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At least 379 records · Page 21
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

The Interface of Safety and Security in Transport: A Regional Perspective

Security of nuclear and other radioactive material in transport continues to be a challenge for States that are working on strengthening their nuclear security regime. One reason for this is that State regulatory agencies and other organizations lack the resources and trained personnel to dedicate to this field. For over 50 years safety has been a major focus in the use, storage and transport of radioactive material. Only recently, since the late 1990s, has dedicated focus been given to the field of security. One way to assist States to advance nuclear security is to reach out to safety workers (regulators, inspectors, and safety compliance personnel) and showcase the need to better integrate safety and security practices. A recent IAEA regional workshop in Bangkok, Thailand (June 2015) yielded profound results when subject matter experts lectured on both the safety and the security of radioactive material in transport. These experts presented and discussed experiences and best practices for: 1) developing and implementing safety requirements and security recommendations for radioactive material in transport; 2) national and international cooperation; and 3) preventing shipment delays/denials of radioactive material. The workshop participants, who were predominantly from safety organizations, shared that they received the following from this event: 1. A clear understanding of the objectives of the IAEA safety requirements and security recommendations for radioactive material in transport. 2. A general understanding of and appreciation for the similarities and differences between safety requirements and security recommendations for radioactive material in transport. 3. A greater appreciation of the interface between transport safety and security and potential impacts of this interface on the efforts to strengthen the compliance assurance regime for the safe transport of radioactive material. 4. A general understanding of assessing the transport security scenarios and developing transport security plans. Many participants also reported their appreciation of the workshop exercises that specifically focused on practical aspects of safety and security of transport of radioactive material. These workshop outcomes highlight the important role professionals can offer when they receive additional safety training and education for radioactive material in transport. Moreover, these professionals can help to increase capacity in countries with developing nuclear security regimes. This paper explores workshop outcomes and transportation regulations and guidelines for radioactive material.

Welch, Tim↗
DOE OSTI2013

Lessons Learned in Over a Decade of Technical Support for U.S. Nuclear Cyber Security Programmes

Pacific Northwest National Laboratory’s (PNNL) nuclear cyber security team has been providing technical support to the U.S. Nuclear Regulatory Commission (NRC) since 2002. This team has provided cyber security technical experties in conducting cyber security inspections, developing of regulatory rules and guidance, reviewing facility cyber security plans, developing inspection guidance, and developing and teaching NRC inspectors how to conduct cyber security assessments. The extensive experience the PNNL team has gathered has allowed them to compile a lenghty list of recommendations on how to improve cyber security programs and conduct assessments. A selected set of recommendations are presented, including the need to: integrate an array of defenisve strategies into a facility’s cyber security program, coordinate physical and cyber security activities, train phycial security forces to resist a cyber-enabled physical attack, improve estimates of the consequences of a cyber attack, properly resource cyber security assessments, appropropriately account for insider threats, routinely monitor security devices for potential attacks, supplement compliance-based requirements with risk-based decision making, and introduce the concept of resilience into cyber security programs.

Glantz, Clifford S.↗
DOE OSTI2017

Enhanced Data Authentication System v. 2.0

EDAS is a system, comprised on hardware and software, that plugs in to an existing data stream, and branches all data for transmission to a secondary observer computer. The EDAS Junction box, which inserts into the data stream, has Java software that forms these data into packets, digitally signs, encrypts, and sends these packets to a safeguards inspector computer. Further, there is a second Java program running on the secondary observer computer that receives data from the EDAS Junction Box to decrypt, authenticate, and store incoming packets. Also, there is a stand-alone Java program that is used to configure the EDAS Junction Box.

