Free Energy and Heat Capacity

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kurata, Masaki; Devanathan, Ramaswami

2015-10-13

Free energy and heat capacity of actinide elements and compounds are important properties for the evaluation of the safety and reliable performance of nuclear fuel. They are essential inputs for models that describe complex phenomena that govern the behaviour of actinide compounds during nuclear fuel fabrication and irradiation. This chapter introduces various experimental methods to measure free energy and heat capacity to serve as inputs for models and to validate computer simulations. This is followed by a discussion of computer simulation of these properties, and recent simulations of thermophysical properties of nuclear fuel are briefly reviewed.

A HISTORICAL PERSPECTIVE OF NUCLEAR THERMAL HYDRAULICS

DOE Office of Scientific and Technical Information (OSTI.GOV)

D’Auria, F; Rohatgi, Upendra S.

The nuclear thermal-hydraulics discipline was developed following the needs for nuclear power plants (NPPs) and, to a more limited extent, research reactors (RR) design and safety. As in all other fields where analytical methods are involved, nuclear thermal-hydraulics took benefit of the development of computers. Thermodynamics, rather than fluid dynamics, is at the basis of the development of nuclear thermal-hydraulics together with the experiments in complex two-phase situations, namely, geometry, high thermal density, and pressure.

10 CFR 1703.112 - Computation of time.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Computation of time. 1703.112 Section 1703.112 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PUBLIC INFORMATION AND REQUESTS § 1703.112 Computation of time. In... until the end of the next working day. Whenever a person has the right or is required to take some...

10 CFR 1703.112 - Computation of time.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 4 2011-01-01 2011-01-01 false Computation of time. 1703.112 Section 1703.112 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PUBLIC INFORMATION AND REQUESTS § 1703.112 Computation of time. In... until the end of the next working day. Whenever a person has the right or is required to take some...

10 CFR 1703.112 - Computation of time.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 4 2013-01-01 2013-01-01 false Computation of time. 1703.112 Section 1703.112 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PUBLIC INFORMATION AND REQUESTS § 1703.112 Computation of time. In... until the end of the next working day. Whenever a person has the right or is required to take some...

10 CFR 1703.112 - Computation of time.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 4 2014-01-01 2014-01-01 false Computation of time. 1703.112 Section 1703.112 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PUBLIC INFORMATION AND REQUESTS § 1703.112 Computation of time. In... until the end of the next working day. Whenever a person has the right or is required to take some...

10 CFR 1703.112 - Computation of time.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 4 2012-01-01 2012-01-01 false Computation of time. 1703.112 Section 1703.112 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PUBLIC INFORMATION AND REQUESTS § 1703.112 Computation of time. In... until the end of the next working day. Whenever a person has the right or is required to take some...

Structural integrity of materials in nuclear service: a bibliography

DOE Office of Scientific and Technical Information (OSTI.GOV)

Heddleson, F.A.

This report contains 679 abstracts from the Nuclear Safety Information Center (NSIC) computer file dated 1973 through 1976 covering material properties with respect to structural integrity. All materials important to the nuclear industry (except concrete) are covered for mechanical properties, chemical properties, corrosion, fracture or failure, radiation damage, creep, cracking, and swelling. Keyword, author, and permuted-title indexes are included for the convenience of the user.

Impact of nuclear data uncertainty on safety calculations for spent nuclear fuel geological disposal

NASA Astrophysics Data System (ADS)

Herrero, J. J.; Rochman, D.; Leray, O.; Vasiliev, A.; Pecchia, M.; Ferroukhi, H.; Caruso, S.

2017-09-01

In the design of a spent nuclear fuel disposal system, one necessary condition is to show that the configuration remains subcritical at time of emplacement but also during long periods covering up to 1,000,000 years. In the context of criticality safety applying burn-up credit, k-eff eigenvalue calculations are affected by nuclear data uncertainty mainly in the burnup calculations simulating reactor operation and in the criticality calculation for the disposal canister loaded with the spent fuel assemblies. The impact of nuclear data uncertainty should be included in the k-eff value estimation to enforce safety. Estimations of the uncertainty in the discharge compositions from the CASMO5 burn-up calculation phase are employed in the final MCNP6 criticality computations for the intact canister configuration; in between, SERPENT2 is employed to get the spent fuel composition along the decay periods. In this paper, nuclear data uncertainty was propagated by Monte Carlo sampling in the burn-up, decay and criticality calculation phases and representative values for fuel operated in a Swiss PWR plant will be presented as an estimation of its impact.

Improved Safety Margin Characterization of Risk from Loss of Offsite Power

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nelson, Paul

Original intent: The original intent of this task was “support of the Risk-Informed Safety Margin Characteristic (RISMC) methodology in order” “to address … efficiency of computation so that more accurate and cost-effective techniques can be used to address safety margin characterizations” (S. M. Hess et al., “Risk-Informed Safety Margin Characterization,” Procs. ICONE17, Brussels, July 2009, CD format). It was intended that “in Task 1 itself this improvement will be directed toward upon the very important issue of Loss of Offsite Power (LOOP) events,” more specifically toward the challenge of efficient computation of the multidimensional nonrecovery integral that has been discussedmore » by many previous contributors to the theory of nuclear safety. It was further envisioned that “three different computational approaches will be explored,” corresponding to the three subtasks listed below; deliverables were tied to the individual subtasks.« less

Computation of Thermodynamic Equilibria Pertinent to Nuclear Materials in Multi-Physics Codes

NASA Astrophysics Data System (ADS)

Piro, Markus Hans Alexander

Nuclear energy plays a vital role in supporting electrical needs and fulfilling commitments to reduce greenhouse gas emissions. Research is a continuing necessity to improve the predictive capabilities of fuel behaviour in order to reduce costs and to meet increasingly stringent safety requirements by the regulator. Moreover, a renewed interest in nuclear energy has given rise to a "nuclear renaissance" and the necessity to design the next generation of reactors. In support of this goal, significant research efforts have been dedicated to the advancement of numerical modelling and computational tools in simulating various physical and chemical phenomena associated with nuclear fuel behaviour. This undertaking in effect is collecting the experience and observations of a past generation of nuclear engineers and scientists in a meaningful way for future design purposes. There is an increasing desire to integrate thermodynamic computations directly into multi-physics nuclear fuel performance and safety codes. A new equilibrium thermodynamic solver is being developed with this matter as a primary objective. This solver is intended to provide thermodynamic material properties and boundary conditions for continuum transport calculations. There are several concerns with the use of existing commercial thermodynamic codes: computational performance; limited capabilities in handling large multi-component systems of interest to the nuclear industry; convenient incorporation into other codes with quality assurance considerations; and, licensing entanglements associated with code distribution. The development of this software in this research is aimed at addressing all of these concerns. The approach taken in this work exploits fundamental principles of equilibrium thermodynamics to simplify the numerical optimization equations. In brief, the chemical potentials of all species and phases in the system are constrained by estimates of the chemical potentials of the system components at each iterative step, and the objective is to minimize the residuals of the mass balance equations. Several numerical advantages are achieved through this simplification. In particular, computational expense is reduced and the rate of convergence is enhanced. Furthermore, the software has demonstrated the ability to solve systems involving as many as 118 component elements. An early version of the code has already been integrated into the Advanced Multi-Physics (AMP) code under development by the Oak Ridge National Laboratory, Los Alamos National Laboratory, Idaho National Laboratory and Argonne National Laboratory. Keywords: Engineering, Nuclear -- 0552, Engineering, Material Science -- 0794, Chemistry, Mathematics -- 0405, Computer Science -- 0984

ASC FY17 Implementation Plan, Rev. 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hamilton, P. G.

The Stockpile Stewardship Program (SSP) is an integrated technical program for maintaining the safety, surety, and reliability of the U.S. nuclear stockpile. The SSP uses nuclear test data, computational modeling and simulation, and experimental facilities to advance understanding of nuclear weapons. It includes stockpile surveillance, experimental research, development and engineering programs, and an appropriately scaled production capability to support stockpile requirements. This integrated national program requires the continued use of experimental facilities and programs, and the computational capabilities to support these programs. The Advanced Simulation and Computing Program (ASC) is a cornerstone of the SSP, providing simulation capabilities and computationalmore » resources that support annual stockpile assessment and certification, study advanced nuclear weapons design and manufacturing processes, analyze accident scenarios and weapons aging, and provide the tools to enable stockpile Life Extension Programs (LEPs) and the resolution of Significant Finding Investigations (SFIs). This requires a balance of resources, including technical staff, hardware, simulation software, and computer science solutions.« less

Performance assessment of KORAT-3D on the ANL IBM-SP computer

DOE Office of Scientific and Technical Information (OSTI.GOV)

Alexeyev, A.V.; Zvenigorodskaya, O.A.; Shagaliev, R.M.

1999-09-01

The TENAR code is currently being developed at the Russian Federal Nuclear Center (VNIIEF) as a coupled dynamics code for the simulation of transients in VVER and RBMK systems and other nuclear systems. The neutronic module in this code system is KORAT-3D. This module is also one of the most computationally intensive components of the code system. A parallel version of KORAT-3D has been implemented to achieve the goal of obtaining transient solutions in reasonable computational time, particularly for RBMK calculations that involve the application of >100,000 nodes. An evaluation of the KORAT-3D code performance was recently undertaken on themore » Argonne National Laboratory (ANL) IBM ScalablePower (SP) parallel computer located in the Mathematics and Computer Science Division of ANL. At the time of the study, the ANL IBM-SP computer had 80 processors. This study was conducted under the auspices of a technical staff exchange program sponsored by the International Nuclear Safety Center (INSC).« less

77 FR 50727 - Configuration Management Plans for Digital Computer Software Used in Safety Systems of Nuclear...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-22

... comments received on or before this date. Although a time limit is given, comments and suggestions in... guides are encouraged at any time. ADDRESSES: You may access information and comment submissions related... Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-251-7494 or email...

Advanced Simulation and Computing Fiscal Year 2016 Implementation Plan, Version 0

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCoy, M.; Archer, B.; Hendrickson, B.

2015-08-27

The Stockpile Stewardship Program (SSP) is an integrated technical program for maintaining the safety, surety, and reliability of the U.S. nuclear stockpile. The SSP uses nuclear test data, computational modeling and simulation, and experimental facilities to advance understanding of nuclear weapons. It includes stockpile surveillance, experimental research, development and engineering programs, and an appropriately scaled production capability to support stockpile requirements. This integrated national program requires the continued use of experimental facilities and programs, and the computational capabilities to support these programs. The purpose of this IP is to outline key work requirements to be performed and to control individualmore » work activities within the scope of work. Contractors may not deviate from this plan without a revised WA or subsequent IP.« less

Current global and Korean issues in radiation safety of nuclear medicine procedures.

PubMed

Song, H C

2016-06-01

In recent years, the management of patient doses in medical imaging has evolved as concern about radiation exposure has increased. Efforts and techniques to reduce radiation doses are focussed not only on the basis of patient safety, but also on the fundamentals of justification and optimisation in cooperation with international organisations such as the International Commission on Radiological Protection, the International Atomic Energy Agency, and the World Health Organization. The Image Gently campaign in children and Image Wisely campaign in adults to lower radiation doses have been initiated in the USA. The European Association of Nuclear Medicine paediatric dosage card, North American consensus guidelines, and Nuclear Medicine Global Initiative have recommended the activities of radiopharmaceuticals that should be administered in children. Diagnostic reference levels (DRLs), developed predominantly in Europe, may be an important tool to manage patient doses. In Korea, overexposure to radiation, even from the use of medical imaging, has become a public issue, particularly since the accident at the Fukushima nuclear power plant. As a result, the Korean Nuclear Safety and Security Commission revised the technical standards for radiation safety management in medical fields. In parallel, DRLs for nuclear medicine procedures have been collected on a nationwide scale. Notice of total effective dose from positron emission tomography-computed tomography for cancer screening has been mandatory since mid-November 2014. © The International Society for Prosthetics and Orthotics.

Test Facilities and Experience on Space Nuclear System Developments at the Kurchatov Institute

NASA Astrophysics Data System (ADS)

Ponomarev-Stepnoi, Nikolai N.; Garin, Vladimir P.; Glushkov, Evgeny S.; Kompaniets, George V.; Kukharkin, Nikolai E.; Madeev, Vicktor G.; Papin, Vladimir K.; Polyakov, Dmitry N.; Stepennov, Boris S.; Tchuniyaev, Yevgeny I.; Tikhonov, Lev Ya.; Uksusov, Yevgeny I.

2004-02-01

The complexity of space fission systems and rigidity of requirement on minimization of weight and dimension characteristics along with the wish to decrease expenditures on their development demand implementation of experimental works which results shall be used in designing, safety substantiation, and licensing procedures. Experimental facilities are intended to solve the following tasks: obtainment of benchmark data for computer code validations, substantiation of design solutions when computational efforts are too expensive, quality control in a production process, and ``iron'' substantiation of criticality safety design solutions for licensing and public relations. The NARCISS and ISKRA critical facilities and unique ORM facility on shielding investigations at the operating OR nuclear research reactor were created in the Kurchatov Institute to solve the mentioned tasks. The range of activities performed at these facilities within the implementation of the previous Russian nuclear power system programs is briefly described in the paper. This experience shall be analyzed in terms of methodological approach to development of future space nuclear systems (this analysis is beyond this paper). Because of the availability of these facilities for experiments, the brief description of their critical assemblies and characteristics is given in this paper.

Sensitivity-Uncertainty Based Nuclear Criticality Safety Validation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Forrest B.

2016-09-20

These are slides from a seminar given to the University of Mexico Nuclear Engineering Department. Whisper is a statistical analysis package developed to support nuclear criticality safety validation. It uses the sensitivity profile data for an application as computed by MCNP6 along with covariance files for the nuclear data to determine a baseline upper-subcritical-limit for the application. Whisper and its associated benchmark files are developed and maintained as part of MCNP6, and will be distributed with all future releases of MCNP6. Although sensitivity-uncertainty methods for NCS validation have been under development for 20 years, continuous-energy Monte Carlo codes such asmore » MCNP could not determine the required adjoint-weighted tallies for sensitivity profiles. The recent introduction of the iterated fission probability method into MCNP led to the rapid development of sensitivity analysis capabilities for MCNP6 and the development of Whisper. Sensitivity-uncertainty based methods represent the future for NCS validation – making full use of today’s computer power to codify past approaches based largely on expert judgment. Validation results are defensible, auditable, and repeatable as needed with different assumptions and process models. The new methods can supplement, support, and extend traditional validation approaches.« less

Computer codes for checking, plotting and processing of neutron cross-section covariance data and their application

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sartori, E.; Roussin, R.W.

This paper presents a brief review of computer codes concerned with checking, plotting, processing and using of covariances of neutron cross-section data. It concentrates on those available from the computer code information centers of the United States and the OECD/Nuclear Energy Agency. Emphasis will be placed also on codes using covariances for specific applications such as uncertainty analysis, data adjustment and data consistency analysis. Recent evaluations contain neutron cross section covariance information for all isotopes of major importance for technological applications of nuclear energy. It is therefore important that the available software tools needed for taking advantage of this informationmore » are widely known as hey permit the determination of better safety margins and allow the optimization of more economic, I designs of nuclear energy systems.« less

Radiochemical data collected on events from which radioactivity escaped beyond the borders of the Nevada test range complex. [NONE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hicks, H.G.

1981-02-12

This report identifies all nuclear events in Nevada that are known to have sent radioactivity beyond the borders of the test range complex. There have been 177 such tests, representing seven different types: nuclear detonations in the atmosphere, nuclear excavation events, nuclear safety events, underground nuclear events that inadvertently seeped or vented to the atmosphere, dispersion of plutonium and/or uranium by chemical high explosives, nuclear rocket engine tests, and nuclear ramjet engine tests. The source term for each of these events is given, together with the data base from which it was derived (except where the data are classified). Themore » computer programs used for organizing and processing the data base and calculating radionuclide production are described and included, together with the input and output data and details of the calculations. This is the basic formation needed to make computer modeling studies of the fallout from any of these 177 events.« less

SCALE Code System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rearden, Bradley T.; Jessee, Matthew Anderson

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 deterministicmore » 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.« less

SCALE Code System 6.2.1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rearden, Bradley T.; Jessee, Matthew Anderson

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 deterministicmore » 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.« less

Methods and benefits of experimental seismic evaluation of nuclear power plants. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1979-07-01

This study reviews experimental techniques, instrumentation requirements, safety considerations, and benefits of performing vibration tests on nuclear power plant containments and internal components. The emphasis is on testing to improve seismic structural models. Techniques for identification of resonant frequencies, damping, and mode shapes, are discussed. The benefits of testing with regard to increased damping and more accurate computer models are oulined. A test plan, schedule and budget are presented for a typical PWR nuclear power plant.

Bibliography of reports on research sponsored by the NRC Office of Nuclear Regulatory Research, July--December 1976

DOE Office of Scientific and Technical Information (OSTI.GOV)

Buchanan, J.R.

A bibliography of 148 reports published by contractors of the NRC Office of Nuclear Regulatory Research during the period July through December 1976 is presented along with abstracts from the Nuclear Safety Information Center computer file. The bibliography has been sorted into the subject categories used by NRC to organize the research program. Within the subject categories, the reports are sorted by contractor organization and then chronologically. A brief description of the NRC research program precedes the bibliography.

Parallel computation of multigroup reactivity coefficient using iterative method

NASA Astrophysics Data System (ADS)

Susmikanti, Mike; Dewayatna, Winter

2013-09-01

One of the research activities to support the commercial radioisotope production program is a safety research target irradiation FPM (Fission Product Molybdenum). FPM targets form a tube made of stainless steel in which the nuclear degrees of superimposed high-enriched uranium. FPM irradiation tube is intended to obtain fission. The fission material widely used in the form of kits in the world of nuclear medicine. Irradiation FPM tube reactor core would interfere with performance. One of the disorders comes from changes in flux or reactivity. It is necessary to study a method for calculating safety terrace ongoing configuration changes during the life of the reactor, making the code faster became an absolute necessity. Neutron safety margin for the research reactor can be reused without modification to the calculation of the reactivity of the reactor, so that is an advantage of using perturbation method. The criticality and flux in multigroup diffusion model was calculate at various irradiation positions in some uranium content. This model has a complex computation. Several parallel algorithms with iterative method have been developed for the sparse and big matrix solution. The Black-Red Gauss Seidel Iteration and the power iteration parallel method can be used to solve multigroup diffusion equation system and calculated the criticality and reactivity coeficient. This research was developed code for reactivity calculation which used one of safety analysis with parallel processing. It can be done more quickly and efficiently by utilizing the parallel processing in the multicore computer. This code was applied for the safety limits calculation of irradiated targets FPM with increment Uranium.

Nuclear cask testing films misleading and misused

DOE Office of Scientific and Technical Information (OSTI.GOV)

Audin, L.

In 1977 and 1978, Sandia National Laboratories, located in Albuquerque, New Mexico, and operated for the US Department of Energy (DOE), filmed a series of crash and fire tests performed on three casks designed to transport irradiated nuclear fuel assemblies. While the tests were performed to assess the applicability of scale and computer modeling techniques to actual accidents, films of them were quickly pressed into service by the DOE and nuclear utilities as proof'' to the public of the safety of the casks. In the public debate over the safety of irradiated nuclear fuel transportation, the films have served asmore » the mainstay for the nuclear industry. Although the scripts of all the films were reviewed by USDOE officials before production, they contain numerous misleading concepts and images, and omit significant facts. The shorter versions eliminated qualifying statements contained in the longer version, and created false impressions. This paper discusses factors which cast doubt on the veracity of the films and the results of the tests.« less

Nuclear cask testing films misleading and misused

DOE Office of Scientific and Technical Information (OSTI.GOV)

Audin, L.

In 1977 and 1978, Sandia National Laboratories, located in Albuquerque, New Mexico, and operated for the US Department of Energy (DOE), filmed a series of crash and fire tests performed on three casks designed to transport irradiated nuclear fuel assemblies. While the tests were performed to assess the applicability of scale and computer modeling techniques to actual accidents, films of them were quickly pressed into service by the DOE and nuclear utilities as ``proof`` to the public of the safety of the casks. In the public debate over the safety of irradiated nuclear fuel transportation, the films have served asmore » the mainstay for the nuclear industry. Although the scripts of all the films were reviewed by USDOE officials before production, they contain numerous misleading concepts and images, and omit significant facts. The shorter versions eliminated qualifying statements contained in the longer version, and created false impressions. This paper discusses factors which cast doubt on the veracity of the films and the results of the tests.« less

The Operation of a Specialized Scientific Information and Data Analysis Center With Computer Base and Associated Communications Network.

