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DOE OSTI2013

Polyvalent fuel treatment facility (TCP): shearing and dissolution of used fuel at La Hague facility

Although many used nuclear fuel types have already been recycled, recycling plants are generally optimized for Light Water Reactor (LWR) UO{sub x} fuel. Benefits of used fuel recycling are consequently restricted to those fuels, with only limited capacity for the others like LWR MOX, Fast Reactor (FR) MOX or Research and Test Reactor (RTR) fuel. In order to recycle diverse fuel types, an innovative and polyvalent shearing and dissolving cell is planned to be put in operation in about 10 years at AREVA's La Hague recycling plant. This installation, called TCP (French acronym for polyvalent fuel treatment) will benefit from AREVA's industrial feedback, while taking part in the next steps towards a fast reactor fuel cycle development using innovative treatment solutions. Feasibility studies and R/Development trials on dissolution and shearing are currently ongoing. This new installation will allow AREVA to propose new services to its customers, in particular in term of MOX fuel, Research Test Reactors fuel and Fast Reactor fuel treatment. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

National Reactor Innovation Center (NRIC) Grid Connection Roadmap

In an effort to define the roadmap for establishing the necessary electrical infrastructure at the Idaho National Laboratory to fully support connection of NRIC reactors, this roadmap serves to identify the current forecasted reactors planned for installation at INL and provide a consolidated list of the identified infrastructure improvements that will be necessary to support testing of NRIC reactors. Additionally, the roadmap will leverage this information to establish an integrated timeline to support detailed planning and execution of system improvements. To drive this planning, requirements, as derived from authoritative sources, will be identified. A clear definition of impacted stakeholders will be established, as well as a list of risks that will need to be managed as the program moves forward.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

National Reactor Innovation Center (NRIC) Grid Connection Roadmap (Rev. 1)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

ECAR-2322 As-Run Thermal Analysis Of The AGC-2 Experiment

The second Advanced Graphite Creep (AGC-2) experiment was designed to irradiate various types of graphite specimens at a temperature of 600°C. The specimens were irradiated in an instrumented leadout capsule experiment in the south flux trap of the ATR during cycles 149A, 1498, 1508, 151A, and 1518. Temperature was monitored using twelve thermocouples located at various elevations in the reactor core, and a helium-argon gas mixture was used for gas gap temperature control of the specimens. The purpose of this analysis is to calculate specimen temperature using measured data on reactor power and helium-argon gas flows, and as-run calculations of heating rates and displacement per atom (DPA) in graphite. The accuracy of the model is assessed by comparing measured and calculated thermocouple temperatures. Uncertainty in gas gaps may preclude an accurate temperature calculation. In these cases, adjustments are made to the thermal model in order to reconcile the measured and calculated thermocouple temperature and to ensure the accuracy of the calculated specimen temperature.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2012

Section III, Division 5 - Development and Future Directions

This paper provides commentary on a new division under Section III of the ASME Boiler and Pressure Vessel (BPV) Code. This new Division 5 has an issuance date of November 1, 2011 and is part of the 2011 Addenda to the 2010 Edition of the BPV Code. The new Division covers the rules for the design, fabrication, inspection and testing of components for high temperature nuclear reactors. Information is provided on the scope and need for Division 5, the structure of Division 5, where the rules originated, the various changes made in finalizing Division 5, and the future near-term and long-term expectations for Division 5 development. Portions of this paper were based on Chapter 17 of the Companion Guide to the ASME Boiler & Pressure Vessel Code, Fourth Edition, © ASME, 2012, Reference.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

Long-term performance of a UASB reactor treating acid mine drainage: effects of sulfate loading rate, hydraulic retention time, and COD/SO42- ratio

