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DOE OSTI2019

TREAT Fuel Motion Summary Report. SETH A-E Experiments

In support of the Accident Tolerant Fuels program, experiment devices are being developed to enable transient testing of pressurized water reactor (PWR) type fuel specimens. In fiscal year 2019 several fueled tests were conducted in the Transient Reactor Test Facility (TREAT) using the Separate Effects Test Holder (SETH). Five different SETH capsules (SETH A-E) were assembled, each holding a 10-pellet rodlet with 4.9% enriched, commercially-produced fresh UO2 pellets in zirconium alloy cladding with pellet dimensions having typical PWR radial dimensions and a total pellet stack length measuring 10.16 cm. The rodlet’s length allows the specimen to be axially positioned in the center of the core with minimal flux variation over the height of the rodlet. The fuel motion monitoring system (FMMS) is located on the north side of the TREAT reactor. Having direct line-of-sight of the core enables the FMMS to collect spatially-resolved data for fast neutrons generated due to fission occurring in experiment specimens. The FMMS contains 360 channels that make up a 10 (horizontal) by 36 (vertical) array of viewing slots; for the SETH A-E test series 96 proton-recoil scintillator (PRS) detectors were installed in the system. The fast neutrons are detected in the PRS detectors from light produced by the recoil of protons interacting with ZnS(Ag) grains in each detector's scintillator button. This report focuses on the FMMS data captured in the SETH A-E experiments, with multiple transients varying in reactivity addition and length. The first three experiments had a total reactor energy of 101 MJ each and were critical in understanding the FMMS field of view, lead shielding, and other TREAT data acquisition systems. SETH-B2 and C had a reactivity addition of almost double the previous transients and served as a demonstration of the detector array behavior under increased reactor energy. The final two SETH experiments (D and E) were aimed at reaching cladding melting temperatures with a reactor energy over 500 MJ for each test. In preparation for the higher energy transients, the FMMS lead shielding was reduced to 0.5 inches from the 6 inches used in SETH A-C. This was expected to saturate the detectors at the peak of the transient in exchange for higher sensitivity at lower energies after the peak, when the motion was expected. SETH-D was clipped at approximately 3 seconds, which was too early for the FMMS to capture any rodlet motion. SETH-E had the same reactivity addition as SETH-D but the transient time was extended to 18 seconds. The FMMS captured the downward motion of the rodlet in SETH-E as the count rate significantly changed for the detectors viewing the specimen. Snapshots taken throughout the transient helped identify the approximate time at which the motion initiated. Further analysis of the individual detector data captured on a per-millisecond time basis resulted in a determination that the approximate time for the onset of the downward motion was 2.84 ± 0.02 seconds. Neutron radiographs confirmed that downward motion occurred for SETH-D and SETH-E. Motion in SETH-D could also have been captured by the FMMS if the transient time had been longer for SETH-D. Ultimately, these final two tests served to demonstrate the FMMS's ability to track fuel motion during transient experiments at TREAT.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Preliminary results of calculations for heavy-water nuclear-power-plant reactors employing {sup 235}U, {sup 233}U, and {sup 232}Th as a fuel and meeting requirements of a nonproliferation of nuclear weapons

A physical design is developed for a gas-cooled heavy-water nuclear reactor intended for a project of a nuclear power plant. As a fuel, the reactor would employ thorium with a small admixture of enriched uranium that contains not more than 20% of {sup 235}U. It operates in the open-cycle mode involving {sup 233}U production from thorium and its subsequent burnup. The reactor meets the conditions of a nonproliferation of nuclear weapons: the content of fissionable isotopes in uranium at all stages of the process, including the final one, is below the threshold for constructing an atomic bomb, the amount of product plutonium being extremely small.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Neutronic analysis of a PWR-type SMR core using duplex ThO{sub 2}-UO{sub 2} in TRISO fuel particles