ERIC Educational Resources Information Center

Cottrell, William B.; And Others

The Nuclear Safety Information Center (NSIC) is a highly sophisticated scientific information center operated at Oak Ridge National Laboratory (ORNL) for the U.S. Atomic Energy Commission. Its information file, which consists of both data and bibliographic information, is computer stored and numerous programs have been developed to facilitate the…

Trends in HFE Methods and Tools and Their Applicability to Safety Reviews

DOE Office of Scientific and Technical Information (OSTI.GOV)

O'Hara, J.M.; Plott, C.; Milanski, J.

2009-09-30

The U.S. Nuclear Regulatory Commission's (NRC) conducts human factors engineering (HFE) safety reviews of applicant submittals for new plants and for changes to existing plants. The reviews include the evaluation of the methods and tools (M&T) used by applicants as part of their HFE program. The technology used to perform HFE activities has been rapidly evolving, resulting in a whole new generation of HFE M&Ts. The objectives of this research were to identify the current trends in HFE methods and tools, determine their applicability to NRC safety reviews, and identify topics for which the NRC may need additional guidance tomore » support the NRC's safety reviews. We conducted a survey that identified over 100 new HFE M&Ts. The M&Ts were assessed to identify general trends. Seven trends were identified: Computer Applications for Performing Traditional Analyses, Computer-Aided Design, Integration of HFE Methods and Tools, Rapid Development Engineering, Analysis of Cognitive Tasks, Use of Virtual Environments and Visualizations, and Application of Human Performance Models. We assessed each trend to determine its applicability to the NRC's review by considering (1) whether the nuclear industry is making use of M&Ts for each trend, and (2) whether M&Ts reflecting the trend can be reviewed using the current design review guidance. We concluded that M&T trends that are applicable to the commercial nuclear industry and are expected to impact safety reviews may be considered for review guidance development. Three trends fell into this category: Analysis of Cognitive Tasks, Use of Virtual Environments and Visualizations, and Application of Human Performance Models. The other trends do not need to be addressed at this time.« less

Recent developments: Industry briefs

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1990-09-01

Recent nuclear industry briefs are presented. These briefs include: Bechtel, Westinghouse and consumers form joint venture to operate Palisades; British Goverment to sell PowerGen in public offering; NPT conference opens in Geneva; Soviets buy US computers for nuclear safety; Cameco completes sale of interest in Rabbit Lake; Ebasco and CEGA each win defense reactor contract; East German utility takeover settled; Rio Algom shuts down Quirke and Panel early; and General Atomics buys Beverly mine.

How DARHT Works - the World's Most Powerful X-ray Machine

ScienceCinema

None

2018-06-01

The Dual Axis Radiographic Hydrodynamic Test (DARHT) facility at Los Alamos National Laboratory is an essential scientific tool that supports Stockpile Stewardship at the Laboratory. The World's most powerful x-ray machine, it's used to take high-speed images of mock nuclear devices - data that is used to confirm and modify advanced computer codes in assuring the safety, security, and effectiveness of the U.S. nuclear deterrent.

78 FR 27443 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Digital I&C...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-10

...-1207), ``Test Documentation for Digital Computer Software Used in Safety Systems of Nuclear Power... would result in a major inconvenience. If attending this meeting, please enter through the One White...

Multiphysics Object-Oriented Simulation Environment (MOOSE)

ScienceCinema

None

2017-12-09

Nuclear reactor operators can expand safety margins with more precise information about how materials behave inside operating reactors. INL's new simulation platform makes such studies easier & more informative by letting researchers "plug-n-play" their mathematical models, skipping years of computer code development.

The Nuclear Energy Advanced Modeling and Simulation Enabling Computational Technologies FY09 Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Diachin, L F; Garaizar, F X; Henson, V E

2009-10-12

In this document we report on the status of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Enabling Computational Technologies (ECT) effort. In particular, we provide the context for ECT In the broader NEAMS program and describe the three pillars of the ECT effort, namely, (1) tools and libraries, (2) software quality assurance, and (3) computational facility (computers, storage, etc) needs. We report on our FY09 deliverables to determine the needs of the integrated performance and safety codes (IPSCs) in these three areas and lay out the general plan for software quality assurance to meet the requirements of DOE andmore » the DOE Advanced Fuel Cycle Initiative (AFCI). We conclude with a brief description of our interactions with the Idaho National Laboratory computer center to determine what is needed to expand their role as a NEAMS user facility.« less

Bibliography of reports on research sponsored by the NRC office of nuclear regulatory research, July--December 1977

DOE Office of Scientific and Technical Information (OSTI.GOV)

Buchanan, J.R.

A bibliography of 198 reports published by contractors of the NRC Office of Nuclear Regulatory Research during the period July through December 1977 is presented along with abstracts from the Nuclear Safety Information Center computer file. The bibliography has been sorted into the subject categories used by NRC to organize the research program. Within the subject categories, the reports are arranged first by contractor organization and then chronologically. A brief description of the NRC research program precedes the bibliography.

Bibliography of reports on research sponsored by the NRC Office of Nuclear Regulatory Research, November 1975--June 1976

DOE Office of Scientific and Technical Information (OSTI.GOV)

Buchanan, J.R.

1976-09-30

A bibliography of 152 reports published by contractors of the NRC Office of Nuclear Regulatory Research during the period November 1975 through June 1976 is presented along with abstracts from the Nuclear Safety Information Center computer file. The bibliography has been sorted into the subject categories used by NRC to organize the research program. Within the subject categories, the reports are sorted by contractor organization and then chronologically. A brief description of the NRC research program precedes the bibliography.

Extreme Scale Computing to Secure the Nation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, D L; McGraw, J R; Johnson, J R

2009-11-10

Since the dawn of modern electronic computing in the mid 1940's, U.S. national security programs have been dominant users of every new generation of high-performance computer. Indeed, the first general-purpose electronic computer, ENIAC (the Electronic Numerical Integrator and Computer), was used to calculate the expected explosive yield of early thermonuclear weapons designs. Even the U. S. numerical weather prediction program, another early application for high-performance computing, was initially funded jointly by sponsors that included the U.S. Air Force and Navy, agencies interested in accurate weather predictions to support U.S. military operations. For the decades of the cold war, national securitymore » requirements continued to drive the development of high performance computing (HPC), including advancement of the computing hardware and development of sophisticated simulation codes to support weapons and military aircraft design, numerical weather prediction as well as data-intensive applications such as cryptography and cybersecurity U.S. national security concerns continue to drive the development of high-performance computers and software in the U.S. and in fact, events following the end of the cold war have driven an increase in the growth rate of computer performance at the high-end of the market. This mainly derives from our nation's observance of a moratorium on underground nuclear testing beginning in 1992, followed by our voluntary adherence to the Comprehensive Test Ban Treaty (CTBT) beginning in 1995. The CTBT prohibits further underground nuclear tests, which in the past had been a key component of the nation's science-based program for assuring the reliability, performance and safety of U.S. nuclear weapons. In response to this change, the U.S. Department of Energy (DOE) initiated the Science-Based Stockpile Stewardship (SBSS) program in response to the Fiscal Year 1994 National Defense Authorization Act, which requires, 'in the absence of nuclear testing, a progam to: (1) Support a focused, multifaceted program to increase the understanding of the enduring stockpile; (2) Predict, detect, and evaluate potential problems of the aging of the stockpile; (3) Refurbish and re-manufacture weapons and components, as required; and (4) Maintain the science and engineering institutions needed to support the nation's nuclear deterrent, now and in the future'. This program continues to fulfill its national security mission by adding significant new capabilities for producing scientific results through large-scale computational simulation coupled with careful experimentation, including sub-critical nuclear experiments permitted under the CTBT. To develop the computational science and the computational horsepower needed to support its mission, SBSS initiated the Accelerated Strategic Computing Initiative, later renamed the Advanced Simulation & Computing (ASC) program (sidebar: 'History of ASC Computing Program Computing Capability'). The modern 3D computational simulation capability of the ASC program supports the assessment and certification of the current nuclear stockpile through calibration with past underground test (UGT) data. While an impressive accomplishment, continued evolution of national security mission requirements will demand computing resources at a significantly greater scale than we have today. In particular, continued observance and potential Senate confirmation of the Comprehensive Test Ban Treaty (CTBT) together with the U.S administration's promise for a significant reduction in the size of the stockpile and the inexorable aging and consequent refurbishment of the stockpile all demand increasing refinement of our computational simulation capabilities. Assessment of the present and future stockpile with increased confidence of the safety and reliability without reliance upon calibration with past or future test data is a long-term goal of the ASC program. This will be accomplished through significant increases in the scientific bases that underlie the computational tools. Computer codes must be developed that replace phenomenology with increased levels of scientific understanding together with an accompanying quantification of uncertainty. These advanced codes will place significantly higher demands on the computing infrastructure than do the current 3D ASC codes. This article discusses not only the need for a future computing capability at the exascale for the SBSS program, but also considers high performance computing requirements for broader national security questions. For example, the increasing concern over potential nuclear terrorist threats demands a capability to assess threats and potential disablement technologies as well as a rapid forensic capability for determining a nuclear weapons design from post-detonation evidence (nuclear counterterrorism).« less

Safety of coronary CT angiography and functional testing for stable chest pain in the PROMISE trial: A randomized comparison of test complications, incidental findings, and radiation dose.

PubMed

Lu, Michael T; Douglas, Pamela S; Udelson, James E; Adami, Elizabeth; Ghoshhajra, Brian B; Picard, Michael H; Roberts, Rhonda; Lee, Kerry L; Einstein, Andrew J; Mark, Daniel B; Velazquez, Eric J; Carter, William; Ridner, Michael; Al-Khalidi, Hussein R; Hoffmann, Udo

Coronary computed tomography angiography (CTA) and functional testing strategies for stable chest pain yield similar outcomes; one aspect that may guide test choice is safety. We compared test safety (test complications, incidental findings, and effective radiation dose) between CTA and functional testing as-tested in PROMISE (PROspective Multicenter Imaging Study for Evaluation of Chest Pain). In the subgroup whose physicians intended nuclear stress over other functional tests if randomized to the functional arm, we compared radiation dose of CTA versus nuclear stress and identified characteristics associated with dose. Of 9470 patients, none had major and <1% had minor complications (CTA: 0.8% [37/4633] vs. functional: 0.6% [27/4837]). CTA identified more incidental findings (11.6% [539/4633] vs. 0.7% [34/4837], p < 0.001), most commonly pulmonary nodules (9.4%, 437/4633). CTA had similar 90-day cumulative radiation dose to functional testing. However, in the subgroup whose physicians intended nuclear stress (CTA 3147; nuclear 3203), CTA had lower median index test (8.8 vs. 12.6 mSv, p < 0.001) and 90-day cumulative (11.6 vs. 13.1 mSv, p < 0.001) dose, independent of patient characteristics. The lowest nuclear doses employed 1-day Tc-99m protocols (12.2 mSv). The lowest CTA doses were at sites performing ≥500 CTAs/year (6.9 mSv) and with advanced (latest available) CT scanners (5.5 mSv). Complications were negligibly rare for both CTA and functional testing. CTA detects more incidental findings. Compared to nuclear stress testing, CTA's lower radiation dose, independent of patient characteristics, makes it an attractive test choice. Radiation dose varies with imaging protocol, indicating opportunities to further reduce dose. (ClinicalTrials.gov number, NCT01174550). Copyright © 2017 Society of Cardiovascular Computed Tomography. Published by Elsevier Inc. All rights reserved.

Review of cardiovascular imaging in the Journal of Nuclear Cardiology in 2017. Part 2 of 2: Myocardial perfusion imaging.

PubMed

Hage, Fadi G; AlJaroudi, Wael A

2018-04-16

In 2017, the Journal of Nuclear Cardiology published many high-quality articles. In this review, we will summarize a selection of these articles to provide a concise review of the main advancements that have recently occurred in the field. In the first article of this 2-part series, we focused on publications dealing with positron emission tomography, computed tomography, and magnetic resonance. This review will place emphasis on myocardial perfusion imaging using single-photon emission computed tomography summarizing advances in the field including prognosis, safety and tolerability, the impact of imaging on management, and the use of novel imaging protocols.

Requirements for a multifunctional code architecture

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tiihonen, O.; Juslin, K.

1997-07-01

The present paper studies a set of requirements for a multifunctional simulation software architecture in the light of experiences gained in developing and using the APROS simulation environment. The huge steps taken in the development of computer hardware and software during the last ten years are changing the status of the traditional nuclear safety analysis software. The affordable computing power on the safety analysts table by far exceeds the possibilities offered to him/her ten years ago. At the same time the features of everyday office software tend to set standards to the way the input data and calculational results aremore » managed.« less

Ensuring the validity of calculated subcritical limits

DOE Office of Scientific and Technical Information (OSTI.GOV)

Clark, H.K.

1977-01-01

The care taken at the Savannah River Laboratory and Plant to ensure the validity of calculated subcritical limits is described. Close attention is given to ANSI N16.1-1975, ''Validation of Calculational Methods for Nuclear Criticality Safety.'' The computer codes used for criticality safety computations, which are listed and are briefly described, have been placed in the SRL JOSHUA system to facilitate calculation and to reduce input errors. A driver module, KOKO, simplifies and standardizes input and links the codes together in various ways. For any criticality safety evaluation, correlations of the calculational methods are made with experiment to establish bias. Occasionallymore » subcritical experiments are performed expressly to provide benchmarks. Calculated subcritical limits contain an adequate but not excessive margin to allow for uncertainty in the bias. The final step in any criticality safety evaluation is the writing of a report describing the calculations and justifying the margin.« less

Safety and efficacy for new techniques and imaging using new equipment to support European legislation: an EU coordination action.

PubMed

Zoetelief, J; Faulkner, K

2008-01-01

The past two decades have witnessed a technologically driven revolution in radiology. At the centre of these developments has been the use of computing. These developments have also been driven by the introduction of new detector and imaging devices in radiology and nuclear medicine, as well as the widespread application of computing techniques to enhance and extract information within the images acquired. Further advances have been introduced into digital practice. These technological developments, however, have not been matched by justification and optimisation studies to ensure that these new imaging devices and techniques are as effective as they might be, or performed at the lowest possible dose. The work programme of the SENTINEL Coordination Action was subdivided into eight work packages: functional performance and standards; efficacy and safety in digital radiology, dentistry and nuclear medicine, cardiology, interventional radiology, population screening/sensitive groups; justification, ethics and efficacy; good practice guidance and training; and project management. The intention of the work programme was to underwrite the safety, efficacy and ethical aspects of digital practice as well as to protect and add value to the equipment used in radiology.

Nuclear Cardiology: Are We Using the Right Protocols and Tracers the Right Way?

PubMed

Dondi, Maurizio; Pascual, Thomas; Paez, Diana; Einstein, Andrew J

2017-12-01

The field of nuclear cardiology has changed considerably over recent years, with greater attention paid to safety and radiation protection issues. The wider usage of technetium-99m (Tc-99m)-labeled radiopharmaceuticals for single-photon emission computed tomography (SPECT) imaging using gamma cameras has contributed to better quality studies and lower radiation exposure to patients. Increased availability of tracers and scanners for positron emission tomography (PET) will help further improve the quality of studies and quantify myocardial blood flow and myocardial flow reserve, thus enhancing the contribution of non-invasive imaging to the management of coronary artery disease. The introduction of new instrumentation such as solid state cameras and new software will help reduce further radiation exposure to patients undergoing nuclear cardiology studies. Results from recent studies, focused on assessing the relationship between best practices and radiation risk, provide useful insights on simple measures to improve the safety of nuclear cardiology studies without compromising the quality of results.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yates, K.R.; Schreiber, A.M.; Rudolph, A.W.

The US Nuclear Regulatory Commission has initiated the Fuel Cycle Risk Assessment Program to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. Both the once-through cycle and plutonium recycle are being considered. A previous report generated by this program defines and describes fuel cycle facilities, or elements, considered in the program. This report, the second from the program, describes the survey and computer compilation of fuel cycle risk-related literature. Sources of available information on the design, safety, and risk associated with the defined set of fuel cycle elements were searchedmore » and documents obtained were catalogued and characterized with respect to fuel cycle elements and specific risk/safety information. Both US and foreign surveys were conducted. Battelle's computer-based BASIS information management system was used to facilitate the establishment of the literature compilation. A complete listing of the literature compilation and several useful indexes are included. Future updates of the literature compilation will be published periodically. 760 annotated citations are included.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCoy, Michel; Archer, Bill; Hendrickson, Bruce

The Stockpile Stewardship Program (SSP) is an integrated technical program for maintaining the safety, surety, and reliability of the U.S. nuclear stockpile. The SSP uses nuclear test data, computational modeling and simulation, and experimental facilities to advance understanding of nuclear weapons. It includes stockpile surveillance, experimental research, development and engineering programs, and an appropriately scaled production capability to support stockpile requirements. This integrated national program requires the continued use of experimental facilities and programs, and the computational capabilities to support these programs. The Advanced Simulation and Computing Program (ASC) is a cornerstone of the SSP, providing simulation capabilities and computationalmore » resources that support annual stockpile assessment and certification, study advanced nuclear weapons design and manufacturing processes, analyze accident scenarios and weapons aging, and provide the tools to enable stockpile Life Extension Programs (LEPs) and the resolution of Significant Finding Investigations (SFIs). This requires a balance of resource, including technical staff, hardware, simulation software, and computer science solutions. ASC is now focused on increasing predictive capabilities in a three-dimensional (3D) simulation environment while maintaining support to the SSP. The program continues to improve its unique tools for solving progressively more difficult stockpile problems (sufficient resolution, dimensionality, and scientific details), and quantifying critical margins and uncertainties. Resolving each issue requires increasingly difficult analyses because the aging process has progressively moved the stockpile further away from the original test base. Where possible, the program also enables the use of high performance computing (HPC) and simulation tools to address broader national security needs, such as foreign nuclear weapon assessments and counter nuclear terrorism.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Korsah, K.; Wood, R.T.; Hassan, M.

This document presents the results of studies sponsored by the Nuclear Regulatory Commission (NRC) to provide the technical basis for environmental qualification of computer-based safety equipment in nuclear power plants. The studies were conducted by Oak Ridge National Laboratory (ORNL), Sandia National Laboratories (SNL), and Brookhaven National Laboratory (BNL). The studies address the following: (1) adequacy of the present test methods for qualification of digital I and C systems; (2) preferred (i.e., Regulatory Guide-endorsed) standards; (3) recommended stressors to be included in the qualification process during type testing; (4) resolution of need for accelerated aging for equipment to be locatedmore » in a benign environment; and (5) determination of an appropriate approach for addressing the impact of smoke in digital equipment qualification programs. Significant findings from the studies form the technical basis for a recommended approach to the environmental qualification of microprocessor-based safety-related equipment in nuclear power plants.« less

A Research Roadmap for Computation-Based Human Reliability Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boring, Ronald; Mandelli, Diego; Joe, Jeffrey

2015-08-01

The United States (U.S.) Department of Energy (DOE) is sponsoring research through the Light Water Reactor Sustainability (LWRS) program to extend the life of the currently operating fleet of commercial nuclear power plants. The Risk Informed Safety Margin Characterization (RISMC) research pathway within LWRS looks at ways to maintain and improve the safety margins of these plants. The RISMC pathway includes significant developments in the area of thermalhydraulics code modeling and the development of tools to facilitate dynamic probabilistic risk assessment (PRA). PRA is primarily concerned with the risk of hardware systems at the plant; yet, hardware reliability is oftenmore » secondary in overall risk significance to human errors that can trigger or compound undesirable events at the plant. This report highlights ongoing efforts to develop a computation-based approach to human reliability analysis (HRA). This computation-based approach differs from existing static and dynamic HRA approaches in that it: (i) interfaces with a dynamic computation engine that includes a full scope plant model, and (ii) interfaces with a PRA software toolset. The computation-based HRA approach presented in this report is called the Human Unimodels for Nuclear Technology to Enhance Reliability (HUNTER) and incorporates in a hybrid fashion elements of existing HRA methods to interface with new computational tools developed under the RISMC pathway. The goal of this research effort is to model human performance more accurately than existing approaches, thereby minimizing modeling uncertainty found in current plant risk models.« less

Stockpile Stewardship: How We Ensure the Nuclear Deterrent Without Testing

ScienceCinema

None

2018-01-16

In the 1990s, the U.S. nuclear weapons program shifted emphasis from developing new designs to dismantling thousands of existing weapons and maintaining a much smaller enduring stockpile. The United States ceased underground nuclear testing, and the Department of Energy created the Stockpile Stewardship Program to maintain the safety, security, and reliability of the U.S. nuclear deterrent without full-scale testing. This video gives a behind the scenes look at a set of unique capabilities at Lawrence Livermore that are indispensable to the Stockpile Stewardship Program: high performance computing, the Superblock category II nuclear facility, the JASPER a two stage gas gun, the High Explosive Applications Facility (HEAF), the National Ignition Facility (NIF), and the Site 300 contained firing facility.