Acid mine drainage (AMD) is among the most serious threats to water and the typical alkali-based treatment costs are high. This study's main objective was the establishment of a highly efficient biological process using an upflow anaerobic sludge blanket (UASB) reactor to treat AMD based on a shorter hydraulic retention time (HRT) and lower organic matter input. The process was evaluated for a long-term operation (739days) in terms of the influence of HRT (14-24h), metal addition, sulfate loading rate (0.5-2.6gSO(4)(2-)l(-1)d(-1)), and the COD/SO42- ratio (0.67-1.0) using ethanol as the only electron donor at a pH of 4.0. Neutral effluent pH was achieved throughout the time apart from operational modifications. The reduction in HRT from 24 to 16h and an increase in the sulfate loading rate (SLR) up to 2.25gSO(4)(2-)l(-1)d(-1) improved the sulfate removal to (92.1 +/- 1.8)% with 80% chemical oxygen demand (COD) removal. However, the sulfate reduction was less than 80% when the HRT and SLR was changed to 14h and 2.6gSO(4)(2-)l(-1)d(-1), respectively. The oxidation of organic matter by sulfate reduction was greater than 50% regardless of the conditions imposed but the use of ethanol to treat AMD was more efficient when either the HRT was 16h (1.5gSO(4)(2-)l(-1)d(-1)) in the presence of Fe, Zn, and Cu or the HRT was 14h (2.6gSO(4)(2-)l(-1)d(-1)) but the COD/SO42- ratio was reduced to 0.67. The fully optimized conditions of the UASB reactor were set at an HRT of 16h, SLR of 1.5gSO(4)(2-)l(-1)d(-1), and a COD/SO42- ratio of 1.0.

Acid mine drainage↗

DOE OSTI2021

Preliminary Procedures and Acceptance Criteria for in-situ Structural Materials Surveillance for MSR

This report describes the initial development of the technical basis for a materials surveillance technology that, when fully developed and validated, can be used by stakeholders to develop and implement a materials surveillance program to manage materials degradation during reactor operations. The procedures described in this report aim to bound the detrimental effects of the reactor environment on the creep-fatigue life of the reactor components through materials surveillance. Specifically, the report describes procedures for selecting critical locations within the reactor components for surrogate materials surveillance and the sizing and placements of passively-actuated materials surveillance test articles under development by the US Department of Energy, Advanced Reactor Technologies Program. These test articles apply cyclic thermomechanical load to a surrogate sample of the component structural material passively through thermal expansion mismatch within the articles, without requiring any penetrations in the reactor coolant boundary. The procedures can be used to size materials surveillance test articles so that they will fail before the corresponding structural component. These articles would be exposed to the component operating environment, located to experience a bounding radiation fluence and representative coolant exposure. They can be monitored during reactor operations to ensure that they have not failed under the combination of creep-fatigue damage and environmental degradation. The report describes the key technical decisions and the corresponding rational for using these procedures to provide advanced warning of any impending material failure. An appendix describes the technical basis and the procedure for sizing a family of passively-actuated test articles. Finally, the report describes future work needed to complete the technical basis underlying this materials surveillance technology.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

On the Interpretation and Correlation of High-Temperature Ignition Delays in Reactors with Varying Thermodynamic Conditions

Ignition delay time (IDT) is a useful global metric for fuel performance screening and a major target for kinetic modeling. Measurements of IDT are conceptually straightforward, however interpretation could be complicated, especially for systems with changing temperature and pressure. Some experimental conditions in the high repetition rate miniature shock tube (HRRST) exhibit complex temperature and pressure state histories. To better interpret and correlate IDTs, especially those obtained in reactors with varying thermodynamic conditions, an inverse Livengood-Wu (L-W) integral technique is applied to de-convolve the constant condition IDTs from measured IDTs in the HRRST using information on the varying state history. In this paper, the approximate problem is demonstrated using only the measured pressure history leading up to ignition, where the temperature history is estimated based on an isentropic assumption. The IDTs of several fuels were first measured in the HRRST including isooctane and acetone to represent those fuels without strong low temperature chemistry effects. Based on the measured pressure and approximated temperature history, measurements of IDTs in the HRRST compare very favorably with those measured using more conventional techniques, including conventionally sized shock tubes, via the inverse L-W correlation with relaxed Newton iteration and Genetic Algorithm (GA). Furthermore, this study demonstrates the feasibility of using a high throughput ignition testing facility, like the HRRST to extract the constant state IDT measured in regular shock tubes, assisted by the inverse L-W correlation. It is expected with an independent temperature measurement available in the future, IDTs in a broader range of thermodynamic conditions and effects of heat loss could be better resolved.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2019

