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At least 865 records · Page 48

Hadron-argon cross-section measurements in ProtoDUNE-SP

The single-phase liquid argon ProtoDUNE detector (ProtoDUNE-SP) is located at CERN’s neutrino platform facility and serves as a prototype to validate the technology for the 10-kton fiducial mass liquid argon detectors for the DUNE experiment. ProtoDUNE-SP was exposed to a variety of test-beam particles (electrons, muons, pions, kaons, and protons) towards the end of 2018 collecting data in a broad range of momenta, from 0.3 - 7 GeV/c. The results will improve the understanding of final-state interactions in neutrino-argon cross-sections. The progress of the analyses towards the cross-section measurements on argon from the test-beam data collected at protoDUNE-SP will be presented.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

NASA-Lewis experiences with multigroup cross sections and shielding calculations

The nuclear reactor shield analysis procedures employed at NASA-Lewis are described. Emphasis is placed on the generation, use, and testing of multigroup cross section data. Although coupled neutron and gamma ray cross section sets are useful in two dimensional Sn transport calculations, much insight has been gained from examination of uncoupled calculations. These have led to experimental and analytic studies of areas deemed to be of first order importance to reactor shield calculations. A discussion is given of problems encountered in using multigroup cross sections in the resolved resonance energy range. The addition to ENDF files of calculated and/or measured neutron-energy-dependent capture gamma ray spectra for shielding calculations is questioned for the resonance region. Anomalies inherent in two dimensional Sn transport calculations which may overwhelm any cross section discrepancies are illustrated.

Lahti, G. P.↗

Current and future uses of accelerators in particle astrophysics

Beams of artificially accelerated heavy ions, protons, antiprotons, electrons, and positrons currently available at (and planned for) numerous facilities around the world are a valuable resource to the Cosmic Ray community. Such beams have been used to test detector concepts, calibrate balloon-borne and space flight experiments and to measure fundamental nuclear physics parameter necessary for the interpretation of Cosmic Ray data. As new experiments are flown the quality and extent of Cosmic Ray measurements will continue to improve. It will be necessary to increase activity at ground based accelerators in order to test/calibrate these new instruments and to maintain (or possibly improve) the ability to interpret these data. In this area, the newly formed Transport Collaboration, supported by NASA, will be providing new nuclear interaction cross section measurements for beams with Z less than or = 58 and supporting new instrument calibrations at the Lawrence Berkeley Laboratory Bevalac accelerator.

Guzik, T. G.↗

TREAT Multi-SERTTA Neutronics Design and Analysis Support

Transient Reactor Experiment Test (TREAT) experiment vehicle design is necessary in preparation for TREAT facility restart and the resumption of transient testing to support current Accident Tolerant Fuel (ATF) characterization needs and future fuels testing requirements. Currently the most mature vehicle design is the Multi-SERTTA (Static Environments Rodlet Transient Test Apparatuses), which can accommodate up to four concurrent rodlet-sized specimens under separate environmental conditions. Development of this robust test vehicle design requires neutronics analyses to support optimization of the Power Coupling Factor (PCF) to efficiently maximize energy generation in the fuel rodlets and to perform essential safety analyses. Current Multi-SERTTA calculations support final design of this vehicle, the design process for other TREAT vehicles, and the development of analyses practices for upcoming transient testing experiments. Models of the preliminary design Multi-SERTTA vehicle with rodlets were prepared and neutronics calculations were performed using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. Discussion of results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Strength Oxidation of NBG-17 Graphite

This poster is to show the strength after oxidation of NBG-17 graphite at various mass loss percentages. This data is compared to previous data that had no standards of testing to establish an ASTM standard for future application.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

The MELCOR Analysis of Chinshan Nuclear Power Plant Spent Fuel Pool for Fukushima-like Accident

