Suppression of Richtmyer-Meshkov instability via special pairs of shocks and phase transitions

The classical Richtmyer-Meshkov instability is a hydrodynamic instability characterizing the evolution of an interface following shock loading. In contrast to other hydrodynamic instabilities such as Rayleigh-Taylor, it is known for being unconditionally unstable: regardless of the direction of shock passage, any deviations from a flat interface will be amplified. In this article, we show that for negative Atwood numbers, there exist special sequences of shocks which result in a nearly perfectly suppressed instability growth. We demonstrate this principle computationally and experimentally with stepped fliers and phase transition materials. A fascinating immediate corollary is that in specific instances a phase transitioning material may self-suppress RMI

NGNP Point Design - Results of the Initial Neutronics and Thermal-Hydraulic Assessments During FY-03, Rev. 1

This report presents the preliminary preconceptual designs for two possible versions of the Next Generation Nuclear Plant (NGNP), one for a prismatic fuel type helium gas-cooled reactor and one for a pebble bed fuel helium gas reactor. Both designs are to meet three basic requirements: a coolant outlet temperature of 1000 °C, passive safety, and a total power output consistent with that expected for commercial high-temperature gas-cooled reactors. The two efforts are discussed separately below. The analytical results presented in this report are very promising, however, we wish to caution the reader that future, more detailed, design work will be needed to provide final answers to a number of key questions including the allowable power level, the inlet temperature, the power density, the optimum fuel form, and others. The point design work presented in this report provides a starting point for other evaluations, and directions for the detailed design, but not final answers

The Next Generation Nuclear Plant - Insights Gained from the INEEL Point Design Studies

This paper provides the results of an assessment of two possible versions of the Next Generation Nuclear Plant (NGNP), a prismatic fuel type helium gas-cooled reactor and a pebble-bed fuel helium gas reactor. Insights gained regarding the strengths and weaknesses of the two designs are also discussed. Both designs will meet the three basic requirements that have been set for the NGNP: a coolant outlet temperature of 1000 C, passive safety, and a total power output consistent with that expected for commercial high-temperature gas-cooled reactors. Two major modifications of the current Gas Turbine- Modular Helium Reactor (GT-MHR) design were needed to obtain a prismatic block design with a 1000 C outlet temperature: reducing the bypass flow and better controlling the inlet coolant flow distribution to the core. The total power that could be obtained for different core heights without exceeding a peak transient fuel temperature of 1600 °C during a high or low-pressure conduction cooldown event was calculated. With a coolant inlet temperature of 490 °C and 10% nominal core bypass flow, it is estimated that the peak power for a 10-block high core is 686 MWt, for a 12-block high core is 786 MWt, and for a 14-block core is about 889 MWt. The core neutronics calculations showed that the NGNP will exhibit strongly negative Doppler and isothermal temperature coefficients of reactivity over the burnup cycle. In the event of rapid loss of the helium gas, there is negligible core reactivity change. However, water or steam ingress into the core coolant channels can produce a relatively large reactivity effect. Two versions of an annular pebble-bed NGNP have also been developed, a 300 and a 600 MWt module. From this work we learned how to design passively safe pebble bed reactors that produce more than 600 MWt. We also found a way to improve both the fuel utilization and safety by modifying the pebble design (by adjusting the fuel zone radius in the pebble to optimize the fuel-to-moderator ratio). We also learned how to perform design optimization calculations by using a genetic algorithm that automatically selects a sequence of design parameter sets to meet specified fitness criteria increasingly well. In the pebble-bed NGNP design work, we use the genetic algorithm to direct the INEEL’s PEBBED code to perform hundreds of code runs in less than a day to find optimized design configurations. And finally, we learned how to calculate cross sections more accurately for pebble bed reactors, and we identified research needs for the further refinement of the cross section calculations