SCALE Code System 6.2.2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rearden, Bradley T.; Jessee, Matthew Anderson

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 physics, radiation shielding, 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 3 deterministic and 3 Monte Carlomore » 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. SCALE 6.2 represents one of the most comprehensive revisions in the history of SCALE, providing several new capabilities and significant improvements in many existing features.« less

Needs analysis of a flexible computerized management infrastructure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Usman, S.; Hajek, B. K.; Ali, S. F.

2006-07-01

The United States' Energy Policy Act of 2005 is expected to facilitate construction of new commercial nuclear power plants. In the meanwhile, current plants are in the process of obtaining licenses for extended operation beyond their predetermined design life. In this beneficial yet challenging situation, it seems desirable to develop a strategic plan for smooth and seamless transition from paper based procedure systems to computer based procedure systems for improved performance and safety of the existing nuclear power plants. Many utilities already maintain procedures using word processing software, but it is common to print paper copies for daily use. Atmore » this time it is highly desirable to better understand the collective as well as individual document management needs of a commercial nuclear power plant as they migrate to a computer based system. As a contributory role in initiating a strategic plan, this paper offers a comprehensive questionnaire that is suitable for conducting a survey to determine the related needs of the utilities. The questionnaire covers three major areas: Formatting and User Friendly Features; Technical and Environmental Considerations; and Safety, System Integrity and Regulatory Considerations. A plan to conduct the proposed survey is also outlined in the future work section of this paper. (authors)« less

Environment, Safety, and Health: Status of DOE’s Reorganization of it’s Safety Oversight Function

DTIC Science & Technology

1990-01-01

facilities. After deliberation, the Congress in late 1988 directed that the Defense Nuclear Facilities Safety Board be established to provide...nuclear safety matters will be conducted by either the Advisory Committee on Nuclear Facility Safety or the recently mandated Defense Nuclear Facilities Safety...the facilities under the statutory purview of the Defense Nuclear Facilities Safety Board once the board determines it is ready to assume independent

Availability of Neutronics Benchmarks in the ICSBEP and IRPhEP Handbooks for Computational Tools Testing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.; Briggs, J. Blair; Ivanova, Tatiana

2017-02-01

In the past several decades, numerous experiments have been performed worldwide to support reactor operations, measurements, design, and nuclear safety. Those experiments represent an extensive international investment in infrastructure, expertise, and cost, representing significantly valuable resources of data supporting past, current, and future research activities. Those valuable assets represent the basis for recording, development, and validation of our nuclear methods and integral nuclear data [1]. The loss of these experimental data, which has occurred all too much in the recent years, is tragic. The high cost to repeat many of these measurements can be prohibitive, if not impossible, to surmount.more » Two international projects were developed, and are under the direction of the Organisation for Co-operation and Development Nuclear Energy Agency (OECD NEA) to address the challenges of not just data preservation, but evaluation of the data to determine its merit for modern and future use. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) was established to identify and verify comprehensive critical benchmark data sets; evaluate the data, including quantification of biases and uncertainties; compile the data and calculations in a standardized format; and formally document the effort into a single source of verified benchmark data [2]. Similarly, the International Reactor Physics Experiment Evaluation Project (IRPhEP) was established to preserve integral reactor physics experimental data, including separate or special effects data for nuclear energy and technology applications [3]. Annually, contributors from around the world continue to collaborate in the evaluation and review of select benchmark experiments for preservation and dissemination. The extensively peer-reviewed integral benchmark data can then be utilized to support nuclear design and safety analysts to validate the analytical tools, methods, and data needed for next-generation reactor design, safety analysis requirements, and all other front- and back-end activities contributing to the overall nuclear fuel cycle where quality neutronics calculations are paramount.« less

Index to Nuclear Safety: a technical progress review by chronology, permuted title, and author, Volume 18 (1) through Volume 22 (6)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cottrell, W.B.; Passiakos, M.

This index to Nuclear Safety covers articles published in Nuclear Safety, Volume 18, Number 1 (January-February 1977) through Volume 22, Number 6 (November-December 1981). The index is divided into three section: a chronological list of articles (including abstracts), a permuted-title (KWIC) index, and an author index. Nuclear Safety, a bimonthly technical progress review prepared by the Nuclear Safety Information Center, covers all safety aspects of nuclear power reactors and associated facilities. Over 300 technical articles published in Nuclear Safety in the last 5 years are listed in this index.

Nuclear safety policy working group recommendations on nuclear propulsion safety for the space exploration initiative

NASA Technical Reports Server (NTRS)

Marshall, Albert C.; Lee, James H.; Mcculloch, William H.; Sawyer, J. Charles, Jr.; Bari, Robert A.; Cullingford, Hatice S.; Hardy, Alva C.; Niederauer, George F.; Remp, Kerry; Rice, John W.

1993-01-01

An interagency Nuclear Safety Working Group (NSPWG) was chartered to recommend nuclear safety policy, requirements, and guidelines for the Space Exploration Initiative (SEI) nuclear propulsion program. These recommendations, which are contained in this report, should facilitate the implementation of mission planning and conceptual design studies. The NSPWG has recommended a top-level policy to provide the guiding principles for the development and implementation of the SEI nuclear propulsion safety program. In addition, the NSPWG has reviewed safety issues for nuclear propulsion and recommended top-level safety requirements and guidelines to address these issues. These recommendations should be useful for the development of the program's top-level requirements for safety functions (referred to as Safety Functional Requirements). The safety requirements and guidelines address the following topics: reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, safeguards, risk/reliability, operational safety, ground testing, and other considerations.

Stockpile Stewardship: How We Ensure the Nuclear Deterrent Without Testing

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

2014-09-04

In the 1990s, the U.S. nuclear weapons program shifted emphasis from developing new designs to dismantling thousands of existing weapons and maintaining a much smaller enduring stockpile. The United States ceased underground nuclear testing, and the Department of Energy created the Stockpile Stewardship Program to maintain the safety, security, and reliability of the U.S. nuclear deterrent without full-scale testing. This video gives a behind the scenes look at a set of unique capabilities at Lawrence Livermore that are indispensable to the Stockpile Stewardship Program: high performance computing, the Superblock category II nuclear facility, the JASPER a two stage gas gun,more » the High Explosive Applications Facility (HEAF), the National Ignition Facility (NIF), and the Site 300 contained firing facility.« less

Research on the management and endorsement of nuclear safety standards in the United States and its revelation for China

NASA Astrophysics Data System (ADS)

Liu, Ting; Tian, Yu; Yang, Lili; Gao, Siyi; Song, Dahu

2018-01-01

This paper introduces the American standard system, the Nuclear Regulatory Commission (NRC)’s responsibility, NRC nuclear safety regulations and standards system, studies on NRC’s standards management and endorsement mode, analyzes the characteristics of NRC standards endorsement management, and points out its disadvantages. This paper draws revelation from the standard management and endorsement model of NRC and points suggestion to China’s nuclear and radiation safety standards management.The issue of the “Nuclear Safety Law”plays an important role in China’s nuclear and radiation safety supervision. Nuclear and radiation safety regulations and standards are strong grips on the implementation of “Nuclear Safety Law”. This paper refers on the experience of international advanced countriy, will effectively promote the improvement of the endorsed management of China’s nuclear and radiation safety standards.

Probabilistic assessment of dynamic system performance. Part 3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Belhadj, Mohamed

1993-01-01

Accurate prediction of dynamic system failure behavior can be important for the reliability and risk analyses of nuclear power plants, as well as for their backfitting to satisfy given constraints on overall system reliability, or optimization of system performance. Global analysis of dynamic systems through investigating the variations in the structure of the attractors of the system and the domains of attraction of these attractors as a function of the system parameters is also important for nuclear technology in order to understand the fault-tolerance as well as the safety margins of the system under consideration and to insure a safemore » operation of nuclear reactors. Such a global analysis would be particularly relevant to future reactors with inherent or passive safety features that are expected to rely on natural phenomena rather than active components to achieve and maintain safe shutdown. Conventionally, failure and global analysis of dynamic systems necessitate the utilization of different methodologies which have computational limitations on the system size that can be handled. Using a Chapman-Kolmogorov interpretation of system dynamics, a theoretical basis is developed that unifies these methodologies as special cases and which can be used for a comprehensive safety and reliability analysis of dynamic systems.« less

48 CFR 923.7001 - Nuclear safety.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 48 Federal Acquisition Regulations System 5 2010-10-01 2010-10-01 false Nuclear safety. 923.7001... Efficiency, Renewable Energy Technologies, and Occupational Safety Programs 923.7001 Nuclear safety. The DOE regulates the nuclear safety of its major facilities under its own statutory authority derived from the...

Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. Kokkinos

2005-04-28

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophymore » on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.« less

Manned space flight nuclear system safety. Volume 5: Nuclear System safety guidelines. Part 1: Space base nuclear safety

NASA Technical Reports Server (NTRS)

1972-01-01

The design and operations guidelines and requirements developed in the study of space base nuclear system safety are presented. Guidelines and requirements are presented for the space base subsystems, nuclear hardware (reactor, isotope sources, dynamic generator equipment), experiments, interfacing vehicles, ground support systems, range safety and facilities. Cross indices and references are provided which relate guidelines to each other, and to substantiating data in other volumes. The guidelines are intended for the implementation of nuclear safety related design and operational considerations in future space programs.

Index to Nuclear Safety. A technical progress review by chronology, permuted title, and author. Vol. 11, No. 1--Vol. 17, No. 6

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cottrell, W.B.; Klein, A.

1977-02-23

This index to Nuclear Safety covers articles in Nuclear Safety Vol. 11, No. 1 (Jan.-Feb. 1970), through Vol. 17, No. 6 (Nov.-Dec. 1976). The index includes a chronological list of articles (including abstract) followed by KWIC and Author Indexes. Nuclear Safety, a bimonthly technical progress review prepared by the Nuclear Safety Information Center, covers all safety aspects of nuclear power reactors and associated facilities. The index lists over 350 technical articles in the last six years of publication.

SCALE: A modular code system for performing Standardized Computer Analyses for Licensing Evaluation. Volume 1, Part 2: Control modules S1--H1; Revision 5

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

SCALE--a modular code system for Standardized Computer Analyses Licensing Evaluation--has been developed by Oak Ridge National Laboratory at the request of the US Nuclear Regulatory Commission. The SCALE system utilizes well-established computer codes and methods within standard analysis sequences that (1) allow an input format designed for the occasional user and/or novice, (2) automated the data processing and coupling between modules, and (3) provide accurate and reliable results. System development has been directed at problem-dependent cross-section processing and analysis of criticality safety, shielding, heat transfer, and depletion/decay problems. Since the initial release of SCALE in 1980, the code system hasmore » been heavily used for evaluation of nuclear fuel facility and package designs. This revision documents Version 4.3 of the system.« less

Evaluation of radiological dispersion/consequence codes supporting DOE nuclear facility SARs

DOE Office of Scientific and Technical Information (OSTI.GOV)

O`Kula, K.R.; Paik, I.K.; Chung, D.Y.

1996-12-31

Since the early 1990s, the authorization basis documentation of many U.S. Department of Energy (DOE) nuclear facilities has been upgraded to comply with DOE orders and standards. In this process, many safety analyses have been revised. Unfortunately, there has been nonuniform application of software, and the most appropriate computer and engineering methodologies often are not applied. A DOE Accident Phenomenology and Consequence (APAC) Methodology Evaluation Program was originated at the request of DOE Defense Programs to evaluate the safety analysis methodologies used in nuclear facility authorization basis documentation and to define future cost-effective support and development initiatives. Six areas, includingmore » source term development (fire, spills, and explosion analysis), in-facility transport, and dispersion/ consequence analysis (chemical and radiological) are contained in the APAC program. The evaluation process, codes considered, key results, and recommendations for future model and software development of the Radiological Dispersion/Consequence Working Group are summarized in this paper.« less

Nuclear safety

NASA Technical Reports Server (NTRS)

Buden, D.

1991-01-01

Topics dealing with nuclear safety are addressed which include the following: general safety requirements; safety design requirements; terrestrial safety; SP-100 Flight System key safety requirements; potential mission accidents and hazards; key safety features; ground operations; launch operations; flight operations; disposal; safety concerns; licensing; the nuclear engine for rocket vehicle application (NERVA) design philosophy; the NERVA flight safety program; and the NERVA safety plan.

SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

DOE PAGES

Michel-Sendis, F.; Gauld, I.; Martinez, J. S.; ...

2017-08-02

SFCOMPO-2.0 is the new release of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) database of experimental assay measurements. These measurements are isotopic concentrations from destructive radiochemical analyses of spent nuclear fuel (SNF) samples. We supplement the measurements with design information for the fuel assembly and fuel rod from which each sample was taken, as well as with relevant information on operating conditions and characteristics of the host reactors. These data are necessary for modeling and simulation of the isotopic evolution of the fuel during irradiation. SFCOMPO-2.0 has been developed and is maintained by the OECDmore » NEA under the guidance of the Expert Group on Assay Data of Spent Nuclear Fuel (EGADSNF), which is part of the NEA Working Party on Nuclear Criticality Safety (WPNCS). Significant efforts aimed at establishing a thorough, reliable, publicly available resource for code validation and safety applications have led to the capture and standardization of experimental data from 750 SNF samples from more than 40 reactors. These efforts have resulted in the creation of the SFCOMPO-2.0 database, which is publicly available from the NEA Data Bank. Our paper describes the new database, and applications of SFCOMPO-2.0 for computer code validation, integral nuclear data benchmarking, and uncertainty analysis in nuclear waste package analysis are briefly illustrated.« less

SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michel-Sendis, F.; Gauld, I.; Martinez, J. S.

SFCOMPO-2.0 is the new release of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) database of experimental assay measurements. These measurements are isotopic concentrations from destructive radiochemical analyses of spent nuclear fuel (SNF) samples. We supplement the measurements with design information for the fuel assembly and fuel rod from which each sample was taken, as well as with relevant information on operating conditions and characteristics of the host reactors. These data are necessary for modeling and simulation of the isotopic evolution of the fuel during irradiation. SFCOMPO-2.0 has been developed and is maintained by the OECDmore » NEA under the guidance of the Expert Group on Assay Data of Spent Nuclear Fuel (EGADSNF), which is part of the NEA Working Party on Nuclear Criticality Safety (WPNCS). Significant efforts aimed at establishing a thorough, reliable, publicly available resource for code validation and safety applications have led to the capture and standardization of experimental data from 750 SNF samples from more than 40 reactors. These efforts have resulted in the creation of the SFCOMPO-2.0 database, which is publicly available from the NEA Data Bank. Our paper describes the new database, and applications of SFCOMPO-2.0 for computer code validation, integral nuclear data benchmarking, and uncertainty analysis in nuclear waste package analysis are briefly illustrated.« less

From Lift-Off to Light-Off

NASA Technical Reports Server (NTRS)

1982-01-01

Shuttle's propellant measurement system is produced by Simmonds Precision. Company has extensive experience in fuel management systems and other equipment for military and commercial aircraft. A separate corporate entity, Industrial Controls Division was formed due to a number of non-aerospace spinoffs. One example is a "custody transfer" system for measuring and monitoring liquefied natural gas (LNG). LNG is transported aboard large tankers at minus 260 degrees Fahrenheit. Value of a single shipload may reach $15 million. Precision's LNG measurement and monitoring system aids accurate financial accounting and enhances crew safety. Custody transfer systems have been provided for 10 LNG tankers, built by Owing Shipbuilding. Simmonds also provided measurement systems for several liquefied petroleum gas (LPG) production and storage installations. Another spinoff developed by Simmonds Precision is an advanced ignition system for industrial boilers that offers savings of millions of gallons of fuel, and a computer based monitoring and control system for improving safety and reliability in electrical utility applications. Simmonds produces a line of safety systems for nuclear and non-nuclear electrical power plants.

U.S. Nuclear Weapons Modernization - the Stockpile Life Extension Program

NASA Astrophysics Data System (ADS)

Cook, Donald

2016-03-01

Underground nuclear testing of U.S. nuclear weapons was halted by President George H.W. Bush in 1992 when he announced a moratorium. In 1993, the moratorium was extended by President Bill Clinton and, in 1995, a program of Stockpile Stewardship was put in its place. In 1996, President Clinton signed the Comprehensive Nuclear Test Ban Treaty (CTBT). Twenty years have passed since then. Over the same time, the average age of a nuclear weapon in the stockpile has increased from 6 years (1992) to nearly 29 years (2015). At its inception, achievement of the objectives of the Stockpile Stewardship Program (SSP) appeared possible but very difficult. The cost to design and construct several large facilities for precision experimentation in hydrodynamics and high energy density physics was large. The practical steps needed to move from computational platforms of less than 100 Mflops/sec to 10 Teraflops/sec and beyond were unknown. Today, most of the required facilities for SSP are in place and computational speed has been increased by more than six orders of magnitude. These, and the physicists and engineers in the complex of labs and plants within the National Nuclear Security Administration (NNSA) who put them in place, have been the basis for underpinning an annual decision, made by the weapons lab directors for each of the past 20 years, that resort to underground nuclear testing is not needed for maintaining confidence in the safety and reliability of the U.S stockpile. A key part of that decision has been annual assessment of the physical changes in stockpiled weapons. These weapons, quite simply, are systems that invariably and unstoppably age in the internal weapon environment of radioactive materials and complex interfaces of highly dissimilar organic and inorganic materials. Without an ongoing program to rebuild some components and replace other components to increase safety or security, i.e., life extending these weapons, either underground testing would again be required to assess many changes at once, or confidence in these weapons would be reduced. The strategy and details of the U.S. Stockpile Life Extension Program will be described in this talk. In brief, the strategy is to reduce the number of weapons in the stockpile while increasing confidence in the weapons that remain and, where possible, increase their safety, increase their security, and reduce their nuclear material quantities and yields. A number of ``myths'' pertaining to nuclear weapons, the SSP, and the Stockpile Life Extension Program will be explored.

An interagency space nuclear propulsion safety policy for SEI - Issues and discussion

NASA Technical Reports Server (NTRS)

Marshall, A. C.; Sawyer, J. C., Jr.

1991-01-01

An interagency Nuclear Safety Policy Working Group (NSPWG) was chartered to recommend nuclear safety policy, requirements, and guidelines for the Space Exploration Initiative nuclear propulsion program to facilitate the implementation of mission planning and conceptual design studies. The NSPWG developed a top level policy to provide the guiding principles for the development and implementation of the nuclear propulsion safety program and the development of Safety Functional Requirements. In addition, the NSPWG reviewed safety issues for nuclear propulsion and recommended top level safety requirements and guidelines to address these issues. Safety topics include reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, safeguards, risk/reliability, operational safety, ground testing, and other considerations. In this paper the emphasis is placed on the safety policy and the issues and considerations that are addressed by the NSPWG recommendations.

NSPWG-recommended safety requirements and guidelines for SEI nuclear propulsion

NASA Technical Reports Server (NTRS)

Marshall, Albert C.; Sawyer, J. C., Jr.; Bari, Robert A.; Brown, Neil W.; Cullingford, Hatice S.; Hardy, Alva C.; Lee, James H.; Mcculloch, William H.; Niederauer, George F.; Remp, Kerry

1992-01-01

An interagency Nuclear Safety Policy Working Group (NSPWG) was chartered to recommend nuclear safety policy, requirements, and guidelines for the Space Exploration Initiative (SEI) nuclear propulsion program to facilitate the implementation of mission planning and conceptual design studies. The NSPWG developed a top-level policy to provide the guiding principles for the development and implementation of the nuclear propulsion safety program and the development of safety functional requirements. In addition, the NSPWG reviewed safety issues for nuclear propulsion and recommended top-level safety requirements and guidelines to address these issues. Safety requirements were developed for reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, and safeguards. Guidelines were recommended for risk/reliability, operational safety, flight trajectory and mission abort, space debris and meteoroids, and ground test safety. In this paper the specific requirements and guidelines will be discussed.

Methodology, status, and plans for development and assessment of the RELAP5 code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, G.W.; Riemke, R.A.

1997-07-01

RELAP/MOD3 is a computer code used for the simulation of transients and accidents in light-water nuclear power plants. The objective of the program to develop and maintain RELAP5 was and is to provide the U.S. Nuclear Regulatory Commission with an independent tool for assessing reactor safety. This paper describes code requirements, models, solution scheme, language and structure, user interface validation, and documentation. The paper also describes the current and near term development program and provides an assessment of the code`s strengths and limitations.

