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At least 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Waste management strategy for cost effective and environmentally friendly NPP decommissioning

Decommissioning of nuclear power plants generates large volumes of radioactive or potentially radioactive waste. The proper management of the dismantling waste plays an important role for the time needed for the dismantling phase and thus is critical to the decommissioning cost. An efficient and thorough process for inventorying, characterization and categorization of the waste provides a sound basis for the planning process. As part of comprehensive decommissioning studies for Nordic NPPs, Westinghouse has developed the decommissioning inventories that have been used for estimations of the duration of specific work packages and the corresponding costs. As part of creating the design basis for a national repository for decommissioning waste, the total production of different categories of waste packages has also been predicted. Studsvik has developed a risk based concept for categorization and handling of the generated waste using six different categories with a span from extremely small risk for radiological contamination to high level waste. The two companies have recently joined their skills in the area of decommissioning on selected market in a consortium named ndcon to further strengthen the proposed process. Depending on the risk for radiological contamination or the radiological properties and other properties of importance for waste management, treatment routes are proposed with well-defined and proven methods for on-site or off-site treatment, activity determination and conditioning. The system is based on a graded approach philosophy aiming for high confidence and sustainability, aiming for re-use and recycling where found applicable. The objective is to establish a process where all dismantled material has a pre-determined treatment route. These routes should through measurements, categorization, treatment, conditioning, intermediate storage and final disposal be designed to provide a steady, un-disturbed flow of material to avoid interruptions. Bottle-necks in the process causes increased space requirements and will have negative impact on the project schedule, which increases not only the cost but also the dose exposure to personnel. For these reasons it is critical to create a process that transfers material into conditioned waste ready for disposal as quickly as possible. To a certain extent the decommissioning program should be led by the waste management process. With the objective to reduce time for handling of dismantled material at site and to efficiently and environmental-friendly use waste management methods (clearance for re-use followed by clearance for recycling), the costs for the plant decommissioning could be reduced as well as time needed for performing the decommissioning project. Also, risks for delays would be reduced with a well-defined handling scheme which limits surprises. Delays are a major cost driver for decommissioning projects. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Waste Management Strategy for Dismantling Waste to Reduce Costs for Power Plant Decommissioning - 13543

Decommissioning of nuclear power plants generates large volumes of radioactive or potentially radioactive waste. The proper management of the dismantling waste plays an important role for the time needed for the dismantling phase and thus is critical to the decommissioning cost. An efficient and thorough process for inventorying, characterization and categorization of the waste provides a sound basis for the planning process. As part of comprehensive decommissioning studies for Nordic NPPs, Westinghouse has developed the decommissioning inventories that have been used for estimations of the duration of specific work packages and the corresponding costs. As part of creating the design basis for a national repository for decommissioning waste, the total production of different categories of waste packages has also been predicted. Studsvik has developed a risk based concept for categorization and handling of the generated waste using six different categories with a span from extremely small risk for radiological contamination to high level waste. The two companies have recently joined their skills in the area of decommissioning on selected market in a consortium named 'ndcon' to further strengthen the proposed process. Depending on the risk for radiological contamination or the radiological properties and other properties of importance for waste management, treatment routes are proposed with well-defined and proven methods for on-site or off-site treatment, activity determination and conditioning. The system is based on a graded approach philosophy aiming for high confidence and sustainability, aiming for re-use and recycling where found applicable. The objective is to establish a process where all dismantled material has a pre-determined treatment route. These routes should through measurements, categorization, treatment, conditioning, intermediate storage and final disposal be designed to provide a steady, un-disturbed flow of material to avoid interruptions. Bottle-necks in the process causes increased space requirements and will have negative impact on the project schedule, which increases not only the cost but also the dose exposure to personnel. For these reasons it is critical to create a process that transfers material into conditioned waste ready for disposal as quickly as possible. To a certain extent the decommissioning program should be led by the waste management process. With the objective to reduce time for handling of dismantled material at site and to efficiently and environmental-friendly use waste management methods (clearance for re-use followed by clearance for recycling), the costs for the plant decommissioning could be reduced as well as time needed for performing the decommissioning project. Also, risks for delays would be reduced with a well-defined handling scheme which limits surprises. Delays are a major cost driver for decommissioning projects. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Crossflow Filtration of Hanford Tank AP-105 Supernatant