The safety analysis of the nuclear power plant (NPP) is very important work in the NPP safety. Especially after the Fukushima NPP accident occurred, the importance of NPP safety analysis has been raised and there is more concern for the safety of NPPs in Taiwan. Because the earthquake and tsunami occurred, the cooling system of spent fuel pool failed and the safety issue of spent fuel pool generated in Japan's Fukushima NPP. Chinshan NPP is the first NPP in Taiwan which is BWR/4 plant and OLTP (Original Licensed Thermal Power) for each unit is 1775 MWt. Chinshan NPP finished SPU (stretch power uprate) and the operating power is 103.66% of the OLTP, which is 1840 MWt now. After Fukushima NPP accident occurred, in order to concern the safety of Chinshan NPP spent fuel pool, we performed the safety analysis of spent fuel pool by using MELCOR/SNAP which also assumed the cooling system of spent fuel pool failed. The geometry of Chinshan NPP spent fuel pool is 12.17 m x 7.87 m x 11.61 m and the initial condition is 60 deg. (water temperature) / 1.013 x 10{sup 5} Pa. And, the total power of the fuels is roughly 8.9 MWt initially. MELCOR is developed by Sandia National Laboratories. MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor NPPs, including containment response and leakage of activity to the environment. A broad spectrum of severe accident phenomena in both BWRs and PWRs is treated in MELCOR in a unified framework. These include thermalhydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. According to Dr. Carbajo's paper, MELCOR was used to perform the study of spent fuel pool of Fukushima Daiichi Unit 4. Niina E. Koenoenen used MELCOR to establish the spent fuel pool model of a Nordic BWR. The loss-of-pool-cooling accidents was simulated and analyzed by this model. The above studies indicate that MELCOR is capable of handling the simulation of the spent fuel pool. An increasing number of researchers are using MELCOR code to analyze test facilities and nuclear power plants. Jun Wang et al. used MELCOR to perform core degradation simulation and MELCOR results were compared with CORA experimental data. Through this work, they reviewed the performance of MELCOR COR package in detail. Tae Woon Kim et al. developed MELCOR model of APR1400 (an advanced pressurized water reactor). They used this model to perform the LBLOCA transient simulation and sensitivity study. Longze Li et al. established the MELCOR model of CPR1000 (a Chinese pressurized reactor 1000-MW power plant). The station blackout (SBO) with failure of the steam generator (SG) safety relief valve (SRV) transient was simulated and analyzed by this model. SNAP is a graphic user interface program that processes the inputs and outputs of MELCOR. In addition, there is an animation function in SNAP which can present the animation of analysis results. There are two main steps in this research. The first step was the establishment of Chinshan NPP spent fuel pool MELCOR/SNAP model. And the transient analysis of Chinshan NPP spent fuel pool MELCOR/SNAP model under the cooling system failure condition was performed. The next step was the comparison among the MELCOR, TRACE and CFD data. TRACE and CFD data were from INER reports. In addition, the animation model of Chinshan NPP spent fuel pool was presented by using the animation function of SNAP with MELCOR/SNAP analysis results. This study has developed MELCOR/SNAP model of Chinshan NPP spent fuel pool successfully. By using the above model, the safety analysis of the spent fuel pool was performed under the cooling system of spent fuel pool failed condition. The analysis results of MELCOR, TRACE and CFD were similar in this case. It indicated that there was a respectable accuracy in MELCOR/SNAP model. The analysis results depicted that the uncovered of the fuels occurred at 2.6 day and the metal-water reaction of fuels occurred roughly at 3.7 day after the cooling system failed. The above results indicated that the failure of cladding occurred after 3.7 day. This study's results can help to evaluate the safety issue of Chinshan NPP spent fuel pool. In addition, the cooling system and some mitigation measures of spent fuel pool is not simulated in the above model. However, we will do it in the future. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Korean Localization Program for Performance Verification Test of APR1400's RCP