76 FR 42686 - DOE Response to Recommendation 2011-1 of the Defense Nuclear Facilities Safety Board, Safety...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-07-19

... DEPARTMENT OF ENERGY DOE Response to Recommendation 2011-1 of the Defense Nuclear Facilities... Nuclear Facilities Safety Board, Office of Health, Safety and Security, U.S. Department of Energy, 1000... Department of Energy (DOE) acknowledges receipt of Defense Nuclear Facilities Safety Board (Board...

Validation of MCNP6 Version 1.0 with the ENDF/B-VII.1 Cross Section Library for Plutonium Metals, Oxides, and Solutions on the High Performance Computing Platform Moonlight

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chapman, Bryan Scott; Gough, Sean T.

This report documents a validation of the MCNP6 Version 1.0 computer code on the high performance computing platform Moonlight, for operations at Los Alamos National Laboratory (LANL) that involve plutonium metals, oxides, and solutions. The validation is conducted using the ENDF/B-VII.1 continuous energy group cross section library at room temperature. The results are for use by nuclear criticality safety personnel in performing analysis and evaluation of various facility activities involving plutonium materials.

NUCLEAR NONPROLIFERATION AND SAFETY: Challenges Facing the International Atomic Energy Agency.

DTIC Science & Technology

1993-09-01

safeguards), and the Chernobyl nuclear power plant accident have focused greater attention on nuclear proliferation and the safety of nuclear power... Chernobyl , IAEA has placed increasing emphasis on assisting member states in improving the safety of nuclear power plants. Despite funding shortfalls...report language, GAO has incorporated their comments where appropriate. 2Nuclear Power Safety: Chernobyl Accident Prompted Worldwide Actions but

Activities of the DOE Nuclear Criticality Safety Program (NCSP) at the Oak Ridge Electron Linear Accelerator (ORELA)

NASA Astrophysics Data System (ADS)

Valentine, Timothy E.; Leal, Luiz C.; Guber, Klaus H.

2002-12-01

The Department of Energy established the Nuclear Criticality Safety Program (NCSP) in response to the Recommendation 97-2 by the Defense Nuclear Facilities Safety Board. The NCSP consists of seven elements of which nuclear data measurements and evaluations is a key component. The intent of the nuclear data activities is to provide high resolution nuclear data measurements that are evaluated, validated, and formatted for use by the nuclear criticality safety community to provide improved and reliable calculations for nuclear criticality safety evaluations. High resolution capture, fission, and transmission measurements are performed at the Oak Ridge Electron Linear Accelerator (ORELA) to address the needs of the criticality safety community and to address known deficiencies in nuclear data evaluations. The activities at ORELA include measurements on both light and heavy nuclei and have been used to identify improvements in measurement techniques that greatly improve the measurement of small capture cross sections. The measurement activities at ORELA provide precise and reliable high-resolution nuclear data for the nuclear criticality safety community.

Right Size Determining the Staff Necessary to Sustain Simulation and Computing Capabilities for Nuclear Security

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nikkel, Daniel J.; Meisner, Robert

The Advanced Simulation and Computing Campaign, herein referred to as the ASC Program, is a core element of the science-based Stockpile Stewardship Program (SSP), which enables assessment, certification, and maintenance of the safety, security, and reliability of the U.S. nuclear stockpile without the need to resume nuclear testing. The use of advanced parallel computing has transitioned from proof-of-principle to become a critical element for assessing and certifying the stockpile. As the initiative phase of the ASC Program came to an end in the mid-2000s, the National Nuclear Security Administration redirected resources to other urgent priorities, and resulting staff reductions inmore » ASC occurred without the benefit of analysis of the impact on modern stockpile stewardship that is dependent on these new simulation capabilities. Consequently, in mid-2008 the ASC Program management commissioned a study to estimate the essential size and balance needed to sustain advanced simulation as a core component of stockpile stewardship. The ASC Program requires a minimum base staff size of 930 (which includes the number of staff necessary to maintain critical technical disciplines as well as to execute required programmatic tasks) to sustain its essential ongoing role in stockpile stewardship.« less

Seismic risk management solution for nuclear power plants

DOE PAGES

Coleman, Justin; Sabharwall, Piyush

2014-12-01

Nuclear power plants should safely operate during normal operations and maintain core-cooling capabilities during off-normal events, including external hazards (such as flooding and earthquakes). Management of external hazards to expectable levels of risk is critical to maintaining nuclear facility and nuclear power plant safety. Seismic risk is determined by convolving the seismic hazard with seismic fragilities (capacity of systems, structures, and components). Seismic isolation (SI) is one protective measure showing promise to minimize seismic risk. Current SI designs (used in commercial industry) reduce horizontal earthquake loads and protect critical infrastructure from the potentially destructive effects of large earthquakes. The benefitmore » of SI application in the nuclear industry is being recognized and SI systems have been proposed in American Society of Civil Engineer Standard 4, ASCE-4, to be released in the winter of 2014, for light water reactors facilities using commercially available technology. The intent of ASCE-4 is to provide criteria for seismic analysis of safety related nuclear structures such that the responses to design basis seismic events, computed in accordance with this standard, will have a small likelihood of being exceeded. The U.S. nuclear industry has not implemented SI to date; a seismic isolation gap analysis meeting was convened on August 19, 2014, to determine progress on implementing SI in the U.S. nuclear industry. The meeting focused on the systems and components that could benefit from SI. As a result, this article highlights the gaps identified at this meeting.« less

48 CFR 923.7001 - Nuclear safety.

Code of Federal Regulations, 2012 CFR

2012-10-01

... 48 Federal Acquisition Regulations System 5 2012-10-01 2012-10-01 false Nuclear safety. 923.7001 Section 923.7001 Federal Acquisition Regulations System DEPARTMENT OF ENERGY SOCIOECONOMIC PROGRAMS... Programs 923.7001 Nuclear safety. The DOE regulates the nuclear safety of its major facilities under its...

48 CFR 923.7001 - Nuclear safety.

Code of Federal Regulations, 2014 CFR

2014-10-01

... 48 Federal Acquisition Regulations System 5 2014-10-01 2014-10-01 false Nuclear safety. 923.7001 Section 923.7001 Federal Acquisition Regulations System DEPARTMENT OF ENERGY SOCIOECONOMIC PROGRAMS... Programs 923.7001 Nuclear safety. The DOE regulates the nuclear safety of its major facilities under its...

48 CFR 923.7001 - Nuclear safety.

Code of Federal Regulations, 2013 CFR

2013-10-01

... 48 Federal Acquisition Regulations System 5 2013-10-01 2013-10-01 false Nuclear safety. 923.7001 Section 923.7001 Federal Acquisition Regulations System DEPARTMENT OF ENERGY SOCIOECONOMIC PROGRAMS... Programs 923.7001 Nuclear safety. The DOE regulates the nuclear safety of its major facilities under its...

48 CFR 923.7001 - Nuclear safety.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 48 Federal Acquisition Regulations System 5 2011-10-01 2011-10-01 false Nuclear safety. 923.7001 Section 923.7001 Federal Acquisition Regulations System DEPARTMENT OF ENERGY SOCIOECONOMIC PROGRAMS... Programs 923.7001 Nuclear safety. The DOE regulates the nuclear safety of its major facilities under its...

Primary system fission product release and transport: A state-of-the-art report to the committee on the safety of nuclear installations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wright, A.L.

This report presents a summary of the status of research activities associated with fission product behavior (release and transport) under severe accident conditions within the primary systems of water-moderated and water-cooled nuclear reactors. For each of the areas of fission product release and fission product transport, the report summarizes relevant information on important phenomena, major experiments performed, relevant computer models and codes, comparisons of computer code calculations with experimental results, and general conclusions on the overall state of the art. Finally, the report provides an assessment of the overall importance and knowledge of primary system release and transport phenomena andmore » presents major conclusions on the state of the art.« less

Proof-of-Concept Demonstrations for Computation-Based Human Reliability Analysis. Modeling Operator Performance During Flooding Scenarios

DOE Office of Scientific and Technical Information (OSTI.GOV)

Joe, Jeffrey Clark; Boring, Ronald Laurids; Herberger, Sarah Elizabeth Marie

The United States (U.S.) Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) program has the overall objective to help sustain the existing commercial nuclear power plants (NPPs). To accomplish this program objective, there are multiple LWRS “pathways,” or research and development (R&D) focus areas. One LWRS focus area is called the Risk-Informed Safety Margin and Characterization (RISMC) pathway. Initial efforts under this pathway to combine probabilistic and plant multi-physics models to quantify safety margins and support business decisions also included HRA, but in a somewhat simplified manner. HRA experts at Idaho National Laboratory (INL) have been collaborating with othermore » experts to develop a computational HRA approach, called the Human Unimodel for Nuclear Technology to Enhance Reliability (HUNTER), for inclusion into the RISMC framework. The basic premise of this research is to leverage applicable computational techniques, namely simulation and modeling, to develop and then, using RAVEN as a controller, seamlessly integrate virtual operator models (HUNTER) with 1) the dynamic computational MOOSE runtime environment that includes a full-scope plant model, and 2) the RISMC framework PRA models already in use. The HUNTER computational HRA approach is a hybrid approach that leverages past work from cognitive psychology, human performance modeling, and HRA, but it is also a significant departure from existing static and even dynamic HRA methods. This report is divided into five chapters that cover the development of an external flooding event test case and associated statistical modeling considerations.« less

Revitalizing Nuclear Safety Research.

ERIC Educational Resources Information Center

National Academy of Sciences - National Research Council, Washington, DC.

This report covers the general issues involved in nuclear safety research and points out the areas needing detailed consideration. Topics included are: (1) "Principles of Nuclear Safety Research" (examining who should fund, who should conduct, and who should set the agenda for nuclear safety research); (2) "Elements of a Future…

The Organization and Management of the Nuclear Weapons Program.

DTIC Science & Technology

1997-03-01

over operations include the Defense Nuclear Facilities Safety Board, the Environmental Protection Agency, the Occupational Safety and Health...Safety, and Health. Still more guidance is received from the Defense Nuclear Facilities Safety Board and other external bodies such as the...state regulatory agencies, and the Defense Nuclear Facilities Safety Board. This chapter briefly reviews the most recent decade of this history, describes

77 FR 51943 - Procedures for Safety Investigations

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-28

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD 10 CFR Part 1708 Procedures for Safety Investigations AGENCY: Defense Nuclear Facilities Safety Board. ACTION: Proposed rule; extension of comment period. SUMMARY: The Defense Nuclear Facilities Safety Board is extending the time for comments on its proposed...

78 FR 12042 - Public Availability of Defense Nuclear Facilities Safety Board FY 2011 Service Contract Inventory...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-21

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Public Availability of Defense Nuclear Facilities Safety Board FY 2011 Service Contract Inventory Analysis/FY 2012 Service Contract Inventory AGENCY: Defense Nuclear Facilities Safety Board (DNFSB). ACTION: Notice of Public Availability of FY 2011 Service Contract...

76 FR 5354 - Public Availability of Defense Nuclear Facilities Safety Board FY 2010 Service Contract Inventory

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-31

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Public Availability of Defense Nuclear Facilities Safety Board FY 2010 Service Contract Inventory AGENCY: Defense Nuclear Facilities Safety Board (Board). ACTION: Notice of public availability of FY 2010 Service Contract Inventories. SUMMARY: In accordance with...

77 FR 7139 - Public Availability of Defense Nuclear Facilities Safety Board; FY 2010 Service Contract...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-10

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Public Availability of Defense Nuclear Facilities Safety Board; FY 2010 Service Contract Inventory Analysis/FY 2011 Service Contract Inventory AGENCY: Defense Nuclear Facilities Safety Board (DNFSB). ACTION: Notice of Public Availability of FY 2010 Service Contract...

Autonomous Control of Space Nuclear Reactors

NASA Technical Reports Server (NTRS)

Merk, John

2013-01-01

Nuclear reactors to support future robotic and manned missions impose new and innovative technological requirements for their control and protection instrumentation. Long-duration surface missions necessitate reliable autonomous operation, and manned missions impose added requirements for failsafe reactor protection. There is a need for an advanced instrumentation and control system for space-nuclear reactors that addresses both aspects of autonomous operation and safety. The Reactor Instrumentation and Control System (RICS) consists of two functionally independent systems: the Reactor Protection System (RPS) and the Supervision and Control System (SCS). Through these two systems, the RICS both supervises and controls a nuclear reactor during normal operational states, as well as monitors the operation of the reactor and, upon sensing a system anomaly, automatically takes the appropriate actions to prevent an unsafe or potentially unsafe condition from occurring. The RPS encompasses all electrical and mechanical devices and circuitry, from sensors to actuation device output terminals. The SCS contains a comprehensive data acquisition system to measure continuously different groups of variables consisting of primary measurement elements, transmitters, or conditioning modules. These reactor control variables can be categorized into two groups: those directly related to the behavior of the core (known as nuclear variables) and those related to secondary systems (known as process variables). Reliable closed-loop reactor control is achieved by processing the acquired variables and actuating the appropriate device drivers to maintain the reactor in a safe operating state. The SCS must prevent a deviation from the reactor nominal conditions by managing limitation functions in order to avoid RPS actions. The RICS has four identical redundancies that comply with physical separation, electrical isolation, and functional independence. This architecture complies with the safety requirements of a nuclear reactor and provides high availability to the host system. The RICS is intended to interface with a host computer (the computer of the spacecraft where the reactor is mounted). The RICS leverages the safety features inherent in Earth-based reactors and also integrates the wide range neutron detector (WRND). A neutron detector provides the input that allows the RICS to do its job. The RICS is based on proven technology currently in use at a nuclear research facility. In its most basic form, the RICS is a ruggedized, compact data-acquisition and control system that could be adapted to support a wide variety of harsh environments. As such, the RICS could be a useful instrument outside the scope of a nuclear reactor, including military applications where failsafe data acquisition and control is required with stringent size, weight, and power constraints.

Human reliability and plant operating efficiency: Are 12-hour work schedules cause for concern

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baker, T.L.

1992-01-01

Since the introduction of 12-h shifts to the US nuclear power industry only 8 yr ago, compressed workweek schedules have proliferated among operations departments at a phenomenal rate. Many plants that continue to use 8-h shifts during normal operations routinely change to scheduled 12-h shifts during refueling or maintenance outages. The most critical issue in the use of extended work shifts is whether alertness, physical stamina, or mental performance are compromised to the point of reducing safety or efficiency of nuclear power plant operation. Laboratory and field research sponsored by the National Institute of Occupational Safety and Health suggests thatmore » alertness, measured by self-ratings, and mental performance, measured by computer-based performance tests, are impaired on 12-h shifts compared with 8-h shifts. In contrast to these findings, plant operating efficiency and operator performance have been rated as improved in two field studies conducted in operating nuclear power plants (Fast Flux Test Facility, Washington and Ontario Hydro, Canada). A recent Electric Power Research Institute review of nuclear industry experience with 12-h shifts also suggests an overwhelmingly positive rating of 12-h schedules from both control room operators and management.« less

Sandia National Laboratories: National Security Missions: Nuclear Weapons:

Science.gov Websites

Safety & Security Sandia National Laboratories Exceptional service in the national interest & Figures Programs Nuclear Weapons About Nuclear Weapons Safety & Security Weapons Science & Twitter YouTube Flickr RSS Top Nuclear Weapons About Nuclear Weapons at Sandia Safety & Security

Evaluation of the concrete shield compositions from the 2010 criticality accident alarm system benchmark experiments at the CEA Valduc SILENE facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miller, Thomas Martin; Celik, Cihangir; Dunn, Michael E

In October 2010, a series of benchmark experiments were conducted at the French Commissariat a l'Energie Atomique et aux Energies Alternatives (CEA) Valduc SILENE facility. These experiments were a joint effort between the United States Department of Energy Nuclear Criticality Safety Program and the CEA. The purpose of these experiments was to create three benchmarks for the verification and validation of radiation transport codes and evaluated nuclear data used in the analysis of criticality accident alarm systems. This series of experiments consisted of three single-pulsed experiments with the SILENE reactor. For the first experiment, the reactor was bare (unshielded), whereasmore » in the second and third experiments, it was shielded by lead and polyethylene, respectively. The polyethylene shield of the third experiment had a cadmium liner on its internal and external surfaces, which vertically was located near the fuel region of SILENE. During each experiment, several neutron activation foils and thermoluminescent dosimeters (TLDs) were placed around the reactor. Nearly half of the foils and TLDs had additional high-density magnetite concrete, high-density barite concrete, standard concrete, and/or BoroBond shields. CEA Saclay provided all the concrete, and the US Y-12 National Security Complex provided the BoroBond. Measurement data from the experiments were published at the 2011 International Conference on Nuclear Criticality (ICNC 2011) and the 2013 Nuclear Criticality Safety Division (NCSD 2013) topical meeting. Preliminary computational results for the first experiment were presented in the ICNC 2011 paper, which showed poor agreement between the computational results and the measured values of the foils shielded by concrete. Recently the hydrogen content, boron content, and density of these concrete shields were further investigated within the constraints of the previously available data. New computational results for the first experiment are now available that show much better agreement with the measured values.« less

Radiation safety audit of a high volume Nuclear Medicine Department.

PubMed

Jha, Ashish Kumar; Singh, Abhijith Mohan; Shetye, Bhakti; Shah, Sneha; Agrawal, Archi; Purandare, Nilendu Chandrakant; Monteiro, Priya; Rangarajan, Venkatesh

2014-10-01

Professional radiation exposure cannot be avoided in nuclear medicine practices. It can only be minimized up to some extent by implementing good work practices. The aim of our study was to audit the professional radiation exposure and exposure rate of radiation worker working in and around Department of nuclear medicine and molecular imaging, Tata Memorial Hospital. We calculated the total number of nuclear medicine and positron emission tomography/computed tomography (PET/CT) procedures performed in our department and the radiation exposure to the radiation professionals from year 2009 to 2012. We performed an average of 6478 PET/CT scans and 3856 nuclear medicine scans/year from January 2009 to December 2012. The average annual whole body radiation exposure to nuclear medicine physician, technologist and nursing staff are 1.74 mSv, 2.93 mSv and 4.03 mSv respectively. Efficient management and deployment of personnel is of utmost importance to optimize radiation exposure in a high volume nuclear medicine setup in order to work without anxiety of high radiation exposure.

Real Time Conference 2016 Overview

NASA Astrophysics Data System (ADS)

Luchetta, Adriano

2017-06-01

This is a special issue of the IEEE Transactions on Nuclear Science containing papers from the invited, oral, and poster presentation of the 20th Real Time Conference (RT2016). The conference was held June 6-10, 2016, at Centro Congressi Padova “A. Luciani,” Padova, Italy, and was organized by Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA) and the Istituto Nazionale di Fisica Nucleare. The Real Time Conference is multidisciplinary and focuses on the latest developments in real-time techniques in high-energy physics, nuclear physics, astrophysics and astroparticle physics, nuclear fusion, medical physics, space instrumentation, nuclear power instrumentation, general radiation instrumentation, and real-time security and safety. Taking place every second year, it is sponsored by the Computer Application in Nuclear and Plasma Sciences technical committee of the IEEE Nuclear and Plasma Sciences Society. RT2016 attracted more than 240 registrants, with a large proportion of young researchers and engineers. It had an attendance of 67 students from many countries.

76 FR 26716 - Sunshine Act Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2011-05-09

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Sunshine Act Meeting AGENCY: Defense Nuclear Facilities... Defense Nuclear Facilities Safety Board's (Board) public meeting and hearing. FEDERAL REGISTER CITATIONS... Defense Nuclear Facilities Safety Board, Public Hearing Room, 625 Indiana Avenue, NW., Suite 300...

Mapping the Risks. Assessing the Homeland Security Implications of Publicly Available Geospatial Information

DTIC Science & Technology

2004-01-01

Defense Nuclear Facilities Safety Board 1 0.2 Export-Import Bank 1 0.2 National Archives and Records Administration 1 0.2 Supreme Court of the United...Agency Commodity Futures Trading Commission Consumer Product Safety Commission Defense Nuclear Facilities Safety Board Environmental Protection Agency...Intelligence www.cia.gov Defense Nuclear Facilities Safety Board Defense Nuclear Facilities Safety Board www.dnfsb.gov Department of

78 FR 25488 - Qualification Tests for Safety-Related Actuators in Nuclear Power Plants

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-01

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0079] Qualification Tests for Safety-Related Actuators in... regulatory guide (DG), DG-1235, ``Qualification Tests for Safety-Related Actuators in Nuclear Power Plants...-251- 7495, email: [email protected] . Both of the Office of Nuclear Regulatory Research, U.S. Nuclear...