A small-scale test platform to demonstrate the solids filtration, Cs removal, and low-activity waste (LAW) vitrification was constructed and installed at the Pacific Northwest National Laboratory (PNNL) 325 Building. Bench scale testing was conducted to demonstrate cross flow filtration using the cells unit filter (CUF) on approximately 14 liters of 241-AP-105 feed to support the confirmation of the Low-Activity Waste Pretreatment System (LAWPS) equipment performance and design basis for entrained solids removal process. Despite the low solids content (~16 ppm) of the AP-105 feed, significant fouling of the filter was observed that required backpulsing of the filter to maintain the permeate flux within the LAWPS baseline design rate. The improvement from backpulsing was significant, but transitory, and lasted a few hours. Increasing the transmembrane pressure was shown to have limited impact on the permeate flux in terms of magnitude and duration, especially compared with backpulsing. Because LAWPS must operate three to eight days to throughput requirements, testing demonstrated that backpulsing or chemical cleaning may need to be done every 7 to 14 hours when processing feed from AP-105. Because of this, backpulsing capability is strongly recommended for the baseline design. The LAWPS prototypic filter cleaning was demonstrated at the end of testing and was shown to be very effective as it largely restored the filter permeability to near pre-test conditions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Cyber-Physical Tabletop Exercise for Small Modular Reactor Facilities

U.S. nuclear power facilities face increasing challenges in meeting dynamic security requirements caused by evolving and expanding threats while keeping costs reasonable to make nuclear energy competitive. This evolving threat landscape includes adversaries having offensive cyber capabilities to attack information technology (IT) systems and operation technology (OT) systems. These adversaries may have the ability to attack the physical protection system (PPS) networks with potential consequential impacts that could degrade the effectiveness of the PPS. These cyber attacks may also be used to attack the safety and operational systems used to operate and ensure the safety of the reactor. Additionally, adversaries may gain access to unmanned aerial systems (UAS) that may be used to provide reconnaissance and surveillance of the facility, provide information to the adversaries, and be equipped with kinetic capabilities such as explosives or weapons that can be used to directly attack the facility. The Department of Energy’s Office of Nuclear Energy’s Advanced Reactor Safeguards and Security (ARSS) program funded Sandia National Laboratories (SNL) and Idaho National Laboratory (INL) to develop a cyber-physical tabletop exercise (TTX). This exercise was conducted on a hypothetical small modular reactor (SMR) facility, and only considered a potential adversary cyber attack on the PPS to a physical attack on the hypothetical facility to achieve a radiological release. This cyber-physical TTX is meant to provide lessons learned to integrate the cyber security system design and the physical protection system (PPS) design to decrease design, operation, and maintenance costs as well as increase effectiveness for defending against design basis threat attacks at the facility. This TTX will also provide a framework and method for SMR and microreactor vendors to conduct their own cyber-physical TTX and gain impactful insights to improving the cyber and physical protection system design for their SMR or microreactor facility design.

