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At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Stand-Alone Dynamic System Simulation of a Fissile Solution System

Dynamic system simulations (DSS) of fissile solution systems (FSS) have been performed at Los Alamos National Laboratory (LANL) for several years. DSS have been designed and executed using the DESIRE (Direct Executing Simulation in Real Time) system, a platform ideal for expert system modelers, as it allows for direct mathematical specification of a model, and facilities for performing experiments and analyses with the model. Models built in DESIRE have been compared to historical data from several historic aqueous homogeneous reactors (AHR) showing close agreement between experimental data and model estimates. Historical systems modeled include: SUPO (Super Power), an AHR that operated at Los Alamos National Laboratory from 1951 to 1974; SILENE, which was operated by CEA at Valduc, France until 2011; KEWB (Kinetics Experiment Water Boiler) A-2 and B-5 cores, operated by North American Atomics during the 1960’s and 1970’s. SUPO is considered a benchmark for steady-state operations, SILENE for pulse, and KEWB as confirmatory for both modes. However, there are some simulation tasks that are not well-suited to DESIRE model implementation. These tasks include engineering design and analysis and providing a hardware simulator backend. It has been determined that executable simulation models would better suit these purposes, hence a process for converting DESIRE simulation models to executable simulation models was designed and implemented, and several models have been converted. DESIRE simulation models are converted to C++ executable models by an automated process. The conversion process is quick, accurate and repeatable, giving confidence that converted models will perform as well as the original DESIRE models. Extensive testing on the SUPO and accelerator driven subcritical system models has shown that the executable models perform almost identically to the DESIRE models. All major trends and events are simulated with excellent agreement and the steady-state results typically agree within a small fraction of a percent (always better than 0.1% and generally better than 0.01% agreement).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Stability of Fissile Solution Systems

A technique has been developed to model the dynamics of fissile solution systems. This technique, which can be characterized as Dynamic System Simulation (DSS), is a set of coupled nonlinear differential equations that may be solved in the time domain to simulate the dynamic behavior of the overall system. The details of model development and results have previously been documented, hence, herein only a summary is presented. Herein an extension of the modeling technique is described that uses output of the simulation models to examine stability. Simplified versions of the simulation model have been linearized and transformed into the complex plane. Transfer functions for each of the system components were developed and combined to generate overall system closed- and open-loop transfer functions. Standard linear stability analysis techniques, such as frequency response and pole configuration, were used to evaluate the stability of the system.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Extensions to Dynamic System Simulation of Fissile Solution Systems

Previous reports have documented the results of applying dynamic system simulation (DSS) techniques to model a variety of fissile solution systems. The SUPO (Super Power) aqueous homogeneous reactor (AHR) was chosen as the benchmark for comparison of model results to experimental data for steadystate operation.1 Subsequently, DSS was applied to additional AHR to verify results obtained for SUPO and extend modeling to prompt critical excursions, ramp reactivity insertions of various magnitudes and rate, and boiling operations in SILENE and KEWB (Kinetic Experiment Water Boiler).2 Additional models for pressurized cores (HRE: Homogeneous Reactor Experiment), annular core geometries, and accelerator-driven subcritical systems (ADAHR) were developed and results reported.3 The focus of each of these models is core dynamics; neutron kinetics, thermal hydraulics, radiolytic gas generation and transport are coupled to examine the time-based evolution of these systems from start-up through transition to steady-state. A common characteristic of these models is the assumption that (a) core cooling system inlet temperature and flow and (b) plenum gas inlet pressure and flow are held constant; no external (to core) component operations that may result in dynamic change to these parameters are considered. This report discusses extension of models to include explicit reference to cooling structures and radiolytic gas handling. The accelerator-driven subcritical generic system model described in References 3 and 4 is used as a basis for this extension.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Tank 40 final sludge batch 9 chemical and fissile radionuclide characterization results