Influence of heat treatment on the size of sodium lignosulfonate particles in water—ethanol media

Dynamic light scattering and size exclusion chromatography were used to investigate the size of sodium lignosulfonate particles in water—ethanol media and the changes occurring at temperatures of 115–440 °C (13 MPa). It was found that the characteristic size (diameter) of particles of the starting sodium lignosulfonate in solutions with a concentration of 1.5 and 13.3 g L{sup −1} was 2.2−5.2 nm. Treatment in a flow reactor at temperature of 235–325 °C led to a disintegration of larger particles with an increase in the fraction of particles with size of 2.2 nm and a simultaneous formation of particles with a size of ~100 nm. Increasing the treatment temperature (>415 °C) led to a complete disappearance of particles with size of 2.2–5.2 nm and the formation of particles 100–130 nm in diameter, as well as the precipitation of large soot-like particles. IR spectroscopy showed that increasing the treatment temperature above 275 °C led to a decrease in the specific content of oxygen-containing and sulfoxyl groups in the sodium lignosulfonate structure, with the exception of carboxyl groups, whose fraction remained almost unchanged, as well as CH{sub 3} groups (methoxy and aliphatic), whose content increased slightly. The temperature range of 235–275 °C was optimal for hydrolysis of sodium lignosulfonate and preparation of aromatic monomers.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

NASA NTRS1997

Effects of Bioreactor Retention Time on Aerobic Microbial Decomposition of CELSS Crop Residues

The focus of resource recovery research at the KSC-CELSS Breadboard Project has been the evaluation of microbiologically mediated biodegradation of crop residues by manipulation of bioreactor process and environmental variables. We will present results from over 3 years of studies that used laboratory- and breadboard-scale (8 and 120 L working volumes, respectively) aerobic, fed-batch, continuous stirred tank reactors (CSTR) for recovery of carbon and minerals from breadboard grown wheat and white potato residues. The paper will focus on the effects of a key process variable, bioreactor retention time, on response variables indicative of bioreactor performance. The goal is to determine the shortest retention time that is feasible for processing CELSS crop residues, thereby reducing bioreactor volume and weight requirements. Pushing the lower limits of bioreactor retention times will provide useful data for engineers who need to compare biological and physicochemical components. Bioreactor retention times were manipulated to range between 0.25 and 48 days. Results indicate that increases in retention time lead to a 4-fold increase in crop residue biodegradation, as measured by both dry weight losses and CO2 production. A similar overall trend was also observed for crop residue fiber (cellulose and hemicellulose), with a noticeable jump in cellulose degradation between the 5.3 day and 10.7 day retention times. Water-soluble organic compounds (measured as soluble TOC) were appreciably reduced by more than 4-fold at all retention times tested. Results from a study of even shorter retention times (down to 0.25 days), in progress, will also be presented.

Strayer, R. F.↗

DOE OSTI2012

Calculation of the Phenix end-of-life test in natural circulation with the CATHARE code

The Inst. of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. As such, IRSN is in charge of safety assessment of operating and under construction reactors, as well as future projects. In this framework, one current objective of IRSN is to evaluate the ability and accuracy of numerical tools to foresee consequences of accidents. One of the advantages pointed up for fast reactors cooled by heavy liquid metal is the possibility of decay heat removal based on natural convection. The promotion of this passive cooling mode in future safety demonstrations will involve the use of adapted and validated numerical codes. After the final shutdown of the Phenix sodium cooled fast reactor in 2009, a set of tests covering different areas was conducted for code validation, including a natural circulation test in the primary circuit. Experimental data were issued by CEA to organize a benchmark exercise in the frame of an IAEA Coordinated Research Project (CRP), with the objective to assess the system-codes capability in simulating the thermal-hydraulics behavior of sodium cooled fast reactors in such accidental conditions. IRSN participated to this benchmark with the CATHARE code. This code, co-developed by CEA, EDF, AREVA and IRSN and widely used for PWR safety studies, was recently extended for sodium applications. This paper presents the CATHARE modeling of the Phenix primary circuit and the results obtained. A relatively good agreement was found with available measurements considering the experimental uncertainties. This work stressed the local aspects of phenomena occurring during the natural convection establishment and the limits of a 0D/1D approach. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