The Korean government together with the Korean nuclear society completed the development of technologies for the design, manufacturing, and performance verification test of the APR1400's Reactor Coolant Pump (RCP) as one of the national programs for the localization of deficient nuclear technology. The RCP localization program was driven by KAERI, which was responsible for the development of the performance verification and evaluation technology, and by DOOSAN heavy industrial company, which was in charge of the design and manufacturing of the RCP. The test facility for the localized APR1400's RCP was constructed at the KAERI site at the end of 2012. The thermal hydraulic and electric capability of the RCP test facility (RCPTF) covers up to 18.5 MPa, 343 deg. C, 11.7 m{sup 3}/s, and 14.0 MW for the design pressure, temperature, flow rate, and maximum electric power, respectively. In 2013, a commissioning test was performed to verify its designed capability, followed by several modifications in the RCPTF including signal processing and control logic to enhance the verification and evaluation capability of the RCP performance. After finishing the commissioning and modification of the RCPTF, a type test for the new-type of RCP was performed successfully. In 2014 and 2015, eight production tests for the localized RCPs, which will be installed in a commercial nuclear power plant in Korea, were successfully completed. As a final stage of the localization program of the RCP, KAERI will perform a seal LOCA test to verify the leakage characteristics and produce experimental data on the localized seal assembly for a safety analysis during a station black out (SBO) condition. In this summary, several technical issues developed through a series of research processes for a performance verification test of the localized RCP are described.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Using Information Automation and Human Technology Integration to Implement Integrated Operations for Nuclear

The purpose of the research effort described in this report is to develop and demonstrate an approach to the design and implementation of advanced, automated systems intended to increase operational efficiencies at nuclear power plants. In particular, we describe methods for considering human-technology integration (HTI) issues throughout the various phases of system design, test, and implementation and how these considerations help promote effective design. This research project will develop planning tools and comprehensive guidance on how HTI principles and methods, in combination with information automation technologies, can enable effective data integration and coordination for full nuclear plant modernization. Specifically, this research project will develop an approach to automate the mapping of data from plant systems and processes to application needs, thereby significantly reducing the amount of human workload currently required for the execution of these tasks. In addition to developing an automated solution as a replacement for these tasks, this research will also analyze digitalization’s effectiveness in reducing operational costs of compliance related activities. Compliance activities are estimated to account for as much as 50% of operations and maintenance (i.e., non-fuel and non-capital) costs of plant operation.

97 MATHEMATICS AND COMPUTING↗

FY2018 ILAW Glass SPFT Testing for Disposal at IDF: Glasses ORLEC46 and ORLEC52 (Final Report, Rev. 0)

About 50 million gallons of high-level mixed waste is currently stored in underground tanks at The United States Department of Energy’s (DOE’s) Hanford site in the State of Washington. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) will provide DOE’s Office of River Protection (ORP) with a means of treating this waste by vitrification for subsequent disposal. The tank waste will be separated into low- and high-activity waste fractions, which will then be vitrified respectively into Immobilized Low Activity Waste (ILAW) and Immobilized High Level Waste (IHLW) products. The ILAW product will be disposed in an engineered facility – the Integrated Disposal Facility (IDF) – on the Hanford site, while the IHLW product is designed for deep geological disposal in a national facility for high level nuclear waste. The ILAW and IHLW products must meet a variety of requirements with respect to protection of the environment before they can be accepted for disposal. The objective of the work described in this report is to perform testing, data collection, and analyses for the ILAW glass product for subsequent use in the performance assessment (PA) of the IDF to assess potential environmental risks associated with long-term storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Searching for Flavor Dependence in Nuclear Quark Behavior

The observed correlation between the EMC effect and the contribution of short-range correlations (SRCs) in nuclei suggests that the modification of the quark distributions of bound protons and neutrons might occur within SRCs. This raises the possibility that the EMC effect may have an isospin dependence arising from the np dominance of SRCs. We discuss previous attempts to test this possibility and perform a new analysis of existing data. We find no experimental support for the observation of an isospin dependence of the EMC effect.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Collision of Physics and Software in the Monte Carlo Application Toolkit (MCATK)