A sample of Sludge Batch (SB) 9 was pulled from Tank 40 in order to obtain radionuclide inventory analyses necessary for compliance with the Waste Acceptance Product Specifications (WAPS)i. The SB9 WAPS sample was also analyzed for chemical composition, including noble metals, and fissile constituents, and these results are reported here. These analyses along with the WAPS radionuclide analyses will help define the composition of the sludge in Tank 40 that is fed to the Defense Waste Processing Facility (DWPF) as SB9. At the Savannah River National Laboratory (SRNL), the 3-L Tank 40 SB9 sample was transferred from the shipping container into a 4-L high density polyethylene bottle and solids were allowed to settle. Supernate was then siphoned off and circulated through the shipping container to complete the transfer of the sample. Following thorough mixing of the 3-L sample, a 547 g sub-sample was removed. This sub-sample was then utilized for all subsequent slurry sample preparations. Eight separate aliquots of the slurry were digested, four with HNO3/HCl (aqua regiaii) in sealed Teflon® vessels and four with NaOH/Na2O2 (alkali or peroxide fusioniii) using Zr crucibles. Three Analytical Reference Glass – 1iv (ARG-1) standards were digested along with a blank for each preparation. Each aqua regia digestion and blank was diluted to 1:100 with deionized water and submitted to Analytical Development (AD) for inductively coupled plasma – atomic emission spectroscopy (ICP-AES) analysis, inductively coupled plasma – mass spectrometry (ICP-MS) analysis, atomic absorption spectroscopy (AA) for As and Se, and cold vapor atomic absorption spectroscopy (CV-AA) for Hg. Equivalent dilutions of the alkali fusion digestions and blank were submitted to AD for ICP-AES analysis. Tank 40 SB9 supernate was collected from a mixed slurry sample in the SRNL Shielded Cells and submitted to AD for ICP-AES, ion chromatography (IC), total base/free OH-/other base, total inorganic carbon/total organic carbon (TIC/TOC) analyses. Weighted dilutions of slurry were submitted for IC, TIC/TOC, and total base/free OH-/other base analyses. Activities for U-233, U-235, and Pu-239 were determined from the ICP-MS data for the aqua regia digestions of the SB9 WAPS slurry using the specific activity of each isotope. The Pu-241 value was determined from a Pu-238/-241 method developed by SRNL AD and previously described.v

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

INDEN Fissile Actinide Issues [Slides]

This presentation addresses the outlook for the International Nuclear Data Evaluation Network (INDEN), specifically regarding general issues that have been solved, including Thermal Neutron Constants (TNC) and Prompt Fission Neutron Spectra (PFNS). Other topics discussed include outstanding issues in fissile evaluations, comments on the RPI quasi-differential experiment, and reaction rate testing of trial evaluations in fast assemblies. The presentation concludes by discussing how INDEN interactions strongly helped to find deficiencies in existing evaluations and highlighted potential solutions to existing challenges, that updated trial evaluations for U-235, U-233, Pu-239 are available for testing, that further work is expected on U-235, Pu-239, and U-233, and that performance was tested on the ICSBEP and/or SINBAD benchmarks, during which significant improvement was demonstrated. The INDEN scheme of international collaboration on nuclear data evaluation is working well.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2012

DEVELOPMENT OF THE HS99 AIR TRANSPORT TYPE A FISSILE PACKAGE

An air-transport Type A Fissile radioactive shipping package for the transport of special form uranium sources has been developed by the Savannah River National Laboratory (SRNL) for the Department of Homeland Security. The Package model number is HS99 for Homeland Security Model 99. This paper presents the major design features of the HS99 and highlights engineered materials necessary for meeting the design requirements for this light-weight Type AF packaging. A discussion is provided demonstrating how the HS99 complies with the regulatory safety requirements of the Nuclear Regulatory Commission. The paper summarizes the results of structural testing to specified in 10 CFR 71 for Normal Conditions of Transport and Hypothetical Accident Conditions events. Planned and proposed future missions for this packaging are also addressed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

TYPE A FISSILE PACKAGING FOR AIR TRANSPORT PROJECT OVERVIEW

This paper presents the project status of the Model 9980, a new Type A fissile packaging for use in air transport. The Savannah River National Laboratory (SRNL) developed this new packaging to be a light weight (<150-lb), drum-style package and prepared a Safety Analysis for Packaging (SARP) for submission to the DOE/EM. The package design incorporates unique features and engineered materials specifically designed to minimize packaging weight and to be in compliance with 10CFR71 requirements. Prototypes were fabricated and tested to evaluate the design when subjected to Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC). An overview of the design details, results of the regulatory testing, and lessons learned from the prototype fabrication for the 9980 will be presented.

42 ENGINEERING↗
DOE OSTI2017

Validation of the Fast-Neutron Multiplicity Expressions for Fissile Mass Estimation

