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All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Development of Ultra-Fine Multigroup Cross Section Library of the AMPX/SCALE Code Packages

The Consortium for Advanced Simulation of Light Water Reactors Virtual Environment for Reactor Applications (VERA) neutronic simulator MPACT is being developed by Oak Ridge National Laboratory and the University of Michigan for various reactor applications. The MPACT and simplified MPACT 51- and 252-group cross section libraries have been developed for the MPACT neutron transport calculations by using the AMPX and Standardized Computer Analyses for Licensing Evaluations (SCALE) code packages developed at Oak Ridge National Laboratory. It has been noted that the conventional AMPX/SCALE procedure has limited applications for fast-spectrum systems such as boiling water reactor (BWR) fuels with very high void fractions and fast reactor fuels because of its poor accuracy in unresolved and fast energy regions. This lack of accuracy can introduce additional error sources to MPACT calculations, which is already limited by the Bondarenko approach for resolved resonance self-shielding calculation. To enhance the prediction accuracy of MPACT for fast-spectrum reactor analyses, the accuracy of the AMPX/SCALE code packages should be improved first. The purpose of this study is to identify the major problems of the AMPX/SCALE procedure in generating fast-spectrum cross sections and to devise ways to improve the accuracy. For this, various benchmark problems including a typical pressurized water reactor fuel, BWR fuels with various void fractions, and several fast reactor fuels were analyzed using the AMPX 252-group libraries. Isotopic reaction rates were determined by SCALE multigroup (MG) calculations and compared with continuous energy (CE) Monte Carlo calculation results. This reaction rate analysis revealed three main contributors to the observed differences in reactivity and reaction rates: (1) the limitation of the Bondarenko approach in coarse energy group structure, (2) the normalization issue of probability tables, and (3) neglect of the self-shielding effect of resonance-like cross sections at high energy range such as (n,p) cross section of Cl35. The first error source can be eliminated by an ultra-fine group (UFG) structure in which the broad scattering resonances of intermediate-weight nuclides can be represented accurately by a piecewise constant function. A UFG AMPX library was generated with modified probability tables and tested against various benchmark problems. The reactivity and reaction rates determined with the new UFG AMPX library agreed very well with respect to Monte Carlo Neutral Particle (MCNP) results. To enhance the lattice calculation accuracy without significantly increasing the computational time, performing the UFG lattice calculation in two steps was proposed. In the first step, a UFG slowing-down calculation is performed for the corresponding homogenized composition, and UFG cross sections are collapsed into an intermediate group structure. In the second step, the lattice calculation is performed for the intermediate group level using the condensed group cross sections. A preliminary test showed that the condensed library reproduces the results obtained with the UFG cross section library. This result suggests that the proposed two-step lattice calculation approach is a promising option to enhance the applicability of the AMPX/SCALE system to fast system analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Influence of Multi-Dimension Heat Conduction on Heat Flux Calculation for HFIR LEU Analysis

Steady-state safety margins calculations are an important step in the evaluation of research reactor design. Margins for phenomena such as onset of nucleate boiling and critical heat flux are evaluated by comparing the heat flux leading to those phenomena to a local heat flux. Evaluating properly the magnitude of the heat flux is therefore important and can affect the design. Plate-type research reactors often exhibit significant power peaking on the edges of the plate, near the fuel/cladding interface. The magnitude of the heat flux calculated in these regions can be influenced by the size of the region in which they are evaluated but also by the ability to model heat transport in one or multiple dimensions. The present analysis studies the influence of these two parameters on the heat flux magnitude for the HFIR reactor. From this analysis, a fuel discretization grid is proposed that can generate one-dimension heat flux values relatively close to best estimate values obtained with a high-resolution grid and complex three-dimensional CFD modeling.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Updated Verification and Validation Assessment for VERA