42 ENGINEERING↗
DOE OSTI2021

The development of a Sustained High Power Density (SHPD) facility

A Sustained High Power Density (SHPD) facility is being planned within the U.S. as a non DT device which can be a bridge to a Compact Fusion Pilot Plant (CFPP) DT operated facility with the future capacity to produce a levelised cost of electricity at a rate competitive with current power generating systems. To accomplish this a review of ITER cost data and recent next step physics/engineering studies that promoted technology development, physics enhancements and innovated configuration improvements that increase operating availability have been injected into a proposed PPPL SHPD design. Much of the physics and engineering design basis for this effort centers on a series of ST studies carried out in recent years attempting to maximize the low aspect ratio (AR) ST performance within realistic engineering constraints of this compact fusion device. One issue found in the low AR design is the lack of volt-seconds generated by a small central solenoid that resides in the available space left after sizing the TF structure to support the high current density/HTS TF coil. Although not required in a larger pilot plant device, a solenoid wrapped around the plasma side of the SC TF coil was considered for the small low aspect ratio SHPD device. Adding a solenoid to the plasma side of the TF imposes a major change to the device configuration – a change which has been found to have cost and assembly issues when compared with an OH solenoid located at the machine center. Finally, the inclusion of a liquid metal first wall divertor system, the integration of local pairs of outboard DCLL blanket segments and a summary of physics performance conditions will be presented.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Pilot-Scale Algal Oil Production

The main objective of the project is complete: development of a preliminary planning and design for a pilot-scale algal oil cultivation and processing facility, FEL-3 design with -5% / +15% cost estimate accuracy, and conversion of the algal oil to biofuel in an off-site existing bio-oil refinery. The design basis includes 10 tons per day of dried algae cultivated with CO 2 supplied by direct air capture, electricity supplied solar power, well water supply, zero liquid discharge, off-site extraction, and off-site conversion of oil to biofuel. The design and permitting package is an important milestone in the path to commercialization of algal biofuels and bioproducts as it provides the preliminary design, permitting path, long-term land lease, planning documents, and team needed for success in future engineering, construction, start-up and operations of a pilot-scale farm at a site in Paso Robles, CA. Outcomes of the business assessment include (i) identification of a product spectrum for economical algal biofuels using co-products with markets that are commensurate with production of 6-7 billion gallons per year of sustainable aviation fuel (SAF), renewable diesel, and renewable gasoline, (ii) a path toward near-term contribution of algae oil to SAF, and (iii) an approach for long-term operation of a pilot-scale farm.

09 BIOMASS FUELS↗
DOE OSTI2016

Advanced Nuclear Technology. Using Technology for Small Modular Reactor Staff Optimization, Improved Effectiveness, and Cost Containment, 3002007071

Through this grant, the U.S. Department of Energy (DOE) will review several functional areas within a nuclear power plant, including fire protection, operations and operations support, refueling, training, procurement, maintenance, site engineering, and others. Several functional areas need to be examined since there appears to be no single staffing area or approach that alone has the potential for significant staff optimization at new nuclear power plants. Several of the functional areas will require a review of technology options such as automation, remote monitoring, fleet wide monitoring, new and specialized instrumentation, human factors engineering, risk informed analysis and PRAs, component and system condition monitoring and reporting, just in time training, electronic and automated procedures, electronic tools for configuration management and license and design basis information, etc., that may be applied to support optimization. Additionally, the project will require a review key regulatory issues that affect staffing and could be optimized with additional technology input. Opportunities to further optimize staffing levels and staffing functions by selection of design attributes of physical systems and structures need also be identified. A goal of this project is to develop a prioritized assessment of the functional areas, and R&D actions needed for those functional areas, to provide the best optimization

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2020

Modernization of Technical Requirements for Licensing of Advanced Non-Light Water Reactors: Risk-Informed and Performance-Based Evaluation of Defense-in-Depth Adequacy

