Dry spent fuel cask monitoring by {sup 252}Cf-source-driven frequency analysis measurements

If developed, a nondestructive method would be useful for verifying canister contents without requiring the canister to be opened. This paper addresses the application of the {sup 252}Cf-source-driven frequency analysis measurements for verification of the fissile material content of sealed spent fuel canisters. The cross-power spectral density (CPSD) between the {sup 252}Cf source in an ionization chamber and external neutron detectors depends only on the induced fission rate in the fissile system and is independent of inherent sources. Thus the source-to-detector CPSD is ideal for determination of fissile material content of the spent fuel. This paper evaluates the application of this method to a 125 ton spent fuel canister that contained 21 pressurized-water reactor fuel elements. The results demonstrate that the fissile materials content of a sealed spent fuel canister could be obtained using the {sup 252}Cf frequency analysis method if calibration standards were available. The results also indicate that a measurement could be performed in less than a day for burnups up to 36 GWd/MTU and in less time for lower burnups

Characterization of an enriched uranyl fluoride deposit in a valve and pipe intersection using time-of-flight transmission measurements with {sup 252}Cf

A method was developed and successfully applied to characterize large uranyl fluoride (UO{sub 2}F{sub 2}) deposits at the former Oak Ridge Gaseous Diffusion Plant. These deposits were formed by a wet air in-leakage into the UF{sub 6} process gas lines over a period of years. The resulting UO{sub 2}F{sub 2} is hygroscopic, readily absorbing moisture from the air to form hydrates as UO{sub 2}F{sub 2}-nH{sub 2}O. The ratio of hydrogen to uranium can vary from 0--16, and has significant nuclear criticality safety impacts for large deposits. In order to properly formulate the required course of action, a non-intrusive characterization of the distribution of the fissile material within the pipe, its total mass, and amount of hydration was necessary. The Nuclear Weapons Identification System (NWIS) previously developed at the Oak Ridge Y-12 Plant for identification of uranium weapons components in storage containers was used to successfully characterize these deposits

Within-Household Selection Methods: A Critical Review and Experimental Examination

Probability samples are necessary for making statistical inferences to the general population (Baker et al. 2013). Some countries (e.g. Sweden) have population registers from which to randomly select samples of adults. The U.S. and many other countries, however, do not have population registers. Instead, researchers (i) select a probability sample of households from lists of areas, addresses, or telephone numbers and (ii) select an adult within these sampled households. The process by which individuals are selected from sampled households to obtain a probability-based sample of individuals is called within-household (or within-unit) selection (Gaziano 2005).Within-household selection aims to provide each member of a sampled household with a known, nonzero chance of being selected for the survey (Gaziano 2005; Lavrakas 2008). Thus, it helps to ensure that the sample represents the target population rather than only those most willing and available to participate and, as such, reduces total survey error (TSE). In interviewer-administered surveys, trained interviewers can implement a prespecified within-household selection procedure, making the selection process relatively straightforward. In self-administered surveys, within-household selection is more challenging because households must carry out the selection task themselves. This can lead to errors in the selection process or nonresponse, resulting in too many or too few of certain types of people in the data (e.g. typically too many female, highly educated, older, and white respondents), and may also lead to biased estimates for other items. We expect the smallest biases in estimates for items that do not differ across household members (e.g. political views, household income) and the largest biases for items that do differ across household members (e.g. household division of labor). In this chapter, we review recent literature on within-household selection across survey modes, identify the methodological requirements of studying within-household selection methods experimentally, provide an example of an experiment designed to improve the quality of selecting an adult within a household in mail surveys, and summarize current implications for survey practice regarding within-household selection. We focus on selection of one adult out of all possible adults in a household; screening households for members who have particular characteristics has additional complications (e.g. Tourangeau et al. 2012; Brick et al. 2016; Brick et al. 2011), although designing experimental studies for screening follows the same principles