78 FR 4393 - Sunshine Act Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-22

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Sunshine Act Notice AGENCY: Defense Nuclear Facilities... given of the Defense Nuclear Facilities Safety Board's (Board) public meeting and hearing described... Session II, the Board will receive testimony concerning safety at Pantex defense nuclear facilities. The...

75 FR 74022 - Safety Analysis Requirements for Defining Adequate Protection for the Public and the Workers

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-30

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD [Recommendation 2010-1] Safety Analysis Requirements for Defining Adequate Protection for the Public and the Workers AGENCY: Defense Nuclear Facilities Safety Board... Nuclear Facilities Safety Board has made a recommendation to the Secretary of Energy requesting an...

Psychology in nuclear power plants: an integrative approach to safety - general statement

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shikiar, R.

Since the accident at the Three Mile Island nuclear power plant on March 28, 1979, the commercial nuclear industry in the United States has paid increasing attention to the role of humans in overall plant safety. As the regulatory body with primary responsibility for ensuring public health and safety involving nuclear operations, the United States Nuclear Regulatory Commission (NRC) has also become increasingly involved with the ''human'' side of nuclear operations. The purpose of this symposium is to describe a major program of research and technical assistance that the Pacific Northwest Laboratory is performing for the NRC that deals withmore » the issues of safety at nuclear power plants (NPPs). This program addresses safety from several different levels of analysis, which are all important within the context of an integrative approach to system safety.« less

77 FR 479 - Sunshine Act Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-05

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Sunshine Act Notice AGENCY: Defense Nuclear Facilities... of the Defense Nuclear Facilities Safety Board's (Board) public hearing and meeting described below... Nuclear Facilities Safety Board, 625 Indiana Avenue NW., Suite 700, Washington, DC 20004-2901, (800) 788...

77 FR 48970 - Sunshine Act Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-15

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Sunshine Act Notice AGENCY: Defense Nuclear Facilities... given of the Defense Nuclear Facilities Safety Board's (Board) public meeting and hearing described... (NNSA) efforts to mitigate risks to public and worker safety posed by aging defense nuclear facilities...

Integral nuclear data validation using experimental spent nuclear fuel compositions

DOE PAGES

Gauld, Ian C.; Williams, Mark L.; Michel-Sendis, Franco; ...

2017-07-19

Measurements of the isotopic contents of spent nuclear fuel provide experimental data that are a prerequisite for validating computer codes and nuclear data for many spent fuel applications. Under the auspices of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) and guidance of the Expert Group on Assay Data of Spent Nuclear Fuel of the NEA Working Party on Nuclear Criticality Safety, a new database of expanded spent fuel isotopic compositions has been compiled. The database, Spent Fuel Compositions (SFCOMPO) 2.0, includes measured data for more than 750 fuel samples acquired from 44 different reactors andmore » representing eight different reactor technologies. Measurements for more than 90 isotopes are included. This new database provides data essential for establishing the reliability of code systems for inventory predictions, but it also has broader potential application to nuclear data evaluation. Furthermore, the database, together with adjoint based sensitivity and uncertainty tools for transmutation systems developed to quantify the importance of nuclear data on nuclide concentrations, are described.« less

Integral nuclear data validation using experimental spent nuclear fuel compositions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gauld, Ian C.; Williams, Mark L.; Michel-Sendis, Franco

Measurements of the isotopic contents of spent nuclear fuel provide experimental data that are a prerequisite for validating computer codes and nuclear data for many spent fuel applications. Under the auspices of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) and guidance of the Expert Group on Assay Data of Spent Nuclear Fuel of the NEA Working Party on Nuclear Criticality Safety, a new database of expanded spent fuel isotopic compositions has been compiled. The database, Spent Fuel Compositions (SFCOMPO) 2.0, includes measured data for more than 750 fuel samples acquired from 44 different reactors andmore » representing eight different reactor technologies. Measurements for more than 90 isotopes are included. This new database provides data essential for establishing the reliability of code systems for inventory predictions, but it also has broader potential application to nuclear data evaluation. Furthermore, the database, together with adjoint based sensitivity and uncertainty tools for transmutation systems developed to quantify the importance of nuclear data on nuclide concentrations, are described.« less

Manned space flight nuclear system safety. Volume 4: Space shuttle nuclear system transportation. Part 1: Space shuttle nuclear safety

NASA Technical Reports Server (NTRS)

1972-01-01

An analysis of the nuclear safety aspects (design and operational considerations) in the transport of nuclear payloads to and from earth orbit by the space shuttle is presented. Three representative nuclear payloads used in the study were: (1) the zirconium hydride reactor Brayton power module, (2) the large isotope Brayton power system and (3) small isotopic heat sources which can be a part of an upper stage or part of a logistics module. Reference data on the space shuttle and nuclear payloads are presented in an appendix. Safety oriented design and operational requirements were identified to integrate the nuclear payloads in the shuttle mission. Contingency situations were discussed and operations and design features were recommended to minimize the nuclear hazards. The study indicates the safety, design and operational advantages in the use of a nuclear payload transfer module. The transfer module can provide many of the safety related support functions (blast and fragmentation protection, environmental control, payload ejection) minimizing the direct impact on the shuttle.

Fluid Flow Investigations within a 37 Element CANDU Fuel Bundle Supported by Magnetic Resonance Velocimetry and Computational Fluid Dynamics

DOE PAGES

Piro, M.H.A; Wassermann, F.; Grundmann, S.; ...

2017-05-23

The current work presents experimental and computational investigations of fluid flow through a 37 element CANDU nuclear fuel bundle. Experiments based on Magnetic Resonance Velocimetry (MRV) permit three-dimensional, three-component fluid velocity measurements to be made within the bundle with sub-millimeter resolution that are non-intrusive, do not require tracer particles or optical access of the flow field. Computational fluid dynamic (CFD) simulations of the foregoing experiments were performed with the hydra-th code using implicit large eddy simulation, which were in good agreement with experimental measurements of the fluid velocity. Greater understanding has been gained in the evolution of geometry-induced inter-subchannel mixing,more » the local effects of obstructed debris on the local flow field, and various turbulent effects, such as recirculation, swirl and separation. These capabilities are not available with conventional experimental techniques or thermal-hydraulic codes. Finally, the overall goal of this work is to continue developing experimental and computational capabilities for further investigations that reliably support nuclear reactor performance and safety.« less

Fluid Flow Investigations within a 37 Element CANDU Fuel Bundle Supported by Magnetic Resonance Velocimetry and Computational Fluid Dynamics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Piro, M.H.A; Wassermann, F.; Grundmann, S.

The current work presents experimental and computational investigations of fluid flow through a 37 element CANDU nuclear fuel bundle. Experiments based on Magnetic Resonance Velocimetry (MRV) permit three-dimensional, three-component fluid velocity measurements to be made within the bundle with sub-millimeter resolution that are non-intrusive, do not require tracer particles or optical access of the flow field. Computational fluid dynamic (CFD) simulations of the foregoing experiments were performed with the hydra-th code using implicit large eddy simulation, which were in good agreement with experimental measurements of the fluid velocity. Greater understanding has been gained in the evolution of geometry-induced inter-subchannel mixing,more » the local effects of obstructed debris on the local flow field, and various turbulent effects, such as recirculation, swirl and separation. These capabilities are not available with conventional experimental techniques or thermal-hydraulic codes. Finally, the overall goal of this work is to continue developing experimental and computational capabilities for further investigations that reliably support nuclear reactor performance and safety.« less

Manned space flight nuclear system safety. Voluem 5: Nuclear system safety guidelines. Part 2: Space shuttle/nuclear payloads safety

NASA Technical Reports Server (NTRS)

1972-01-01

The design and operations guidelines and requirements developed in the study of space shuttle nuclear system transportation are presented. Guidelines and requirements are presented for the shuttle, nuclear payloads (reactor, isotope-Brayton and small isotope sources), ground support systems and facilities. Cross indices and references are provided which relate guidelines to each other, and to substantiating data in other volumes. The guidelines are intended for the implementation of nuclear safety related design and operational considerations in future space programs.

2015 Stewardship Science Academic Programs Annual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stone, Terri; Mischo, Millicent

The Stockpile Stewardship Academic Programs (SSAP) are essential to maintaining a pipeline of professionals to support the technical capabilities that reside at the National Nuclear Security Administration (NNSA) national laboratories, sites, and plants. Since 1992, the United States has observed the moratorium on nuclear testing while significantly decreasing the nuclear arsenal. To accomplish this without nuclear testing, NNSA and its laboratories developed a science-based Stockpile Stewardship Program to maintain and enhance the experimental and computational tools required to ensure the continued safety, security, and reliability of the stockpile. NNSA launched its academic program portfolio more than a decade ago tomore » engage students skilled in specific technical areas of relevance to stockpile stewardship. The success of this program is reflected by the large number of SSAP students choosing to begin their careers at NNSA national laboratories.« less

Nuclear Warheads: The Reliable Replacement Warhead Program and the Life Extension Program

DTIC Science & Technology

2007-07-16

The Defense Nuclear Facilities Safety Board was created by Congress 1988 “as an independent oversight organization within the Executive Branch charged... nuclear facilities .” U.S. Defense Nuclear Facilities Safety Board. “Who We Are,” at [http://www.dnfsb.gov/about/index.html]. beginning, addressed safety...approach, if successful, would “reduce or eliminate the need for ESD controls.”55 Kent Fortenberry, Technical Director of the Defense Nuclear Facilities Safety

The combined hybrid system: A symbiotic thermal reactor/fast reactor system for power generation and radioactive waste toxicity reduction

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hollaway, W.R.

1991-08-01

If there is to be a next generation of nuclear power in the United States, then the four fundamental obstacles confronting nuclear power technology must be overcome: safety, cost, waste management, and proliferation resistance. The Combined Hybrid System (CHS) is proposed as a possible solution to the problems preventing a vigorous resurgence of nuclear power. The CHS combines Thermal Reactors (for operability, safety, and cost) and Integral Fast Reactors (for waste treatment and actinide burning) in a symbiotic large scale system. The CHS addresses the safety and cost issues through the use of advanced reactor designs, the waste management issuemore » through the use of actinide burning, and the proliferation resistance issue through the use of an integral fuel cycle with co-located components. There are nine major components in the Combined Hybrid System linked by nineteen nuclear material mass flow streams. A computer code, CHASM, is used to analyze the mass flow rates CHS, and the reactor support ratio (the ratio of thermal/fast reactors), IFR of the system. The primary advantages of the CHS are its essentially actinide-free high-level radioactive waste, plus improved reactor safety, uranium utilization, and widening of the option base. The primary disadvantages of the CHS are the large capacity of IFRs required (approximately one MW{sub e} IFR capacity for every three MW{sub e} Thermal Reactor) and the novel radioactive waste streams produced by the CHS. The capability of the IFR to burn pure transuranic fuel, a primary assumption of this study, has yet to be proven. The Combined Hybrid System represents an attractive option for future nuclear power development; that disposal of the essentially actinide-free radioactive waste produced by the CHS provides an excellent alternative to the disposal of intact actinide-bearing Light Water Reactor spent fuel (reducing the toxicity based lifetime of the waste from roughly 360,000 years to about 510 years).« less

Renovated Korean nuclear safety and security system: A review and suggestions to successful settlement

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chung, W. S.; Yun, S. W.; Lee, D. S.

2012-07-01

Questions of whether past nuclear regulatory body of Korea is not a proper system to monitor and check the country's nuclear energy policy and utilization have been raised. Moreover, a feeling of insecurity regarding nuclear safety after the nuclear accident in Japan has spread across the public. This has stimulated a renovation of the nuclear safety regime in Korea. The Nuclear Safety and Security Commission (NSSC) was launched on October 26, 2011 as a regulatory body directly under the President in charge of strengthening independence and nuclear safety. This was a meaningful event as the NSSC it is a muchmore » more independent regulatory system for Korea. However, the NSSC itself does not guarantee an enhanced public acceptance of the nuclear policy and stable use nuclear energy. This study introduces the new NSSC system and its details in terms of organization structure, appropriateness of specialty, budget stability, and management system. (authors)« less

Nuclear Safety

DOE Office of Scientific and Technical Information (OSTI.GOV)

Silver, E G

This document is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

Nuclear Safety. Technical progress journal, April--June 1996: Volume 37, No. 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Muhlheim, M D

1996-01-01

This journal covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities.

Nuclear Safety. Technical progress journal, January--March 1994: Volume 35, No. 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Silver, E G

1994-01-01

This is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities.

Sandia technology: Engineering and science applications

NASA Astrophysics Data System (ADS)

Maydew, M. C.; Parrot, H.; Dale, B. C.; Floyd, H. L.; Leonard, J. A.; Parrot, L.

1990-12-01

This report discusses: protecting environment, safety, and health; Sandia's quality initiative; Sandia vigorously pursues technology transfer; scientific and technical education support programs; nuclear weapons development; recognizing battlefield targets with trained artificial neural networks; battlefield robotics: warfare at a distance; a spinning shell sizes up the enemy; thwarting would-be nuclear terrorists; unattended video surveillance system for nuclear facilities; making the skies safer for travelers; onboard instrumentation system to evaluate performance of stockpile bombs; keeping track with lasers; extended-life lithium batteries; a remote digital video link acquires images securely; guiding high-performance missiles with laser gyroscopes; nonvolatile memory chips for space applications; initiating weapon explosives with lasers; next-generation optoelectronics and microelectronics technology developments; chemometrics: new methods for improving chemical analysis; research team focuses ion beam to record-breaking intensities; standardizing the volt to quantum accuracy; new techniques improve robotic software development productivity; a practical laser plasma source for generating soft x-rays; exploring metal grain boundaries; massively parallel computing; modeling the amount of desiccant needed for moisture control; attacking pollution with sunshine; designing fuel-conversion catalysts with computers; extending a nuclear power plant's useful life; plasma-facing components for the International Thermonuclear Experimental Reactor.

HIGH-FIDELITY SIMULATION-DRIVEN MODEL DEVELOPMENT FOR COARSE-GRAINED COMPUTATIONAL FLUID DYNAMICS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hanna, Botros N.; Dinh, Nam T.; Bolotnov, Igor A.

Nuclear reactor safety analysis requires identifying various credible accident scenarios and determining their consequences. For a full-scale nuclear power plant system behavior, it is impossible to obtain sufficient experimental data for a broad range of risk-significant accident scenarios. In single-phase flow convective problems, Direct Numerical Simulation (DNS) and Large Eddy Simulation (LES) can provide us with high fidelity results when physical data are unavailable. However, these methods are computationally expensive and cannot be afforded for simulation of long transient scenarios in nuclear accidents despite extraordinary advances in high performance scientific computing over the past decades. The major issue is themore » inability to make the transient computation parallel, thus making number of time steps required in high-fidelity methods unaffordable for long transients. In this work, we propose to apply a high fidelity simulation-driven approach to model sub-grid scale (SGS) effect in Coarse Grained Computational Fluid Dynamics CG-CFD. This approach aims to develop a statistical surrogate model instead of the deterministic SGS model. We chose to start with a turbulent natural convection case with volumetric heating in a horizontal fluid layer with a rigid, insulated lower boundary and isothermal (cold) upper boundary. This scenario of unstable stratification is relevant to turbulent natural convection in a molten corium pool during a severe nuclear reactor accident, as well as in containment mixing and passive cooling. The presented approach demonstrates how to create a correction for the CG-CFD solution by modifying the energy balance equation. A global correction for the temperature equation proves to achieve a significant improvement to the prediction of steady state temperature distribution through the fluid layer.« less

Modeling and simulation challenges pursued by the Consortium for Advanced Simulation of Light Water Reactors (CASL)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Turinsky, Paul J., E-mail: turinsky@ncsu.edu; Kothe, Douglas B., E-mail: kothe@ornl.gov

The Consortium for the Advanced Simulation of Light Water Reactors (CASL), the first Energy Innovation Hub of the Department of Energy, was established in 2010 with the goal of providing modeling and simulation (M&S) capabilities that support and accelerate the improvement of nuclear energy's economic competitiveness and the reduction of spent nuclear fuel volume per unit energy, and all while assuring nuclear safety. To accomplish this requires advances in M&S capabilities in radiation transport, thermal-hydraulics, fuel performance and corrosion chemistry. To focus CASL's R&D, industry challenge problems have been defined, which equate with long standing issues of the nuclear powermore » industry that M&S can assist in addressing. To date CASL has developed a multi-physics “core simulator” based upon pin-resolved radiation transport and subchannel (within fuel assembly) thermal-hydraulics, capitalizing on the capabilities of high performance computing. CASL's fuel performance M&S capability can also be optionally integrated into the core simulator, yielding a coupled multi-physics capability with untapped predictive potential. Material models have been developed to enhance predictive capabilities of fuel clad creep and growth, along with deeper understanding of zirconium alloy clad oxidation and hydrogen pickup. Understanding of corrosion chemistry (e.g., CRUD formation) has evolved at all scales: micro, meso and macro. CFD R&D has focused on improvement in closure models for subcooled boiling and bubbly flow, and the formulation of robust numerical solution algorithms. For multiphysics integration, several iterative acceleration methods have been assessed, illuminating areas where further research is needed. Finally, uncertainty quantification and data assimilation techniques, based upon sampling approaches, have been made more feasible for practicing nuclear engineers via R&D on dimensional reduction and biased sampling. Industry adoption of CASL's evolving M&S capabilities, which is in progress, will assist in addressing long-standing and future operational and safety challenges of the nuclear industry. - Highlights: • Complexity of physics based modeling of light water reactor cores being addressed. • Capability developed to help address problems that have challenged the nuclear power industry. • Simulation capabilities that take advantage of high performance computing developed.« less

Permanent Disposal of Nuclear Waste in Salt

NASA Astrophysics Data System (ADS)

Hansen, F. D.

2016-12-01

Salt formations hold promise for eternal removal of nuclear waste from our biosphere. Germany and the United States have ample salt formations for this purpose, ranging from flat-bedded formations to geologically mature dome structures. Both nations are revisiting nuclear waste disposal options, accompanied by extensive collaboration on applied salt repository research, design, and operation. Salt formations provide isolation while geotechnical barriers reestablish impermeability after waste is placed in the geology. Between excavation and closure, physical, mechanical, thermal, chemical, and hydrological processes ensue. Salt response over a range of stress and temperature has been characterized for decades. Research practices employ refined test techniques and controls, which improve parameter assessment for features of the constitutive models. Extraordinary computational capabilities require exacting understanding of laboratory measurements and objective interpretation of modeling results. A repository for heat-generative nuclear waste provides an engineering challenge beyond common experience. Long-term evolution of the underground setting is precluded from direct observation or measurement. Therefore, analogues and modeling predictions are necessary to establish enduring safety functions. A strong case for granular salt reconsolidation and a focused research agenda support salt repository concepts that include safety-by-design. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000. Author: F. D. Hansen, Sandia National Laboratories

Modernization at the Y-12 National Security Complex: A Case for Additional Experimental Benchmarks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Thornbury, M. L.; Juarez, C.; Krass, A. W.

Efforts are underway at the Y-12 National Security Complex (Y-12) to modernize the recovery, purification, and consolidation of un-irradiated, highly enriched uranium metal. Successful integration of advanced technology such as Electrorefining (ER) eliminates many of the intermediate chemistry systems and processes that are the current and historical basis of the nuclear fuel cycle at Y-12. The cost of operations, the inventory of hazardous chemicals, and the volume of waste are significantly reduced by ER. It also introduces unique material forms and compositions related to the chemistry of chloride salts for further consideration in safety analysis and engineering. The work hereinmore » briefly describes recent investigations of nuclear criticality for 235UO2Cl2 (uranyl chloride) and 6LiCl (lithium chloride) in aqueous solution. Of particular interest is the minimum critical mass of highly enriched uranium as a function of the molar ratio of 6Li to 235U. The work herein also briefly describes recent investigations of nuclear criticality for 235U metal reflected by salt mixtures of 6LiCl or 7LiCl (lithium chloride), KCl (potassium chloride), and 235UCl3 or 238UCl3 (uranium tri-chloride). Computational methods for analysis of nuclear criticality safety and published nuclear data are employed in the absence of directly relevant experimental criticality benchmarks.« less

Information Services at the Nuclear Safety Analysis Center.