This document supports the work contained in Nuclear Energy Institute (NEI) 18-04 “Risk-Informed Performance-Based Technology Inclusive Guidance for Advanced Reactor Licensing Basis Development” Revision 0. NEI 18-04 presents a modern, technology-inclusive, risk-informed, and performance-based (TI-RIPB) process for selection of Licensing Basis Events (LBEs); safety classification of structures, systems, and components (SSCs) and associated risk-informed special treatments; and determination of defense-in-depth (DID) adequacy for non-LWRs. The NEI guidance document provides one acceptable means for addressing the aforementioned topics as part of demonstrating a specific design provides reasonable assurance of adequate radiological protection. This report provides the framework and associated methodology guidelines and discussion for establishing, then evaluating, confirming, and documenting the adequacy of defense-in-depth (DID) for advanced non-light-water reactor technologies. It was developed as part of the Licensing Modernization Project led by Southern Company and cost-shared by the United States Department of Energy and has benefited from considerable NRC formal reviews and public workshops. The methodology converts the DID philosophy into a structured process that is implementable, embraces existing United States and international definitions and philosophies of DID that set the foundation for the process. It builds on the DID framework developed in the Department of Energy Next Generation Nuclear Plant Project and earlier works on this subject. The approach to establishing DID adequacy involves the incorporation of DID attributes into the plant capabilities and programmatic elements of DID. The integrated evaluation of DID adequacy includes both quantitative elements to incorporate risk-informed and performance-based (RIPB) considerations and qualitative elements that address uncertainties and limitations in the quantitative models and supporting data. Demonstration of DID adequacy ensures that there are multiple layers of defense for risk-significant challenges to the design and that the plant capabilities and programs that support each layer are provided in a manner that minimizes dependencies among these layers. The focus of this report is assurance of DID adequacy with respect to protection of the public from radiological exposures resulting from accidental releases of radioactive material. While other hazards are not specifically addressed, this methodology is expected to be beneficial for determining DID adequacy for them as well. Risk-informed evaluation of DID considers the integrated performance of all plant SSCs and associated programs to manage daily operational activities, transients, and accidents, including the evaluation of strategies for accident prevention and mitigation. The RIPB LBE scenario methodology used in this evaluation defines the challenges to the plant safety features included in the plant design basis and beyond, and the scope of all deterministic and probabilistic safety evaluations. By examining event sequences across the whole spectrum of LBEs, a systematic assessment of DID can be accomplished.

42 ENGINEERING↗
DOE OSTI2010

Idaho National Laboratory Materials and Fuels Complex Natural Phenomena Hazards Flood Assessment

This report presents the results of flood hazards analyses performed for the Materials and Fuels Complex (MFC) and the adjacent Transient Reactor Experiment and Test Facility (TREAT) located at Idaho National Laboratory. The requirements of these analyses are provided in the U.S. Department of Energy Order 420.1B and supporting Department of Energy (DOE) Natural Phenomenon Hazard standards. The flood hazards analyses were performed by Battelle Energy Alliance and Pacific Northwest National Laboratory. The analyses addressed the following: • Determination of the design basis flood (DBFL) • Evaluation of the DBFL versus the Critical Flood Elevations (CFEs) for critical existing structures, systems, and components (SSCs).

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2015

Status Report on the Water-Based NSTF RCCS Capability: Preparations and Design for the Transformation of the Natural Convection Shutdown Heat Removal Test Facility (NSTF) From Air to Water-based Cooling

