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At least 613 records · Page 34

A safety and licensing roadmap to identify the research and development gaps of commercial molten salt reactors

Molten salt reactors (MSRs) both inherently and through good design possess highly desirable safety attributes. Successful commercial deployment of MSRs is heavily dependent on establishing their safety characteristics in a well-coordinated licensing requirements program. The approach outlined is based on adapting the MSR licensing process to the current LWR–based NRC process that exists today, which has been the traditional approach used by new technologies such as the Pebble Bed Reactor licensing application, the Clinch River Breeder Reactor licensing application, and more recently the NuScale licensing application. NRC has recognized some of the unique passively safe aspects of advanced non-LWR reactors and has introduced an advanced reactor strategic plan which will address policy issues that need to be addressed to accommodate advanced reactor concepts within the existing regulatory framework. Examples of some of these policy issues are the use of functional containment, the use of mechanistic source terms, and more reliance on performance-based and risk-informed regulatory processes, such as the Licensing Modernization Project, which emphasizes the use of risk-informed processes to select licensing basis events, safety categorization of SSC, and well-defined defense-in-depth systems. Applying performance-based, risk-informed regulatory processes is, consequently, vital to substantially decreasing the cost, time, and uncertainty of commercial MSR reactor licensing while providing for the health and safety of the public and protecting the environment. This document is intended to provide a roadmap for the MSR licensing and safety activities necessary to provide a success path to commercialization.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Thermal Test Suite for Code Maintenance and Validation of Diablo

This document describes a series of test problems that ensure that the Thermal capabilities of Diablo are working correctly. Diablo utilizes the Finite Element Method to discretize the problem in space and a trapezoidal rule finite difference scheme to discretize the problem in time. This suite will test the codes capabilities to handle various boundary conditions, source terms, and initial conditions. It will also test the coupling between thermal and solid effects, and the use of various linear and nonlinear solvers.

42 ENGINEERING↗

Rad-NESHAP Review of TA-21-257 Remediation Activities

Environmental Protection and Compliance (EPC) Division personnel reviewed the remediation work at Technical Area 21, Building 257 (TA-21-257) for compliance with Radionuclide NESHAP1 requirements. Remediation activities include removal of Building 257, the associated waste lines, and waste storage tanks. We used project data and the most recent CAP88 plume dispersion model to estimate the emissions source term and the resulting off-site dose to members of the public. Emissions monitoring requirements from this work are being met by measuring ambient air concentrations at public receptor locations using LANL’s Airnet system. All requirements under 40 CFR 61 Subpart H are being met for the currently planned scope of operations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Inventory Data Package for the Hanford Site Composite Analysis

The disposed inventory estimate is perhaps the most important component of the composite analysis (CA), as it directly affects the future radiological impacts following site closure. The goal is to estimate radionuclide inventory from site inception to closure. Because of the inclusive nature of a CA, all relevant contaminants are identified and initially considered. Then, subsets of contaminants appropriate for quantitative analysis are selected. Inventory is reported within the geographical boundary of the Hanford Site 200 Areas and includes contributing sources on the Central Plateau. The inventory covers the radiological inventories of historical discharges and residuals that have the potential to release after closure. The reported inventory is focused on the source terms needed for transport modeling and the evaluation of risk associated with the ground water pathway.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Directional Software Algorithms and Sensor Evaluations

This report evaluates the relative performance of two directional gamma-ray spectrometers and processing algorithms that are used to construct images and spatially resolved spectra. Polaris, which was developed by H3D Inc., uses 18 pixelated CZT crystals to construct gamma-ray images in either Compton camera(CC) or coded aperture (CA) mode. The other sensor that is referenced in this report incorporates a commercial high-purity germanium based imager, called GeGI, with a coded aperture mask and processing software developed by Oak Ridge National Laboratory (ORNL). H3D and the University of Michigan provided several algorithms that can be used to process data collected by Polaris in CC mode. This evaluation compares the performance of these algorithms with a Directional Unfolded Source Term (DUST) approach developed by Sandia National Laboratories (SNL). DUST differs from the other algorithms because its primary objective is synthesis of spatially-resolved gamma ray spectra as opposed to image reconstruction.