The Virtual Environment for Reactor Applications (VERA) code suite is assessed in terms of capability and credibility against the Consortium for Advanced Simulation of Light Water Reactors (CASL) Verification and Validation Plan (presented herein) in the context of three selected challenge problems: CRUD-Induced Power Shift (CIPS), Departure from Nucleate Boiling (DNB), and Pellet-Clad Interaction (PCI). Capability refers to evidence of required functionality for capturing phenomena of interest while credibility refers to the evidence that provides confidence in the calculated results. For this assessment, each challenge problem defines a set of phenomenological requirements against which the VERA software is assessed. This approach, in turn, enables the focused assessment of only those capabilities relevant to the challenge problem. The evaluation of VERA against the challenge problem requirements represents a capability assessment. The mechanism for assessment is the Sandia-developed Predictive Capability Maturity Model (PCMM) that, for this assessment, evaluates VERA on 8 major criteria: (1) Representation and Geometric Fidelity, (2) Physics and Material Model Fidelity, (3) Software Quality Assurance and Engineering, (4) Code Verification, (5) Solution Verification, (6) Separate Effects Model Validation, (7) Integral Effects Model Validation, and (8) Uncertainty Quantification. For each attribute, a maturity score from zero to three is assigned in the context of each challenge problem. The evaluation of these eight elements constitutes the credibility assessment for VERA.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

External Cooling of the BWR Mark I and II Drywell Head as a Potential Accident Mitigation Measure – Expanded Scoping Assessment

This report documents a scoping assessment of a potential accident mitigation action applicable to the US fleet of boiling water reactors with Mark I and II containments. The mitigation action is to externally flood the primary containment vessel drywell head using portable pumps or other means. A scoping assessment of the potential benefits of this mitigation action was conducted focusing on the ability to (1) passively remove heat from containment, (2) prevent or delay leakage through the drywell head seal (due to high temperatures and/or pressure), and (3) scrub radionuclide releases if the drywell head seal leaks. This report is an expanded revision of ORNL/TM-2017/457 and includes an analysis of (1) the delay in initiating drywell head cavity flooding from 2 hours to 8 hours after accident scenario initiation and (2) the effectiveness of including additional drywell head heat transfer area.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Physics based Reliability Models for Supercritical CO 2 Turbomachinery Components

CSP plants using sCO 2 power cycle can potentially achieve high thermal conversion efficiency at low capital cost due to compact turbo-machinery and other components. An sCO 2 expander and improved heat exchanger is expected to provide a major stepping stone for achieving CSP power at $0.06/kW-hr LCOE, energy conversion efficiency > 50%, and total power block cost < $1,200/kW installed. Hybrid gas bearings (HGB) and Dry gas seals (DGS) are two key components that are critical for robust and efficient operation of future sCO 2 power cycles. From a system perspective, HGBs can provide substantial benefits (~3 to 5%) to the modular CSP operation, while use of DGSs limits the leakage losses as well as the windage losses in the generator and can increase overall generator efficiency by ~ 8%. The life limiting mechanisms of HGBs and DGSs in high pressure, high temperature sCO2 environment were not well understood and hence this program aims at developing physics-based performance prediction and life prediction models for these two components. The output of the performance models serves as input to the life models. The performance models were validated using experimental data and using data from literature. DGS and HGB models have shed light on some risks associated with its operation in sCO2 expander and will guide more robust designs tailored towards sCO 2 cycle applications. A DGS was tested in 1MWe sCO2 turbo expander to study its performance and validate the performance models. To understand the effect of high pressures, high temperatures and sCO 2 environment on crack initiation, crack propagation and LCF/HCF life of these components, a novel experimental setup was developed. The LCF test data was used to calibrate and validate the life prediction models. The effect of sCO 2 corrosion on the life of Ni-based super-alloys is quantified using long duration experiments. Advanced microstructure and spectroscopic analyses conducted shed light on some key differences between IN718 and IN617 in terms of oxidation morphology, chemical species diffusion and trapping, the formation of protective corrosion resistant layers and changes in surface properties. A novel crack initiation model was developed which takes into account strain distribution, temperature, hold times and activation energy at the gas-metal interface and predicts the number of cycles to initiate the crack in the Inconel alloy. An advanced 3D fracture mechanics-based crack propagation model was also developed which takes into account the transient loading mission of the component and predict how fast the crack will grow and what will be its final shape and size. High energy X-Ray tomography measurements were conducted to study and validate 3D morphology of crack growth inside the specimen. These two models along with chemical kinetics models obtained from corrosion tests can be combined to predict the total life of the component under given long term operating conditions. A Bayesian Hybrid Modeling (BHM) probabilistic framework was developed to quantify uncertainty in physics-based performance models and in crack initiation and crack propagation models, to calibrate the models and to validate the models with statistical confidence. Advanced coupled physics multi-phase flow models were developed to study condensation, boiling and phase change behavior of CO 2 under various sCO 2 turbomachinery and heat exchanger conditions. A novel experimental setup was also designed to observe phase change of CO 2 during its transition across liquid-gas-supercritical state lines.

