doses in this room would require two more matrices like that in Table 1 - one for plutonium in the salt form (all three processes encounter plutonium chloride) and one for plutonium in the oxide form (primarily used in MC-DOR). 4
References 1. DOOLEY, D. E. and D. E. Kornreich, A MCNP Model of Gloveboxes in a Room, Trans. Am. NUC.SOC.19,255 (1998). 2. MCNP - A General Monte Carlo Code for Neutron and Photon Transport, Version 4B, LA-7396-M, Rev. 2, J. F. BRIESMEISTER, Ed., Los Alamos National Laboratory (1995). 3. MALENFANT, R. E., QAD: A Series of Point-Kernel General-Purpose Shielding Programs, LA-3573, Los Alamos Scientific Laboratory (1967). 4. KORNREICH, D.E., Efficient Dose Calculations for Glovebox Operations, 2 1St Annual Actinide Separations Conference, Charleston, S.C., 55 (1997). 5. Visio TechnicalTM, Developing Visio Solutons, version 5.0, for PCs using WindowsTM NT 4.0, Visio Corporation (1997). 6. American National Standard for Neutron and Gamma-Ray Fluence-to-Dose Factors, ANSVANS-6.1.1-1991, American Nuclear Society (1991). 7. R. E. Alcouffe, et al., DANTSYS: A Diffusion Accelerated Neutral Particle Transport Code System, LA-12969-M, Los Alamos National Laboratory (1995). 8. WILSON, W. B., et al., SOURCES-3A: A Code for Calculating (a,n) Spontaneous Fission, and Delayed Neutron Sources and Spectra, LA-UR-97-4365, Los Alamos National Laboratory (1997). 9. American National Standard for Gamma-Ray Attenuation Coefficients and Buildup Factors for Engineering Materials, ANSI/ANS-6.4.3-1991, American Nuclear Society (1991). 5
Table 1. EDE Matrix for Process Dose Calculations for Metal. Dose Receivers Dose Providers (mrem/hr-kgpu) MSE 1 ER I MCDOR 1 Controlpanel MSE 2.14E-1 2.74E-4 5.90E-4 ER 2.43E-4 2.3OE-1 5.68E-4 MCDOR 4.72E-4 4.49E-4 2.30E-1 Control Panel 8.11E-5 6.58E-5 8.72E-4 Dose Providers (me&-kgam) Dose Receivers MSE ER MCDOR Control Panel MSE 5.15E-1 1.13E-4 7.5OE-4 0 ER 8.35E-5 3.89E+0 1.19E-3 0 MCDOR 5.22E-4 8.75E-5 3.90E+O 0 Control Panel 2.18E-4 3.93E-4 4.82E-2 0 6
-2 MSE Glovebox 0 lf8"-1/8"-1/8" SS-Pb-SS All other containment 1/4"SS Gloveboxes are 4'by 8 MC-DOR Glovebox 4" Polyethylene Shield 0 Control Panel 0 Source 0 Detector Figure 1. Sample pyrochemical processing room layout. 7