Impacts of Model Fidelity on Simulated Gamma Spectra in Estimating Nuclear Safeguards Systems Performance (original) (raw)
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Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 2011
The International Atomic Energy Agency will require the development of advanced technologies to effectively safeguard nuclear material at increasingly large-scale nuclear recycling facilities. Ideally, the envisioned technologies would be capable of nondestructive, near-real-time, autonomous process monitoring. This paper describes recent results from model simulations designed to test the Multi-Isotope Process (MIP) monitor, a novel addition to a safeguards system for reprocessing facilities. The MIP monitor combines the detection of intrinsic gamma ray signatures emitted from process solutions with multivariate analysis to detect off-normal conditions in process streams nondestructively and in nearreal-time. Three computer models including ORIGEN-ARP, AMUSE, and SYNTH were used in series to predict spent nuclear fuel composition, estimate element partitioning during separation, and simulate spectra from product and raffinate streams using a variety of gamma detectors, respectively. Simulations were generated for fuel with various irradiation histories and under a variety of plant operating conditions. Principal component analysis was applied to the simulated gamma spectra to investigate pattern variations as a function of acid concentration, burnup, and cooling time. Hierarchical cluster analysis and partial least squares (PLS) were also used in the analysis. The MIP monitor was found to be sensitive to induced variations of several operating parameters including distinguishing 7 2.5% variation from normal process acid concentrations. The ability of PLS to predict burnup levels from simulated spectra was also demonstrated to be within 3.5% of measured values.
The Multi-Isotope Process (MIP) Monitor Project: FY13 Final Report
2013
The Multi-Isotope Process (MIP) Monitor provides an efficient approach to monitoring the process conditions in reprocessing facilities in support of the goal of "… (minimization of) the risks of nuclear proliferation and terrorism." The MIP Monitor measures the distribution of the radioactive isotopes in product and waste streams of a nuclear reprocessing facility. These isotopes are monitored online by gamma spectrometry and compared, in near-real-time, to spectral patterns representing "normal" process conditions using multivariate analysis and pattern recognition algorithms. The combination of multivariate analysis and gamma spectroscopy allows us to detect small changes in the gamma spectrum, which may indicate changes in process conditions. By targeting multiple gamma-emitting indicator isotopes, the MIP Monitor approach is compatible with the use of small, portable, relatively high-resolution gamma detectors that may be easily deployed throughout an existing facility. The automated multivariate analysis can provide a level of data obscurity, giving a built-in information barrier to protect sensitive or proprietary operational data. Proof-of-concept simulations and experiments have been performed in previous years to demonstrate the validity of this tool in a laboratory setting for systems representing aqueous reprocessing facilities. However, pyroprocessing is emerging as an alternative to aqueous reprocessing techniques. This report describes research to evaluate the applicability of the MIP Monitor approach to pyroprocessing systems during fiscal year 2013 (FY13). Specific aspects of research completed in FY13 include:
The Multi-Isotope Process Monitor Project: FY11 Progress and Accomplishments
2012
The Multi-Isotope Process (MIP) Monitor represents a potentially new and efficient approach to monitoring process conditions in reprocessing facilities with the high-level goal of aiding in the "...(minimization of) the risks of nuclear proliferation and terrorism" (Office of Technology Assessment 1995). This approach relies on multivariate analysis and gamma spectroscopy of spent fuel product and waste streams to automatically and simultaneously monitor a variety of process conditions (e.g., acid concentrations, burnup, cooling time, etc.) in near real-time (NRT). While the conceptual basis for the MIP Monitor has been shown to be effective in an aqueous reprocessing system, the fundamental approach should also be viable in a pyro-processing recycle system. The MIP Monitor may be calibrated to provide online quantitative information about process variables for process control or domestic safeguards applications; or it can simply monitor, with a built-in information barrier, for off-normal conditions in process streams, making the approach well-suited for applications were it is necessary to respect proprietary information or for international safeguards applications. Proof-of-concept simulations and experiments were performed in previous years demonstrating the validity of this tool in a laboratory setting. This report details follow-on research and development efforts sponsored by the U.
Nuclear safeguards at inspected facilities aims to deter or detect special nuclear material (SNM) diversion and to do so is increasingly relying on process monitoring (PM) to augment nuclear material accounting (NMA). In NMA, SNM material balances are computed approximately every 30 days, and modeling and simulation are used to predict detector performance, to model SNM flows and inventory, and predict overall NMA performance as measured by the measurement error standard deviation of the material balance, MB. In PM, much more frequent and often shortcut measurements (less than full SNM accountability) are used, and modeling and simulation are increasingly used to predict the effects of SNM diversion on normal operating data under various scenarios. This paper reviews traditional modeling and simulation roles in NMA, describes new roles in PM, and illustrates using a case study.