97 MATHEMATICS AND COMPUTING↗

Advanced Reactor Siting Policy Considerations

This report describes considerations for and challenges to revising the US Nuclear Regulatory Commission (NRC) guidance regarding population for siting an advanced reactor. Current regulations in 10 Code of Federal Regulations (CFR) Part 100, Reactor Site Criteria, and guidance support in Regulatory Guide (RG) 4.7, Revision 3, General Site Suitability Criteria for Nuclear Power Stations, are focused on large light water reactors (LWRs), addressing the distance that a nuclear reactor should be sited from a densely populated center. This report analyzes the alternatives to using a formula distance as the primary figure of merit for siting small modular reactors (SMRs) or non-light water reactors (non-LWRs). Some reactor vendors use a business plan that sites reactors close to an industrial partner to supply heat or to back-fit a reactor at a fossil plant site to take advantage of existing infrastructure. Factors such as smaller source terms (fission product release to the environment), passive safety systems, advances in barrier technology, advances in simulation and modeling, and improved understanding of societal risk may allow for siting these types of reactors closer to densely populated centers than has historically been accepted for large LWRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Completion of Level 2 Milestone – M2AT 19IN030401044, Complete construction of the gas and moisture supply systems and receive the custom-designed tube furnace from the vendor

The Advanced Gas Reactor (AGR) program has designed and is now constructing the Air/Moisture Ingress Experiment (AMIX) facility to test irradiated tristructural isotropic (TRISO) fuels under air or moisture-ingress accidents postulated for high-temperature gas-cooled reactors. AMIX will ultimately be installed in the Fuel Conditioning Facility at the Materials and Fuels Complex (MFC) at Idaho National Laboratory. AMIX consists of a gas supply (capable of supplying mixtures of air/helium and steam/helium), a tube furnace capable of accepting samples up to 6 cm in diameter and heating them to 1650°C, gamma detectors for measuring the release of condensable, gamma-emitting fission products, and cold traps with gamma detectors for measuring the release of fission gases. Data obtained from AMIX experiments will be used to qualify this new generation of TRISO fuels with the U.S. Nuclear Regulatory Commission, and inform models and simulations of TRISO fuel accident performance and source term.

42 ENGINEERING↗

Monitoring an Electro-Refining Process (Rev. 1)

In this report we examine measurements from two different sensors for detecting a diversion of plutonium in an electro-refining (ER) process. One method is based on counts from a high dose neutron detector (HDND). The other is based on spectra from a micro-calorimeter sensor. The spectra come from ER-salt, fission product waste, and from a source term. The goal is to differentiate between normal and off normal operating conditions using neutron counts or changes in counts in the spectra. Unfortunately, there is no electro-refining process that we can actually monitor to obtain real data from various operating conditions so we will use simulated data to illustrate our approach.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Modeling of Pyroprocessing Hot Cell for Process Monitoring Evaluation

In this report, we examine studies undertaken to determine the applicability of applying measurements with a neutron counter within a pyroprocessing hot cell. The studies focus on the modeling of the hot cell and neutron counter in Monte Carlo N- Particle Code (MCNP) to determine expected radiation effects and applicability of neutron counters. The neutron counter of focus was the High Dose Neutron Detector (HDND) developed at Los Alamos National Laboratory (LANL). We studied the HDND to determine what would be the detector response under normal background conditions with operating process equipment as well as the background’s potential effect on counting applications involving a U/TRU ingot produced by the pyroprocessing electrorefiner. To perform this task, we were required to develop models of the pyroprocessing process cell and unit operations and generate source terms for various different times in the process. The data for the pyroprocessing materials came from SNL’s Safeguards and Security Process Model (SSPM) as part of MPACT’s advanced integration mission. This report will discuss the development of the pyroprocessing process cell and unit operations models in MCNP as well as various simulations to determine the applicability of the HDND for process monitoring efforts in the hot cell.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Failures and Implications of Heat Pipe Systems