Currently, studies on small modular reactors (SMRs) present an important development due to the potential they represent in terms of safety, operational flexibility, economy, and non-energy applications. Furthermore, there is increasing interest in studying the use of thorium as fuel, as an effective way to solve problems such as the shortage of uranium reserves, reduction of nuclear waste and nuclear proliferation. Also, the use of thorium combined with highly enriched uranium TRISO particles has been studied, showing proper performance. In this work, the concept of ThO{sub 2}-UO{sub 2} duplex fuel is used, for the core configuration of a PWR type SMR that uses TRISO fuel, designed to achieve extended fuel cycles. Three distribution cases of ThO{sub 2} and UO{sub 2} in TRISO particles inside the fuel rods are compared. First, particles composed by a homogeneous mixture of ThO{sub 2} and UO{sub 2} are distributed inside of fuel elements. Second, the fuel zone of the fuel elements is divided into two radially, an internal one where the TRISO particles composed of ThO{sub 2} are distributed and the external one where the TRISO particles composed of UO{sub 2} are distributed. The third case is like the previous one, except that the particles that contain UO{sub 2} are distributed in the inner zone and those that contain ThO{sub 2} in the outer zone of the fuel elements. The comparison of the cases is carried out in terms of cycle main isotopes' mass transmutation, power distribution and temperature reactivity coefficients. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

ESBWR - Robust design for natural circulation and stability performance effectiveness

ESBWR is a 4500 MWt Generation III+ natural circulation reactor with an array of robust design features and passive safety systems to deliver highly effective plant performance during normal operation and to keep the reactor safe during postulated transients and accidents. With the submittal of the latest revision of the Design Control Document (DCD) to US Nuclear Regulatory Commission, ESBWR is nearing the completion of the US design certification process. This paper focuses on the natural circulation-driven plant performance aspects during normal operation, and stability evaluation of the robust ESBWR design. The TRACG computer code is used for the analysis of ESBWR plant performance, safety analysis, and stability margins. The paper describes the evaluation of ESBWR stability performance during normal power operation including operation in the Core Power-Feed Water Temperature Operating Domain. For ESBWR the normal power operation condition has the highest power/flow ratio and is limiting from the perspective of stability. The paper includes results from detailed evaluation of the most limiting decay ratio for out-of-phase regional oscillations calculated by perturbing the core inlet flow rate in this out-of-phase mode about the line of symmetry for the azimuthal harmonic mode. The paper also summarizes the ESBWR regional mode stability evaluations during a limiting transient (Loss of Feedwater Heating), and during ATWS (Anticipated Transient without Scram). Nominal decay ratios of limiting Channel oscillation, Core wide oscillation and Regional oscillation are within the maximum acceptance criterion of 0.8, at 95% content and 95% confidence. These stability evaluation results indicate decay ratio is within design limits. The paper also describes the evaluation of ESBWR stability performance during plant startup, and summarizes the defense-in-depth stability solution for ESBWR. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Measurements of SNAC2 area dosimeters placed in different configurations around the PROSPERO reactor and comparison with TRIPOLI-4 calculations

In the context of criticality accident alarm system tests, several experiments were carried out in 2013 on the PROSPERO reactor to study the response to neutron and gamma of different devices and dosimeters, particularly on the SNAC2 dosimeter. This article presents the results of this criticality dosimeter in different configurations, and compares the experimental measurements with the results of calculation performed with the TRIPOLI-4 Monte-Carlo Neutral Particles transport code. PROSPERO is a metallic critical assembly managed by the Criticality, Neutron Science and Measurement Department located at the French CEA Research Center of Valduc. The core, surrounded by a reflector of depleted uranium, is composed of 2 horizontal cylindrical blocks made of a highly enriched uranium alloy which can be placed in contact, and of 4 depleted uranium control rods which allow the reactor to be driven. This reactor, placed in a cell 10 m x 8 m x 6 m high, with 1.4-meter-thick concrete walls, is used as a fast neutron spectrum source and is operated at stable power level in delayed critical state, which can vary from 3 mW to 3 kW. PROSPERO is extensively used for electronic hardening or to study the effect of the neutrons on various materials. The SNAC2 criticality dosimeter is a zone dosimeter allowing the off line measurement of criticality accident neutron doses. This dosimeter consists of the pile up of seven activation foils embedded into a 23 mm diameter x 21 mm height cadmium container. The activation measurement of each foil, using a gamma spectroscopy technique, gives information about the neutron reaction rates. The SNAC2 software allows the spectrum unfolding from these values, taking into account the hypothesis of a particular spectrum shape, in three components: a Maxwell spectrum component for the thermal range, a 1/E component for the epithermal range, and a Watt spectrum component for the high energy range. Moreover, from the neutron spectrum, the SNAC software can calculate the neutron fluence integrated by the dosimeter and the neutron dose. During the 3 weeks measurement campaign many radioprotection devices were used. To modify the spectrum seen by these devices, several shields of various thicknesses made of concrete or polyethylene, with or without cadmium covers, were placed in the PROSPERO cell. These devices allow the study of criticality accident spectra in several environments: from metal to pseudo liquid. The fluxes measured by the SNAC2 devices were compared with TRIPOLI-4 calculations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

