@article{davis_dudziak_yim_mcnelis_wooten_2011, title={PHOTON BUILDUP FACTORS IN LAMINATED DUAL-LAYER SHIELDS}, volume={173}, ISSN={["1943-7471"]}, DOI={10.13182/nt11-110}, abstractNote={Abstract In radiation protection, photon buildup factors provide a convenient method for calculating dose and exposure response after various shielding configurations, as well as information about the behavior of radiation in these configurations. Though many situations call for multilayer shields, few databases and derived analytical formulas for photon buildup in multilayer shields exist. This research develops buildup factors and analytical fits to these data for slab-geometric, dual-layer shields composed of various materials. The photon buildup factors were calculated for monoenergetic photon sources incident on two-layer shields of various combinations of lead, polyethylene, aluminum, and stainless steel for thicknesses varying between 2 and 20 mean free paths using the Parallel Time Independent Sn (PARTISN) discrete ordinates code. The Gauss-Lobatto S100 quadrature was used with a 244-energy-group structure and coupled photon and electron cross sections. Data from PARTISN calculations were then benchmarked for representative cases using MCNP5, and fits to a new analytical formula were developed using Mathematica 6.0.}, number={3}, journal={NUCLEAR TECHNOLOGY}, author={Davis, Adam and Dudziak, Donald J. and Yim, Man-Sung and McNelis, David and Wooten, H. Omar}, year={2011}, month={Mar}, pages={270–288} }
 @misc{li_yim_mcnelis_2010, title={Model-based calculations of the probability of a country's nuclear proliferation decisions}, volume={52}, ISSN={["0149-1970"]}, DOI={10.1016/j.pnucene.2010.07.001}, abstractNote={Abstract This paper presents an attempt to project a country’s nuclear proliferation-related behaviors by using quantitative models with the use of open source information. The approach is based on the combined use of data on a country’s economic status, security environment, political development, nuclear technological capability, and commitment to nuclear nonproliferation. Projections of country’s proliferation-related behaviors were made by using the multinomial logit regression and the Weibull and Cox event history modeling for 189 countries. Results from the developed models were compared with the historical records from 1945 through 2000 with respect to “explore”, “pursue”, and “acquire” decisions. Overall, this study indicated that quantitative models could be useful in providing warnings against potential nuclear proliferation attempts. Key variables of importance in quantitative modeling of proliferation-related behaviors were identified and discussed.}, number={8}, journal={PROGRESS IN NUCLEAR ENERGY}, author={Li, Jun and Yim, Man-Sung and McNelis, David N.}, year={2010}, month={Nov}, pages={789–808} }
 @article{liu_yim_mcnelis_2009, title={ACCELERATOR-BASED TARGET DESIGN AND OPTIMIZATION DRIVEN BY HIGH-ENERGY ELECTRON BEAMS AND PROTON BEAMS FOR THE ANALYSIS OF NEUTRON GENERATION CHARACTERISTICS}, volume={165}, ISSN={["0029-5450"]}, DOI={10.13182/NT09-A4064}, abstractNote={Abstract Accelerator-based target design and optimization are presented in this paper as an approach for the analysis of neutron generation and characteristics. Electron-based targets and proton-based targets driven by high-energy accelerator beams are investigated. The target plays an important role in the external neutron sources in which the target was driven by high-energy accelerator beams to generate neutrons. The optimization of target design in this work is to obtain the maximum generation of neutrons out of targets considering target material and geometry, accelerator beam energy, and beam size. A three-dimensional particle detection methodology and a surface matrix arithmetic technique were used to determine the spatial distribution of the source particles (electron and proton) and the total neutron generation from the target outer surfaces. Neutron generation and characteristics were analyzed based on the optimized targets regarding neutron spectrum, average energy, and average flux. Monte Carlo calculations were performed by using MCNPX to estimate the particle interaction inside the target and to calculate the neutrons escaping out of the target surfaces. Results in this work indicated that a high-energy (1-GeV) electron accelerator beam is capable of producing high-intensity neutron flux at the range of 1.60 × 1013 n/cm2·s of 1-mA electron. Compared to an electron accelerator beam, a proton beam (1 GeV) generates higher-intensity neutron flux at the level of 4.83 × 1013 n/cm2·s of 1-mA proton. The neutron generation ratio (neutron per incident particle escaping from the target) was computed as 0.76 neutrons per electron and 38.8 neutrons per proton for the selected targets. In the electron accelerator-based target, neutron generation was mostly through photonuclear reactions (88%), followed by prompt fission (12%). Neutron production in the target of the proton accelerator-based target was mainly due to spallation reactions (40%) and prompt fissions (48%). The optimized size of the target for the electron accelerator-based target, in terms of the volume, was about 16 times smaller than that for the proton accelerator-based target. The estimated neutron energy distribution was much narrower, with the electron accelerator target ranging from 1.0 × 10−3 to 30 MeV. In the proton accelerator target, the neutron energies ranged between 1.0 × 10−5 MeV and 1 GeV.}, number={1}, journal={NUCLEAR TECHNOLOGY}, author={Liu, Yaxi and Yim, Man-Sung and McNelis, David}, year={2009}, month={Jan}, pages={111–123} }