14 SOLAR ENERGY↗
DOE OSTI2017

Initiating Event Rates at U.S. Nuclear Power Plants: 1988–2016

Analyzing initiating event rates is important because it indicates performance among plants and also provides inputs to several U.S. Nuclear Regulatory Commission (NRC) risk-informed regulatory activities. This report presents an analysis of initiating event frequencies at U.S. commercial nuclear power plants from calendar year 1988 through 2016 as reported in licensee event reports. Engineers with nuclear power plant experience reviewed each event report since the last update to this report for the presence of valid scrams or reactor trips at power. To be included in the study, an event had to meet all of the following criteria: (1) the event includes an unplanned reactor trip (not a scheduled reactor trip on the daily operations schedule), (2) the sequence of events started when reactor is critical and at or above the point of adding heat, (3) the event occurred at a U.S. commercial nuclear power plant (excluding Fort St. Vrain and LaCrosse), and (4) the event was reported by a licensee event report. Sixteen initiating event groupings are trended and displayed. For some of the categories as relevant initiators are plotted separately for boiling water reactors and pressurized water reactors. P values are given for the possible presence of a trend over the most recent 10 years. Statistically significant decreasing trends were identified for BWR loss of heat sink and general transients.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Irradiation Testing of a SiC/SiC Channel Box in the High Flux Isotope Reactor

This report provides a summary of the experiment design and the neutronics and structural analyses of a miniature SiC/SiC channel box that will be irradiated in the High Flux Isotope Reactor (HFIR). The goal of this experiment is to evaluate the lateral bowing in the channel box during irradiation under a radial neutron flux gradient. The concern is that lateral bowing could potentially block coolant flow or interfere with control blade movements in boiling water reactors (BWRs). The experiments described in this work will help validate thermomechanical models of channel box bowing. The experiment design allows for one channel box specimen to be irradiated in the HFIR while being directly cooled by the reactor coolant. Monte Carlo neutronics analyses are performed to determine the three-dimensional profile of the dose rate in the channel box. Dose is then converted to volumetric swelling, using literature data, and the resulting swelling profile is used as an input to structural finite element calculations that predict lateral bowing.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Design and analysis of oxidation tests to inform FeCrAl ATF severe accident models

Iron-chromium-aluminum (FeCrAl) alloys are being considered as advanced fuel cladding concepts with enhan ced accident tolerance. At high temperatures, FeCrAl alloys have slower oxidation kinetics and higher strength compared with zirconium-based alloys. FeCrAl could be used for fuel cladding and spacer or mixing vane grids in light water reactors and/or as ch annel box material in boiling water reactors (BWRs). There is a need to assess the potential gains afforded by the FeCrAl accident-tolerant-fuel (ATF) concept over the existing zirconium-based materials employed today. To accurately assess the response of FeCrAl aloys under severe accident conditions, a number of FeCrAl properties and characteristics are required. These include thermophysical properties as well as burst characteristics, oxidation kinetics, possible eutectic interactions, and failure temperatures. These properties can vary among different FeCrAl alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2018

U.S. Efforts in Support of Examinations at Fukushima Daiichi - 2018 Evaluations

Much is still not known about the end-state of core materials in each unit that was operating on March 11, 2011 at the Fukushima Daiichi Nuclear Power Station [Daiichi]. Clearly, the accident signatures for each of these units differed. Some of this uncertainty about debris end-state can be attributed to a lack of information related to cooling system operation and cooling water injection during the events. There is also uncertainty in our understanding of phenomena affecting: (a) in-vessel core damage progression during severe accidents in boiling water reactors (BWRs) [compared with Pressurized Water Reactors (PWRs)], and (b) accident progression after vessel failure (ex-vessel progression) for BWRs and PWRs. These uncertainties arise due to limited full scale prototypic data. Similar to what occurred after the accident at Three Mile Island Unit 2, these Daiichi units offer the international community a means to reduce such uncertainties by obtaining prototypic data from multiple full-scale BWR severe accidents.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Initiating Event Rates at U.S. Nuclear Power Plants 1988–2013