Thomas, Maikael A.↗
DOE OSTI2016

Monte Carlo Uncertainty Quantification for an Unattended Enrichment Monitor

As a case study for uncertainty analysis, we consider a model flow monitor for measuring enrichment in gas centrifuge enrichment plants (GCEPs) that could provide continuous monitoring of all declared gas flow and provide high-accuracy gas enrichment estimates as a function of time. The monitor system could include NaI(Tl) gamma-ray spectrometers, a pressure signal-sharing device to be installed on an operator\rq{}s pressure gauge or a dedicated inspector pressure sensor, and temperature sensors attached to the outside of the header pipe, to provide pressure, temperature, and gamma-ray spectra measurements of UF$_6$ gas flow through unit header pipes. Our study builds on previous modeling and analysis methods development for enrichment monitor concepts and a software tool that was developed at Oak Ridge National Laboratory to generate and analyze synthetic data.

international safeguards↗
DOE OSTI2018

Safeguards by Design in Nuclear Engineering Design Courses: Engaging Universities in Safeguards by Design

Engaging professionals across the breadth of the nuclear industry to educate them on international safeguards objectives and methodology is logistically challenging, especially in the United States, where individuals may not interact with the IAEA and are not as familiar with implementing international safeguards. However, incorporating international safeguards concepts into the mandatory curriculum of nuclear-related university degree programs around the world, complementary to nuclear safety and security concepts, could reach individuals who will go on to be professionals acting in a variety of roles across the nuclear industry (e.g., developing nuclear-related advanced technologies, designing or operating nuclear facilities, serving as State regulators). This would ultimately lead to more effective and efficient safeguards, including helping to move safeguards by design from a concept to a standard practice, as future designers, operators, and State regulators would be more familiar with international safeguards. This could also potentially address human resource challenges by creating a broader base of individuals knowledgeable about the IAEA mission from which to draw future inspectors and analysts. In order to implement this idea, the Y-12 National Security Complex in Oak Ridge, Tennessee, in conjunction with the Center for Nuclear Security Science & Policy Initiatives (a research institute associated with Texas A&M University), targeted nuclear engineering undergraduate students taking the mandatory design course at Texas A&M and the University of Tennessee. The project objective was to introduce safeguards by design concepts to individuals who are planning to work in a broad range of nuclear-related careers. The paper describes the project methodology, which included the project leads creating and presenting lectures on safeguards to the design class and mentoring teams working on design team projects that incorporated safeguards. The paper also identifies lessons learned while incorporating safeguards topics into nuclear engineering university degree programs.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2014

Determining Remaining Useful Life of Aging Cables in Nuclear Power Plants using Non-destructive Evaluation

License renewal of existing nuclear power plants (NPPs) to extend operation from 40 to 60 to 80 years will require effective aging management programs for the 100’s of kilometers of electric power, control and instrumentation cables installed in each plant. Degradation of jacket, insulation and other cable components can reduce the ability of cables to withstand design-basis accidents. The groundwork for a research and development roadmap to address aging cable management in NPPs was laid at a July 2012 workshop1 that brought together subject matter experts from the U.S. Nuclear Regulatory Commission (NRC), the U.S. Department of Energy (DOE) national laboratories, the Electric Power Research Institute (EPRI), universities, and cable manufacturers and inspectors. The most common nondestructive evaluation (NDE) methods for gauging cable status include indentation and visual inspection. These methods are limited to evaluation of the surface easily accessed cable assemblies. Other current methods (such as time and frequency domain analysis, inductance and capacitance measurements, tan d, etc.) provide flaw detection, but are not predictive. The workshop identified the need to advance and develop new NDE methods to enable in-situ cable condition assessment. Key chemical, physical, and electrical indicators of cable aging with exposure to heat, radiation, moisture, chemicals, and mechanical stress need to be determined. Finally, models correlating the key cable indicators with NDE signals need to be developed that can aging cable remaining useful life.