Multivariate analysis of gamma spectra to characterize used nuclear fuel
Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment
The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gamma spectra were used to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. This approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to
Journal of Environmental Radioactivity, 2018
When post-irradiation materials from the nuclear fuel cycle are released to the environment, certain isotopes of actinides and fission products carry signatures of irradiation history that can potentially aid a nuclear forensic investigation into the material's provenance. In this study, combinations of Pu, Cs, and Ba isotope ratios that produce position (in the reactor core) independent monitors of irradiation history in spent light water reactor fuel are identified and explored. These position independent monitors (PIMs) are modeled for various irradiation scenarios using automated depletion codes as well as ordinary differential equation solutions to approximate nuclear physics models. Experimental validation was performed using irradiated low enriched uranium oxide fuel from a light water reactor, which was sampled at 8 axial positions from a single rod. Plutonium, barium and cesium were chemically separated and isotope ratio measurements of the separated solutions were made by quadrupole and multi-collector inductively coupled mass spectrometry (Cs and Pu, respectively) and thermal ionization mass spectrometry (Ba). The effect of axial variations in neutron fluence and energy spectrum are evident in the measured isotope ratios. Two versions of a combined Pu and Cs based PIM are developed. A linear PIM model, which can be used to solve for irradiation time is found to work well for natural U fuel with < 10% 240 Pu and known or short cooling times. A non-linear PIM model, which cannot be solved explicitly for irradiation time without additional information, can nonetheless still group samples by irradiation history, including high burnup LEU fuel with unknown cooling time. 137 Ba/ 138 Ba is also observed to act as a position independent monitor; it is nearly single valued across the sampled fuel rod, indicating that samples sharing an irradiation history (same irradiation time and cooling time) in a reactor despite experiencing different neutron fluxes will have a common 137 Ba/ 138 Ba ratio. Modeling of this Ba PIM shows it increases monotonically with irradiation and cooling time, and a confirmatory first order analytical solution is also presented.
Needs of Accurate Prompt and Delayed γ-spectrum and Multiplicity for Nuclear Reactor Designs
Physics Procedia, 2012
The local energy photon deposit must be accounted accurately for Gen-IV fast reactors, advanced light-water nuclear reactors (Gen-III + ) and the new experimental Jules Horowitz Reactor (JHR). The energy accounts for about 10% of the total energy released in the core of a thermal or fast reactor. The -energy release is much greater in the core of the reactor than in its structural sub-assemblies (such as reflector, control rod followers, dummy sub-assemblies). However, because of the propagation of from the core regions to the neighboring fuel-free assemblies, the contribution of energy to the total heating can be dominant. For reasons related to their performance, power reactors require a 7.5% (1 ) uncertainty for the energy deposition in non-fuelled zones. For the JHR material-testing reactor, a 5% (1s) uncertainty is required in experimental positions. In order to verify the adequacy of the calculation of γ-heating, TLD and γ-fission chambers were used to derive the experimental heating values. Experimental programs were and are still conducted in different Cadarache facilities such as MASURCA (for SFR), MINERVE and EOLE (for JHR and Gen-III + reactors). The comparison of calculated and measured γ-heating values shows an underestimation in all experimental programs indicating that for the most γ-production data from 239 Pu in current nuclear-data libraries is highly suspicious. The first evaluation priority is for prompt -multiplicity for 235 U and 239 Pu fission but similar values for other actinides such as 241 Pu and 238 U are also required. The nuclear data library JEFF3.1.1 contains most of the photon production data. However, there are some nuclei for which there are missing or erroneous data which need to be completed or modified. A review of the data available shows a lack of measurements for conducting serious evaluation efforts. New measurements are needed to guide new evaluation efforts which benefit from consolidated modeling techniques.
Revisiting Statistical Aspects of Nuclear Material Accounting
Science and Technology of Nuclear Installations
Nuclear material accounting (NMA) is the only safeguards system whose benefits are routinely quantified. Process monitoring (PM) is another safeguards system that is increasingly used, and one challenge is how to quantify its benefit. This paper considers PM in the role of enabling frequent NMA, which is referred to as near-real-time accounting (NRTA). We quantify NRTA benefits using period-driven and data-driven testing. Period-driven testing makes a decision to alarm or not at fixed periods. Data-driven testing decides as the data arrives whether to alarm or continue testing. The difference between period-driven and datad-riven viewpoints is illustrated by using one-year and two-year periods. For both one-year and two-year periods, period-driven NMA using once-per-year cumulative material unaccounted for (CUMUF) testing is compared to more frequent Shewhart and joint sequential cusum testing using either MUF or standardized, independently transformed MUF (SITMUF) data. We show tha...
Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 2014
Statistical analyses have been performed to develop bounding estimates of the expected performance of a 13 conceptual fast-neutron multiplicity system (FNMS) for assaying plutonium. The conceptual FNMS design includes 32 cubic liquid scintillator detectors, measuring 7.62 cm per side, configured into 4 stacked rings of 8 detectors each. Expected response characteristics for the individual FNMS detectors, as well as the response characteristics of the entire FNMS, were determined using Monte Carlo simulations based on prior validation experiments. The results from these simulations were then used to estimate the Pu assay capabilities of the FNMS in terms of counting time, assay mass, and assay mass variance, using assay mass variance as a figure of merit. The analysis results are compared against a commonly used thermal-neutron coincidence counter. The advantages of using a fast-neutron counting system versus a thermal-neutron counting system are significant. Most notably, the time required to perform an assay to an equivalent assay mass variance is greatly reduced with a fast-neutron system, by more than an order of magnitude compared with a thermal-neutron system, due to the reduced probability