Initial development for the introduction of provisions for irradiation effects in Division 5 rules

Heat pipe microreactor concept involves the use of a solid core block to encapsulate the nuclear fuel and the passive extraction of heat from the core block to the heat exchanger block by advanced heat pipes. The core block is a key component of the heat pipe microreactor design. It integrates the functions of reactor vessel, core structural components and fuel cladding of traditional Gen IV advanced reactor designs into one single structure. The effects of irradiation on the structural failure modes are outside the scope of the rules of the ASME Boiler and Pressure Vessel Code, Section III, Division 5, Subsection HB, Subpart B for metallic Class A components. Methods to account for the irradiation effect were investigated in order to develop design methods that can be applied to the core block design in the heat pipe microreactor. This report presents an initial development of an approach to account for the irradiation effects in the design rules.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Photoreduction of mercury metal (Hg) using catalyst of oxalic acid from cellulose of rice husks (Oryza sativa L.)

Photoreduction of mercury metal using catalyst of oxalic acid from cellulose of rice husks (Oryza sativa L.) is one of methods to reduce toxicity properties of the mercury metal in the society. The purpose of this research is to enhance photoreduction of mercury metal using catalyst of oxalic acid from cellulose of rice husks (Oryza sativa L.) at various concentrations. Photoreduction process is carried out in a closed reactor equipped with UV light and magnetic stirrer. Analysis of the influence of oxalic acid is determined by adding 25 mL of Hg (II) 5 ppm without oxalic acid, 25 mL of Hg (II) 5 ppm + 25 mL of oxalic acid 3 ppm, 25 mL of Hg (II) 5 ppm + 25 mL of oxalic acid 6 ppm, 25 mL of Hg (II) 5 ppm + 25 mL of oxalic acid 9 ppm, 25 mL of Hg (II) 5 ppm + 25 mL of oxalic acid 12 ppm and 25 mL of Hg (II) 5 ppm + 25 mL of oxalic acid 15 ppm. All treatments are followed by centrifugation for 15 minutes, then the concentration of Hg residual in the solution is measured by mercury analyzer. The research results showed that addition of oxalic acid concentration from the cellulose of rice husks (Oryza sativa L.) can enhance photoreduction of mercury metal. Optimum concentration reduction of mercury metal with addition of oxalic acid is obtained as many as 9-12 ppm. It can reduce the concentration of mercury metal (II) by 68.8% to 88.6%.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

DOE OSTI2015

Non-Contact Measurement of Thermal Diffusivity in Ion-Implanted Nuclear Materials

Knowledge of mechanical and physical property evolution due to irradiation damage is essential for the development of future fission and fusion reactors. Ion-irradiation provides an excellent proxy for studying irradiation damage, allowing high damage doses without sample activation. Limited ion-penetration-depth means that only few-micron-thick damaged layers are produced. Substantial effort has been devoted to probing the mechanical properties of these thin implanted layers. Yet, whilst key to reactor design, their thermal transport properties remain largely unexplored due to a lack of suitable measurement techniques. Here we demonstrate non-contact thermal diffusivity measurements in ion-implanted tungsten for nuclear fusion armour. Alloying with transmutation elements and the interaction of retained gas with implantation-induced defects both lead to dramatic reductions in thermal diffusivity. These changes are well captured by our modelling approaches. Our observations have important implications for the design of future fusion power plants.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Latest Results from DIII-D and Their Implications for Future Devices

The DIII-D Tokamak has become one of the major sources of physics and technology data for the design of future large machines such as ITER and CIT. In large part this is because of its extremely flexible design and the ability to run tests or add new diagnostic devices with a minimum of down-time and expense. Fundamental plasma physics studies are still the major focus of our experimental program, but increasingly, we are looking at new ways to answer the complex fusion technology questions emerging from the design studies for the next-generation devices. Recent results have demonstrated world record beta plasmas, long quiescent H-mode operation, partial noninductive current drive using both neutral beam and rf power injection, single- and double-null divertor operation, and divertor particle- and heat-load management. The recently installed Advanced Divertor hardware will allow critical transport experiments to be run with independent control of the density. It will also permit the concept of dc helicity injection current drive to be tested. This paper summarizes our most important recent findings and also outlines a few of the many interesting problems that are now under study to address fusion technology questions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2015

