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In this paper, we discuss an accurate and fast software tool (INSPCT-S, Inspection of Nuclear Spent fuel-Pool Calculation Tool, version Spreadsheet) developed for calculation of the response of fission chambers placed in a spent fuel pool, such as Atucha-I. INSPCT-S is developed for identification of suspicious regions of the pool that may have missing or substitute assemblies. INSPCT-S uses a hybrid algorithm based on the adjoint function methodology. The neutron source is comprised of spontaneous fission, (α, n) interactions, and subcritical multiplication. The former is evaluated using the ORIGEN-ARP code, and the latter is obtained with the fission matrix (FM) formulation. The FM coefficients are determined using the MCNP Monte Carlo code, and the importance function is determined using the PENTRAN 3-D parallel Sn code. Three databases for the neutron source, FM elements, and adjoint flux are prepared as functions of different parameters including burnup, cooling time, enrichment, and pool lattice size. INSPCT-S uses the aforementioned databases and systems of equations to calculate detector responses, which are subsequently compared with normalized experimental data.

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IntroductionThe contents approved for shipment in a Type B radioactive material transportation package have historically been descriptions of discrete items, or groupings of well-defined similar items, in the package safety basis documentation. The need for a comprehensive functional content envelope of both gamma and neutron emitting nuclides compliant with regulatory limits has become necessary as the DOE complex requires shipments of unique mixtures of radionuclides and impurities.Recent publications [1,2] have presented a calculational model and a corresponding content envelope intended to be compliant with federal regulatory external radiation limits as well as design decay heat limits based on the Model 9977 Packaging. The methodology used to develop this content envelope consisted of determining the external radiation dose rates based on one gram of a given isotope combined, in the case of actinides, with various levels of light element impurities and determining the allowable mass to the regulatory limits based on these dose rates [1,2]. The method of ratioing from these calculations to the masses that meet the regulatory limits fails in the case of some combinations of actinides and light element impurities because of the effect of subcritical multiplication [3], resulting in non-conservative and non-compliant dose rates in some cases. This is particularly true for many actinides combined with beryllium, boron, fluorine, lithium, and sodium.In this study, the source was modeled as a sphere with the appropriate dimensions based on the actinide or gamma source density and placed at the bottom of the containment vessel for maximum conservatism. The results presented in this report rectify the above-noted deficiencies by adjusting the previously determined mass limits so that the resulting dose rates are compliant with regulatory limits. Some amount of iteration was required in some instances to adjust the masses for dose compliance. In some instances the masses were increased since they were too conservative. In addition, the method of ratioing the masses when applied to several gamma emitting isotopes also resulted in either non-conservative or overly conservative data. The mass limits presented here are for bare sources and are complaint with the design decay heat limit of 19W. The masses are based on package surface dose rates (the limiting dose rate) that fall in the range 185-195 mrem/h, thus giving an additional margin of between 7.5% and 2.5% to the regulatory limit of 200 mrem/h. This report presents a revised set of mass limits (i.e., content envelope) for several neutron emitting actinides with varying levels of light element impurities compliant with both external radiation and design decay heat limits. In addition, revised limits for gamma sources are also presented. The neutron emitter mass limits are in Tables 1 through 9 and the gamma emitter mass limits are in Table 10.The revised content envelope covering a wide range of materials present in the DOE complex, was developed for the Model 9 ...

Abstract. Analyses were performed to characterize the radiation field in the vicinity of the Final Optics Assemblies (FOAs) at the National Ignition Facility (NIF) due to neutron activation following Deuterium-Deuterium (DD), Tritium-Hydrogen-Deuterium (THD), and Deuterium-Tritium (DT) shots associated with different phases of the NIF operations. The activation of the structural components of the FOAs produces one of the larger sources of gamma radiation and is a key factor in determining the stay out time between shots to ensure worker protection. This study provides estimates of effective dose rates in the vicinity of a single FOA and concludes that the DD and THD targets produce acceptable dose rates within10 minutes following a shot while about 6-days of stay out time is suggested following DT shots. Studies are ongoing to determine the combined effects of multiple FOAs and other components present in the Target Bay on stay-out time and worker dose.