The topic is presented in a series of slides organized as follows: MCATK overview, development strategy, available algorithms, problem modeling (sources, geometry, data, tallies), parallelism, miscellaneous tools/features, example MCATK application, recent areas of research, and summary and future work. MCATK is a C++ component-based Monte Carlo neutron-gamma transport software library with continuous energy neutron and photon transport. Designed to build specialized applications and to provide new functionality in existing general-purpose Monte Carlo codes like MCNP, it reads ACE formatted nuclear data generated by NJOY. The motivation behind MCATK was to reduce costs. MCATK physics involves continuous energy neutron & gamma transport with multi-temperature treatment, static eigenvalue (k eff and α) algorithms, time-dependent algorithm, and fission chain algorithms. MCATK geometry includes mesh geometries and solid body geometries. MCATK provides verified, unit-test Monte Carlo components, flexibility in Monte Carlo application development, and numerous tools such as geometry and cross section plotters.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Identification of nuclear effects in neutrino-carbon interactions at low three-momentum transfer

Two different nuclear-medium effects are isolated using a low three-momentum transfer subsample of neutrino-carbon scattering data from the MINERvA neutrino experiment. The observed hadronic energy in charged-current νμ interactions is combined with muon kinematics to permit separation of the quasielastic and Δ(1232) resonance processes. First, we observe a small cross section at very low energy transfer that matches the expected screening effect of long-range nucleon correlations. Second, additions to the event rate in the kinematic region between the quasielastic and Δ resonance processes are needed to describe the data. The data in this kinematic region also have an enhanced population of multiproton final states. Contributions predicted for scattering from a nucleon pair have both properties; the model tested in this analysis is a significant improvement but does not fully describe the data. We present the results as a double-differential cross section to enable further investigation of nuclear models. Furthermore, improved description of the effects of the nuclear environment are required by current and future neutrino oscillation experiments.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

Development and tests of molybdenum armored copper components for MITICA ion source

In order to prevent detrimental material erosion of components impinged by back-streaming positive D or H ions in the megavolt ITER injector and concept advancement beam source, a solution based on explosion bonding technique has been identified for producing a 1 mm thick molybdenum armour layer on copper substrate, compatible with ITER requirements. Prototypes have been recently manufactured and tested in the high heat flux test facility Garching Large Divertor Sample Test Facility (GLADIS) to check the capability of the molybdenum-copper interface to withstand several thermal shock cycles at high power density. This paper presents both the numerical fluid-dynamic analyses of the prototypes simulating the test conditions in GLADIS as well as the experimental results.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Benchmark Specifications for Select Experiments Conducted at the Kansas State University Gallium Thermal-hydraulic Experiment Facility

The Department of Energy (DOE) – Nuclear Energy University Programs (NEUP) supported the creation and operation of the Gallium Thermal-hydraulic Experiment (GaTE) facility at Kansas State University (KSU) as part of a larger effort to understand thermal stratification behavior in liquid-metal-cooled reactors. GaTE was designed to simulate transients in a reactor plenum that are known to cause thermal stratification. High-reliability and high-resolution measurements describing stratification behavior in the coolant were collected for use as experimental benchmarks in validation efforts for computational models. The results of these tests contribute to a greater understanding of thermal stratification behavior of liquid metal under various configurations and operating conditions. This report provides a complete description of the benchmark problem, including all necessary details and description of a set of four forced flow and four natural circulation tests and measured data for comparison with model results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Simulation, Model Verification and Controls Development of Brayton Cycle PM Alternator: Testing and Simulation of 2 KW PM Generator with Diode Bridge Output

Professor Stankovic will be developing and refining Simulink based models of the PM alternator and comparing the simulation results with experimental measurements taken from the unit. Her first task is to validate the models using the experimental data. Her next task is to develop alternative control techniques for the application of the Brayton Cycle PM Alternator in a nuclear electric propulsion vehicle. The control techniques will be first simulated using the validated models then tried experimentally with hardware available at NASA. Testing and simulation of a 2KW PM synchronous generator with diode bridge output is described. The parameters of a synchronous PM generator have been measured and used in simulation. Test procedures have been developed to verify the PM generator model with diode bridge output. Experimental and simulation results are in excellent agreement.