H-mode grade confinement in L-mode edge plasmas at negative triangularity on DIII-D

Plasmas with a negative triangularity shape have been created on the DIII-D tokamak that, despite maintaining standard L-mode edge radial profiles, reach volume averaged pressure levels typical of H-mode scenarios. Within the auxiliary power available for these experiments, plasmas exhibit near-zero power degradation while sustaining β N = 2.7 and H 98, y 2 = 1.2 for several energy confinement times. Detailed comparison with matched discharges at positive triangularity indicates that Trapped Electron Modes are weakened at negative triangularity, consistent with increased confinement and reduced intensity of fluctuations in electron density, electron temperature, and ion density. Furthermore, these results indicate that a negative triangularity plasma operating without an edge pedestal might provide an attractive scenario for operations in future reactors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Taxonomy for Common-Cause Failure Vulnerability and Mitigation

Applying current guidance and practices for common-cause failure (CCF) mitigation to digital instrumentation and control (I&C) systems has proven problematic, and the regulatory environment has been unpredictable. The potential for CCF vulnerability inhibits I&C modernization, thereby challenging the long-term sustainability of existing plants. For new plants and advanced reactor concepts, concern about CCF vulnerability in highly integrated digital I&C systems imposes a design burden that results in higher costs and increased complexity. The regulatory uncertainty in determining which mitigation strategies will be acceptable (e.g., what diversity is needed and how much is sufficient) drives designers to adopt complicated, costly solutions devised for existing plants. To address the conditions that constrain the transition to digital I&C technology by the US nuclear industry, crosscutting research is needed to resolve uncertainty, demonstrate necessary characteristics, and establish an objective basis for qualification of digital technology for nuclear power plant (NPP) I&C applications. To fulfill this research need, Oak Ridge National Laboratory is investigating mitigation of CCF vulnerability for nuclear-qualified applications. The outcome of this research is expected to contribute to a fundamentally sound, comprehensive basis to qualify digital technology for nuclear power applications. This report documents the development of a CCF taxonomy. The basis for the CCF taxonomy was generated by determining consistent terminology and establishing a classification approach. The terminology is based on definitions from standards, guides, and relevant nuclear power industry technical reports. The classification approach is derived from identified classification schemes focused on I&C systems and key characteristics, including failure modes. The CCF taxonomy provides the basis for a systematic organization of key systems aspects relevant to analyzing the potential for CCF vulnerability and the suitability of mitigation techniques. Development of an effective CCF taxonomy will help to provide a framework for establishing the objective analysis and assessment capabilities desired to facilitate rigorous identification of fault types and triggers that are the fundamental elements of CCF.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

Evaluations of University of Wisconsin Silicon Carbide Temperature Monitors 300 LO and 400 LO B