The following report provides an overview of the current status of the water-based transformation, including details of the design, geometry, and fabrication of the test section and network piping, for the ½ scale water-based NSTF program at Argonne National Laboratory. In early FY2015, efforts towards a water-based design began in parallel with on-going air-based work, and began by identifying top level objectives of the transformation. Scaling studies and preparation tasks with computational modeling were carried out to guide the design decisions, minimize distortions, and ensure relevant data among scales. This entailed close collaboration with AREVA, whose water-based RCCS (included as part of their 625 MWt SC-HTGR) served as the primary design basis for incorporation into the NSTF. A literature review of previous works was made, with an emphasis placed on similar facilities, relevance to passive decay heat removal, and details pertaining to two-phase flow and measurement techniques. The study focused on input from key designers and a thermal analysis of existing cooling panels (e.g. at TAMU and UW-Madison) and two proposed AREVA cooling panels. A base case was used as a reference point for parametric studies of effects from varying tube spacing (pitch), tube diameter, fin thickness, and materials. Then, a structural analysis was performed to ensure safe material stresses during high temperature operation. Details of the pipe network geometry and guidelines for design of the water storage tank are finally presented, followed by engineering drawings in the appendix. The results of this study have identified a suitable configuration to support both the ANL/DOE project goals and the DOE vision to provide AREVA with data suitable for characterizing the RCCS of their full scale HTGR design. The dimensions were not selected for optimal RCCS performance, but instead to serve as a representative yet bounding configuration for future implementation into a full scale design. Scaling distortions are unavoidable; however, they can be accurately predicted based on earlier works of derived similarity solutions. Flexibility remained a primary design philosophy, and will allow the test facility to easily accommodate future alterations. The final design for the test section and network piping is as follows: riser tubes: 1.5” Schedule 160, 5.91” (150-mm) pitch, 304L stainless, heat transfer panels: 5/16” plates, 4.01” (102-mm) width, full penetration weld to risers, 304L stainless, split into 4 panels per riser separated by horizontal 1/8” gaps, test section: Eight (x8) riser tubes and nine (x9) heat transfer panels, fabricated into banks of two (x2) riser tubes and three (x3) fins, joined to form single section, and network geometry: 4.0” Schedule 40, 304L stainless. Details of the water storage tank will reflect competing effects such as thermal hydraulic phenomena during single and two-phase operation. These considerations include mixing during single-phase jet penetration, condensation of rising bubbles during two-phase discharge, bubble entrainment into the liquid outlet, and liquid carry-over into the steam outlet.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Multinational underground nuclear parks

Newcomer countries expected to develop new nuclear power programs by 2030 are being encouraged by the International Atomic Energy Agency to explore the use of shared facilities for spent fuel storage and geologic disposal. Multinational underground nuclear parks (M-UNPs) are an option for sharing such facilities. Newcomer countries with suitable bedrock conditions could volunteer to host M-UNPs. M-UNPs would include back-end fuel cycle facilities, in open or closed fuel cycle configurations, with sufficient capacity to enable M-UNP host countries to provide for-fee waste management services to partner countries, and to manage waste from the M-UNP power reactors. M-UNP potential advantages include: the option for decades of spent fuel storage; fuel-cycle policy flexibility; increased proliferation resistance; high margin of physical security against attack; and high margin of containment capability in the event of beyond-design-basis accidents, thereby reducing the risk of Fukushima-like radiological contamination of surface lands. A hypothetical M-UNP in crystalline rock with facilities for small modular reactors, spent fuel storage, reprocessing, and geologic disposal is described using a room-and-pillar reference-design cavern. Underground construction cost is judged tractable through use of modern excavation technology and careful site selection. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

V31 Test Report

The V31 containment vessel was procured by the US Army Recovered Chemical Material Directorate (RCMD) as a third - generation EDS containment vessel. It is the fifth EDS vessel to be fabricated under Code Case 2564 of the 2019 ASME Boiler and Pressure Vessel Code, which provides rules for the design of impulsively loaded vessels. The explosive rating for the vessel, based on the code case, is twenty-four (24) pounds TNT - equivalent for up to 1092 detonations. This report documents the results of explosive tests that were performed on the vessel at Sandia National Laboratories in Albuquerque, New Mexico to qualify the vessel for field operations use. There were three design basis configurations for qualification testing. Qualification test (1) consisted of a simulated M55 rocket motor and warhead assembly of 24lbs of Composition C-4 (30 lb TNT equivalent). This test was considered the maximum load case, based on modeling and simulation methods performed by Sandia prior to the vessel design phase. Qualification test (2) consisted of a regular, right circular cylinder, unitary charge, located central to the vessel interior of 19.2 lb of Composition C-4 (24 lb TNT equivalent). Qualification test (3) consisted of a 12-pack of regular, right circular cylinders of 2 lb each, distributed evenly inside the vessel (totaling 19.2 lb of C-4, or 24 lb TNT equivalent). All vessel acceptance criteria were met.