Report on FY15 Two-Bar Thermal Ratcheting Test Results

Alloy 617 is a reference structural material for very high temperature components of advanced-gas cooled reactors with outlet temperatures in the range of 900-950°C. In order for designers to be able to use Alloy 617 for these high temperature components, Alloy 617 has to be approved for use in Section III (the nuclear section) of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. A plan has been developed to submit a draft code for Alloy 617 to ASME Section III by 2015. However, the current rules in Subsection NH* for the evaluation of strain limits and creep-fatigue damage using simplified methods based on elastic analysis have been deemed inappropriate for Alloy 617 at temperatures above 1200°F(650°C). The rationale for this exclusion is that at higher temperatures it is not feasible to decouple plasticity and creep deformation, which is the basis for the current simplified rules. This temperature, 1200°F(650°C), is well below the temperature range of interest for this material in High Temperature Gas Cooled Reactor (HTGR) applications. The only current alternative is, thus, a full inelastic analysis which requires sophisticated material models which have been formulated but not yet verified. To address this issue, proposed code rules have been developed which are based on the use of elastic-perfectly plastic (EPP) analysis methods and which are expected to be applicable to very high temperatures. These newly proposed rules also address a long-term objective to provide an option for more simple, comprehensive and easily applied rules than the current so called “simplified” rules.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Turbulence and E × B flow correlations across the L-H transition in DIII-D deuterium and hydrogen plasmas

The isotope dependence of the low- to high confinement-mode (L- to H-mode) transition power threshold P LH presents significant challenges for the initial (non-nuclear) hydrogen operations phase of the international thermonuclear experimental reactor (ITER). Here, we examine the isotope dependence of turbulence and E × B flow correlation properties in the L-mode edge plasma, leading up to the L-H transition. At marginal auxiliary power (near P LH ), turbulence is initially suppressed periodically during limit cycle oscillations (LCO) that precede the transition to sustained H-mode confinement. Here, we present evidence that the long-range (toroidal) correlation of the E × B edge plasma flow across the LCO phase is much weaker in hydrogen than in deuterium in the DIII-D tokamak, congruent with the higher threshold power P LH required to access LCO and H-mode in hydrogen. Concomitantly, the time required to initially quench edge turbulence via localized edge E × B flow shear is significantly longer in hydrogen (1–1.5 ms) than in deuterium (approx. 100 µs). No toroidal long-range correlation of the turbulence amplitude is observed, in agreement with expectations based on the relatively short poloidal turbulence correlation length. Radial edge turbulence and flow correlation lengths are longer in deuterium than in hydrogen plasmas as one would expect from ‘naïve’ gyro-Bohm isotope transport scaling, despite the substantially higher thermal flux across the last closed flux surface in hydrogen before the L-H transition. This article is part of a discussion meeting issue 'H-mode transition and pedestal studies in fusion plasmas'.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2016

HTTR-GT/H2 test plant - conceptual design study on new intermediate heat exchanger - 18747

Japan Atomic Energy Agency is developing a nuclear cogeneration demonstration test facility named HTTR-GT/H2 that is connected to the existing test reactor HTTR. In the facility, an intermediate heat exchanger (IHX) at maximum service temperature of 900 deg. C is to be newly constructed to supply the high temperature helium flows to both the helium gas turbine system and the hydrogen production system. The conceptual design study on the IHX is conducted considering the future application to the Gas Turbine High Temperature Reactor 300 Cogeneration (GTHTR300C). Several new concepts are proposed and incorporated in the demonstrating IHX design in order to (1) ease integration of the IHX with the gas turbine system to simplify the cogeneration plant layout; (2) allow conventional low-cost steel for construction of the IHX pressure vessel and nozzles; (3) flexibility to accommodate thermal expansion of the bundle and internal structures.; (4) maintaining the structural integrity at elevated service temperature range. This paper describes the conceptual design study results for the IHX and technical issues to be solved in the future design phases. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