NDE, Nuclear Cables, Polymer Aging↗
DOE OSTI2019

Hafnium Isotope Ratio Method for Enhanced Safeguards at Research Reactors

Pacific Northwest National Laboratory (PNNL) is developing a Hafnium (Hf)-based, enhanced safeguards measure for the verification of energy production in a research reactor. Historically, verification of nuclear reactor use has been based on extracting and measuring physical samples from non-fuel components of the reactor. The Hafnium Isotope Ratio Method for Enhanced Safeguards (HIRMES) provides a means of verifying reactor operation based upon measuring the changes in isotopic ratios in purpose-built coupons. The first activity is installing a purpose-built, integrated sample collection coupon containing hafnium which undergoes isotopic changes when irradiated in the reactor. The second activity is analysis of the isotopic changes of the irradiated coupon after removal of the coupon using the isotope ratio method (IRM). The energy, or plutonium production, in a fission reactor can be estimated using IRM by measuring and correlating stable isotope ratios in non-fuel reactor components using standard reactor modelling methods. This paper presents HIRMES, a technology intended to provide an inexpensive, accurate, and robust measurement capability for validating operator declarations at low-power research reactors. Deploying HIRMES in research reactors would provide the IAEA with an in-core reactor “odometer” on total energy production with higher accuracy than other methods. HIRMES could be integrated in new fuel or reactor core designs as part of a protocol that is simple to implement and minimizes effort required of the operator or inspector. This paper also discusses early demonstrations at U.S. research reactors to-date, which demonstrate HIRMES as a practical tool for independent verification of low-power research reactor utilization. Specifically, the measurement uncertainty during an irradiation campaign of 39766 MWh was 22 MWh, representing two hours at a 10 MW reactor for a total irradiation campaign over the course of approximately six months.

research reactor safeguards, isotope ratio method↗
DOE OSTI2020

In-Field Alpha Spectrometry to Assess Uranium Enrichment In Uranium Hexafluoride

A new In-Field Alpha Spectrometry (IFAS) system is being developed to allow quick turnaround measurements of uranium enrichment in uranium hexafluoride (UF6) samples. Samples are collected using specially-designed Single-Use Destructive Assay (SUDA) wafers which have an absorptive film of zeolite deposited onto a quartz or metal substrate. The SUDA samples, along with a special holder, are designed to be attached directly to a sampling tap at a gas centrifuge enrichment plant or a uranium conversion facility, allowing gaseous UF6 to come into direct contact with the zeolite. The zeolite absorbs this material and converts it into stable, safe UO2F2 within a few minutes. Once the sample has been collected it is transferred to the IFAS system where a solid-state, silicon alpha spectrometer is used to quantitatively measure the alpha emissions from 234U, 235U, and 238U in the sample. The IFAS system is a small, light-weight device that can be quickly turned on and put into use. The SUDA samples and IFAS hardware are designed so that thin-film spectra are produced, allowing a complete measurement with a final 235U:Utotal determination having a precision <±2% to be made in less than eight hours. This short time frame will allow an inspector to get results on-site within one working shift; further innovations may reduce this to even shorter times.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Validation of ORIGEN for VVER-440 Spent Fuel with Application to Fork Detector Safeguards Measurements

The US Department of Energy National Nuclear Security Administration and the European Atomic Energy Community (Euratom) are collaborating with the Radiation and Nuclear Safety Authority in Finland to assess spent fuel verification methods for potential application to measurements at the planned spent fuel encapsulation plant and geological repository prior to final disposal in Finland. The fork detector (FDET) used to measure the neutron and gamma radiation from a given spent fuel assembly is an existing technology widely used by the International Atomic Energy Agency (IAEA) and Euratom for the safeguards verification of spent fuel operator declaration. Recently, an FDET data analysis software, referred to as the ORIGEN module, was developed and incorporated into the IAEA/Euratom Integrated Review and Analysis Program. This module uses the ORIGEN burnup code to calculate the nuclide concentrations and the neutron and gamma source terms in the spent fuel based on operator declarations. Then, to predict the expected FDET signals for a given fuel assembly, the module combines the source terms calculated by ORIGEN with FDET response functions that were pre-generated using MCNP models. The inspector can compare the calculated signals to the measured values in real time to identify possible anomalies in fuel assembly’s operator declaration, integrity, or measurement. In this work, the capability of the ORIGEN module was extended from application to typical light water reactor fuels to use for VVER-440 assembly types. The accuracy of ORIGEN for this type of fuel assembly calculations was assessed by comparing calculated nuclide concentrations against destructive assay measurements for VVER-440 spent fuel. The performance of the ORIGEN module for FDET safeguards verification in routine inspections was assessed using FDET measurement data for 13 spent VVER-440 assemblies that were measured at the Loviisa Nuclear Power Plant in Finland, by comparing the calculated signals to the measured quantities. The results show that the calculated FDET neutron and gamma detector signals are generally within 12% of the measurements except for one assembly. Thus, these results are applicable to future safeguards verification in the planned Finland encapsulation plant, such as for passive neutron albedo reactivity safeguards measurements.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2021