42 ENGINEERING↗
DOE OSTI2023

V31 Test Report

The V31 containment vessel was procured by the US Army Recovered Chemical Materiel Directorate (RCMD) as a third-generation EDS containment vessel. It is the fifth EDS vessel to be fabricated under Code Case 2564 of the 2019 ASME Boiler and Pressure Vessel Code, which provides rules for the design of impulsively loaded vessels. The explosive rating for the vessel, based on the code case, is twenty-four (24) pounds TNT-equivalent for up to 1092 detonations. This report documents the results of explosive tests that were performed on the vessel at Sandia National Laboratories in Albuquerque, New Mexico to qualify the vessel for field operations use. There were three design basis configurations for qualification testing. Qualification test (1) consisted of a simulated M55 rocket motor and warhead assembly of 24lbs of Composition C-4 (30 lb TNT equivalent). This test was considered the maximum load case, based on modeling and simulation methods performed by Sandia prior to the vessel design phase. Qualification test (2) consisted of a regular, right circular cylinder, unitary charge, located central to the vessel interior of 19.2 lb of Composition C-4 (24 lb TNT equivalent). Qualification test (3) consisted of a 12-pack of regular, right circular cylinders of 2 lb each, distributed evenly inside the vessel (totaling 19.2 lb of C-4, or 24 lb TNT equivalent). All vessel acceptance criteria were met .

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2010

Commerical-Scale CO2 Capture and Sequestration for the Cement Industry

On June 8, 2009, DOE issued Funding Opportunity Announcement (FOA) Number DE-FOA-000015 seeking proposals to capture and sequester carbon dioxide from industrial sources. This FOA called for what was essentially a two-tier selection process. A number of projects would receive awards to conduct front-end engineering and design (FEED) studies as Phase I. Those project sponsors selected would be required to apply for Phase II, which would be the full design, construction, and operation of their proposed technology. Over forty proposals were received, and ten were awarded Phase I Cooperative Agreements. One of those proposers was CEMEX. CEMEX proposed to capture and sequester carbon dioxide (CO2) from one of their existing cement plants and either sequester the CO2 in a geologic formation or use it for enhanced oil recovery. The project consisted of evaluating their plants to identify the plant best suited for the demonstration, identify the best available capture technology, and prepare a design basis. The project also included evaluation of the storage or sequestration options in the vicinity of the selected plant.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2023

Advanced Modeling and Simulation to Characterize Advanced Boiling Water Reactor Source Terms to Support a Regulatory Approval Pathway for Right-Sized Emergency Planning Zone. Advanced Boiling Water Reactor (A-BWR) Project Phase 2

The United States Department of Energy (US DOE) is currently supporting the development of various small modular reactor (SMR), microreactor and advanced reactor designs. These reactors have improved safety features as compared to conventional Large Light Water Reactors (LLWRs), which include features that improve potential reduction of radiological source terms in the event of design and beyond-design basis accidents. Specifically, some reactors feature a relatively smaller containment volume with respect to the available surface area for fission product deposition; these include Integrated Pressurized Water Reactors (iPWRs), such as the NuScale SMR and SMR-type Boiling Water Reactors (BWRs) such as General Electric-Hitachi’s BWRX-300 design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Source Term Reduction for Advanced and Small Modular Boiling Water Reactors

The United States Department of Energy (US DOE) is currently supporting the development of various advanced and small modular reactor (SMR) designs. Several of these designs have commenced license application with the US Nuclear Regulatory Commission (US NRC). These reactors have improved safety features that may significantly reduce radiological source terms in the event of design and beyond-design basis accidents. Specifically, some reactors feature a smaller containment volume relative to the available fission product depositional surface area, which supports increased fission product retention in the containment vessel. Pressurized water reactors (PWR) and boiling water reactors (BWR) with this feature include the integral pressurized water reactor (iPWR) and the BWRX-300 design by General Electric. A prior research program supported by the US DOE quantified the source term reduction associated with light water iPWRs and developed iPWR-specific theoretical models for fission product deposition rates. This program included a sequence of research projects that started with a feasibility study, development of theoretical models that predict higher deposition rates, and finally, development of empirical data for verification and validation of the theoretical models. The current project, which is a feasibility study, is the first step in a similar program to quantify the source term reduction associated with small and advanced light water BWRs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