Analyzing initiating event rates is important because it indicates performance among plants and also provides inputs to several U.S. Nuclear Regulatory Commission (NRC) risk-informed regulatory activities. This report presents an analysis of initiating event frequencies at U.S. commercial nuclear power plants since each plant’s low-power license date. The evaluation is based on the operating experience from fiscal year 1988 through 2013 as reported in licensee event reports. Engineers with nuclear power plant experience staff reviewed each event report since the last update to this report for the presence of valid scrams or reactor trips at power. To be included in the study, an event had to meet all of the following criteria: includes an unplanned reactor trip (not a scheduled reactor trip on the daily operations schedule), sequence of events starts when reactor is critical and at or above the point of adding heat, occurs at a U.S. commercial nuclear power plant (excluding Fort St. Vrain and LaCrosse), and is reported by a licensee event report. This report displays occurrence rates (baseline frequencies) for the categories of initiating events that contribute to the NRC’s Industry Trends Program. Sixteen initiating event groupings are trended and displayed. Initiators are plotted separately for initiating events with different occurrence rates for boiling water reactors and pressurized water reactors. p-values are given for the possible presence of a trend over the most recent 10 years.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Physics based Reliability Models for Supercritical CO 2 Turbomachinery Components

CSP plants using sCO 2 power cycle can potentially achieve high thermal conversion efficiency at low capital cost due to compact turbo-machinery and other components. An sCO 2 expander and improved heat exchanger is expected to provide a major stepping stone for achieving CSP power at 0.06/kW-hr LCOE, energy conversion efficiency > 50 % ,and total power block cost< $1,200/kW installed. Hybrid gas bearings (HGB) and Dry gas seals (DGS) are two key components that are critical for robust and efficient operation of future sCO 2 power cycles. From a system perspective, HGBs can provide substantial benefits (~3 to 5%) to the modular CSP operation, while use of DGSs limits the leakage losses as well as the windage losses in the generator and can increase overall generator efficiency by ~ 8%. The life limiting mechanisms of HGBs and DGSs in high pressure, high temperature sCO 2 environment were not well understood and hence this program aims at developing physics-based performance prediction and life prediction models for these two components. The output of the performance models serves as input to the life models. The performance models were validated using experimental data and using data from literature. DGS and HGB models have shed light on some risks associated with its operation in sCO 2 expander and will guide more robust designs tailored towards sCO 2 cycle applications. A DGS was tested in 1MWe sCO 2 turbo expander to study its performance and validate the performance models. To understand the effect of high pressures, high temperatures and sCO 2 environment on crack initiation, crack propagation and LCF/HCF life of these components, a novel experimental setup was developed. The LCF test data was used to calibrate and validate the life prediction models. The effect of sCO 2 corrosion on the life of Ni-based super-alloys is quantified using long duration experiments. Advanced microstructure and spectroscopic analyses conducted shed light on some key differences between IN718 and IN617 in terms of oxidation morphology, chemical species diffusion and trapping, the formation of protective corrosion resistant layers and changes in surface properties. A novel crack initiation model was developed which takes into account strain distribution, temperature, hold times and activation energy at the gas-metal interface and predicts the number of cycles to initiate the crack in the Inconel alloy. An advanced 3D fracture mechanics-based crack propagation model was also developed which takes into account the transient loading mission of the component and predict how fast the crack will grow and what will be its final shape and size. High energy X-Ray tomography measurements were conducted to study and validate 3D morphology of crack growth inside the specimen. These two models along with chemical kinetics models obtained from corrosion tests can be combined to predict the total life of the component under given long term operating conditions. A Bayesian Hybrid Modeling (BHM) probabilistic framework was developed to quantify uncertainty in physics-based performance models and in crack initiation and crack propagation models, to calibrate the models and to validate the models with statistical confidence. Advanced coupled physics multi-phase flow models were developed to study condensation, boiling and phase change behavior of CO 2 under various sCO 2 turbomachinery and heat exchanger conditions. A novel experimental setup was also designed to observe phase change of CO 2 during its transition across liquid-gas-supercritical state lines.