Optimization Of In-field Alpha Spectrometry For Uranium Enrichment Determination In Uranium Hexafluoride

In response to needs identified by the International Atomic Energy Agency (IAEA) research is underway to develop In-Field Alpha Spectrometry (IFAS) as a method to allow IAEA safeguards inspectors to collect samples of uranium hexafluoride (UF6) at processing facilities to assess and verify uranium enrichment. For sample collection, the IFAS method uses Single-Use Destructive Assay (SUDA) samplers, which contain thin zeolite coatings that trap UF6 gas and convert it to the safer, more stable form uranyl fluoride (as a dihydrate, UO2F2&middot;2H2O). For alpha spectrometry, the IFAS instrument employs a large area silicon semiconductor transducer to detect and record alpha particle energy-deposition events. Over the past year optimization work has significantly increased the diameter of useful SUDA samples (from 12.7 mm to 48 mm), improved the manufacturability and reproducibility of SUDA samples, increased the area of the IFAS alpha spectrometer sensor from 1.2 cm to 3.1 cm, and improved source positioning within the IFAS. This paper will report on this optimization work, its impacts on IFAS performance, and future plans for IFAS miniaturization, improvements, and testing.

Chichester, David↗
DOE OSTI2021

2021 ROMANIA TRANSPORT SECURITY TABLE-TOP EXERCISE (TTX) – LESSONS LEARNED AND GOOD PRACTICES

During the period of 15-18 March 2021, the IAEA conducted a national workshop addressing the security of radioactive material in transport. The workshop was jointly hosted by the Romanian regulatory authority and the General Inspectorate of Romanian Gendarmerie. The workshop assisted Romania in improving its capability to evaluate the ability of response forces to counter a malicious act against radioactive material in transport. The event consisted of briefings, exercises, a demonstration of transport security equipment, and a transport security table-top exercise (TTX). The workshop was conducted in a virtual-hybrid format in such a way as to meet the COVID-19 related requirements in compliance with Romania’s guidelines. This paper provides an overview of the workshop’s activities, lessons-learned, and good practices related to the transport security TTX exercises.

Maican, Catalin↗
DOE OSTI2017

IAEA Domestic Inspections in the US

The Voluntary Offer Agreement between the United States and the IAEA for the implementation of IAEA safeguards in the United States, INFCIRC/288, entered into force in 1980. Under this Agreement, the United States provides to the IAEA an Eligible Facilities List, which lists facilities eligible for IAEA inspection and verification of compliance through either a reporting-only process or for full-scope safeguards (to include inspections). Approximately 300 facilities are currently listed on the Eligible Facilities List. Since 1981, the IAEA has conducted over 700 inspections on ~20 of these facilities. Currently, four NRC-licensed facilities have been selected under the Initial (Reporting) Protocol, and one DOE facility, KAMS at the SRS, has been selected for full-scope safeguards. These IAEA inspections are documented in the ID/288 database managed by staff at ORNL and the Y-12 National Security Complex for the NNSA Office of Nonproliferation and Arms Control. This database also maintains records of US designated inspectors, the status of indices checks, and inspection schedules. The database was most recently updated in FY 2014 to include information related to the IAEA visit to URENCO-USA in New Mexico and will be upgraded again via proposal in FY 2017 to include the Small Quantities Protocol (SQP) inspections in the US Caribbean territories, associated with INFCIRC/366. During the FY 2017 upgrade, the system will also be overhauled to allow for quicker updates and to improve performance using a modern software package.

White-Horton, Jessica↗
DOE OSTI2023

WAP Workforce Needs Analysis

As identified through stakeholder dialogue, Weatherization Assistance Program (WAP) Grantees and Subgrantees face challenges in hiring and retaining field staff in the following positions: Retrofit Installers/Technicians (RITs), Crew Leaders (CLs), Energy Auditors (EAs), and Quality Control Inspectors (QCIs). To gain a baseline understanding of workforce needs, NREL collected voluntary workforce data from May 4-June 30, 2022. This presentation includes a qualitative analysis of this data, key takeaways, and an introduction to new tools supporting WAP workforce development.

ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2023

Photovoltaic (PV) Quick Reference Guide

This Photovoltaic (PV) and Energy Storage System (EES) Quick Reference Guide is for inspectors inspecting projects permitted with SolarAPP+.

ENERGY PLANNING, POLICY, AND ECONOMY,SOLAR ENERGY↗
DOE OSTI2023

Direct analysis of cotton swipes for uranium and plutonium isotopic determination by microextraction-ICP-MS

The determination of uranium and plutonium isotopic abundance on environmental samples collected by International Atomic Energy Agency (IAEA) inspectors is vital for the detection of undeclared nuclear activities and material under the international nuclear safeguards regime. Current analytical protocols require time-consuming sample preparation steps prior to subsequent measurement by inorganic mass spectrometry (MS). Recent efforts from this laboratory have focused on developing sample preparation methods for faster analysis, potentially allowing higher sample throughput[1]. Alternative methods including microextraction sampling in conjunction with inductively coupled plasma-mass spectrometry (ICP-MS) have been recently explored. This methodology, microextraction-ICP-MS, was developed such that uranium and plutonium could be extracted from the swipe surface and directed into the ICP-MS for an in-situ measurement, eliminating the need for swipe ashing and digestion. A commercial off-the-shelf microextraction system was customized with an automated movable XY stage that can be programmed to save sampling locations, allowing for automated rapid sampling of swipe surfaces. Additional efforts have focused on the utilization of collision cell technology to the microextraction ICP-MS method. This would eliminate the need for lengthy column chemistry procedures to purify separated uranium and plutonium fractions before analysis. Here, the extracted U/Pu analyte is measured by reacting the uranium ions with CO2 in the collision cell of an ICP-MS, shifting the uranium to UO+, which will not interfere with the plutonium isotopic determination. The developed method utilizing collision cell – ICP-MS technology has demonstrated the ability to measure plutonium isotope ratios in the presence of high uranium concentration on the transient signal from the microextraction system utilizing certified reference materials from JRC-Geel and the New Brunswick Laboratory Program Office.

Bradley, Veronica↗
DOE OSTI2023

Characterizing Reactor Operations from Realistic Simulated Environmental Samples: Combining High-Performance Computing and Data Analytics

Environmental sampling is a common technique employed by inspectors and facility operators in nuclear safeguards, proliferation detection, and process monitoring contexts. Interpreting measurements performed on samples or collections of samples and ensuring the information extracted is accurate and precise is difficult. To date, these analyses have relied on simulated data to enable systematic studies; however, these models are inherently limited by the fidelity of the models and the implicit spatial averaging of isotopic composition or other signatures of interest. To advance this capability, we have refined the spatial discretization and expanded the range of physics in the simulation codes we use to perform reactor simulations and depletion calculations. This allows us to generate data that are more representative of real environmental samples, especially for the length scale of the isotopic composition and associated variation. Accordingly, these new data allow a more realistic assessment of traditional and new data analytic analysis methods. Here we present motivation for developing reactor simulations using high-performance computing methods and resources, impacts of these new simulations on our assessment of data analysis and interpretation methods, and initial results of developing and systematically testing data analytic methods designed to overcome the challenges expected of real-world samples. We also quantify the performance of these analyses using defensible statistical methods.

Dayman, Ken J.↗
DOE OSTI2023

INTEGRATION OF THE H3D-M400 DETECTOR AND SPOT ROBOT FOR AUTOMATED AREA SURVEY MISSIONS

Office of Nuclear Smuggling Detection and Deterrence (NSDD) is charged to identify and develop technologies to detect, disrupt, and investigate smuggling of radiological and nuclear materials •Border protection involves primary inspections using Radiation Portal Monitors (RPM), complimented with secondary and area survey inspections, usually performed manually using portable radiation detectors (PRDs) -RPM rely on fast technologies which provide quick scans of passing cargo/vehicles -Secondary inspections rely on trained, field deployed staff responsible for additional interdiction of screened cargo/vehicle putting them in potentially hazardous environment -Area survey inspections rely on trained, field deployed staff responsible for scanning and identifying presence of radiation while patrolling through a facility or public space •This project explores the use of technology which could reduce risks to field inspectors and add efficiency •The work presented here is a proof of principle of detector-robot integration to perform area survey inspections

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