Silicon carbide (SiC) temperature monitors 05R4-02-A KG1403 (300 LO) and 05R4-01-A KG1415 (400 LO B) were evaluated at the High Temperature Test Lab (HTTL) to determine their peak irradiation temperatures. HTTL measurements indicate that the peak irradiation temperature for the 300 LO monitor was 295 {+-} 20 °C and the peak irradiation temperature for the 400 LO B monitor was 294 {+-} 25 °C. Two silicon carbide (SiC) temperature monitors irradiated in the Advanced Test Reactor (ATR) were evaluated at the High Temperature Test Lab (HTTL) to determine their peak temperature during irradiation. These monitors were irradiated as part of the University of Wisconsin Pilot Project with a target dose of 3 dpa. Temperature monitors were fabricated from high density (3.203 g/cm 3 ) SiC manufactured by Rohm Haas with a nominal size of 12.5 mm x 1.0 mm x 0.75 mm (see Attachment A). Table 1 provides identification for each monitor with an expected peak irradiation temperature range based on preliminary thermal analysis (see Attachment B). Post irradiation calculations are planned to reduce uncertainties in these calculated temperatures. Since the early 1960s, SiC has been used as a post-irradiation temperature monitor. As noted in Reference 2, several researchers have observed that neutron irradiation induced lattice expansion of SiC annealed out when the post-irradiation annealing temperature exceeds the peak irradiation temperature. As noted in Reference 3, INL uses resistivity measurements to infer peak irradiation temperature from SiC monitors. Figure 1 depicts the equipment at the HTTL used to evaluate the SiC monitors. The SiC monitors are heated in the annealing furnace using isochronal temperature steps that, depending on customer needs, can range from 50 to 800 °C. This furnace is located under a ventilation hood within the stainless steel enclosure. The ventilation system is activated during heating so that any released vapors are vented through this system. Annealing temperatures are recorded using a National Institute of Standards and Technology (NIST) traceable thermocouple inserted into an alumina tube in the furnace. After each isochronal annealing, the specimens are placed in a specialized fixture located in the constant temperature chamber (maintained at 30 °C) for a minimum of 30 minutes. After the 30 minute wait time, each specimen's resistance is measured using the specialized fixture and a calibrated DC power analyzer. This report discusses the evaluation of the SiC monitors and presents the results. Testing was conducted in accordance with Reference 3. Sections 2 and 3 present the data collected for each monitor and provide interpretation of the data. Section 4 presents the evaluated temperature results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

3D Simulation of Missing Pellet Surface Defects in Light Water Reactor Fuel Rods

The cladding on light water reactor (LWR) fuel rods provides a stable enclosure for fuel pellets and serves as a first barrier against fission product release. Consequently, it is important to design fuel to prevent cladding failure due to mechanical interactions with fuel pellets. Cladding stresses can be effectively limited by controlling power increase rates. However, it has been shown that local geometric irregularities caused by manufacturing defects known as missing pellet surfaces (MPS) in fuel pellets can lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. Nuclear fuel performance codes commonly use a 1.5D (axisymmetric, axially-stacked, one-dimensional radial) or 2D axisymmetric representation of the fuel rod. To study the effects of MPS defects, results from 1.5D or 2D fuel performance analyses are typically mapped to thermo-mechanical models that consist of a 2D plane-strain slice or a full 3D representation of the geometry of the pellet and clad in the region of the defect. The BISON fuel performance code developed at Idaho National Laboratory employs either a 2D axisymmetric or 3D representation of the full fuel rod. This allows for a computational model of the full fuel rod to include local defects. A 3D thermo-mechanical model is used to simulate the global fuel rod behavior, and includes effects on the thermal and mechanical behavior of the fuel due to accumulation of fission products, fission gas production and release, and the effects of fission gas accumulation on thermal conductivity across the fuel-clad gap. Local defects can be modeled simply by including them in the 3D fuel rod model, without the need for mapping between two separate models. This allows for the complete set of physics used in a fuel performance analysis to be included naturally in the computational representation of the local defect, and for the effects of the local defect to be coupled with the global fuel rod model. This approach for modeling fuel with MPS defects is demonstrated and compared with alternative techniques. The effects of varying parameters of the MPS defect are studied using this technique and presented here.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Thermal Evaluations for Hypothetically Drain-Down Spent Fuel Storage Facility at SRS