14 SOLAR ENERGY↗
DOE OSTI2018

Validation of k eff Calculations for Extended BWR Burnup Credit

The US Nuclear Regulatory Commission and consensus standards recommend validation of the numerical methods used in criticality safety analyses. This validation requires the comparison of computational results with measurements of physical systems which are neutronically similar to those used in the safety analysis being performed. To this end, this document examines the methods available to generate sensitivity data to help identify systems similar to spent boiling- water reactor (BWR) fuel in a flooded spent nuclear fuel transportation or storage cask. A large number of critical benchmark experiments are surveyed using sensitivity/uncertainty (S/U) techniques to assess their applicability to BWR burnup credit (BUC) beyond the burnup of peak reactivity. Multiple burnups of BWR assemblies are considered herein, as well as the actinide-only (AO) and actinide-and-major-fission-product (AFP) isotope sets. Sample validations are completed for representative application models to demonstrate that appropriate validation is possible and to indicate the bias and bias uncertainty values expected for related applications.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Test Plan for the Horizontal Dry Cask Simulator

The thermal performance of commercial spent nuclear fuel dry storage casks is evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission(NRC). Canistered dry storage cask systems rely on ventilation between the inner canister and the overpack to convect heat away from the canister to the surrounding environment for both horizontal and vertical configurations. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. While several testing programs have been previously conducted, these earlier validation studies did not integrate all the physics or components important in a modern, horizontal dry cask system. The purpose of the investigation described in this test plan is to produce data sets that can be used to benchmark the codes and best practices presently used to determine cladding temperatures and induced cooling air flows in modern horizontal dry storage systems. The horizontal dry cask simulator(HDCS) has been designed to generate this benchmark data and add to the existing knowledgebase. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure(MAWP)rating of 2,400 kPa at400 °C. An existing electrically heated but otherwise prototypic boiling water reactor(BWR), Incoloy-clad test assembly will be deployed inside of a representative storage basket and canister. An insulated sheet metal enclosure will be used to mimic the thermal properties of the concrete vault enclosure used in a modern horizontal storage system. Radial and axial temperature profiles along with induced cooling air flow will be measured for a wide range of decay powers and representative(and higher)cask pressures using various backfills of helium, argon, or air. The single assembly geometry with well-controlled boundary conditions simplifies computational requirements while preserving relevant physics. The proposed test apparatus integrates all the underlying thermal-hydraulics important to defining the performance of a modern horizontal storage system. These include combined-mode heat transfer from the electrically-heated assembly to the canister walls and the primarily natural-convective heat transfer from the canister to the cooling air flow passing through the horizontal vault enclosure. The objective of the HDCS is not to reproduce the performance of a commercial dry storage system for any given set of operational parameters. Rather ,the objective is to capture the dominant physics in a well-characterized test apparatus. The close coupling between the thermal response of the canister system and the resulting induced cooling air flow rate is of particular importance. While incorporating the best available information based on thermal-hydraulic scaling arguments as well as previous vertical testing, this test plan is subject to changes due to improved understanding or from as built deviations to designs. As-built conditions and actual procedures will be documented in the final test report.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Modeling a SERTTA-type vessel using RELAP-7

A transient in a SERTTA vessel geometry was modeled using RELAP-7. This transient included a significant amount of boiling in the fuel region, and a resulting transport of vapor into a large expansion tank. A preliminary gas mixture model and transport PDE was used for the presence of non-condensable gases. Results demonstrated that the solution algorithm and/or mixture model needs modification to avoid evaluation of unphysical mixture fluid properties.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Estimation of Dose from Potential Radionuclide Emissions from the Sample Preparation Laboratory at the Materials and Fuels Complex

The Sample Preparation Laboratory (SPL) at the Materials and Fuels Complex will perform sample preparation, examination, and mechanical-property testing of highly irradiated materials. The purpose of this report is to document calculation of the estimated dose at public receptor locations from potential radionuclide releases to the atmosphere from SPL. Atmospheric dispersion and dose calculations were performed with the computer model CAP88-PC Version 4 in accordance with the requirements of Code of Federal Regulations (CFR), Title 40, “Protection of the Environment,” Part 61, “National Emission Standards for Hazardous Air Pollutants (NESHAPs),” Subpart H, “National Emission Standards for Emissions of Radionuclides Other than Radon from Department of Energy Facilities” (40 CFR 61, Subpart H 2010). Emission potentials were calculated based on the methodology in 40 CFR Part 61, Appendix D and additional guidance/approval by EPA Region 10 [see letter from Donald Dossett (EPA Region 10) to Tim Safford (DOE-ID), Oct 19, 2017 (CCN 241475)] for solid materials that undergo heating. Estimates of total effective dose are based on low-level chronic exposure. The potential maximum dose to a member of the public was calculated to be 0.02 mrem/year based on an estimated 16 g of the 30 kg source term (0.053%) being exposed to very high temperatures associated with electrical-discharge machining (EDM) and the remaining material heated to 100°C. Calculations show the maximum dose could be as high as 0.05 mrem/year if the remaining material were heated to 1400°C (i.e., the melting point of stainless steel). The potential maximum dose to a public receptor from emissions at SPL is not expected to exceed 0.1 mrem/year (1% of the standard) so long as the material impacted by EDM does not exceed 56 g (3.5 times the expected mass of 16 g and less than 0.2% of the total source term), and the remaining material is not heated to temperatures greater than 1400°C. It should be noted that credit for HEPA filtration does not impact the dose results because the dose is dominated by radionuclides that have exceeded their boiling point; thus, no credit for HEPA filtration may be applied to the emission factor.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2019