Preliminary Evaluation of NMIS for Interrogation of Pu and HEU in AT400-RContainers at Mayak

Preliminary Monte Carlo simulations have demonstrated the sensitivity of the NMIS active interrogation method to the amount of uranium fissile material stored in AT400-R containers that are being proposed for use in the Mayak facility. Properties of the time-of-flight signature can be used to determine the absence of one of the uranium metal spheres or to determine if a different enrichment sphere is present in the container. The tail of the time-of-flight signature from induced fission is directly dependent on the amount of uranium 235 present in the containers, and a particular ratio of the correlated counts due to fission to the correlated counts due to transmission is nearly linear with 235U mass. These simulations demonstrate that the NMIS active method can be used to assay the amount of {sup 235}U with a sensitivity of coefficient of {approx}0.1 per kg {sup 235}U (approximately 10% change in the ratio per kilogram of {sup 235}U). These calculations have shown that these active measurements with a {sup 252}Cf source of 1 x 10{sup 6} fission per sec would require the order of a few minutes of data accumulation time for a container with two 8 kg spheres, and since NMIS operates in real time, 1.6 minutes of measurement time is required. This measurement time is short. The calculations for Pu have shown that NMIS in the passive mode (no Cf source) can determine the mass of Pu in AT400-R containers with short measurement times of a few minutes. The sensitivity of the proposed detectors to gamma rays should enhance this measurement method since the gamma rays from fission, induced or spontaneous, escape the container more easily than neutrons. In addition to the time correlation measurements, the multiplicity options of NMIS allow conventional multiplicity measurements that, depending on the type of detector, can include prompt gamma rays from fission. If gamma ray spectrometry is included in the NMIS processor, then one NMIS system can perform the desired NMC&A measurements for use at Mayak (i.e., HEU, Pu, and Pu isotopics). Higher order NMIS correlation measurements have promise for determining the Pu shape passively also if desired, but this requires four detectors. Active higher order measurements with Pu can separate out the effects of induced fission and spontaneous fission and thus may yield information on the ratio of {sup 239}Pu to {sup 240}Pu. Incorporation of gamma-ray spectrometry is discussed in Appendix A. These capabilities should be verified by measurements both with uranium and plutonium

Review of Subcritical Source-Driven Noise Analysis Measurements

Subcritical source-driven noise measurements are simultaneous Rossia and randomly pulsed neutron measurements that provide measured quantities that can be related to the subcritical neutron multiplication factor. In fact, subcritical source-driven noise measurements should be performed in lieu of Rossia measurements because of the additional information that is obtained from noise measurements such as the spectral ratio and the coherence functions. The basic understanding of source-driven noise analysis measurements can be developed from a point reactor kinetics model to demonstrate how the measured quantities relate to the subcritical neutron multiplication factor