The L-Area Spent Nuclear Fuel (SNF) underwater storage facility at Savannah River Site (SRS), referred to as L-Basin SNF facility, has an array of Vertical Tube Storage (VTS) racks and a group of High Flux Irradiation Reactor (HFIR) fuel cores. The underwater facility is integrally designed as a wet SNF storage system with a chemically-controlled water pool. The SNF storage racks are seismically qualified. The L-Basin pool holds about 3.4 million gallons of water with concrete wall thickness of about 5 ft and water depths of 17 ft to 30 ft. The facility has mainly two different elevation levels of 30 ft and 17 ft below the ground level. At the level 30 ft below the ground, a maximum of 110 racks of VTS and 5 racks of Dry Cave can be stored. At the level of 17 ft below the ground, 120 HFIR cores can be stored, considering that each rack contains two HFIR cores. In case of a hypothetical accident, all cooling water in the storage pool may be drained away, resulting in dry SNF storage condition. The primary objective of the work is to estimate the maximum fuel temperature in a conservative manner when the L-Basin facility is completely dry due to the hypothetical drain-down of the pool water. A three-dimensional Computational Fluid Dynamics (CFD) modeling approach was taken for the thermal evaluations of the water-drained SNF storage facility. The modeling calculations were performed, assuming that all of the storage racks in L-Basin are fully loaded with SNF assemblies containing decay heat sources. In the case of full loading, the L-Basin facility stores 14,600 SNF assemblies in VTS racks and 120 cores in HFIR racks. The thermal analysis results of the L-Basin model yielded maximum temperatures of VTS and HFIR fuel assemblies for different fuel loading scenarios and boundary conditions of the SNF storage racks.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Predicting deviations in ex-core detector response in Krsko NPP during an earthquake with Monte-Carlo neutron transport methods

To study the response of ex-core detectors during an earthquake, an approach using Monte Carlo code MCNP was developed. As a first step to predict the response of the ex-core detectors a detailed MCNP model of the Krsko reactor core was created. A detailed MCNP core model was used to calculate 3D power distributions inside the core. Calculated power distributions were verified by comparison to the CORD-2 calculations, which is currently used for core design calculation verification of the Krsko nuclear power plant. Due to the outside the core position of the neutron detectors, the hybrid code ADVANTG is used to generate weight windows to speed up neutron transport outside the reactor core. To be able to use ADVANTG, fixed neutron source had to be reconstructed from the criticality core calculation. It was found that the power range ex-core detector movement has a negligible effect on the value of the thermal neutron flux at the active part of the detector. However, the 5 cm radial movement of the intermediate-range detector leads to 7 % - 8 % change in thermal neutron flux within the active intermediate-range detector region. The analysis continued with the evaluation of the effect of core barrel movement on the ex-core detector response. It was determined that the 2 mm core barrel radial oscillation can lead to ∼ 4% change in thermal neutron flux within the active detector region. It was evaluated that reactor core movement inside the baffle can lead up to ∼ 1.9% and ∼ 1.3% deviation in power-range and intermediate-range detector signal respectively. Analysis showed that the mechanical movement of ex-core neutron detectors could not explain the fluctuations in the ex-core detector signal. However, combined core barrel and reactor core inside baffle oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Reliability and Integrity Management Program Implementation Approach

Nuclear energy is the most reliable and environmentally sustainable energy source available today. In the United States, nuclear-generated power accounts for approximately 20% of total electricity and over 55% of clean energy. New advanced reactors have enormous potential to help further decarbonize the energy market, enhance grid resiliency, create new jobs, and build a stronger economy. More than 50 new reactors are being developed in the United States, and the federal government realizes an urgent need to deploy nuclear technologies to meet the country’s energy, environmental, and national security goals. As such, the U.S. Department of Energy launched multiple programs to support advanced reactor deployment. The research described in this paper explores implementation strategies for the Reliability and Integrity Management Program that directly supports the U.S. Department of Energy goal to enable the near-term deployment of the advanced reactor technologies. The project is conducted under the Regulatory Development Program for advanced reactors sponsored by the U.S. Department of Energy.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