Neutron multiplicity analysis is a non-destructive assay technique that involves measurement of the multiplets of neutrons that are emitted from fission of fissile material. Typically, multiplicity analysis is conducted by means of a neutron multiplicity counting (NMC) system that uses capture-based thermal-neutron detectors (e.g., 3He-gas detectors). To relieve the reliance on such detectors, we have developed a fast-neutron multiplicity counting (FNMC) system that uses scatter-based, organic scintillators to directly detect the fast fission neutrons emitted from the sample. One challenge with FNMC systems, however, is that they are prone to effects of neutron cross-talk, where a single neutron may scatter and cause more than one detection event. This leads to an overestimate of the fission rate (F) and inaccurate estimation of other sample parameters such as the leakage multiplication (M) and the alpha-ratio (a) when using the traditional capture-based NMC equations. We have formulated fully generalized multiplicity expressions that account for cross-talk effects for any arbitrary order N. The validity of these expressions were verified with MCNPX-PoliMi simulations and experimental measurements; this paper will present the validation of the proposed generalized multiplicity expressions. Californium-252 and Pu metal plates were measured using the FNMC system and the data was used for neutron multiplicity analysis to estimate sample parameters F, M, and a. We demonstrate that accounting for neutron cross-talk effects of order N = 2 and 3 improves the accuracy of the estimated sample parameters.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2013

Random effects of fissile lumps in molten salt reactors

The problem of the effect of fissile lumps spatially appearing in a random fashion inside a fluid fuel reactor is addressed. The effect on reactivity is evaluated by means of first-order perturbation theory. The analysis is carried out in diffusion theory with the presence of delayed neutron emissions in one dimensional plane geometry. The estimation of the mean value and standard deviation of the reactivity inserted is performed by Monte Carlo simulations and a deterministic quadrature approach, to compare the methods in terms of computational effort and the accuracy of the results. The results presented show that the effects constitute an important issue in the assessment of these innovative systems. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

A setup for active neutron analysis of the fissile material content in fuel assemblies of nuclear reactors

An active neutron method for measuring the residual mass of {sup 235}U in spent fuel assemblies (FAs) of the IRT MEPhI research reactor is presented. The special measuring stand design and uniform irradiation of the fuel with neutrons along the entire length of the active part of the FA provide high accuracy of determination of the residual {sup 235}U content. AmLi neutron sources yield a higher effect/background ratio than other types of sources and do not induce the fission of {sup 238}U. The proposed method of transfer of the isotope source in accordance with a given algorithm may be used in experiments where the studied object needs to be irradiated with a uniform fluence.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Increased Fissile Loading Flowsheet Review

A request was made by H-Canyon Process Engineering to assess the impact of blending dissolved, neutralized Spent Nuclear Fuel (SNF) with future sludge batches and their impact on downstream processing facilities. The purpose of this change is to accelerate the deinventory of SNF which is currently stored in the L-Area Disassembly Basin.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

APPLICATION OF VACUUM SALT DISTILLATION TECHNOLOGY FOR THE REMOVAL OF FLUORIDE AND CHLORIDE FROM LEGACY FISSILE MATERIALS

Between September 2009 and January 2011, the Savannah River National Laboratory (SRNL) and the Savannah River Site (SRS) HB-Line Facility designed, developed, tested, and successfully deployed a production-scale system for the distillation of sodium chloride (NaCl) and potassium chloride (KCl) from plutonium oxide (PuO{sub 2}). Subsequent efforts adapted the vacuum salt distillation (VSD) technology for the removal of chloride and fluoride from less-volatile halide salts at the same process temperature and vacuum. Calcium chloride (CaCl{sub 2}), calcium fluoride (CaF{sub 2}), and plutonium fluoride (PuF{sub 3}) were of particular concern. To enable the use of the same operating conditions for the distillation process, SRNL employed in situ exchange reactions to convert the less-volatile halide salts to compounds that facilitated the distillation of halide without removal of plutonium. SRNL demonstrated the removal of halide from CaCl{sub 2}, CaF{sub 2} and PuF{sub 3} below 1000 C using VSD technology.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Criticality safety analysis on fissile materials in Fukushima reactor cores

The present study focuses on the criticality analysis for geological disposal of damaged fuels from Fukushima reactor cores. Starting from the basic understanding of behaviors of plutonium and uranium, a scenario sequence for criticality event is considered. Due to the different mobility of plutonium and uranium in geological formations, the criticality safety is considered in two parts: (1) near-field plutonium system and (2) far-field low enriched uranium (LEU) system. For the near-field plutonium system, a mathematical analysis for pure-solute transport was given, assuming a particular buffer material and waste form configuration. With the transport and decay of plutonium accounted, the critical mass of plutonium was compared with the initial load of a single canister. Our calculation leads us to the conclusion that our system with the initial loading being the average mass of plutonium in an assembly just before the accident is very unlikely to become critical over time. For the far-field LEU system, due to the uncertainties in the geological and geochemical conditions, calculations were made in a parametric space that covers the variation of material compositions and different geometries. Results show that the LEU system could not remain sub-critical within the entire parameter space assumed, although in the iron-rich rock, the neutron multiplicity is significantly reduced.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