Anna V Stankovic↗

Environmental Restoration of Diesel-Range Organics from Project Chariot, Cape Thompson, Alaska - 16147

Cape Thompson, Alaska, is located approximately 120 miles north of Kotzebue, Alaska, within the Ogotoruk Creek valley. The remote Alaskan location is also home to the US Atomic Energy Commission (AEC) Project Chariot study area. Project Chariot was part of the Plowshare Program, created in 1957 by the AEC to study peaceful uses for atomic energy. Project Chariot began in 1958 when a scientific field team chose Cape Thompson as a potential site to excavate a harbor using a series of nuclear explosions. Between 1959 and 1962, AEC was responsible for over 40 pretest bio-environmental and geotechnical studies of the Cape Thompson area. Strong public opposition to the tests developed, which caused the project to be cancelled with no nuclear explosions ever conducted at the site. In the early 1960's, several deep boreholes were drilled to obtain subsurface, temperature, and high-explosive testing data. Refrigerated diesel was used as a drilling fluid to keep the boreholes open so that thermistors could measure subsurface temperatures. The test boreholes were named Able, Baker, Charlie, Dog, and X-1. In the summer of 2014, the US Department of Energy, Office of Legacy Management (LM) completed the remediation of remnant diesel at these sites. Prior to the 2014 remediation activities, LM conducted two separate characterization field studies to determine the nature and extent of the contamination from the 1960's drilling activities. In 2013, LM developed a remediation plan that was submitted and approved by the Alaska Department of Environmental Conservation (ADEC). The 2014 Project Chariot remediation had some unique logistical challenges and historic issues with stakeholders to overcome. Many of the historic issues centered on previous AEC scientific actions and a general mistrust of the government by the local stakeholders. Prior to and during the 2014 remediation work, an outreach campaign was implemented to discuss prior activities at the site and include stakeholders from Point Hope, Alaska, in the remediation-planning process. The remoteness of the region also provided unique logistical challenges. Because of the short field season, LM initiated several subcontracts over a year in advance. The field camp, heavy equipment, and soil storage containers were shipped by barge to the site. Personnel, food, and small materials were flown in by bush plane from Kotzebue, Alaska. Previous site characterization had identified diesel-range organics (DROs) as the primary contaminant. The cleanup levels for the site were provided by the Alaska Administrative Code (AAC) for the Arctic Zone region. The AAC cleanup level for DROs on gravel pads is 500 mg/kg and 12,500 mg/kg tundra in the arctic-zone region. To help determine the extent and direct soil excavation in the field, photoionization detectors (PIDs) were used for field screening. While the PID results provided rough limits for excavation, laboratory confirmation sampling was required to provide actual excavation guidance. All excavation locations remained open until laboratory results met AAC criteria and ADEC approved closure. The laboratory sample protocol required the bush plane to pick up samples almost daily and transfer them to Anchorage for analysis. In order to meet holding time requirements, all laboratory samples were analyzed on an expedited basis. Analytical results and ADEC approval notifications were e-mailed back to the site via satellite connection. A total of 202 laboratory analyses were needed to complete the remediation effort. At each location (Able, Baker, Charlie, Dog, and X-1), the boreholes were plugged and abandoned in accordance with ADEC regulations. Contaminated soil was excavated, and the material was placed in super-sacks and hauled to connex boxes located near the sea shore. Approximately 786 tons of material was removed and shipped by barge to the Columbia Ridge Landfill in Arlington, Oregon, for disposal. The well casings were cut and sealed below ground surface. Each location was graded to its original contours and reclaimed with native vegetation, and a geodetic survey was conducted. Tours with leaders of the Village of Point Hope and the land owner (Arctic Slope Regional Corporation) were conducted prior to the final demobilization. All remediation equipment and camp facilities were removed at the completion of the project. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Corrective Action Decision Document/Corrective Action Plan for Corrective Action Unit 413: Clean Slate II Plutonium Dispersion (TTR) Tonopah Test Range, Nevada. Revision 0