U.S. Efforts in Support of Examinations at Fukushima Daiichi - September 2024 Meeting Notes

Information obtained from Fukushima Daiichi Nuclear Power Station (Daiichi) is required to inform future Decontamination and Decommissioning (D&D) activities, improving the ability of the Tokyo Electric Power Company Holdings, Incorporated (TEPCO Holdings) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This information also has important implications for the safety and operation of U.S. Commercial nuclear power plants. A collaborative U.S. and Japanese effort was initiated in 2014 by the Department of Energy Office of Nuclear Energy to identify Daiichi examination needs and evaluate recent Daiichi examination data to address these needs. This document summarizes information presented at and findings, action items, and recommendations by U.S. and Japanese experts in reactor safety and plant operations during the September 2024 Forensics Effort meeting. Significant safety insights were obtained in several areas: system and component performance, radionuclide surveys and sampling, debris end-state location, combustible gas effects, and plant operations and maintenance. In addition to reducing uncertainties and knowledge gaps in severe accident modeling progression, these insights continue to be used to assess whether additional updates are needed in guidance for severe accident prevention, mitigation, and emergency planning. Furthermore, Daiichi-related activities, such as code modeling improvements and analysis, testing, and new technology deployment efforts, have the potential to offer additional safety and economic benefits to the operating fleet and new light water reactor (LWR) and non-LWR designs.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

An Evaluation of Energy Storage Options for Nuclear Power

The dynamics of energy generation, delivery, and consumption in the United States continue to change as new energy sources come online and consumers purchase more energy-intensive devices. The percentage of electricity generation from variable renewable energy sources like wind and solar power has grown year after year. Meanwhile, the evolution of electric vehicles, smart devices, and consumer appliances is changing the way that electricity is used. As a result of these trends, nuclear power plants (NPPs) are being asked to operate more flexibly. Newly constructed, advanced nuclear power plants (A-NPPs) will likely need to adapt their business models with these market shifts in mind. In the future, A-NPPs might operate in an electricity system that values flexibility (e.g. fast ramping) more than stability (e.g. baseload supply). The current fleet of NPPs is struggling to remain economically viable in competitive markets that favor low natural gas prices and renewable energy sources with low marginal costs. Rising construction costs and ambiguous national policies have only added to the challenges facing plant operators. The industry as a whole is in decline, with 19 reactors in the process of being decommissioned and 5 of these plants retiring in the past decade. However, NPPs also provide several performance and environmental benefits by maintaining a high capacity factor and generating almost zero greenhouse gas (GHG) emissions over the life cycle of the plant. As more areas of energy consumption in the United States are electrified, NPPs could potentially replace coal and natural gas as a source of heat and power to help decarbonize the U.S. energy mix. Improving the flexibility of newly constructed A-NPPs to increase revenue through market hedging and reduce costs via stable operation was explored in this analysis. The increasing penetration of variable energy resources could contribute to the uncertainty of electricity demand, which could in turn increase the need for dispatchable electricity generating units. This role is currently filled by inexpensive natural gas combined cycle (NGCC) plants, whereas coal and NPPs have traditionally been called upon to provide baseload power by maintaining a constant power output. As wholesale electricity prices continue to fall due to low natural gas prices and federally subsidized renewable energy, this business model has begun to fail, resulting in incidences of NPPs selling electricity at a loss. Since it has not been economically favorable for NPPs to operate flexibly in this situation in the past, due to their low variable costs compared to fossil fuel power plants, NPPs in the United States were not built with this purpose in mind. Therefore, legacy NPPs have had difficulty adapting to these shifting dynamics due to regulatory and technical constraints, coupled with a lack of meaningful experience. NPPs in France have operated in this manner for years, due to a much higher percentage of baseload generators and a lower percentage of fast-ramping NGCC plants. The integration of energy storage technologies (ESTs) alongside newly constructed A-NPPs will be investigated in this analysis as an alternative method for helping NPPs resolve the economic challenges they currently face. Although the deregulation of energy markets in the United States is responsible for much of the increased competition from natural gas and renewable energy that has negatively impacted NPPs, this process has also led to the creation of ancillary service markets. ESTs can provide many of these services, so an integrated energy storage system could supplement both the reliability of the electricity grid and the economic performance of a NPP by bidding into these markets. However, many technical, financial, and regulatory barriers still prevent ESTs from operating profitably in this space. The purpose of this body of work is to compile a transparent database of metrics for available ESTs and identify barriers to widespread deployment. An analytical selection methodology was also developed to identify favorable technologies for integration with specific NPPs based on information supplied by an interested developer. These tools could help reduce the perceived risk surrounding investment in energy storage systems, which could help both NPPs in need of economic assistance and ESTs in need of further research and development.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Liquid balance monitoring inside conventional, Retrofit, and bio-reactor landfill cells

