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At least 721 records · Page 40

Results for the Aboveground Configuration of the Boiling Water Reactor Dry Cask Simulator

The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and also by increasing the internal convection through greater canister helium pressure. These same canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and belowground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. The purpose of the current investigation was to produce data sets that can be used to test the validity of the assumptions associated with the calculations used to determine steady-state cladding temperatures in modern dry casks that utilize elevated helium pressure in the sealed canister in an aboveground configuration. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly was deployed inside of a representative storage basket and cylindrical pressure vessel that represents a vertical canister system. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. The arrangement of ducting was used to mimic conditions for an aboveground storage configuration in a vertical, dry cask systems with canisters. Transverse and axial temperature profiles were measured for a wide range of decay power and helium cask pressures. Of particular interest was the evaluation of the effect of increased helium pressure on peak cladding temperatures (PCTs) for identical thermal loads. All steady state peak temperatures and induced flow rates increased with increasing assembly power. Peak cladding temperatures decreased with increasing internal helium pressure for a given assembly power, indicating increased internal convection. In addition, the location of the PCT moved from near the top of the assembly to ~1/3 the height of the assembly for the highest (8 bar absolute) to the lowest (0 bar absolute) pressure studied, respectively. This shift in PCT location is consistent with the varying contribution of convective heat transfer proportional with of internal helium pressure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Gamma-ray Analysis of the Pressurized-Water-Reactor and Boiling-Water-Reactor Assemblies

Conclusions: Passive gamma analysis can be used to determine BU, CT, and IE of spent fuel assembly. Assemblies with small CT give best results.; Without any operator declaration, a 10 minutes measurement time of a PWR assembly can give the results for BU, CT, and IE with about 3% accuracy.; BWR analysis is more complicated and less effective than PWR and the results are somewhat less accurate.; A set of 4 measurements at 4 corners of an assembly can average out the variations of the assembly positions in the reactor during the burn cycles and can result in a better accuracy.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Modifications to MELCOR 1.8.6 for modeling Accident Tolerant Fuel in Boiling Water Reactors

This work describes further modifications made to a previously implemented User-Defined Generalized Coating (UDGC) model in MELCOR 1.8.6. The UDGC version of the code allows for the modeling of cladding and canister materials other than Zircaloy (without overwriting Zircaloy properties) as well as coated Zircaloy cladding and canisters, with the thermophysical and oxidation properties of the clad or coating material specified by the user. Specific changes outlined in this report are the extension of the UDGC models to BWR canisters, and generalization of the oxidation model for all UDGC components to accommodate different oxidation rate correlations in two different temperature ranges. User input requirements are specified and an example case is described that illustrates the potential impact of model assumptions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Hydrothermal Corrosion of Second Generation FeCrAl Alloys in Boiling Water Reactor Conditions

As part of an effort to develop advanced fuel cladding for use in light water reactors, several ORNL-developed FeCrAl alloys were tested in BWR conditions to determine their resistance to hydrothermal corrosion. Coupons with alloy compositions Fe-10Cr-6Al-2Mo, Fe-13Cr-5Al-2Mo, Fe-13Cr-6Al-2Mo, Fe-13Cr-7Al-2Mo, and Fe-13Cr-5Al-2Mo-1Nb were tested in BWR-NWC and BWR-HWC for 9 months in continuously refreshing autoclaves. Commercial APMT and Zircaloy-2 were also tested for comparison. Among samples exposed to HWC, Alloy C06M (10Cr-6Al) performed best, with a mass loss of 0.34 mg/cm 2 over 9 months, and C037M performed worst with a mass loss of 1.2 mg/cm 2 over 9 months. This mass loss was similar to Gen-1 FeCrAl alloys, and represents an estimated annual thickness loss of approximately a few µm per year. Samples exposed to NWC had very small mass losses of less than 0.15 mg/cm 2 or mass gains up to 0.05 mg/cm 2 . These represent very small annual losses in thickness. Based on the results of this testing, the Gen-2 FeCrAl alloys tested have suitable resistance to hydrothermal corrosion for use as fuel cladding in BWR-NWC and BWR-HWC conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Materials and Dimensional Reference Handbook for the Boiling Water Reactor Dry Cask Simulator.

The performance of commercial nuclear spent fuel dry storage casks is evaluated through detailed analytical modeling of the system’s thermal performance. A recent investigation has been completed that produced a data set that can be used to test the validity of the modeling presently used to determine cladding temperatures in modern vertical dry casks. To produce these data sets under well-controlled boundary conditions, the dry cask simulator (DCS) was built to study the thermal-hydraulic response of fuel under a variety of heat loads, internal vessel pressures, and external configurations [Durbin and Lindgren, 2017]. The experiments were conducted in Albuquerque, New Mexico where the local ambient atmospheric pressure is typically 83 kPa. The purpose of this handbook is to document the pertinent geometric and material property information needed to perform model validation efforts.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Vendor-Independent Design Requirements for a Boiling Water Reactor Safety System Upgrade

This Vendor-Independent License Amendment Request (LAR) Framework Document is a research product developed for the Idaho National Laboratory Light Water Reactor Sustainability (LWRS) Program. It was developed by MPR Associates, Inc. with technical input from LWRS and Exelon Generation personnel. To accomplish the purpose of this research, an operating plant (Limerick Generating Station [LGS]) was used as a reference to provide a concrete example and to present a "top-down" view of design concepts consistent with functional requirements baseline documents (references X and Y) also developed for this research. While this document is written in a framework to support a complete LAR submittal, its purpose is to communicate research concepts and provide an example of how to present digital information necessary to address expectations with regard to leveraging the Alternate Review Process as provided in Digital Instrumentation and Controls Interim Staff Guidance #06, Revision 2, Licensing Process. To leverage that process, this research document presupposes that a utility would select a vendor platform which has been prequalified for safety-related by the United States Nuclear Regulatory Commission. While this document is written using the Exelon Generation Limerick Generating Station (LGS) Units 1 and 2 as a baseline, this document is a research product. This document contains no commitments and makes no binding design decisions for Exelon Generation. Exelon Generation is leveraging this research to support their in-progress LGS Plant Protection System (PPS) upgrade efforts and plans to leverage it for their Redundant Reactor Control System (RRCS) replacement. The LWRS Program appreciates the research support provided by Exelon Generation in the generation of this document.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Boiling Water Capsule Irradiation Test Device Design Study Report

RELAP5 calculations were conducted to support an experimental design, referred to as water capsule, which can facilitate fuel performance testing within ATR. The main objective is to provide useful insights into the experimental design and operating strategies that can mimic LWR operating conditions within the spatial constraints of the ATR position under consideration. This report summarizes the RELAP5 analysis results based on the most recent water capsule design derived through various tests. To discuss a design under conservative conditions, fuel power is consistently assumed to be 400 W/m in all analyses.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