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In accordance with U.S. Department of Energy requirements in DOE O 435.1 and as implemented by DOE/RL-2009-29, the DOE Richland Operations Office has prepared this annul summary of the Hanford Site Composite Analysis for fiscal year 2014.
This annual review provides the projected dose estimates of radionuclide inventories disposed in the active Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. These estimates are calculated using the original does methodology developed in the performance assessment analyses.
In accordance with U.S. Department of Energy (DOE) requirements in DOE O 435.1, and as implemented by DOE/RL-2000-29, the DOE Richland Operations Office has prepared this annual summary of the Hanford Site Composite Analysis for fiscal year 2015.
This annual review provides the projected dose estimates of radionuclide inventories disposed in the active 200 West Area Low-Level Burial Grounds (LLBGs) since September 26, 1988. These estimates area calculated using the original does methodology developed in the performance assessment (PA) analysis (WHC-EP-0645).
This annual review provides the projected dose estimates of radionuclides inventories disposed in the active 200 East Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988.
In accordance with U.S. Department of Energy (DOE) requirements in DOE O 435.1 and as implemented by DOE/RL-2009-29, the DOE Richland Operations Office has prepared this annual summary of the Hanford Site Composite Analysis for fiscal year 2016.
This annual review provides the projected dose estimates of radionuclide inventories disposed in the active 200 West Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988.
This report provides a summary of the progress made towards validating the multi-physics reactor analysis application MAMMOTH using data from measurements performed at the Transient Reactor Test facility, TREAT. The work completed consists of a series of comparisons of TREAT element types (standard and control rod assemblies) in small geometries as well as slotted mini-cores to reference Monte Carlo simulations to ascertain the accuracy of cross section preparation techniques. After the successful completion of these smaller problems, a full core model of the half slotted core used in the M8 Calibration series was assembled. Full core MAMMOTH simulations were compared to Serpent reference calculations to assess the cross section preparation process for this larger configuration. As part of the validation process the M8 Calibration series included a steady state wire irradiation experiment and coupling factors for the experiment region. The shape of the power distribution obtained from the MAMMOTH simulation shows excellent agreement with the experiment. Larger differences were encountered in the calculation of the coupling factors, but there is also great uncertainty on how the experimental values were obtained. Future work will focus on resolving some of these differences.
A collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core fission chambers and thermocouple. As discussed within this report, the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for year two of this three year project. Highlights from research accomplishments include: • Continuation of a joint collaboration between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. • An updated parallel wire HT MPFD design was developed. • Program support for HT MPFD deployments was given to Accident Tolerant Fuels (ATF) and Advanced Gas-cooled Reactor (AGR) irradiation test programs. • Quality approved materials for HT MPFD construction were procured by irradiation test programs for upcoming deployments. • KSU improved and performed electrical contact and fissile material plating. • KSU delivered fissile HT MPFD parts to INL for final construction of HT MPFD prototype. • A prototype HT MPFD was constructed and analyzed at INL. • The HT MPFD has been modeled in MCNP to optimize the amount of fissile material deposition. • The HT MPFD has been modeled in MCNP to optimize the sensor location in the irradiation test. • The fissile material deposition is undergoing independent verifications. • Detector amplifier electronics have been revised and tested by KSU. • Several project meetings were held at INL and KSU to discuss the roles and responsibilities between INL, KSU, and CEA for development and deployment of the HT MPFDs. As documented in this report, FY16 funding has allowed the project to meet year two planned accomplishments to develop a HT MPFD. In addition, the accomplishments of this project have attracted independent funding from other Department of Energy Office of Nuclear Energy (DOE-NE) programs for MTR irradiations of the MPFD technology. These are significant opportunities for this NEET Enhanced Micro-Pocket Fission Detector for High Temperature Reactors project because the irradiation expense of these experiments could not be included in the original project scope.
Grade 91 is a candidate structural material for high temperature advanced reactor applications. Existing ASME Section III, Subsection HB, Subpart B simplified design rules based on elastic analysis are setup as conservative screening tools with the intent to supplement these screening rules with full inelastic analysis when required. The Code provides general guidelines for suitable inelastic models but does not provide constitutive model implementations. This report describes the development of an inelastic constitutive model for Gr. 91 steel aimed at fulfilling the ASME Code requirements and being included into a new Section III Code appendix, HBB-Z. A large database of over 300 experiments on Gr. 91 was collected and converted to a standard XML form. Five families of Gr. 91 material models were identified in the literature. Of these five, two are potentially suitable for use in the ASME code. These two models were implemented and evaluated against the experimental database. Both models have deficiencies so the report develops a framework for developing and calibrating an improved model. This required creating a new modeling method for representing changes in material rate sensitivity across the full ASME allowable temperature range for Gr. 91 structural components: room temperature to 650° C. On top of this framework for rate sensitivity the report describes calibrating a model for work hardening and softening in the material using genetic algorithm optimization. Future work will focus on improving this trial model by including tension/compression asymmetry observed in experiments and necessary to capture material ratcheting under zero mean stress and by improving the optimization and analysis framework.