Highlights: • The Retrofit, Control, and As-Built cells received 48, 14, and 213 L Mg{sup −1} (liters of liquids per metric ton of waste). • The leachate collection system yielded 60, 57 and 198 L Mg{sup −1} from the Retrofit, Control, and As-Built cells. • The head on liner in all cells was below regulatory limits. • Measured moisture content of the waste samples was consistent with that calculated from accumulated liquid by balance. • The in-place saturated hydraulic conductivity of the MSW was calculated to be in the range of 10{sup −8} to 10{sup −7} m s{sup −1}. - Abstract: The Outer Loop landfill bioreactor (OLLB) in Louisville, KY, USA has been the site of a study to evaluate long-term bioreactor performance at a full-scale operational landfill. Three types of landfill units were studied including a conventional landfill (Control cell), a new landfill area that had an air addition and recirculation piping network installed as waste was being placed (As-Built cell), and a conventional landfill that was modified to allow for liquids recirculation (Retrofit cell). During the monitoring period, the Retrofit, Control, and As-Built cells received 48, 14, and 213 L Mg{sup −1} (liters of liquids per metric ton of waste), respectively. The leachate collection system yielded 60, 57 and 198 L Mg{sup −1} from the Retrofit, Control, and As-Built cells, respectively. The head on liner in all cells was below regulatory limits. In the Control and As-Built cells, leachate head on liner decreased once waste placement stopped. The measured moisture content of the waste samples was consistent with that calculated from the estimate of accumulated liquid by the liquid balance. Additionally, measurements on excavated solid waste samples revealed large spatial variability in waste moisture content. The degree of saturation in the Control cells decreased from 85% to 75%. The degree of saturation increased from 82% to 83% due to liquids addition in the Retrofit cells and decreased back to 80% once liquid addition stopped. In the As-Built cells, the degree of saturation increased from 87% to 97% during filling activities and then started to decrease soon after filling activities stopped to reach 92% at the end of the monitoring period. The measured leachate generation rates were used to estimate an in-place saturated hydraulic conductivity of the MSW in the range of 10{sup −8} to 10{sup −7} m s{sup −1} which is lower than previous reports. In the Control and Retrofit cells, the net loss in liquids, 43 and 12 L Mg{sup −1}, respectively, was similar to the measured settlement of 15% and 5–8% strain, respectively (Abichou et al., 2013). The increase in net liquid volume in the As-Built cells indicates that the 37% (average) measured settlement strain in these cells cannot be due to consolidation as the waste mass did not lose any moisture but rather suggests that settlement was attributable to lubrication of waste particle contacts, softening of flexible porous materials, and additional biological degradation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

U.S. Effort Support to Examinations at Fukushima - Meeting Notes with Updated Information Requests (FY2020)