ERIC Educational Resources Information Center

Simard, Ronald

This paper describes the operations of the Nuclear Safety Analysis Center. Established soon after an accident at the Three Mile Island nuclear power plant near Harrisburg, Pennsylvania, its efforts were initially directed towards a detailed analysis of the accident. Continuing functions include: (1) the analysis of generic nuclear safety issues,…

78 FR 4404 - DOE Response to Recommendation 2012-2 of the Defense Nuclear Facilities Safety Board, Hanford...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-22

... DEPARTMENT OF ENERGY DOE Response to Recommendation 2012-2 of the Defense Nuclear Facilities Safety Board, Hanford Tank Farms Flammable Gas Safety Strategy AGENCY: Department of Energy. ACTION: Notice. SUMMARY: On September 28, 2012 the Defense Nuclear Facilities Safety Board submitted...

The Advanced Software Development and Commercialization Project

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gallopoulos, E.; Canfield, T.R.; Minkoff, M.

1990-09-01

This is the first of a series of reports pertaining to progress in the Advanced Software Development and Commercialization Project, a joint collaborative effort between the Center for Supercomputing Research and Development of the University of Illinois and the Computing and Telecommunications Division of Argonne National Laboratory. The purpose of this work is to apply techniques of parallel computing that were pioneered by University of Illinois researchers to mature computational fluid dynamics (CFD) and structural dynamics (SD) computer codes developed at Argonne. The collaboration in this project will bring this unique combination of expertise to bear, for the first time,more » on industrially important problems. By so doing, it will expose the strengths and weaknesses of existing techniques for parallelizing programs and will identify those problems that need to be solved in order to enable wide spread production use of parallel computers. Secondly, the increased efficiency of the CFD and SD codes themselves will enable the simulation of larger, more accurate engineering models that involve fluid and structural dynamics. In order to realize the above two goals, we are considering two production codes that have been developed at ANL and are widely used by both industry and Universities. These are COMMIX and WHAMS-3D. The first is a computational fluid dynamics code that is used for both nuclear reactor design and safety and as a design tool for the casting industry. The second is a three-dimensional structural dynamics code used in nuclear reactor safety as well as crashworthiness studies. These codes are currently available for both sequential and vector computers only. Our main goal is to port and optimize these two codes on shared memory multiprocessors. In so doing, we shall establish a process that can be followed in optimizing other sequential or vector engineering codes for parallel processors.« less

Advanced research workshop: nuclear materials safety

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jardine, L J; Moshkov, M M

The Advanced Research Workshop (ARW) on Nuclear Materials Safety held June 8-10, 1998, in St. Petersburg, Russia, was attended by 27 Russian experts from 14 different Russian organizations, seven European experts from six different organizations, and 14 U.S. experts from seven different organizations. The ARW was conducted at the State Education Center (SEC), a former Minatom nuclear training center in St. Petersburg. Thirty-three technical presentations were made using simultaneous translations. These presentations are reprinted in this volume as a formal ARW Proceedings in the NATO Science Series. The representative technical papers contained here cover nuclear material safety topics on themore » storage and disposition of excess plutonium and high enriched uranium (HEU) fissile materials, including vitrification, mixed oxide (MOX) fuel fabrication, plutonium ceramics, reprocessing, geologic disposal, transportation, and Russian regulatory processes. This ARW completed discussions by experts of the nuclear materials safety topics that were not covered in the previous, companion ARW on Nuclear Materials Safety held in Amarillo, Texas, in March 1997. These two workshops, when viewed together as a set, have addressed most nuclear material aspects of the storage and disposition operations required for excess HEU and plutonium. As a result, specific experts in nuclear materials safety have been identified, know each other from their participation in t he two ARW interactions, and have developed a partial consensus and dialogue on the most urgent nuclear materials safety topics to be addressed in a formal bilateral program on t he subject. A strong basis now exists for maintaining and developing a continuing dialogue between Russian, European, and U.S. experts in nuclear materials safety that will improve the safety of future nuclear materials operations in all the countries involved because of t he positive synergistic effects of focusing these diverse backgrounds of nuclear experience on a common objectiveÑthe safe and secure storage and disposition of excess fissile nuclear materials.« less

The Future of U.S. Nuclear Forces: Boom or Bust

DTIC Science & Technology

2007-03-30

materials, and nuclear waste.45 The Defense Nuclear Facilities Safety Board (DNFSB) was established by Congress in 1988 as an independent federal...adequate protection of public health and safety" at DOE’s defense nuclear facilities .46 This 100- person agency looks at four areas of the nuclear weapons...47 A.J. Eggenberger, Sixteenth Annual Report to Congress (Washington DC: Defense Nuclear Facilities Safety Board, February 2006), 13; available

Radiation Safety in Nuclear Medicine Procedures.

PubMed

Cho, Sang-Geon; Kim, Jahae; Song, Ho-Chun

2017-03-01

Since the nuclear disaster at the Fukushima Daiichi Nuclear Power Plant in 2011, radiation safety has become an important issue in nuclear medicine. Many structured guidelines or recommendations of various academic societies or international campaigns demonstrate important issues of radiation safety in nuclear medicine procedures. There are ongoing efforts to fulfill the basic principles of radiation protection in daily nuclear medicine practice. This article reviews important principles of radiation protection in nuclear medicine procedures. Useful references, important issues, future perspectives of the optimization of nuclear medicine procedures, and diagnostic reference level are also discussed.

Seismic performance assessment of base-isolated safety-related nuclear structures

USGS Publications Warehouse

Huang, Y.-N.; Whittaker, A.S.; Luco, N.

2010-01-01

Seismic or base isolation is a proven technology for reducing the effects of earthquake shaking on buildings, bridges and infrastructure. The benefit of base isolation has been presented in terms of reduced accelerations and drifts on superstructure components but never quantified in terms of either a percentage reduction in seismic loss (or percentage increase in safety) or the probability of an unacceptable performance. Herein, we quantify the benefits of base isolation in terms of increased safety (or smaller loss) by comparing the safety of a sample conventional and base-isolated nuclear power plant (NPP) located in the Eastern U.S. Scenario- and time-based assessments are performed using a new methodology. Three base isolation systems are considered, namely, (1) Friction Pendulum??? bearings, (2) lead-rubber bearings and (3) low-damping rubber bearings together with linear viscous dampers. Unacceptable performance is defined by the failure of key secondary systems because these systems represent much of the investment in a new build power plant and ensure the safe operation of the plant. For the scenario-based assessments, the probability of unacceptable performance is computed for an earthquake with a magnitude of 5.3 at a distance 7.5 km from the plant. For the time-based assessments, the annual frequency of unacceptable performance is computed considering all potential earthquakes that may occur. For both assessments, the implementation of base isolation reduces the probability of unacceptable performance by approximately four orders of magnitude for the same NPP superstructure and secondary systems. The increase in NPP construction cost associated with the installation of seismic isolators can be offset by substantially reducing the required seismic strength of secondary components and systems and potentially eliminating the need to seismically qualify many secondary components and systems. ?? 2010 John Wiley & Sons, Ltd.

Nuclear Safety Information Center, Its Products and Services

ERIC Educational Resources Information Center

Buchanan, J. R.

1970-01-01

The Nuclear Safety Information Center (NSIC) serves as a focal point for the collection, analysis and dissemination of information related to safety problems encountered in the design, analysis, and operation of nuclear facilities. (Author/AB)

Total safety management: An approach to improving safety culture

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blush, S.M.

A little over 4 yr ago, Admiral James D. Watkins became Secretary of Energy. President Bush, who had appointed him, informed Watkins that his principal task would be to clean up the nuclear weapons complex and put the US Department of Energy (DOE) back in the business of producing tritium for the nation's nuclear deterrent. Watkins recognized that in order to achieve these objectives, he would have to substantially improve the DOE's safety culture. Safety culture is a relatively new term. The International Atomic Energy Agency (IAEA) used it in a 1986 report on the root causes of the Chernobylmore » nuclear accident. In 1990, the IAEA's International Nuclear Safety Advisory Group issued a document focusing directly on safety culture. It provides guidelines to the international nuclear community for measuring the effectiveness of safety culture in nuclear organizations. Safety culture has two principal aspects: an organizational framework conducive to safety and the necessary organizational and individual attitudes that promote safety. These obviously go hand in hand. An organization must create the right framework to foster the right attitudes, but individuals must have the right attitudes to create the organizational framework that will support a good safety culture. The difficulty in developing such a synergistic relationship suggests that achieving and sustaining a strong safety culture is not easy, particularly in an organization whose safety culture is in serious disrepair.« less

The Department of Energy Nuclear Criticality Safety Program

NASA Astrophysics Data System (ADS)

Felty, James R.

2005-05-01

This paper broadly covers key events and activities from which the Department of Energy Nuclear Criticality Safety Program (NCSP) evolved. The NCSP maintains fundamental infrastructure that supports operational criticality safety programs. This infrastructure includes continued development and maintenance of key calculational tools, differential and integral data measurements, benchmark compilation, development of training resources, hands-on training, and web-based systems to enhance information preservation and dissemination. The NCSP was initiated in response to Defense Nuclear Facilities Safety Board Recommendation 97-2, Criticality Safety, and evolved from a predecessor program, the Nuclear Criticality Predictability Program, that was initiated in response to Defense Nuclear Facilities Safety Board Recommendation 93-2, The Need for Critical Experiment Capability. This paper also discusses the role Dr. Sol Pearlstein played in helping the Department of Energy lay the foundation for a robust and enduring criticality safety infrastructure.

Nuclear Weapons: NNSA Needs to Establish a Cost and Schedule Baseline for Manufacturing a Critical Nuclear Weapon Component

DTIC Science & Technology

2008-05-01

building up to and beyond the 2013 time frame. However, in October 2007, the Defense Nuclear Facilities Safety Board, which monitors safety...manufacturing. They said that NNSA is still working through this process with the Defense Nuclear Facilities Safety Board. Processing of waste

Experimental Study and Computational Simulations of Key Pebble Bed Thermo-mechanics Issues for Design and Safety

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tokuhiro, Akira; Potirniche, Gabriel; Cogliati, Joshua

2014-07-08

An experimental and computational study, consisting of modeling and simulation (M&S), of key thermal-mechanical issues affecting the design and safety of pebble-bed (PB) reactors was conducted. The objective was to broaden understanding and experimentally validate thermal-mechanic phenomena of nuclear grade graphite, specifically, spheres in frictional contact as anticipated in the bed under reactor relevant pressures and temperatures. The contact generates graphite dust particulates that can subsequently be transported into the flowing gaseous coolent. Under postulated depressurization transients and with the potential for leaked fission products to be adsorbed onto graphite 'dust', there is the potential for fission products to escapemore » from the primary volume. This is a design safety concern. Furthermore, earlier safety assessment identified the distinct possibility for the dispersed dust to combust in contact with air if sufficient conditions are met. Both of these phenomena were noted as important to design review and containing uncertainty to warrant study. The team designed and conducted two separate effects tests to study and benchmark the potential dust-generation rate, as well as study the conditions under which a dust explosion may occure in a standardized, instrumented explosion chamber.« less

76 FR 17460 - South Texas Project Nuclear Operating Company; Establishment of Atomic Safety and Licensing Board

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-29

...] South Texas Project Nuclear Operating Company; Establishment of Atomic Safety and Licensing Board..., 2.318, and 2.321, notice is hereby given that an Atomic Safety and Licensing Board (Board) is being...: Ronald M. Spritzer, Chair, Atomic Safety and Licensing Board Panel, U.S. Nuclear Regulatory Commission...

76 FR 3678 - FirstEnergy Nuclear Operating Company; Establishment of Atomic Safety and Licensing Board

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-20

... Nuclear Operating Company; Establishment of Atomic Safety and Licensing Board Pursuant to delegation by... hereby given that an Atomic Safety and Licensing Board (Board) is being established to preside over the... administrative judges: William J. Froehlich, Chair, Atomic Safety and Licensing Board Panel, U.S. Nuclear...

77 FR 43583 - DOE Response to Recommendation 2012-1 of the Defense Nuclear Facilities Safety Board, Savannah...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-25

... DEPARTMENT OF ENERGY DOE Response to Recommendation 2012-1 of the Defense Nuclear Facilities Safety Board, Savannah River Site Building 235-F Safety AGENCY: Department of Energy. ACTION: Notice. SUMMARY: On May 8, 2012, the Defense Nuclear Facilities Safety Board submitted Recommendation 2012-1...

Computed tomography of radioactive objects and materials

NASA Astrophysics Data System (ADS)

Sawicka, B. D.; Murphy, R. V.; Tosello, G.; Reynolds, P. W.; Romaniszyn, T.

1990-12-01

Computed tomography (CT) has been performed on a number of radioactive objects and materials. Several unique technical problems are associated with CT of radioactive specimens. These include general safety considerations, techniques to reduce background-radiation effects on CT images and selection criteria for the CT source to permit object penetration and to reveal accurate values of material density. In the present paper, three groups of experiments will be described, for objects with low, medium and high levels of radioactivity. CT studies on radioactive specimens will be presented. They include the following: (1) examination of individual ceramic reactor-fuel (uranium dioxide) pellets, (2) examination of fuel samples from the Three Mile Island reactor, (3) examination of a CANDU (CANada Deuterium Uraniun: registered trademark) nuclear-fuel bundle which underwent a simulated loss-of-coolant accident resulting in high-temperature damage and (4) examination of a PWR nuclear-reactor fuel assembly.

The American nuclear construction craftsmen: Will we be ready to build again

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bravo, R.

1990-01-01

The present state of nuclear plant maintenance and operations support reflects sexual, ethnic, and radical integration; continued educational advances; some computer literacy; mixed trades in maintenance; detailed training for maintenance and operations work in the operating plant; plant safety awareness and respect; need for top-quality, take the time to do it right mentality; and planning. With no new nuclear construction, what will be the specific talents, focus, and contributions that the craftsmen can be expected to bring to the project To be prepared to successfully manage the next generation of nuclear plant construction, the industry must be acutely aware ofmore » the needs of the labor pool. To be aware of the needs requires an intimate knowledge of the present state of the craft talent, the changed expectations of their contributions, and the effects of new technologies, materials, methods, and individuals that will be used to design and build.« less

Evaluation of nuclear-facility decommissioning projects. Summary report: Ames Laboratory Research Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Link, B.W.; Miller, R.L.

1983-07-01

This document summarizes the available information concerning the decommissioning of the Ames Laboratory Research Reactor (ALRR), a five-megawatt heavy water moderated and cooled research reactor. The data were placed in a computerized information retrieval/manipulation system which permits its future utilization for purposes of comparative analysis. This information is presented both in detail in its computer output form and also as a manually assembled summarization which highlights the more important aspects of the decommissioning program. Some comparative information with reference to generic decommissioning data extracted from NUREG/CR 1756, Technology, Safety and Costs of Decommissioning Nuclear Research and Test Reactors, is included.

Covariance Data File Formats for Whisper-1.0 & Whisper-1.1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Forrest B.; Rising, Michael Evan

2017-01-09

Whisper is a statistical analysis package developed in 2014 to support nuclear criticality safety (NCS) validation. It uses the sensitivity profile data for an application as computed by MCNP6 along with covariance files for the nuclear data to determine a baseline upper-subcritical-limit (USL) for the application. Whisper version 1.0 was first developed and used at LANL in 2014. During 2015-2016, Whisper was updated to version 1.1 and is to be included with the upcoming release of MCNP6.2. This report describes the file formats used for the covariance data in both Whisper-1.0 and Whisper-1.1.

Nuclear power: Siting and safety

DOE Office of Scientific and Technical Information (OSTI.GOV)

Openshaw, S.

1986-01-01

By 2030, half, or even two-thirds, of all electricity may be generated by nuclear power. Major reactor accidents are still expected to be rare occurrences, but nuclear safety is largely a matter of faith. Terrorist attacks, sabotage, and human error could cause a significant accident. Reactor siting can offer an additional, design-independent margin of safety. Remote geographical sites for new plants would minimize health risks, protect the industry from negative changes in public opinion concerning nuclear energy, and improve long-term public acceptance of nuclear power. U.K. siting practices usually do not consider the contribution to safety that could be obtainedmore » from remote sites. This book discusses the present trends of siting policies of nuclear power and their design-independent margin of safety.« less

Radiation safety audit of a high volume Nuclear Medicine Department

PubMed Central

Jha, Ashish Kumar; Singh, Abhijith Mohan; Shetye, Bhakti; Shah, Sneha; Agrawal, Archi; Purandare, Nilendu Chandrakant; Monteiro, Priya; Rangarajan, Venkatesh

2014-01-01

Introduction: Professional radiation exposure cannot be avoided in nuclear medicine practices. It can only be minimized up to some extent by implementing good work practices. Aim and Objectives: The aim of our study was to audit the professional radiation exposure and exposure rate of radiation worker working in and around Department of nuclear medicine and molecular imaging, Tata Memorial Hospital. Materials and Methods: We calculated the total number of nuclear medicine and positron emission tomography/computed tomography (PET/CT) procedures performed in our department and the radiation exposure to the radiation professionals from year 2009 to 2012. Results: We performed an average of 6478 PET/CT scans and 3856 nuclear medicine scans/year from January 2009 to December 2012. The average annual whole body radiation exposure to nuclear medicine physician, technologist and nursing staff are 1.74 mSv, 2.93 mSv and 4.03 mSv respectively. Conclusion: Efficient management and deployment of personnel is of utmost importance to optimize radiation exposure in a high volume nuclear medicine setup in order to work without anxiety of high radiation exposure. PMID:25400361

75 FR 56080 - Sunshine Act Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-15

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Sunshine Act Notice AGENCY: Defense Nuclear Facilities... Facilities Safety Board's public hearing and meeting. FEDERAL REGISTER CITATION OF PREVIOUS ANNOUNCEMENT: 75... INFORMATION: Brian Grosner, General Manager, Defense Nuclear Facilities Safety Board, 625 Indiana Avenue, NW...

Autonomous Control of Nuclear Power Plants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Basher, H.

2003-10-20

A nuclear reactor is a complex system that requires highly sophisticated controllers to ensure that desired performance and safety can be achieved and maintained during its operations. Higher-demanding operational requirements such as reliability, lower environmental impacts, and improved performance under adverse conditions in nuclear power plants, coupled with the complexity and uncertainty of the models, necessitate the use of an increased level of autonomy in the control methods. In the opinion of many researchers, the tasks involved during nuclear reactor design and operation (e.g., design optimization, transient diagnosis, and core reload optimization) involve important human cognition and decisions that maymore » be more easily achieved with intelligent methods such as expert systems, fuzzy logic, neural networks, and genetic algorithms. Many experts in the field of control systems share the idea that a higher degree of autonomy in control of complex systems such as nuclear plants is more easily achievable through the integration of conventional control systems and the intelligent components. Researchers have investigated the feasibility of the integration of fuzzy logic, neural networks, genetic algorithms, and expert systems with the conventional control methods to achieve higher degrees of autonomy in different aspects of reactor operations such as reactor startup, shutdown in emergency situations, fault detection and diagnosis, nuclear reactor alarm processing and diagnosis, and reactor load-following operations, to name a few. With the advancement of new technologies and computing power, it is feasible to automate most of the nuclear reactor control and operation, which will result in increased safety and economical benefits. This study surveys current status, practices, and recent advances made towards developing autonomous control systems for nuclear reactors.« less

Nuclear regulation: an evaluative study. [Monograph

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wood, W.C.

This paper gives an overview of nuclear regulation, followed by an account of the legislative rationale and the economic rationale for safety regulation of nuclear technology. It goes on to review the structure and safety programs of the Nuclear Regulatory Commission (NRC) and to show how many of the NRC's problems are unlike those facing other regulatory agencies. In assessing the results of nuclear regulation, it examines whether regulation has produced an acceptable level of safety and the nature of the evidence on acceptability. It concludes that the existing level of safety achievement is not cost-effective, and identifies some sourcesmore » of regulatory failure as a failure to make hard decisions on risk, diffuse responsibility for safety, and economic dissincentives to cost-effective safety. Recommendations for reforming NRC are to separate technical issues from social ones and to correct the incentives for safety. 153 references, 6 figures, 4 tables.« less

One in a Million Given the Accident: Assuring Nuclear Weapon Safety

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weaver, Jason

2015-08-25

Since the introduction of nuclear weapons, there has not been a single instance of accidental or unauthorized nuclear detonation, but there have been numerous accidents and “close calls.” As the understanding of these environments has increased, the need for a robust nuclear weapon safety philosophy has grown. This paper describes some of the methods used by the Nuclear Weapon Complex today to assure nuclear weapon safety, including testing, modeling, analysis, and design features. Lastly, it also reviews safety’s continued role in the future and examines how nuclear safety’s present maturity can play a role in strengthening security and other areasmore » and how increased coordination can improve safety and reduce long-term cost.« less

78 FR 33449 - FirstEnergy Nuclear Operating Company; Establishment of Atomic Safety and Licensing Board

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-04

... Nuclear Operating Company; Establishment of Atomic Safety and Licensing Board Pursuant to delegation by... 2.104, 2.105, 2.300, 2.309, 2.313, 2.318, and 2.321, notice is hereby given that an Atomic Safety... administrative judges: Paul S. Ryerson, Chairman, Atomic Safety and Licensing Board Panel, U.S. Nuclear...