 @article{schirmers_davis_wooten_dudziak_yim_mcnelis_2009, title={CALCULATION OF PHOTON EXPOSURE AND AMBIENT DOSE SLANT-PATH BUILDUP FACTORS FOR RADIOLOGICAL ASSESSMENT}, volume={167}, ISSN={["0029-5450"]}, DOI={10.13182/NT09-1}, abstractNote={Abstract Slant-path photon buildup factors for nine radiation shielding materials (air, aluminum, concrete, iron, lead, leaded glass, polyethylene, stainless steel, and water) are calculated with the most recent cross-section data available using Monte Carlo and discrete ordinates methods. Discrete ordinates calculations use a 244-group energy structure based on previous research at Los Alamos National Laboratory (LANL) and focus on the effects of group widths in multigroup calculations for low-energy photons. Buildup-factor calculations in discrete ordinates benefit from coupled photon/electron cross sections to account for secondary photon effects. Also, ambient dose equivalent buildup factors were analyzed at lower energies where corresponding response functions do not exist in the literature. The results of these studies are directly applicable to radiation safety at LANL, where the dose-modeling code PANDEMONIUM is used to estimate worker dose in plutonium-handling facilities. Buildup factors determined in this work will be used to enhance the code’s modeling capabilities but also should be of general interest to the radiation shielding community.}, number={3}, journal={NUCLEAR TECHNOLOGY}, author={Schirmers, Fritz G. and Davis, Adam and Wooten, H. Omar and Dudziak, Donald J. and Yim, Man-Sung and McNelis, David}, year={2009}, month={Sep}, pages={395–409} }
 @article{li_yim_piet_mcnelis_2009, title={Integrated decay heat load method to analyze repository capacity impact of a fuel cycle}, volume={36}, ISSN={["0306-4549"]}, DOI={10.1016/j.anucene.2009.06.015}, abstractNote={Abstract Assessing the needs for repository capacity from nuclear waste disposal is essential for fuel cycle development or repository development planning. As the repository capacity is mainly constrained by thermal design limits on the repository rocks, a detailed mountain-scale heat transfer calculation is needed for repository capacity impact analysis. In this paper, a simplified repository capacity impact analysis method is proposed as an alternative to performing repository scale heat transfer analysis. The method is based on the use of integrated decay heat load (IDHL) limits. The derived integrated decay heat loads were found to appropriately represent the drift wall temperature limit (200 °C) and the midway between adjacent drifts temperature limit (96 °C) under the high temperature operating mode as long as the wastes are uniformly loaded into the repository. Results indicated that the long-term integrated decay heat load (IDHLL) and the short-term integrated decay heat load (IDHLS) can be effectively used to represent the repository capacity impact for SNFs and HLWs, respectively. Comparisons indicated good agreement between the proposed IDHL method and the repository heat transfer analysis-based approach.}, number={9}, journal={ANNALS OF NUCLEAR ENERGY}, author={Li, Jun and Yim, Man-Sung and Piet, Steven and McNelis, David}, year={2009}, month={Sep}, pages={1366–1373} }
 @article{li_yim_mcnelis_2008, title={Assessing the proliferation resistance of nuclear fuel cycle systems using a fuzzy logic-based barrier method}, volume={162}, DOI={10.13182/nt08-a3957}, abstractNote={Abstract The development of a fuzzy logic-based barrier (FLB) method for the evaluation of the proliferation resistance of nuclear fuel cycle systems is described in this paper. The method is based on using a group of system-dependent, measurable, or quantifiable variables to define the proliferation barrier effectiveness of a system as fuzzy numbers. The usefulness of the FLB method and the resulting metric in quantifying the proliferation resistance of fuel cycle systems was also investigated by applying it to three fuel cycles, i.e., light water reactor-once-through, light water reactor with mixed oxide fuel, and direct use of spent pressurized water reactor fuel in CANDU reactor. To address the issue of subjectivity in assigning barrier weighting factors or fuzzy numbers, the sensitivity of the results to the definition of fuzzy numbers and weighting schemes was also investigated.}, number={3}, journal={Nuclear Technology}, author={Li, J. and Yim, M. S. and McNelis, D.}, year={2008}, pages={293–307} }
 @article{stahala_yim_mcnelis_2008, title={Investigation of Yucca Mountain repository capacity for the US spent nuclear fuel inventory}, volume={35}, ISSN={["0306-4549"]}, DOI={10.1016/j.anucene.2007.11.002}, abstractNote={An analytical decay heat model was developed to evaluate the US inventory of commercial spent nuclear fuel (SNF). The model was benchmarked against the results from ORIGEN-ARP 5.01. The new analytical SNF decay heat model was applied to actual (thru 2002) and projected SNF data. The total decay heat from the 63,000 MT commercial SNF at year 2012 was estimated at 182 MW. According to the thermal loading analysis using a mountain-scale heat transfer model, a 4.9 km2 (1165 acre) site designated for SNF disposal was found to have the capacity to store more SNF than the statutory limit of 70,000 MTIHM. The maximum capacity available for SNF disposal at the Yucca Mountain site is dependent upon the thermal loading strategy chosen and SNF cooling time before emplacement. It was also shown that using high burnup SNF and adjusting the drift spacing, the capacity of the repository could be maximized on a per energy production basis, although some additional cooling may be necessary. Future work needs to consider extending the ‘footprint’ of the repository, applying non-uniform SNF loading into the drifts, and the impact of spent fuel reprocessing and other decay heat management strategies.}, number={6}, journal={ANNALS OF NUCLEAR ENERGY}, author={Stahala, Mike P. and Yim, Man-Sung and McNelis, David N.}, year={2008}, month={Jun}, pages={1056–1067} }