Under Department of Energy (DOE), Office of Nuclear Energy (NE), Gateway for Accelerated Innovation in Nuclear (GAIN), Sandia National Laboratories (SNL) was awarded DOE-NE GAIN voucher GA-19SN020107, "Risk-informed mechanistic source term calculations for a sodium fast reactor." Under this GAIN voucher, SNL supported the industry partners development in preparation for licensing and commercialization by providing subject matter expertise on heat pipe technologies, providing computer code training and support, and perform first-of-a-kind experiments demonstrating the safety/risk impacts of heat pipe breach failures. The experiments that were performed had two primary goals: measure the peak heat fluxes that lead to heat pipe dry out and subsequent wall breach; and observe the consequences that result from catastrophic failure of a heat pipe wall. Intentional breaching of the heat pipe walls took advantage of heat pipe physics and operating limits. Large and nearly instantaneous heat fluxes were applied to the heat pipe to first cause localized dry out at the evaporator section which then leads to melting of the heat pipe wall. The hour glass heat pipe (Test 1) experienced dry out at 112 W/cm 2 and after 45 seconds, wall temperatures measure about 1,280°C and intentional failure of the heat pipe wall was achieved. The cylindrical heat pipe (Test 2) experienced dry out at 125 W/cm 2 and after 65 seconds, wall temperatures exceeded 1,400°C and intentional failure of the heat pipe wall was achieved. Both experiments characterize the parameters needed to lead to heat pipe wall failure. Furthermore, the failure of the heat pipes characterizes the safety/risk impacts from sodium-oxygen reactions that occur following the intentional failure. There were two major conclusions of these intentional failure tests: the heat pipes were able to continue operating beyond expected performance limits, and the failure behavior validated decades of operational experience.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Inventory Data Package for the Hanford Site Composite Analysis

The disposed inventory estimate is perhaps the most important component of the composite analysis (CA), as it directly affects the future radiological impacts following site closure. The goal is to estimate radionuclide inventory from site inception to closure. Because of the inclusive nature of a CA, all relevant contaminants are identified and initially considered. Then, subsets of contaminants appropriate for quantitative analysis are selected. Reducing the number of radionuclides for inclusion in the quantitative analysis helps focus budget and resources on simulating only those radionuclides that are likely to contribute to the total dose to the receptor above a threshold value. Inventory is reported within the geographical boundary of the Hanford Site 200 Areas and includes contributing sources on the Central Plateau. The inventory covers the radiological inventories of historical discharges and residuals that have the potential to release after closure. The reported inventory is focused on the source terms needed for transport modeling and the evaluation of risk associated with the ground water pathway. Risk from the air pathway was considered. Inventories of sites with approved performance assessments (PAs) will be integrated using their release curves directly with vadose zone transport modeling.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Final Report on the NA-22 sponsored project from FY17-19: Next-Generation Correlated Fission Event Simulations

Through this multi-organization correlated fission physics research and development project, the CGMF and FREYA fission event generators have been implemented into the MCNP6.2 code and released to the public while a variety of dedicated differential experiments on correlated fission signatures have been performed to help validate the complex fission theory and physics implemented in these codes. In addition to enabling these new capabilities, a multitude of efforts in modeling and transport simulation has been directed toward a variety of applications including general nuclear nonproliferation, subcritical neutron multiplication, and nuclear criticality safety applications. While there is still much more to be done in the area of correlated fission physics, including the study of b-delayed fission observables, more robust validation and an extension into more application spaces, one of the key successes of this project can be summarized as the delivery of a newly demonstrated and consistently integrated simulation capability spanning several complex nuclear physics codes. From the special nuclear materials source term to the radiation transport to the time-coincident detector response modeling, new and existing users around the world, especially throughout the nuclear nonproliferation communities, can now begin to model these kinds of highly complex special nuclear material applications. Another successful aspect of this project included fruitful collaborations, mentoring opportunities and community outreach. One extended collaboration workshop was held in Santa Fe, NM in March 2017. Two other limited collaboration meetings took place including a brief gathering during the mid-project Schubert review at LANL and a single all-day special session on “Correlated Fission Physics, Transport and Applications” held during the 2019 Annual Meeting of the Institute of Nuclear Materials Management (INMM) in Palm Desert, CA. Finally, as part of this project, we have mentored 14 undergraduate students, PhD candidates, and postdoctoral researchers listed in Appendix A. The remainder of this report contains detailed information followed by highlights of the goals and accomplishments in Appendix B broken down by task within this project.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Composite Analysis for Low Level Waste Disposal in the Central Plateau of the Hanford Site (FY2019 Annual Status Report)