The Demands of Nuclear Safety: Mishaps and USSTRATCOM

DTIC Science & Technology

2011-06-01

maintenance operation—the unexpected will occur. Scott D. Sagan On 30 August 2007 the unexpected occurred. Ironically, the safety problem did... Sagan , The Limits of Safety: Organizations, Accidents, and Nuclear Weapons (Princeton, NJ: Princeton University Press, 1993), 14, 48. 39 Sagan , Limits...1 Scott D. Sagan , The Limits of Safety: Organizations, Accidents, and Nuclear Weapons (Princeton, NJ: Princeton

Nuclear Data Activities in Support of the DOE Nuclear Criticality Safety Program

NASA Astrophysics Data System (ADS)

Westfall, R. M.; McKnight, R. D.

2005-05-01

The DOE Nuclear Criticality Safety Program (NCSP) provides the technical infrastructure maintenance for those technologies applied in the evaluation and performance of safe fissionable-material operations in the DOE complex. These technologies include an Analytical Methods element for neutron transport as well as the development of sensitivity/uncertainty methods, the performance of Critical Experiments, evaluation and qualification of experiments as Benchmarks, and a comprehensive Nuclear Data program coordinated by the NCSP Nuclear Data Advisory Group (NDAG). The NDAG gathers and evaluates differential and integral nuclear data, identifies deficiencies, and recommends priorities on meeting DOE criticality safety needs to the NCSP Criticality Safety Support Group (CSSG). Then the NDAG identifies the required resources and unique capabilities for meeting these needs, not only for performing measurements but also for data evaluation with nuclear model codes as well as for data processing for criticality safety applications. The NDAG coordinates effort with the leadership of the National Nuclear Data Center, the Cross Section Evaluation Working Group (CSEWG), and the Working Party on International Evaluation Cooperation (WPEC) of the OECD/NEA Nuclear Science Committee. The overall objective is to expedite the issuance of new data and methods to the DOE criticality safety user. This paper describes these activities in detail, with examples based upon special studies being performed in support of criticality safety for a variety of DOE operations.

NASA safety program activities in support of the Space Exploration Initiatives Nuclear Propulsion program

NASA Technical Reports Server (NTRS)

Sawyer, J. C., Jr.

1993-01-01

The activities of the joint NASA/DOE/DOD Nuclear Propulsion Program Technical Panels have been used as the basis for the current development of safety policies and requirements for the Space Exploration Initiatives (SEI) Nuclear Propulsion Technology development program. The Safety Division of the NASA Office of Safety and Mission Quality has initiated efforts to develop policies for the safe use of nuclear propulsion in space through involvement in the joint agency Nuclear Safety Policy Working Group (NSPWG), encouraged expansion of the initial policy development into proposed programmatic requirements, and suggested further expansion into the overall risk assessment and risk management process for the NASA Exploration Program. Similar efforts are underway within the Department of Energy to ensure the safe development and testing of nuclear propulsion systems on Earth. This paper describes the NASA safety policy related to requirements for the design of systems that may operate where Earth re-entry is a possibility. The expected plan of action is to support and oversee activities related to the technology development of nuclear propulsion in space, and support the overall safety and risk management program being developed for the NASA Exploration Program.

76 FR 17627 - Sunshine Act Meeting Postponed

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-30

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Sunshine Act Meeting Postponed AGENCY: Defense Nuclear Facilities Safety Board. ACTION: Notice of public meeting postponement. SUMMARY: The Defense Nuclear Facilities Safety Board (Board) published a document in the Federal Register of March 3, 2011 (76 FR 11764...

77 FR 14007 - Sunshine Act Meeting Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-08

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Sunshine Act Meeting Notice Federal Register CITATION OF... THE MEETING: The Defense Nuclear Facilities Safety Board (Board) is expanding the matters to be.../ resolution of safety and technical issues across the defense nuclear facilities complex. Since this panel...

Safety Oversight of Decommissioning Activities at DOE Nuclear Sites

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zull, Lawrence M.; Yeniscavich, William

2008-01-15

The Defense Nuclear Facilities Safety Board (Board) is an independent federal agency established by Congress in 1988 to provide nuclear safety oversight of activities at U.S. Department of Energy (DOE) defense nuclear facilities. The activities under the Board's jurisdiction include the design, construction, startup, operation, and decommissioning of defense nuclear facilities at DOE sites. This paper reviews the Board's safety oversight of decommissioning activities at DOE sites, identifies the safety problems observed, and discusses Board initiatives to improve the safety of decommissioning activities at DOE sites. The decommissioning of former defense nuclear facilities has reduced the risk of radioactive materialmore » contamination and exposure to the public and site workers. In general, efforts to perform decommissioning work at DOE defense nuclear sites have been successful, and contractors performing decommissioning work have a good safety record. Decommissioning activities have recently been completed at sites identified for closure, including the Rocky Flats Environmental Technology Site, the Fernald Closure Project, and the Miamisburg Closure Project (the Mound site). The Rocky Flats and Fernald sites, which produced plutonium parts and uranium materials for defense needs (respectively), have been turned into wildlife refuges. The Mound site, which performed R and D activities on nuclear materials, has been converted into an industrial and technology park called the Mound Advanced Technology Center. The DOE Office of Legacy Management is responsible for the long term stewardship of these former EM sites. The Board has reviewed many decommissioning activities, and noted that there are valuable lessons learned that can benefit both DOE and the contractor. As part of its ongoing safety oversight responsibilities, the Board and its staff will continue to review the safety of DOE and contractor decommissioning activities at DOE defense nuclear sites.« less

Confinement of Radioactive Materials at Defense Nuclear Facilities

DTIC Science & Technology

2004-10-01

The design of defense nuclear facilities includes systems whose reliable operation is vital to the protection of the public, workers, and the...final safety-class barrier to the release of hazardous materials with potentially serious public consequences. The Defense Nuclear Facilities Safety...the public at certain defense nuclear facilities . This change has resulted in downgrading of the functional safety classification of confinement

Real-world aspects of the nuclear criticality safety program at the University of Tennessee-Knoxville

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bentley, C.L.; Dunn, M.E.; Goluoglu, S.

1996-12-31

The nuclear criticality safety (NCS) program at the University of Tennessee-Knoxville (UTK) emphasizes the {open_quotes}real world{close_quotes} in the NCS courses that are offered and also the NCS research that is conducted. Two NCS courses are offered at UTK. The first course is an introduction to the NCS field, which uses the text by Knief and includes an overview of criticality accidents that have actually happened, standards that are currently in use and being developed, and state-of-the-art computer methods and codes. The students learn the same codes, including both theory and application, that are used by most professionals in the NCSmore » field. Thus, if a student accepts a job offer in the NCS area after graduation, he or she is capable of doing productive NCS work the first day on the job. Subcritical limits, hand-calculation methods, current regulations [both U.S. Department of Energy (DOE) and U.S. Nuclear Regulatory Commission (NRC)] and current practices are also discussed in the introductory course. The second course emphasizes real world experience and is taught by five instructors with over 100 years of combined experience.« less

75 FR 27228 - Proposed FOIA Fee Schedule Update

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-14

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD 10 CFR Part 1703 Proposed FOIA Fee Schedule Update AGENCY: Defense Nuclear Facilities Safety Board. ACTION: Notice of proposed rulemaking. SUMMARY: Pursuant to 10 CFR 1703.107(b)(6) of the Board's regulations, the Defense Nuclear Facilities Safety Board is...

77 FR 41258 - FOIA Fee Schedule Update

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-13

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD 10 CFR Part 1703 FOIA Fee Schedule Update AGENCY: Defense Nuclear Facilities Safety Board. ACTION: Establishment of FOIA Fee Schedule. SUMMARY: The Defense Nuclear Facilities Safety Board is publishing its Freedom of Information Act (FOIA) Fee Schedule Update pursuant to...

76 FR 28194 - Proposed FOIA Fee Schedule Update

Federal Register 2010, 2011, 2012, 2013, 2014

2011-05-16

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD 10 CFR Part 1703 Proposed FOIA Fee Schedule Update AGENCY: Defense Nuclear Facilities Safety Board. ACTION: Notice of proposed rulemaking. SUMMARY: Pursuant to 10 CFR 1703.107(b)(6) of the Board's regulations, the Defense Nuclear Facilities Safety Board is...

76 FR 43819 - FOIA Fee Schedule Update

Federal Register 2010, 2011, 2012, 2013, 2014

2011-07-22

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD 10 CFR Part 1703 FOIA Fee Schedule Update AGENCY: Defense Nuclear Facilities Safety Board. ACTION: Establishment of FOIA Fee Schedule. SUMMARY: The Defense Nuclear Facilities Safety Board is publishing its Freedom of Information Act (FOIA) Fee Schedule Update pursuant to...

78 FR 20625 - Extension of Hearing Record Closure Date

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-05

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Extension of Hearing Record Closure Date AGENCY: Defense Nuclear Facilities Safety Board. ACTION: Extension of hearing record closure date. SUMMARY: The Defense Nuclear Facilities Safety Board (Board) published a document in the Federal Register on January 22, 2013...

77 FR 65871 - Extension of Hearing Record Closure Date

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-31

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Extension of Hearing Record Closure Date AGENCY: Defense Nuclear Facilities Safety Board. ACTION: Extension of hearing record closure date. SUMMARY: The Defense Nuclear Facilities Safety Board (Board) published a document in the Federal Register on August 15, 2012...

78 FR 1206 - Second Extension of Hearing Record Closure Date

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-08

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Second Extension of Hearing Record Closure Date AGENCY: Defense Nuclear Facilities Safety Board. ACTION: Second extension of hearing record closure date. SUMMARY: The Defense Nuclear Facilities Safety Board (Board) published a document in the Federal Register on...

78 FR 49262 - Sunshine Act Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-13

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Sunshine Act Meeting AGENCY: Defense Nuclear Facilities... given of the Defense Nuclear Facilities Safety Board's (Board) public meeting and hearing described... associated with continued operation of aging defense nuclear [[Page 49263

What You Should Know About Pediatric Nuclear Medicine and Radiation Safety

MedlinePlus

What You Should Know About Pediatric Nuclear Medicine and Radiation Safety www.imagegently.org What is nuclear medicine? Nuclear medicine uses radioactive isotopes to create pictures of the human body. These pictures ...

Nuclear Warheads: The Reliable Replacement Warhead program and the Life Extension Program

DTIC Science & Technology

2007-12-03

eliminate the need for ESD controls.”67 CRS-22 68 The Defense Nuclear Facilities Safety Board was created by Congress 1988 “as an independent oversight...public health and safety’ at DOE’s defense nuclear facilities .” U.S. Defense Nuclear Facilities Safety Board. “Who We Are,” at [http://www.dnfsb.gov...about/index.html]. 69 Personal communication, Kent Fortenberry, Technical Director, Defense Nuclear Facilities Safety Board, September 14, 2006. 70

Nuclear Warheads: The Reliable Replacement Warhead Program and the Life Extension Program

DTIC Science & Technology

2007-04-04

Information provided by Pantex Plant, Sept. 19, 2006. 50 The Defense Nuclear Facilities Safety Board was created by Congress 1988 “as an independent...protection of public health and safety’ at DOE’s defense nuclear facilities .” U.S. Defense Nuclear Facilities Safety Board. “Who We Are,” at [http...www.dnfsb.gov/about/index.html]. 51 Personal communication, Kent Fortenberry, Technical Director, Defense Nuclear Facilities Safety Board, Sept. 14, 2006

Verification and Validation Strategy for LWRS Tools

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carl M. Stoots; Richard R. Schultz; Hans D. Gougar

2012-09-01

One intension of the Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) program is to create advanced computational tools for safety assessment that enable more accurate representation of a nuclear power plant safety margin. These tools are to be used to study the unique issues posed by lifetime extension and relicensing of the existing operating fleet of nuclear power plants well beyond their first license extension period. The extent to which new computational models / codes such as RELAP-7 can be used for reactor licensing / relicensing activities depends mainly upon the thoroughness with which they have been verifiedmore » and validated (V&V). This document outlines the LWRS program strategy by which RELAP-7 code V&V planning is to be accomplished. From the perspective of developing and applying thermal-hydraulic and reactivity-specific models to reactor systems, the US Nuclear Regulatory Commission (NRC) Regulatory Guide 1.203 gives key guidance to numeric model developers and those tasked with the validation of numeric models. By creating Regulatory Guide 1.203 the NRC defined a framework for development, assessment, and approval of transient and accident analysis methods. As a result, this methodology is very relevant and is recommended as the path forward for RELAP-7 V&V. However, the unique issues posed by lifetime extension will require considerations in addition to those addressed in Regulatory Guide 1.203. Some of these include prioritization of which plants / designs should be studied first, coupling modern supporting experiments to the stringent needs of new high fidelity models / codes, and scaling of aging effects.« less

SCALE: A modular code system for performing standardized computer analyses for licensing evaluation. Functional modules F1--F8 -- Volume 2, Part 1, Revision 4

DOE Office of Scientific and Technical Information (OSTI.GOV)

Greene, N.M.; Petrie, L.M.; Westfall, R.M.

SCALE--a modular code system for Standardized Computer Analyses Licensing Evaluation--has been developed by Oak Ridge National Laboratory at the request of the US Nuclear Regulatory Commission. The SCALE system utilizes well-established computer codes and methods within standard analysis sequences that (1) allow an input format designed for the occasional user and/or novice, (2) automate the data processing and coupling between modules, and (3) provide accurate and reliable results. System development has been directed at problem-dependent cross-section processing and analysis of criticality safety, shielding, heat transfer, and depletion/decay problems. Since the initial release of SCALE in 1980, the code system hasmore » been heavily used for evaluation of nuclear fuel facility and package designs. This revision documents Version 4.2 of the system. The manual is divided into three volumes: Volume 1--for the control module documentation; Volume 2--for functional module documentation; and Volume 3--for documentation of the data libraries and subroutine libraries.« less

78 FR 65978 - Draft Revised Strategic Plan for FY 2014-2018

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-04

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Draft Revised Strategic Plan for FY 2014-2018 AGENCY: Defense Nuclear Facilities Safety Board. ACTION: Notice. SUMMARY: In accordance with Office of Management and Budget Circular No. A-11, the Defense Nuclear Facilities Safety Board (DNFSB) is soliciting...

75 FR 4794 - Draft Revised Strategic Plan for FY 2010-2015

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-29

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Draft Revised Strategic Plan for FY 2010-2015 AGENCY: Defense Nuclear Facilities Safety Board. ACTION: Notice. SUMMARY: In accordance with OMB Circular No. A-11, the Defense Nuclear Facilities Safety Board is soliciting comments from all interested and potentially...

The TACIS Nuclear Programme: Assistance in Upgrading Russian Nuclear Power Stations - An Overview of the Individual Projects in the Internet

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bieth, Michel; Schoels, Hubert

2006-07-01

The European Union' TACIS1 programme has been established for the New Independent States (NIS), among them in the Russian Federation since 1991. One priority of TACIS funding is Nuclear Safety. The European Commission has made available a total of 944 Million Euros for nuclear safety programmes covering the period 1991-2003. The TACIS nuclear safety programme is devoted to the improvement of the safety of Soviet designed nuclear installations in providing technology and safety culture transfer. JRC is carrying out works in the following areas: On-Site Assistance for TACIS operating Nuclear Power Plants; Design Safety and Dissemination of TACIS results; Reactormore » Pressure Vessel Embrittlement for VVER; Regulatory Assistance; Industrial Waste Management; Nuclear Safeguards; All TACIS projects, dealing with these areas of activity are now available in so called Project Description Sheets (PDS) or Project Results Sheets (PRS) in the Internet for everybody. JRC has created in the Internet an easy to open and to browse database which contains the result of works in relation to the above mentioned nuclear activities. This presentation gives an on-line overview of the app. 430 projects which have been implemented so far since the outset of the TACIS Nuclear Programme in the Russian Federation, which is representative to the other CIS countries, benefiting from the TACIS. The presentation will mainly consist of an on-line-demonstration of the TACIS Nuclear WEB Page, created by JRC. (authors)« less

Y-12 PLANT NUCLEAR SAFETY HANDBOOK

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wachter, J.W. ed.; Bailey, M.L.; Cagle, T.J.

1963-03-27

Information needed to solve nuclear safety problems is condensed into a reference book for use by persons familiar with the field. Included are a glossary of terms; useful tables; nuclear constants; criticality calculations; basic nuclear safety limits; solution geometries and critical values; metal critical values; criticality values for intermediate, heterogeneous, and interacting systems; miscellaneous and related information; and report number, author, and subject indexes. (C.H.)

Manned space flight nuclear system safety. Volume 1: base nuclear system safety

NASA Technical Reports Server (NTRS)

1972-01-01

The mission and terrestrial nuclear safety aspects of future long duration manned space missions in low earth orbit are discussed. Nuclear hazards of a typical low earth orbit Space Base mission (from natural sources and on-board nuclear hardware) have been identified and evaluated. Some of the principal nuclear safety design and procedural considerations involved in launch, orbital, and end of mission operations are presented. Areas of investigation include radiation interactions with the crew, subsystems, facilities, experiments, film, interfacing vehicles, nuclear hardware and the terrestrial populace. Results of the analysis indicate: (1) the natural space environment can be the dominant radiation source in a low earth orbit where reactors are effectively shielded, (2) with implementation of safety guidelines the reactor can present a low risk to the crew, support personnel, the terrestrial populace, flight hardware and the mission, (3) ten year missions are feasible without exceeding integrated radiation limits assigned to flight hardware, and (4) crew stay-times up to one year are feasible without storm shelter provisions.

Interface design of VSOP'94 computer code for safety analysis

NASA Astrophysics Data System (ADS)

Natsir, Khairina; Yazid, Putranto Ilham; Andiwijayakusuma, D.; Wahanani, Nursinta Adi

2014-09-01

Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects.

Study Gives Good Odds on Nuclear Reactor Safety

ERIC Educational Resources Information Center

Russell, Cristine

1974-01-01

Summarized is data from a recent study on nuclear reactor safety completed by Norman C. Rasmussen and others. Non-nuclear events are about 10,000 times more likely to produce large accidents than nuclear plants. (RH)

Commercial grade item (CGI) dedication of generators for nuclear safety related applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Das, R.K.; Hajos, L.G.

1993-03-01

The number of nuclear safety related equipment suppliers and the availability of spare and replacement parts designed specifically for nuclear safety related application are shrinking rapidly. These have made it necessary for utilities to apply commercial grade spare and replacement parts in nuclear safety related applications after implementing proper acceptance and dedication process to verify that such items conform with the requirements of their use in nuclear safety related application. The general guidelines for the commercial grade item (CGI) acceptance and dedication are provided in US Nuclear Regulatory Commission (NRC) Generic Letters and Electric Power Research Institute (EPRI) Report NP-5652,more » Guideline for the Utilization of Commercial Grade Items in Nuclear Safety Related Applications. This paper presents an application of these generic guidelines for procurement, acceptance, and dedication of a commercial grade generator for use as a standby generator at Salem Generating Station Units 1 and 2. The paper identifies the critical characteristics of the generator which once verified, will provide reasonable assurance that the generator will perform its intended safety function. The paper also delineates the method of verification of the critical characteristics through tests and provide acceptance criteria for the test results. The methodology presented in this paper may be used as specific guidelines for reliable and cost effective procurement and dedication of commercial grade generators for use as standby generators at nuclear power plants.« less

10 CFR 830.202 - Safety basis.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 4 2011-01-01 2011-01-01 false Safety basis. 830.202 Section 830.202 Energy DEPARTMENT OF ENERGY NUCLEAR SAFETY MANAGEMENT Safety Basis Requirements § 830.202 Safety basis. (a) The contractor responsible for a hazard category 1, 2, or 3 DOE nuclear facility must establish and maintain the safety basis...

10 CFR 830.202 - Safety basis.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Safety basis. 830.202 Section 830.202 Energy DEPARTMENT OF ENERGY NUCLEAR SAFETY MANAGEMENT Safety Basis Requirements § 830.202 Safety basis. (a) The contractor responsible for a hazard category 1, 2, or 3 DOE nuclear facility must establish and maintain the safety basis...

Covariance Applications in Criticality Safety, Light Water Reactor Analysis, and Spent Fuel Characterization

DOE PAGES

Williams, M. L.; Wiarda, D.; Ilas, G.; ...