Evaluation of the current state of knowledge for thermolysis of organics within SRS waste forming Volatile Organic Compounds (VOCs)

The 2017 report Flammable Gas Generation Mechanisms for High Level Liquid Waste Facilities, X-ESRG- 00062 Rev. 1, assessing mechanisms for flammable gas generation acknowledged the presence of flammable organics, concluding that available data indicated their concentrations pose a minor contribution (e.g., <5%) to the composite lower flammable limit. Savannah River Remediation (SRR) chartered the Savannah River National Laboratory (SRNL) to perform a more detailed assessment of available and emerging data to assess the current state of knowledge for formation of Volatile Organic Compounds (VOCs) from thermolysis of organics in Savannah River Site (SRS) caustic tank waste. Based on the assessment as detailed herein, SRNL concludes that sufficient evidence exists to demonstrate formation of methane, in addition to hydrogen, does occur during thermolysis of SRS high level waste containing organics under select conditions. This assessment cannot reliably quantify the amount of methane that forms. Nevertheless, the potential rates are sufficient – methane generation rates of approximately 30% and 100% those of hydrogen generation rates for Tank 38 radioactive waste and a high boiling point (HBP) simulant waste, respectively – to advise that ongoing program efforts be enhanced to include additional analysis for off-gas generation for concentration and screening measurements for other possible flammable species.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Investigation of thermolytic hydrogen generation rate of tank farm simulated and actual waste

To support resolution of Potential Inadequacies in the Safety Analysis for the Savannah River Site (SRS) Tank Farm, Savannah River National Laboratory conducted research to determine the thermolytic hydrogen generation rate (HGR) with simulated and actual waste. Gas chromatography methods were developed and used with air-purged flow systems to quantify hydrogen generation from heated simulated and actual waste at rates applicable to the Tank Farm Documented Safety Analysis (DSA). Initial simulant tests with a simple salt solution plus sodium glycolate demonstrated the behavior of the test apparatus by replicating known HGR kinetics. Additional simulant tests with the simple salt solution excluding organics apart from contaminants provided measurement of the detection and quantification limits for the apparatus with respect to hydrogen generation. Testing included a measurement of HGR on actual SRS tank waste from Tank 38. A final series of measurements examined HGR for a simulant with the most common SRS Tank Farm organics at temperatures up to 140 °C. The following conclusions result from this testing. HGR was measured for an actual-waste Tank 38 sample. The HGR measurements at 80, 90, 100, and 110 °C were 1.7×10 -7 , 4.7×10 -7 , 5.9×10 -7 , and 2.6×10 -6 ft 3 hr -1 gal -1 , respectively. The Limit of Detection and Limit of Quantification for the HGR measurements were 2.6×10-8 ft3 hr-1 gal-1 and 5.1×10 -8 ft 3 hr -1 gal -1 , respectively. HGR was measured in a high boiling point (HBP) simulant with major known SRS Tank Farm organic compounds. Testing used fresh organics and did not simulate the long-term exposure of these organics already experienced in the waste. Hence, testing is expected to provide high bias to the hydrogen generation rates. The final HGR measurements at 75, 100, and 140 °C were 6.2×10 -7 , 3.0×10 -6 , and 8.1×10 -5 ft 3 hr -1 gal -1 , respectively. Decreases in the HGR were noted at isothermal conditions with steady gas tracer concentration, which clouds development of a temperature dependent correlation for generation rate. While the simulant contains high concentrations of major SRS organics tested at high temperatures found in the 3H Evaporator, this limited data set may not fully bound values under all possible operating conditions. The HGR measurement apparatus was successfully employed to replicate hydrogen production rate experiments. Simulant without deliberately added organic carbon was found to contain approximately 200 mg/L of organic carbon from tramp organics in the reagents used to prepare the solution. The tests using such simulant generated hydrogen at rates consistent with, but somewhat lower than, those predicted at equivalent carbon and aluminum loadings by methods used at the Hanford Site. Prior to identifying the organic contaminants, potential sources of hydrogen in simulants without deliberately added organic carbon were examined. Testing determined that this hydrogen is not due to caustic reaction of organic materials of construction within the reaction vessel, electrolysis due to current leak, anode/cathode reactions, corrosion, or the method of heating employed in the experiment. During the HBP and Tank 38 testing, a peak was identified at a retention time greater than the retention time of the inert tracer of krypton. This peak has been identified as methane, and the data have been reexamined to obtain methane concentrations. For the HBP testing, methane concentration was constant at the highest temperature studied (140 °C) and higher than hydrogen. At the lower temperatures (100 and 75 °C), methane and hydrogen tracked closely, with methane concentration generally lower than hydrogen. For Tank 38, methane concentration was approximately 1/3 of the hydrogen concentration during testing (100 and 110 °C). At lower temperatures for Tank 38 testing (90, 80, and 27 °C), methane fell below the quantification range.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Development of Critical Experiments to Provide Validation Data for Multiphysics Coupling Methods (Final Report)