In accordance with DOE M 435.1-1 requirements and as implemented by DOE/RL-2000-29, the U.S. Department of Energy (DOE), Richland Operations Office has prepared this annual summary for fiscal year (FY) 2019. Originally reported in PNNL-11800 and PNNL-11800, Addendum 1, the Hanford Site Composite Analysis was approved through issuance of a 2002 memorandum. As required by DOE/RL-2000-29, an annual evaluation of new information and data developed by a number of onsite programs was completed. The reporting period for this annual evaluation is FY 2019 (October 1, 2018, through September 30, 2019). The information provided in this evaluation includes activities performed in FY 2019 that are considered pertinent to the Hanford Site Composite Analysis - information that could change source terms considered, and monitoring, research, and development results.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

SCALE Code System

The SCALE Code System is a widely used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules, including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Technical and Licensing Considerations for Micro-Reactors

The U.S. Nuclear Regulatory Commission (NRC) has interacted with vendors pursuing the commercialization of micro-reactors (i.e., reactors capable of producing about 1 MW(th) to 20 MW(th) of energy from nuclear fission). It is envisioned that micro-reactors could be assembled and fueled in a factory and shipped to a site. Many of the sites are expected to be remote locations requiring off-grid power or in some cases military bases. The objective of this effort is to explore the technical issues and the approach required to reach a finding of "reasonable assurance of public health and safety" for this new and different class of reactors. The analysis performed here leverages available micro-reactor design and testing data available from national laboratory experience as well as commercial design information to explore technical issues. Some factors considered include source term, accidents that would need to be analyzed, and the extent of the probabilistic risk assessment (PRA). The technical evaluation was performed within the framework of the Licensing Modernization Project (LMP) to identify licensing basis events, classification of structures, systems and components, and defense-in-depth needed to provide regulatory certainty. With this framework and technical evaluation in mind, the scope and content of a micro-reactor licensing application is discussed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Molten Salt Reactor Engineering Study for Off-Gas Management

In previous work published by the current team, the waste processing and waste form options were summarized for molten salt reactors (MSR) (Riley et al., 2018b; Riley et al., 2019). The primary types of waste from an MSR are summarized in Figure S1 and include (1) off-gas streams, (2) salt waste streams, (3) separated salt streams, (4) metal waste streams, (5) carbon waste streams, (6) decommissioning and decontaminating (D&D) waste streams, and (7) operating waste streams, or those that are generated from maintenance procedures. The primary focus of this report will be on management of the off-gas stream as it represents the pressure boundary for fuel-salt MSRs and, thus, is required for fission product confinement during reactor operation and reduction of the source term during a reactor accident.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

M4SF-20LL010302042: Process Model for Radionuclide Incorporation into Corrosion Products

This progress report (Level 4 Milestone Number M4SF-20LL010302042) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Crystalline Activity Number SF-20LL010302041. The research is focused on actinide and radionuclide sequestration in steel corrosion products. Fuel matrix degradation models suggest that the near field is likely to be reducing at the time of canister breaching, steel corrosion, and radionuclide release, but more oxidizing conditions may prevail in the far field. The incorporation of radionuclides into corrosion phases may limit the rate of radionuclide release by sequestering a portion of the radionuclide source term. For these reasons there is a need to evaluate the incorporation of Pu and other radionuclides into various Fe-oxide phases, and to understand the behavior of coprecipitated phases during mineral recrystallization processes and during re-oxidation events. Radionuclide coprecipitation with Fe minerals may impact long-term repository performance and is an ongoing research focus at Lawrence Livermore National Laboratory.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