Pulsed Photonuclear Assessment (PPA) Technique: CY 04 Year-end Progress Report

Idaho National Laboratory (INL), along with Los Alamos National Laboratory (LANL) and Idaho State University’s Idaho Accelerator Center (IAC), are developing an electron accelerator-based, photonuclear inspection technology for the detection of smuggled nuclear material within air-, rail-, and especially, maritime-cargo transportation containers. This CY04 report describes the latest developments and progress with the development of the Pulsed, Photonuclear Assessment (PPA) nuclear material inspection ystem, such as: (1) the identification of an optimal range of electron beam energies for interrogation applications, (2) the development of a new “cabinet safe” electron accelerator (i.e., Varitron II) to assess “cabinet safe-type” operations, (3) the numerical and experimental validation responses of nuclear materials placed within selected cargo configurations, 4) the fabrication and utilization of Calibration Pallets for inspection technology performance verification, 5) the initial technology integration of basic radiographic “imaging/mapping” with induced neutron and gamma-ray detection, 6) the characterization of electron beam-generated photon sources for optimal performance, 7) the development of experimentallydetermined Receiver-Operator-Characterization curves, and 8) several other system component assessments. This project is supported by the Department of Homeland Security and is a technology component of the Science & Technology Active Interrogation Portfolio entitled “Photofission-based Nuclear Material Detection and Characterization.

Detection of Shielded Nuclear Material in a Cargo Container

The Idaho National Laboratory, along with Los Alamos National Laboratory and the Idaho State University’s Idaho Accelerator Center, are developing electron accelerator-based, photonuclear inspection technologies for the detection of shielded nuclear material within air-, rail-, and especially, maritime-cargo transportation containers. This paper describes a developing prototypical cargo container inspection system utilizing the Pulsed Photonuclear Assessment (PPA) technology, incorporates interchangeable, well-defined, contraband shielding structures (i.e., "calibration" pallets) providing realistic detection data for induced radiation signatures from smuggled nuclear material, and provides various shielded nuclear material detection results. Using a 4.8-kg quantity of depleted uranium, neutron and gamma-ray detection responses are presented for well-defined shielded and unshielded configurations evaluated in a selected cargo container inspection configuration. © 2001 Elsevier Science. All rights reserve

Next Generation Nuclear Plant Methods Technical Program Plan

One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope of the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR

Pulsed Photonuclear Assessment (PPA) Technique: CY-05 Project Summary Report

Idaho National Laboratory, along with Idaho State University’s Idaho Accelerator Center and Los Alamos National Laboratory, is developing an electron accelerator-based, photonuclear inspection technology, called the Pulsed Photonuclear Assessment (PPA) system, for the detection of nuclear material concealed within air-, rail-, and, primarily, maritime-cargo transportation containers. This report summarizes the advances and progress of the system’s development in 2005. The contents of this report include an overview of the prototype inspection system, selected Receiver-Operator-Characteristic curves for system detection performance characterization, a description of the approach used to integrate the three major detection components of the PPA inspection system, highlights of the gray-scale density mapping technique being used for significant shield material detection, and higher electron beam energy detection results to support an evaluation for an optimal interrogating beam energy. This project is supported by the Department of Homeland Security Office of Research and Development and, more recently, the Domestic Nuclear Detection Office

Next Generation Nuclear Plant Methods Technical Program Plan

One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope of the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR

New Generation Nuclear Plant (NGNP) Project, Preliminary Point Design

This paper provides a preliminary assessment of two possible versions of the Next Generation Nuclear Plant (NGNP), a prismatic fuel type helium gas-cooled reactor and a pebblebed fuel helium gas reactor. Both designs will meet the three basic requirements that have been set for the NGNP: a coolant outlet temperature of 1000 C, passive safety, and a total power output consistent with that expected for commercial high-temperature gas-cooled reactors

Waste disposal options report. Volume 2

Volume 2 contains the following topical sections: estimates of feed and waste volumes, compositions, and properties; evaluation of radionuclide inventory for Zr calcine; evaluation of radionuclide inventory for Al calcine; determination of k{sub eff} for high level waste canisters in various configurations; review of ceramic silicone foam for radioactive waste disposal; epoxides for low-level radioactive waste disposal; evaluation of several neutralization cases in processing calcine and sodium-bearing waste; background information for EFEs, dose rates, watts/canister, and PE-curies; waste disposal options assumptions; update of radiation field definition and thermal generation rates for calcine process packages of various geometries-HKP-26-97; and standard criteria of candidate repositories and environmental regulations for the treatment and disposal of ICPP radioactive mixed wastes