Computational simulation has become one of the most important tools in the modern toolbox for nuclear reactor design. Reactor simulations, like the reactors themselves, involve multiple physical processes including neutronics, fluid flow and heat transfer (thermal-hydraulics), thermo-mechanics, fuel behavior, chemistry, and balance of plant. Each process can affect the behavior of the others, and advanced computational tools have been developed that couple separate physics models together to simulate the feedback effects. The objective of this work is to generate experimental data to support the validation of multi-physics reactor codes. It specifically targets the SHARP toolset from the NEAMS Reactor Product Line, but it is intended to be useful in any set of coded that couples thermal-hydraulics with neutronics simulations. The project takes advantage of the inherent flexibility of the RPI Reactor Critical Facility (RCF), a low power open-pool research reactor. The standard configuration of the RCF is used to explore the effects of moderator temperature on neutronic behavior and reactivity. Because the low power of the reactor generates no appreciable temperature change in the moderator, the moderator temperature is artificially changed with two 18 kW electric immersion heaters. The temperature feedback effects can be observed by recording the change in the reactor power as the reactor goes from a supercritical state, to critical, to a subcritical state. To expand the temperature range and temperature rate of change that can be investigated, an experimental apparatus consisting of a heated water loop was designed and installed in the RCF. In this setup, water from a heated reservoir tank is pumped through a pipe placed through the center of the reactor core. The water in the reservoir can be heated to around 70 °C and pumped through the system. The pump has a variable flow rate and electronic control, so a number of scenarios can be developed. A slug of water can be pushed through the system to generate a reasonably fast change in reactivity. The pump speed can be increased or decreased over the course of a measurement, or even oscillated at different periods, to observe the response of the reactor. To complement the experimental data generated by the RCF, a set of computational models have been developed. They are intended to illustrate the usefulness of the experimental data generated in validating coupled multi-physics simulations. A model previously constructed in MCNP has been used in conjunction with safety analysis at the RCF for many years and was modified to support the design of the heated water loop experiment. This model was translated into Serpent 2, a similar Monte Carlo code, which provided the added advantage of being able to produce collapsed cross-section data files for use in the deterministic neutron transport codes that are part of the SHARP toolset. A complete reconstruction of the RCF model, both with and without the heated fluid loop experiment, were necessary for use with the SHARP codes, PROTEUS (neutronics) and Nek5000 (thermal hydraulics), as these require an unstructured mesh format to operate. The computational requirements for fully coupled three dimensional simulations are significant, and challenges to full implementation and comparison with RCF experiments remain. A wide variety of experiments have been conducted for eventual benchmarking and validation of multi-physics reactor codes. A complete database of detailed time- and temperature-dependent measurements have been generated and stored. Several examples of the experimental results are presented here, with refinement and analysis of experiments ongoing. While these experiments still only address a finite range of operating conditions, in particular being limited to water in the sub-boiling regime, they represent a unique (to date) look at the feedback between temperature and neutronic behavior suited for validation of the coupling routines in multi-physics codes. It is hoped that these will be the foundation of future benchmarks that will provide confidence in the tools that will be used to develop the next generation of nuclear reactor systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