Integration of Several Elements of the DOE Nuclear Criticality Safety Program

The U. S. Department of Energy established the Nuclear Criticality Safety Program (NCSP) to maintain the infrastructure and expertise in nuclear criticality safety to support line criticality safety programs at various DOE sites. The seven tasks of the NCSP include critical experiments, benchmarking, nuclear data, analytical methods, applicable ranges of bounding curves and data, information preservation and dissemination, and training and qualification. The goals of this program are to improve the knowledge, tools, data, guidance, and information available to the nuclear criticality safety community. In addition various elements of the NCSP are integrated together to provide the nuclear criticality safety community with the most precise nuclear data for criticality safety analyses. This paper describes how several elements of the NCSP were integrated together in the evaluation of the silicon nuclear data. Silicon is frequently encountered in decontamination and decommissioning efforts , process sludge and settling tanks, in situ vitrification, and waste remediation efforts (including waste storage, retrieval, characterization, volume reduction, and stabilization). Silicon was also identified as an important isotope for addressing concerns associated with the storage of spent nuclear fuels in a geologic repository. The inadequacy of the silicon nuclear data in the intermediate energy region mandated that additional neutron capture cross-section measurements had to be performed that encompassed the resolved resonance region. An evaluation was performed that included analysis of the most recent neutron capture and existing transmission cross-section measurements performed at the Oak Ridge Electron Linear Accelerator. Critical experiments were performed at the Institute of Physics and Power Engineering in Obninsk, Russia because of the lack of critical experiment data for analysis of storage of nuclear material in a geologic repository. These critical experiments were evaluated and benchmark models were developed and submitted to the International Criticality Safety Benchmark Evaluation Project for review and publication in the ''International Handbook of Evaluated Criticality Safety Benchmark Experiments''. Sensitivity analyses were performed as a part of the benchmark evaluation to determine the sensitivity of the critical experiments to the various constituents of the assembly. The benchmark models were then used to determine the computed k{sub eff} for various cross section data sets. The variation in the computed k{sub eff} value for the new evaluated data set was then used as an indicator to adjust the negative energy capture widths for the capture cross section data. Furthermore, the changes in k{sub eff} were used as an indicator to the inadequacy of previous measured data in the unresolved resonance region. The result of the efforts of the NCSP provided the most precise set of nuclear data for silicon. The resulting ORNL evaluation produced the most consistent evaluation for silicon. This result could only be achieved through integration of many components of the NCSP

Evaluation of prompt fission gamma rays for use in simulating nuclear safeguard measurements

Nondestructive assay methods that rely on measurement of correlated gamma rays from fission have been proposed as a means to determine the mass of fissile materials. Sensitivity studies for such measurements will require knowledge of the multiplicity of prompt gamma rays from fission; however, a very limited number of multiplicity distributions have been measured. A method is proposed to estimate the average number of gamma rays from any fission process by using the correlation of neutron and gamma emission in fission. Using this method, models for the total prompt gamma ray energy from fission adequately reproduce the measured value for thermal neutron induced fission of {sup 233}U. Likewise, the average energy of prompt gamma rays from fission has been adequately estimated using a simple linear model. Additionally, a method to estimate the multiplicity distribution of prompt gamma rays from fission is proposed based on a measured distribution for {sup 252}Cf. These methods are only approximate at best and should only be used for sensitivity studies. Measurements of the multiplicity distribution of prompt gamma rays from fission should be performed to determine the adequacy of the models proposed in this article

Sensitivity of Subcritical Measurement Simulations to Neutron Cross-Section Data

Evaluation of nuclear data typically includes validation of the data through computation of k{sub eff}for critical assemblies. The sensitivity of the computed k{sub eff} values to the nuclear data is used as an indicator in determining the adequacy of an evaluation. Subcritical measurements offer an alternative to critical experiments as a means to evaluate nuclear data through direct computation of subcritical measurement quantities. In some instances, the subcritical measurement quantities are more sensitive to nuclear data changes than the computed k{sub eff} values. Simulations of subcritical source-driven noise measurements were performed for highly enriched uranium metal cylinders and highly enriched uranyl nitrate solutions to demonstrate the sensitivity of the computed subcritical quantities to nuclear data. A particular ratio of spectral quantities from the source-driven noise measurements is of interest because of its independence of detection efficiency and source intensity. These simulations indicate that the spectral ratio is more sensitive to nuclear data evaluations than the computed k{sub eff} values for these systems. Direct simulation of subcritical measurements offers additional means to validate nuclear data evaluations

MCNP-DSP USERS MANUAL

The Monte Carlo code MCNP-DSP was developed from the Los Alamos MCNP4a code to calculate the time and frequency response statistics obtained from subcritical measurements. The code can be used to simulate a variety of subcritical measurements including source-driven noise analysis, Rossi-{alpha}, pulsed source, passive frequency analysis, multiplicity, and Feynman variance measurements. This code can be used to validate Monte Carlo methods and cross section data sets with subcritical measurements and replaces the use of point kinetics models for interpreting subcritical measurements