WIRE-21 Sensor Irradiation Experiment Ready for HFIR Insertion

The ability to deploy new nuclear fuels for current or future reactor concepts requires a wealth of data regarding fuel performance during normal operation, anticipated operational occurrences, and design-basis accidents. Most of these data have historically been collected during experiments in materials test reactors, ideally with online instrumentation to collect as much data as possible. However, advanced instrumentation could also allow for in situ monitoring of fuel operating conditions during commercial reactor operation to maximize fuel utilization, reduce unnecessary conservativism in design margins, and improve operator understanding of limiting peaking factors. The latter approach would complicate fuel handling, particularly during refueling, unless the instrumentation could be placed inside the fuel rods and transmitted wirelessly to a receiver located outside the fuel’s primary pressure boundary. To this end, Westinghouse Electric Company (WEC) developed wireless sensors based on inductive coupling that can transmit information regarding fuel centerline temperatures and rod internal pressures wirelessly from within a fuel rod to a nearby instrument thimble. After testing these sensors in lower-power university research reactors, the next step is to perform high neutron fluence testing to characterize the performance of these wireless sensors under conditions that are more representative of the intended application—in this case, light-water reactors (LWRs). The removable Be (RB) positions of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) provide the neutron flux, experiment volume, and access to instrument leads required to achieve these sensor testing goals. This report summarizes the design, analysis, and assembly of the Wireless Instrumented RB Experiment 2021 (WIRE-21). This is the most highly instrumented irradiation experiment ever performed in HFIR. The experiment will use seven different sensing techniques to measure temperature, pressure, neutron flux, and neutron fluence during reactor operation. In addition to WEC’s wireless temperature and pressure sensors, WIRE-21 includes an array of thermocouples, self-powered neutron detectors, spatially distributed fiber optic temperature sensors, passive SiC temperature monitors, and flux wires. The design of WIRE-21 and the cabling that was installed in HFIR also provide the infrastructure to enable accelerated, economical testing of advanced sensor technologies while leveraging the extremely high neutron flux that is available in HFIR. The containment for WIRE-21 is similar to previous RB irradiation vehicles but includes a few modifications, most notably the use of integrated compression seals to pass a larger number of sensor leads through the experiment’s pressure boundary. In addition to the sensor leads, inert gas lines are passed into the experiment to enable active temperature control and the ability to pneumatically actuate a bellows-driven pressure sensor. WIRE-21 is targeting component temperatures (300–350°C) and neutron fluence levels (~10 22 n/cm 2 ) that would be expected in the plenum region of LWR fuels, except for the active sensing region of the wireless temperature sensor, which is targeting LWR fuel centerline temperatures (~800–1,100°C). WIRE-21 was successfully assembled, passed all nondestructive examination, and was delivered to HFIR for insertion during upcoming cycle 498 (April 2022).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

Augmenting Probabilistic Risk Assesment with Malevolent Initiators

As commonly practiced, the use of probabilistic risk assessment (PRA) in nuclear power plants only considers accident initiators such as natural hazards, equipment failures, and human error. Malevolent initiators are ignored in PRA, but are considered the domain of physical security, which uses vulnerability assessment based on an officially specified threat (design basis threat). This paper explores the implications of augmenting and extending existing PRA models by considering new and modified scenarios resulting from malevolent initiators. Teaming the augmented PRA models with conventional vulnerability assessments can cost-effectively enhance security of a nuclear power plant. This methodology is useful for operating plants, as well as in the design of new plants. For the methodology, we have proposed an approach that builds on and extends the practice of PRA for nuclear power plants for security-related issues. Rather than only considering 'random' failures, we demonstrated a framework that is able to represent and model malevolent initiating events and associated plant impacts.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