This Corrective Action Decision Document/Corrective Action Plan provides the rationale and supporting information for the selection and implementation of corrective actions at Corrective Action Unit (CAU) 413, Clean Slate II Plutonium Dispersion (TTR). CAU 413 is located on the Tonopah Test Range and includes one corrective action site, TA-23-02CS. CAU 413 consists of the release of radionuclides to the surface and shallow subsurface from the Clean Slate II (CSII) storage–transportation test conducted on May 31, 1963. The CSII test was a non-nuclear detonation of a nuclear device located inside a concrete bunker covered with 2 feet of soil. To facilitate site investigation and the evaluation of data quality objectives decisions, the releases at CAU 413 were divided into seven study groups: 1 Undisturbed Areas 2 Disturbed Areas 3 Sedimentation Areas 4 Former Staging Area 5 Buried Debris 6 Potential Source Material 7 Soil Mounds Corrective action investigation (CAI) activities, as set forth in the CAU 413 Corrective Action Investigation Plan, were performed from June 2015 through May 2016. Radionuclides detected in samples collected during the CAI were used to estimate total effective dose using the Construction Worker exposure scenario. Corrective action was required for areas where total effective dose exceeded, or was assumed to exceed, the radiological final action level (FAL) of 25 millirem per year. The results of the CAI and the assumptions made in the data quality objectives resulted in the following conclusions: The FAL is exceeded in surface soil in SG1, Undisturbed Areas; The FAL is assumed to be exceeded in SG5, Buried Debris, where contaminated debris and soil were buried after the CSII test; The FAL is not exceeded at SG2, SG3, SG4, SG6, or SG7. Because the FAL is exceeded at CAU 413, corrective action is required and corrective action alternatives (CAAs) must be evaluated. For CAU 413, three CAAs were evaluated: no further action, clean closure, and closure in place. The CAAs were evaluated on technical merit focusing on performance, reliability, feasibility, safety, and cost. Based on the evaluation of analytical data from the CAI, review of future and current operations at CAU 413, and the detailed and comparative analysis of CAAs, clean closure was selected as the preferred CAA for CAU 413 by the U.S. Air Force, Nevada Division of Environmental Protection, and U.S. Department of Energy at the CAA meeting held on August 24, 2016.

54 ENVIRONMENTAL SCIENCES↗

Phenomena-based Uncertainty Quantification in Predictive Coupled- Physics Reactor Simulations

This project has sought to develop methodologies, tailored to phenomena that govern nuclearreactor behavior, to produce predictions (including uncertainties) for quantities of interest (QOIs) in the simulation of steady-state and transient reactor behavior. Examples of such predictions include, for each QOI, an expected value as well as a distribution around this value and an assessment of how much of the distribution stems from each major source of uncertainty. The project has sought to test its methodologies by comparing against measured experimental outcomes. The main experimental platform has been a 1-MW TRIGA reactor. This is a flexible platform for a wide range of experiments, including steady state with and without temperature feedback, slow transients with and without feedback, and rapid transients with strong feedback. The original plan was for the primary experimental data to come from in-core neutron detectors. We made considerable progress toward this goal but did not get as far along as we had planned. We have designed, developed, installed, and tested vertical guide tubes, each able to accept a detector or stack of detectors that can be moved axially inside the tube, and we have tested several new detector designs. One of these shows considerable promise.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