A series of experiments to monitor the aging performance of Viton® GLT and GLT-S O-rings used in the Model 9975 shipping package has been ongoing since 2004 at the Savannah River National Laboratory. One approach has been to periodically evaluate the leak performance of O-rings being aged in mock-up 9975 Primary Containment Vessels (PCVs) at elevated temperature. Other methods such as compression-stress relaxation (CSR) tests and field surveillance are also on-going to evaluate O-ring behavior. Seventy tests using PCV mock-ups with GLT O-rings were assembled and heated to temperatures ranging from 200 to 450 ºF. They were leak-tested initially and have been tested periodically to determine if they continue to meet the leak-tightness criterion defined in ANSI standard N14.5-97. Due to material substitution, a smaller test matrix with fourteen additional tests was initiated in 2008 with GLT-S O-rings heated to temperatures ranging from 200 to 400 ºF. Leak test failures have been experienced in all of the GLT O-ring fixtures aging at 350 ºF and higher temperatures, and in 8 fixtures aging at 300 ºF. The 300 °F GLT O-ring fixtures failed after 2.8 to 5.7 years at temperature. The remaining GLT O-ring fixtures aging at 300 ºF were retired from testing following more than 5 years at temperature without failure. No failures have yet been observed in GLT O-ring fixtures aging at 200 ºF for 9 to 10.5 years, or in GLT O-ring fixtures aging at 270 ºF for 5.7 years. These aging temperatures bound O-ring temperatures anticipated during normal storage in K-Area Complex (KAC). Leak test failures have been experienced in all of the GLT-S O-ring fixtures aging at 300 ºF and above. No failures have yet been observed in GLT-S O-ring fixtures aging at 200 and 250 ºF for 6.9 to 7.5 years. Data from the O-ring fixtures are generally consistent with results from compression stress relaxation testing, and provide confidence in the predictive models based on those results. However, uncertainty still exists in extrapolating these elevated temperature results to the lower temperatures of interest for normal storage in KAC. Measurement of compression set in O-rings removed from failed fixtures, compared to that from KAC surveillance O-rings, indicates margin remains for O-rings still in service. Aging and periodic leak testing will continue for the remaining PCV fixtures.
This report describes a modification to the elastic-perfectly plastic (EPP) strain limits design method to account for cyclic softening in Gr. 91 steel. The report demonstrates that the unmodified EPP strain limits method described in current ASME code case is not conservative for materials with substantial cyclic softening behavior like Gr. 91 steel. However, the EPP strain limits method can be modified to be conservative for softening materials by using softened isochronous stress-strain curves in place of the standard curves developed from unsoftened creep experiments. The report provides softened curves derived from inelastic material simulations and factors describing the transformation of unsoftened curves to a softened state. Furthermore, the report outlines a method for deriving these factors directly from creep/fatigue tests. If the material softening saturates the proposed EPP strain limits method can be further simplified, providing a methodology based on temperature-dependent softening factors that could be implemented in an ASME code case allowing the use of the EPP strain limits method with Gr. 91. Finally, the report demonstrates the conservatism of the modified method when applied to inelastic simulation results and two bar experiments.
Among long-lived radioactive constituents in the Hanford tank waste, Tc presents a unique challenge in that it exists predominantly in the liquid phase, generally in the anionic form of pertechnetate, TcO 4 - , which is highly volatile at low-activity waste (LAW) vitrification melter temperatures and mobile in the Hanford site’s subsurface environment. The complex behavior of Tc under storage, treatment, and immobilization conditions significantly affects its management options, which to-date remain uncertain. In strongly alkaline environments, Tc exists as pertechnetate, TcO 4 - (oxidation state +7), and in the reduced forms (oxidation state < +7) collectively known as non-pertechnetate species. Pertechnetate is a well-characterized, anionic Tc species that can be removed from LAW by anion exchange or solvent extraction methods. There is no definitive information on the origin of the non-pertechnetate Tc species, nor is there a comprehensive description of their composition and behavior. It has been recently proposed that the non-pertechnetate species can comprise Tc(I) metal center and carbonyl or mixed carbonyl nitrosyl ligands stabilizing low-valent Tc. Recent work by our group has significantly expanded this previous work, generating a series of Tc(I) carbonyl compounds and demonstrating that they can be generated from reduction of TcO 4 - in the simulated Hanford tank waste in presence of CO at elevated temperature (Levitskaia et al. 2014). These results are consistent with the previous proposal that [Tc(CO) 3 ] + species can be present in the Hanford tank waste and suggest that the low Tc(I) oxidation state is stabilized by the π-accepting ability of the CO ligands. The continuation work has been initiated to develop model Tc carbonyl nitrosyl compounds and investigate their potential presence in the Hanford tank wastes. This report summarizes our to-date results.