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). Information obtained from Daiichi is required to inform Decontamination and Decommissioning (D&D) activities, improving the ability of the Tokyo Electric Power Company Holdings, Incorporated (TEPCO Holdings) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This document, which has been updated to include Fiscal Year 2020 (FY2020) information, summarizes results from a U.S. effort to review Daiichi information and extract insights to enhance the safety of existing and future nuclear power plant designs. This U.S. effort, which was initiated in 2014 by the Department of Energy Office of Nuclear Energy (DOE-NE), is completed by a group of experts in reactor safety and plant operations that identify examination needs and evaluate recent Daiichi examination data to address these needs. Since its inception, comprehensive annual reports were issued that document significant safety insights being obtained in areas of special emphasis: system and component performance, radionuclide surveys and sampling, debris end-state location, combustible gas effects, and plant operations and maintenance. In addition to reducing uncertainties related to severe accident modeling progression, these insights have and continue to be used to update guidance for severe accident prevention, mitigation, and emergency planning. Reduced uncertainties in modeling the events at Daiichi improve the realism of reactor safety evaluations that inform future D&D activities. For FY2020, it was decided that the program would gain more benefit from a more concise report that emphasizes new information and insights that affect prior findings and recommendations from the U.S. experts participating in this effort. A key aspect of prior U.S. efforts, the updated list of information requests, is included in this letter report to ensure that they are transmitted to organizations within Japan. In addition, findings and associated recommendations are provided regarding information presented by TEPCO Holdings, Nuclear Damage Compensation and Decommissioning Facilitation Corporation (NDF), and the Japan Atomic Energy Agency (JAEA). This letter report also continues to emphasize how information obtained from the affected reactors at Daiichi has been and will continue to be used to update severe accident management strategies, improve maintenance activities (especially in areas of radiation protection) and reduce uncertainties in systems analysis code models. In addition, recommendations are included that would expand the use of this information to provide insights regarding maintenance, radiation protection, design, and siting activities of existing and new reactors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Enzyme selection, optimization, and production toward biodegradation of post–consumer poly(ethylene terephthalate) at scale

Poly(ethylene terephthalate) (PET) is one of the world's most widely used polyester plastics. Due to its chemical stability, PET is extremely difficult to hydrolyze in a natural environment. Recent discoveries in new polyester hydrolases and breakthroughs in enzyme engineering strategies have inspired enormous research on biorecycling of PET. This study summarizes our research efforts toward large-scale, efficient, and economical biodegradation of post-consumer waste PET, including PET hydrolase selection and optimization, high-yield enzyme production, and high-capacity enzymatic degradation of post-consumer waste PET. First, genes encoding PETase and MHETase from Ideonella sakaiensis and the ICCG variant of leaf-branch compost cutinase (LCC ICCG ) were codon-optimized and expressed in Escherichia coli BL21(DE3) for high-yield production. To further lower the enzyme production cost, a pelB leader sequence was fused to LCC ICCG so that the enzyme can be secreted into the medium to facilitate recovery. To help bind the enzyme on the hydrophobic surface of PET, a substrate-binding module in a polyhydroxyalkanoate depolymerase from Alcaligenes faecalis (PBM) was fused to the C-terminus of LCC ICCG . The resulting four different LCC ICCG variants (LCC, PelB-LCC, LCC-PBM, and PelB-LCC-PBM), together with PETase and MHETase, were compared for PET degradation efficiency. A fed-batch fermentation process was developed to produce the target enzymes up to 1.2 g L –1 . Finally, the best enzyme, PelB-LCC, was selected and used for the efficient degradation of 200 g L –1 recycled PET in a well-controlled, stirred-tank reactor. Furthermore, the results will help develop an economical and scalable biorecycling process toward a circular PET economy.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2018

Liquid Metals in Nuclear Power: An Engineer Looks Into the Past and Future

A brief historical review of the use of liquid metals in nuclear power in the period 1950–2017 and the state of the problem today will make it possible to evaluate the prospects for their use in the first half of the 21st century (to 2050). There is hope that a discussion of these assessments will stimulate new approaches to the development of nuclear power. This presentation does not make predictions and concerns personal assessments by someone who has worked with liquid metals for more than 50 years. The present article is a revised variant of a preprint and report with the same title made at the interdepartmental seminar “Teplofizika-2007. Heat-and-Mass Transfer and Properties of Liquid Metals.”.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