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At least 487 records · Page 27

RELAP-7 Software Verification and Validation Plan

This INL plan comprehensively describes the software for RELAP-7 and documents the software, interface, and software design requirements for the application. The plan also describes the testing-based software verification and validation (SV&V) process—a set of specially designed software models used to test RELAP-7. The RELAP-7 (Reactor Excursion and Leak Analysis Program) code is a nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on the INL’s modern scientific software development framework – MOOSE (Multi-Physics Object-Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5’s capability and extends the analysis capability for all reactor system simulation scenarios.

97 MATHEMATICS AND COMPUTING↗

Toxic Substances Control Act (TSCA) Polychlorinated Biphenyl (PCB)/Radioactive Waste Annual Inventory for Calendar Year 2014

The Toxic Substances Control Act, 40 CFR 761.65(a)(1) provides an exemption from the one year storage time limit for PCB/radioactive waste. PCB/radioactive waste may exceed the one year time limit provided that the provisions at 40 CFR 761.65(a)(2)(ii) and 40 CFR 761.65(a)(2)(iii) are followed. These two subsections require, (ii) "A written record documenting all continuing attempts to secure disposal is maintained until the waste is disposed of" and (iii) "The written record required by subsection (ii) of this section is available for inspection or submission if requested by EPA." EPA Region 10 has requested the Department of Energy (DOE) to submit an inventory of radioactive-contaminated PCB waste in storage at the Idaho National Laboratory (INL) for the previous calendar year. The annual inventory is separated into two parts, INL without Advanced Mixed Waste Treatment Project (AMWTP) (this includes Battelle Energy Alliance, LLC, CH2M-WG Idaho, LLC, and the Naval Reactors Facility), and AMWTP.

99 GENERAL AND MISCELLANEOUS↗

HFIR Dissolution Nuclear Criticality Analysis Information

The purpose of this analysis is to demonstrate the safety of High Flux Isotope Reactor (HFIR) dissolution from a criticality safety perspective and to establish any Criticality Safety Limits (CSLs) that may be necessary. HFIR fuel consists of two annular elements that are designed to be nested together with a beryllium reflector in the reactor configuration. HFIR cores are currently stored in L-Area as separate inner elements and outer elements, each on a unique aluminum carrier. HFIR elements will be shipped to H-Canyon in the 70-ton Cask Railcar and loaded into the dissolver with the HFIR insert installed. The HFIR insert consists of a small well for the inner fuel element and a large well for the outer fuel element.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

A model to evaluate 100-year energy mix scenarios to facilitate deep decarbonization in the southeastern United States

The Southeast United States consumes approximately one billion megawatt-hours of electricity annually; roughly two-thirds from carbon dioxide (CO 2 ) emitting sources. The balance is produced by non-CO 2 emitting sources: nuclear power, hydroelectric power, and other renewables. Approximately 40% of the total CO 2 emissions come from the electric grid. The CO 2 emitting sources, coal, natural gas, and petroleum, produce approximately 372 million metric tons of CO 2 annually. The rest is divided between the transportation sector (36%), the industrial sector (20%), the residential sector (3%), and the commercial sector (2%). An Energy Mix Modeling Analysis (EMMA) tool was developed to evaluate 100-year energy mix strategies to reduce CO 2 emissions in the southeast. Current energy sector data was gathered and used to establish a 2016 reference baseline. The spreadsheet-based calculation runs 100-year scenarios based on current nuclear plant expiration dates, assumed electrical demand changes from the grid, assumed renewable power increases and efficiency gains, and assumed rates of reducing coal generation and deployment of new nuclear reactors. Within the model, natural gas electrical generation is calculated to meet any demand not met by other sources. Thus, natural gas is viewed as a transitional energy source that produces less CO 2 than coal until non-CO 2 emitting sources can be brought online. The annual production of CO 2 and spent nuclear fuel and the natural gas consumed are calculated and summed. A progression of eight preliminary scenarios show that nuclear power can substantially reduce or eliminate demand for natural gas within 100 years if it is added at a rate of only 1000 MWe per year. Any increases in renewable energy or efficiency gains can offset the need for nuclear power. However, using nuclear power to reduce CO 2 will result in significantly more spent fuel. More efficient advanced reactors can only marginally reduce the amount of spent fuel generated in the next 100 years if they are assumed to be available beginning around 2040. Thus closing the nuclear fuel cycle to reduce nuclear spent fuel inventories should be considered. Future work includes the incorporation of economic features into the model and the extension of the evaluation to the industrial sector. It will also be necessary to identify suitable sites for additional reactors.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗

Inelastic analysis procedure based on the Grade 91 unified viscoplastic constitutive model for ASME implementation

This report describes a design by inelastic analysis procedure for Grade 91 to be proposed for incorporation in the ASME Boiler and Pressure Vessel Code. The package of proposed additions includes guidance on inelastic modeling and a reference constitutive model for Grade 91 to be used with the existing Section III, Division 5, Subsection HB, Subpart B rules for design by inelastic analysis. The current Code specifies acceptance criteria for design by inelastic analysis but does not provide an acceptable constitutive model for Grade 91 or any of the other Class A materials. The proposed Nonmandatory Appendix would provide such a reference model, greatly easing the use of design by inelastic analysis. As the inelastic methods produce significantly less overconservative designs than the design by elastic analysis rules this proposed Code change could reduce the design and fabrication cost of future high temperature reactors.

36 MATERIALS SCIENCE↗

Development of a multiaxial deformation measure and creep-fatigue damage summation for multiple load cycle types in support of an improved creep-fatigue design method

This report describes methods for extending the elastic perfectly plastic (EPP) simplified model test (SMT) approach to cover multiaxial load conditions and methods to combine EPP+SMT use fractions for multiple loading cycles. The completion of this work, in conjunction with companion reports on constructing and validating SMT design curves, marks the development of a complete methodology for assessing creep-fatigue damage in high temperature nuclear reactor structural components. This EPP+SMT method aims to better account for elastic follow up effects, simplify the design process, and reduce overconservatism when compared to current simplified bounding approaches in the ASME Boiler and Pressure Vessel Code Section III, Division 5, Subsection HB, Subpart B covering high temperature nuclear reactors.

36 MATERIALS SCIENCE↗

Nuclear Waste Attributes of SMRs Scheduled for Near-Term Deployment

The purpose of this study is to evaluate the nuclear waste attributes of Small Modular Reactors (SMRs) scheduled for deployment within this decade using available data and established nuclear waste metrics, with the results compared to a reference large Pressurized Water Reactor (PWR). The current fleet of commercial nuclear reactors in the U.S. is composed of 92 large Light Water Reactors (LWR) with an average electricity generating capacity of over 1,000 MWe each. These large LWRs built on-site in massive construction projects have been the mainstay of the industry for the last 50 years. However, new construction soon is expected to include several designs of smaller reactors primarily fabricated in factories and installed in the field in modules. Some of these SMRs will also be LWRs, while some will use other coolants such as liquid metals, molten salts or gases, and different types of fuels. The technologies and economics of SMRs have been the focus of many studies, but there has been only minimal information published on the amount of nuclear waste different types of SMRs are expected to generate and no reports focused on near-term-deployable designs. In this study, the nuclear waste attributes of three small reactors scheduled for near-term-deployment, VOYGR TM (from NuScale Power), Natrium TM ab (from TerraPower), and Xe-100 (from X-energy), were assessed by comparing nuclear waste metrics with those of a reference large Pressurized Water Reactor (PWR).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