The overall objective of this project is to develop predictive modeling capabilities for advanced fuel cycle separation processes by gaining a fundamental quantitative understanding of non-ideality effects and speciation in relevant aqueous and organic solutions. Aqueous solutions containing actinides and lanthanides encountered during nuclear fuel reprocessing have high ionic strength and do not behave as ideal solutions. Activity coefficients must be calculated to take into account the deviation from ideality and predict their behavior. In FY 2012-2013, a convenient method for determining activity effects in aqueous electrolyte solutions was developed. Our initial experiments demonstrated that water activity and osmotic coefficients of the electrolyte solutions can be accurately measured by the combination of two techniques, a Water Activity Meter and Vapor Pressure Osmometry (VPO). The water activity measurements have been conducted for binary lanthanide solutions in wide concentration range for all lanthanides (La-Lu with the exception of Pm). The osmotic coefficients and Pitzer parameters for each binary system were obtained by the least squares fitting of the water activity data. However, application of Pitzer model for the quantitative evaluation of the activity effects in the multicomponent mixtures is difficult due to the large number of the required interaction parameters. In FY 2014, the applicability of the Bromley model for the determination of the Ln(NO 3 ) 3 activity coefficients was evaluated. The new Bromley parameters for the binary Ln(NO 3 ) 3 electrolytes were obtained based on the available literature and our experimental data. This allowed for the accurate prediction of the Ln(NO 3 ) 3 activity coefficients for the binary Ln(NO 3 ) 3 electrolytes. This model was then successfully implemented for the determination of the Ln(NO 3 ) 3 activity coefficients in the ternary Nd(NO 3 ) 3 /HNO 3 /H2O, Eu(NO 3 ) 3 /HNO 3 /H 2 O, and Eu(NO 3 ) 3 /NaNO 3 /H 2 O systems. The main achievement of this work is the verified pathway for the estimation of the activity coefficients in the multicomponent aqueous electrolyte systems. The accurate Bromley electrolytes contributions obtained in this work for the entire series of lanthanide(III) nitrates (except Pm) can be applied for predicting activity coefficients and non-ideality effects for multi-component systems containing these species. This work also provides the proof-of-principle of extending the model to more complex multicomponent systems. Moreover, this approach can also be applied to actinide-containing electrolyte systems, for determination of the activity coefficients in concentrated radioactive solutions.
This report summarizes the experiments performed in FY17 on Gr. 91 steels. The testing of Gr. 91 has technical significance because, currently, it is the only approved material for Class A construction that is strongly cyclic softening. Specific FY17 testing includes the following activities for Gr. 91 steel. First, two types of key feature testing have been initiated, including two-bar thermal ratcheting and Simplified Model Testing (SMT). The goal is to qualify the Elastic – Perfectly Plastic (EPP) design methodologies and to support incorporation of these rules for Gr. 91 into the ASME Division 5 Code. The preliminary SMT test results show that Gr. 91 is most damaging when tested with compression hold mode under the SMT creep fatigue testing condition. Two-bar thermal ratcheting test results at a temperature range between 350 to 650o C were compared with the EPP strain limits code case evaluation, and the results show that the EPP strain limits code case is conservative. The material information obtained from these key feature tests can also be used to verify its material model. Second, to provide experimental data in support of the viscoplastic material model development at Argonne National Laboratory, selective tests were performed to evaluate the effect of cyclic softening on strain rate sensitivity and creep rates. The results show the prior cyclic loading history decreases the strain rate sensitivity and increases creep rates. In addition, isothermal cyclic stress-strain curves were generated at six different temperatures, and a nonisothermal thermomechanical testing was also performed to provide data to calibrate the viscoplastic material model.
This progress report describes work done in FY17 at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. Work in FY17 refined our understanding of the chemical and physical environment on canister surfaces, and evaluated the relationship between chemical and physical environment and the form and extent of corrosion that occurs. The SNL corrosion work focused predominantly on pitting corrosion, a necessary precursor for SCC, and process of pit-to-crack transition; it has been carried out in collaboration with university partners. SNL is collaborating with several university partners to investigate SCC crack growth experimentally, providing guidance for design and interpretation of experiments.
This report is a summary of the international collaboration and laboratory work funded by the US Department of Energy Office of Nuclear Energy Spent Fuel and Waste Science & Technology (SFWST) as part of the Sandia National Laboratories Salt R&D work package. This report satisfies milestone levelfour milestone M4SF-17SN010303014. Several stand-alone sections make up this summary report, each completed by the participants. The first two sections discuss international collaborations on geomechanical benchmarking exercises (WEIMOS) and bedded salt investigations (KOSINA), while the last three sections discuss laboratory work conducted on brucite solubility in brine, dissolution of borosilicate glass into brine, and partitioning of fission products into salt phases.