2014-06-15

Recently, we processed a new covariance data library based on ENDF/B-VII.1 for the SCALE nuclear analysis code system. The multigroup covariance data are discussed here, along with testing and application results for critical benchmark experiments. Moreover, the cross section covariance library, along with covariances for fission product yields and decay data, is used to compute uncertainties in the decay heat produced by a burned reactor fuel assembly.

Criticality Safety Evaluation for the TACS at DAF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Percher, C. M.; Heinrichs, D. P.

2011-06-10

Hands-on experimental training in the physical behavior of multiplying systems is one of ten key areas of training required for practitioners to become qualified in the discipline of criticality safety as identified in DOE-STD-1135-99, Guidance for Nuclear Criticality Safety Engineer Training and Qualification. This document is a criticality safety evaluation of the training activities and operations associated with HS-3201-P, Nuclear Criticality 4-Day Training Course (Practical). This course was designed to also address the training needs of nuclear criticality safety professionals under the auspices of the NNSA Nuclear Criticality Safety Program1. The hands-on, or laboratory, portion of the course will utilizemore » the Training Assembly for Criticality Safety (TACS) and will be conducted in the Device Assembly Facility (DAF) at the Nevada Nuclear Security Site (NNSS). The training activities will be conducted by Lawrence Livermore National Laboratory following the requirements of an Integrated Work Sheet (IWS) and associated Safety Plan. Students will be allowed to handle the fissile material under the supervision of an LLNL Certified Fissile Material Handler.« less

78 FR 4477 - Review of Safety Analysis Reports for Nuclear Power Plants, Introduction

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-22

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0268] Review of Safety Analysis Reports for Nuclear Power... Analysis Reports for Nuclear Power Plants: LWR Edition.'' The new subsection is the Standard Review Plan... Nuclear Power Plants: Integral Pressurized Water Reactor (iPWR) Edition.'' DATES: Comments must be filed...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

Plutonium Critical Mass Curve Comparison to Mass at Upper Subcritical Limit (USL) Using Whisper

DOE Office of Scientific and Technical Information (OSTI.GOV)

Alwin, Jennifer Louise; Zhang, Ning

Whisper is computational software designed to assist the nuclear criticality safety analyst with validation studies with the MCNP ® Monte Carlo radiation transport package. Standard approaches to validation rely on the selection of benchmarks based upon expert judgment. Whisper uses sensitivity/uncertainty (S/U) methods to select relevant benchmarks to a particular application or set of applications being analyzed. Using these benchmarks, Whisper computes a calculational margin. Whisper attempts to quantify the margin of subcriticality (MOS) from errors in software and uncertainties in nuclear data. The combination of the Whisper-derived calculational margin and MOS comprise the baseline upper subcritical limit (USL), tomore » which an additional margin may be applied by the nuclear criticality safety analyst as appropriate to ensure subcriticality. A series of critical mass curves for plutonium, similar to those found in Figure 31 of LA-10860-MS, have been generated using MCNP6.1.1 and the iterative parameter study software, WORM_Solver. The baseline USL for each of the data points of the curves was then computed using Whisper 1.1. The USL was then used to determine the equivalent mass for plutonium metal-water system. ANSI/ANS-8.1 states that it is acceptable to use handbook data, such as the data directly from the LA-10860-MS, as it is already considered validated (Section 4.3 4) “Use of subcritical limit data provided in ANSI/ANS standards or accepted reference publications does not require further validation.”). This paper attempts to take a novel approach to visualize traditional critical mass curves and allows comparison with the amount of mass for which the k eff is equal to the USL (calculational margin + margin of subcriticality). However, the intent is to plot the critical mass data along with USL, not to suggest that already accepted handbook data should have new and more rigorous requirements for validation.« less

IMPLEMENTATION OF DEFENSE NUCLEAR FACILITY SAFETY BOARD RECOMMENDATION 2000-2 AT WIPP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jackson, K.; Wu, C.

2002-02-26

The Defense Nuclear Safeties Board (DNFSB) issued Recommendation 2000-2 on March 8, 2000, concerning the degrading conditions of vital safety systems, or systems important to nuclear safety, at DOE sites across the nation. The Board recommended that the DOE take action to assess the condition of its nuclear systems to ensure continued operational readiness of vital safety systems that are important for safely accomplishing the DOE's mission. To verify the readiness of vital safety systems, a two-phased approach was established. Phase I consisted of a qualitative assessment to approved criteria of the defined vital safety systems by operating contractor personnel,more » overseen by Federal field office personnel. Based on Phase I Assessment results, vital safety systems with significant deficiencies would be further assessed in Phase II, a more extensive quantitative assessment, by a contractor and Federal team, using a second set of criteria. In addition, Defense Nuclear Facility Safety Board Recommendation 2000-2 concluded that the degradation of confinement ventilation systems was of major concern, and issued a separate set of criteria to perform a Phase II Assessment on confinement ventilation systems.« less

Integrated Response Time Evaluation Methodology for the Nuclear Safety Instrumentation System

NASA Astrophysics Data System (ADS)

Lee, Chang Jae; Yun, Jae Hee

2017-06-01

Safety analysis for a nuclear power plant establishes not only an analytical limit (AL) in terms of a measured or calculated variable but also an analytical response time (ART) required to complete protective action after the AL is reached. If the two constraints are met, the safety limit selected to maintain the integrity of physical barriers used for preventing uncontrolled radioactivity release will not be exceeded during anticipated operational occurrences and postulated accidents. Setpoint determination methodologies have actively been developed to ensure that the protective action is initiated before the process conditions reach the AL. However, regarding the ART for a nuclear safety instrumentation system, an integrated evaluation methodology considering the whole design process has not been systematically studied. In order to assure the safety of nuclear power plants, this paper proposes a systematic and integrated response time evaluation methodology that covers safety analyses, system designs, response time analyses, and response time tests. This methodology is applied to safety instrumentation systems for the advanced power reactor 1400 and the optimized power reactor 1000 nuclear power plants in South Korea. The quantitative evaluation results are provided herein. The evaluation results using the proposed methodology demonstrate that the nuclear safety instrumentation systems fully satisfy corresponding requirements of the ART.

A Qualitative Assessment of Current CCF Guidance Based on a Review of Safety System Digital Implementation Changes with Evolving Technology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Korsah, Kofi; Muhlheim, Michael David; Wood, Richard

The US Nuclear Regulatory Commission (NRC) is initiating a new rulemaking project to develop a digital system common-cause failure (CCF) rule. This rulemaking will review and modify or affirm the NRC's current digital system CCF policy as discussed in the Staff Requirements Memorandum to the Secretary of the Commission, Office of the NRC (SECY) 93-087, Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light Water Reactor (ALWR) Designs, and Branch Technical Position (BTP) 7-19, Guidance on Evaluation of Defense-in-Depth and Diversity in Digital Computer-Based Instrumentation and Control Systems, as well as Chapter 7, Instrumentation and Controls, in NRCmore » Regulatory Guide (NUREG)-0800, Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants (ML033580677). The Oak Ridge National Laboratory (ORNL) is providing technical support to the NRC staff on the CCF rulemaking, and this report is one of several providing the technical basis to inform NRC staff members. For the task described in this report, ORNL examined instrumentation and controls (I&C) technology implementations in nuclear power plants in the light of current CCF guidance. The intent was to assess whether the current position on CCF is adequate given the evolutions in digital safety system implementations and, if gaps in the guidance were found, to provide recommendations as to how these gaps could be closed.« less

GAO’s Views on DOE’s 1991 Budget for Addressing Problems at the Nuclear Weapons Complex

DTIC Science & Technology

1990-03-02

management, and efforts by DOE to make its contractors more accountable. Also, the Defense Nuclear Facilities Safety Board mandated by the Congress became...and safety matters. 6 Finally, the Defense Nuclear Facilities Safety Board was established. Although not a DOE action, its establishment, nevertheless

78 FR 67206 - Qualification Tests for Safety-Related Actuators in Nuclear Power Plants

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-08

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0079] Qualification Tests for Safety-Related Actuators in..., ``Qualification Tests for Safety-Related Actuators in Nuclear Power Plants.'' This RG is being revised to provide... power plants. This RG is proposed Revision 1 of RG 1.73, ``Qualification Tests of Electric Valve...

33 CFR 165.115 - Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts.

Code of Federal Regulations, 2013 CFR

2013-07-01

... 33 Navigation and Navigable Waters 2 2013-07-01 2013-07-01 false Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts. 165.115 Section 165.115 Navigation and Navigable... Coast Guard District § 165.115 Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth...

33 CFR 165.115 - Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts.

Code of Federal Regulations, 2014 CFR

2014-07-01

... 33 Navigation and Navigable Waters 2 2014-07-01 2014-07-01 false Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts. 165.115 Section 165.115 Navigation and Navigable... Coast Guard District § 165.115 Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth...

33 CFR 165.115 - Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 33 Navigation and Navigable Waters 2 2012-07-01 2012-07-01 false Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts. 165.115 Section 165.115 Navigation and Navigable... Coast Guard District § 165.115 Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth...

33 CFR 165.115 - Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 33 Navigation and Navigable Waters 2 2011-07-01 2011-07-01 false Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts. 165.115 Section 165.115 Navigation and Navigable... Coast Guard District § 165.115 Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth...

33 CFR 165.115 - Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts. 165.115 Section 165.115 Navigation and Navigable... Coast Guard District § 165.115 Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth...

Nuclear Powerplant Safety: Design and Planning.

ERIC Educational Resources Information Center

Department of Energy, Washington, DC. Nuclear Energy Office.

The most important concern in the design, construction and operation of nuclear powerplants is safety. Nuclear power is one of the major contributors to the nation's supply of electricity; therefore, it is important to assure its safe use. Each different type of powerplant has special design features and systems to protect health and safety. One…

Manned space flight nuclear system safety. Volume 6: Space base nuclear system safety plan

NASA Technical Reports Server (NTRS)

1972-01-01

A qualitative identification of the steps required to assure the incorporation of radiological system safety principles and objectives into all phases of a manned space base program are presented. Specific areas of emphasis include: (1) radiological program management, (2) nuclear system safety plan implementation, (3) impact on program, and (4) summary of the key operation and design guidelines and requirements. The plan clearly indicates the necessity of considering and implementing radiological system safety recommendations as early as possible in the development cycle to assure maximum safety and minimize the impact on design and mission plans.

76 FR 11764 - Sunshine Act Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-03

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Sunshine Act Notice AGENCY: Defense Nuclear Facilities... Defense Nuclear Facilities Safety Board's public hearing and meeting described below. Interested persons... the matters to be considered. TIME AND DATE OF MEETING: 9 a.m., March 31, 2011. PLACE: Defense Nuclear...

Function Allocation in Complex Socio-Technical Systems: Procedure usage in nuclear power and the Context Analysis Method for Identifying Design Solutions (CAMIDS) Model

NASA Astrophysics Data System (ADS)

Schmitt, Kara Anne

This research aims to prove that strict adherence to procedures and rigid compliance to process in the US Nuclear Industry may not prevent incidents or increase safety. According to the Institute of Nuclear Power Operations, the nuclear power industry has seen a recent rise in events, and this research claims that a contributing factor to this rise is organizational, cultural, and based on peoples overreliance on procedures and policy. Understanding the proper balance of function allocation, automation and human decision-making is imperative to creating a nuclear power plant that is safe, efficient, and reliable. This research claims that new generations of operators are less engaged and thinking because they have been instructed to follow procedures to a fault. According to operators, they were once to know the plant and its interrelations, but organizationally more importance is now put on following procedure and policy. Literature reviews were performed, experts were questioned, and a model for context analysis was developed. The Context Analysis Method for Identifying Design Solutions (CAMIDS) Model was created, verified and validated through both peer review and application in real world scenarios in active nuclear power plant simulators. These experiments supported the claim that strict adherence and rigid compliance to procedures may not increase safety by studying the industry's propensity for following incorrect procedures, and when it directly affects the outcome of safety or security of the plant. The findings of this research indicate that the younger generations of operators rely highly on procedures, and the organizational pressures of required compliance to procedures may lead to incidents within the plant because operators feel pressured into following the rules and policy above performing the correct actions in a timely manner. The findings support computer based procedures, efficient alarm systems, and skill of the craft matrices. The solution to the problems facing the industry include in-depth, multiple fault failure training which tests the operator's knowledge of the situation. This builds operator collaboration, competence and confidence to know what to do, and when to do it in response to an emergency situation. Strict adherence to procedures and rigid compliance to process may not prevent incidents or increase safety; building operators' fundamental skills of collaboration, competence and confidence will.

A USNRC perspective on the use of commercial-off-shelf software (COTS) in advanced reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stewart, J.C.

1997-12-01

The use of commercially available digital computer systems and components in safety critical systems (nuclear power plant, military, and commercial applications) is increasing rapidly. While this paper focuses on the software aspects of the application most of these continents are applicable to the hardware aspects as well. Commercial dedication (the process of assuring that a commercial grade item will perform its intended safety function) has demonstrated benefits in cost savings and a wide base of user experience, however, care must be taken to avoid difficulties with some aspects of the dedication process such as access to vendor development information, configurationmore » management long term support, and system integration.« less

Proceedings of the international meeting on thermal nuclear reactor safety. Vol. 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

Separate abstracts are included for each of the papers presented concerning current issues in nuclear power plant safety; national programs in nuclear power plant safety; radiological source terms; probabilistic risk assessment methods and techniques; non LOCA and small-break-LOCA transients; safety goals; pressurized thermal shocks; applications of reliability and risk methods to probabilistic risk assessment; human factors and man-machine interface; and data bases and special applications.

IDNS: The Illinois Nuclear Safety Agency

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gallina, C.O.

The Illinois Department of Nuclear Safety (IDNS) is one of only two cabinet-level state agencies in the United States devoted exclusively to nuclear and radiation safety. It was established in 1980 by then Gov. James Thompson in response to the 1979 accident at Three Mile Island-2, so the state would be prepared in case of a similar accident at an Illinois nuclear power facility. There are 13 commercial nuclear reactors at seven sites in Illinois, more than in any other state. If Illinois were a country, it would be seventh in the world in the amount of nuclear-generated electricity, andmore » second in the percentage of electricity produced by nuclear power. The state also has several major nonreactor nuclear facilities. 9 refs.« less

75 FR 77675 - AREVA NP, Inc.; Confirmatory Order (Effective Immediately)

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-13

..., appropriate notification of regulatory authorities, safety culture and safety conscious work environment...., outside the global AREVA organization), safety culture assessment in accordance with an accepted nuclear....e., outside the global AREVA organization), safety culture assessment in accordance with an accepted...

Nuclear-safety institution in France: emergence and development

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vallet, B.M.

1986-01-01

This research work examines the social construction of the nuclear-safety institution in France, and the concurrent increased focus on the nuclear-risk issue. Emphasis on risk and safety, as primarily technical issues, can partly be seen as a strategy. Employed by power elites in the nuclear technostructure, this diverts emphasis away from controversial and normative questions regarding the political and social consequences of technology to questions of technology that appear to be absolute to the technology itself. Nuclear safety, which started from a preoccupation with risk related to the nuclear energy research and development process, is examined using the analytic conceptmore » of field. As a social arena patterned to achieve specific tasks, this field is dominated by a body of state engineers recognized to have high-level scientific and administrative competences. It is structured by procedures and administrative hierarchies as well as by technical rules, norms, and standards. These are formalized and rationalized through technical, economic, political, and social needs; over time; they consolidate the field into an institution. The study documents the nuclear-safety institution as an integral part of the nuclear technostructure, which has historically used the specificity of its expertise as a buffer against outside interference.« less

Manned space flight nuclear system safety. Volume 3: Reactor system preliminary nuclear safety analysis. Part 3: Nuclear Safety Analysis Document (NSAD)

NASA Technical Reports Server (NTRS)

1972-01-01

Nuclear safety analysis as applied to a space base mission is presented. The nuclear safety analysis document summarizes the mission and the credible accidents/events which may lead to nuclear hazards to the general public. The radiological effects and associated consequences of the hazards are discussed in detail. The probability of occurrence is combined with the potential number of individuals exposed to or above guideline values to provide a measure of accident and total mission risk. The overall mission risk has been determined to be low with the potential exposure to or above 25 rem limited to less than 4 individuals per every 1000 missions performed. No radiological risk to the general public occurs during the prelaunch phase at KSC. The most significant risks occur from prolonged exposure to reactor debris following land impact generally associated with the disposal phase of the mission where fission product inventories can be high.

Aluminum Data Measurements and Evaluation for Criticality Safety Applications

NASA Astrophysics Data System (ADS)

Leal, L. C.; Guber, K. H.; Spencer, R. R.; Derrien, H.; Wright, R. Q.

2002-12-01

The Defense Nuclear Facility Safety Board (DNFSB) Recommendation 93-2 motivated the US Department of Energy (DOE) to develop a comprehensive criticality safety program to maintain and to predict the criticality of systems throughout the DOE complex. To implement the response to the DNFSB Recommendation 93-2, a Nuclear Criticality Safety Program (NCSP) was created including the following tasks: Critical Experiments, Criticality Benchmarks, Training, Analytical Methods, and Nuclear Data. The Nuclear Data portion of the NCSP consists of a variety of differential measurements performed at the Oak Ridge Electron Linear Accelerator (ORELA) at the Oak Ridge National Laboratory (ORNL), data analysis and evaluation using the generalized least-squares fitting code SAMMY in the resolved, unresolved, and high energy ranges, and the development and benchmark testing of complete evaluations for a nuclide for inclusion into the Evaluated Nuclear Data File (ENDF/B). This paper outlines the work performed at ORNL to measure, evaluate, and test the nuclear data for aluminum for applications in criticality safety problems.

78 FR 9902 - DOE Response to Recommendation 2012-2 of the Defense Nuclear Facilities Safety Board, Hanford...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-12

... DEPARTMENT OF ENERGY DOE Response to Recommendation 2012-2 of the Defense Nuclear Facilities Safety Board, Hanford Tank Farms Flammable Gas Safety Strategy; Correction AGENCY: Department of Energy... Facilities Safety Board, Hanford Tank Farms Flammable Gas Safety Strategy. This document corrects an error in...

75 FR 69648 - Safety Analysis Requirements for Defining Adequate Protection for the Public and the Workers

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-15

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD [Recommendation 2010-1] Safety Analysis Requirements for Defining Adequate Protection for the Public and the Workers AGENCY: Defense Nuclear Facilities Safety Board... Facilities Safety Board has made a recommendation to the Secretary of Energy requesting an amendment to the...

Technical Guidance from the International Safety Framework for Nuclear Power Source Applications in Outer Space for Design and Development Phases

NASA Astrophysics Data System (ADS)

Summerer, Leopold

2014-08-01

In 2009, the International Safety Framework for Nuclear Power Source Applications in Outer Space [1] has been adopted, following a multi-year process that involved all major space faring nations in the frame of the International Atomic Energy Agency and the UN Committee on the Peaceful Uses of Outer Space. The safety framework reflects an international consensus on best practices. After the older 1992 Principles Relevant to the Use of Nuclear Power Sources in Outer Space, it is the second document at UN level dedicated entirely to space nuclear power sources.This paper analyses aspects of the safety framework relevant for the design and development phases of space nuclear power sources. While early publications have started analysing the legal aspects of the safety framework, its technical guidance has not yet been subject to scholarly articles. The present paper therefore focuses on the technical guidance provided in the safety framework, in an attempt to assist engineers and practitioners to benefit from these.

Overview of Particle and Heavy Ion Transport Code System PHITS

NASA Astrophysics Data System (ADS)

Sato, Tatsuhiko; Niita, Koji; Matsuda, Norihiro; Hashimoto, Shintaro; Iwamoto, Yosuke; Furuta, Takuya; Noda, Shusaku; Ogawa, Tatsuhiko; Iwase, Hiroshi; Nakashima, Hiroshi; Fukahori, Tokio; Okumura, Keisuke; Kai, Tetsuya; Chiba, Satoshi; Sihver, Lembit

2014-06-01

A general purpose Monte Carlo Particle and Heavy Ion Transport code System, PHITS, is being developed through the collaboration of several institutes in Japan and Europe. The Japan Atomic Energy Agency is responsible for managing the entire project. PHITS can deal with the transport of nearly all particles, including neutrons, protons, heavy ions, photons, and electrons, over wide energy ranges using various nuclear reaction models and data libraries. It is written in Fortran language and can be executed on almost all computers. All components of PHITS such as its source, executable and data-library files are assembled in one package and then distributed to many countries via the Research organization for Information Science and Technology, the Data Bank of the Organization for Economic Co-operation and Development's Nuclear Energy Agency, and the Radiation Safety Information Computational Center. More than 1,000 researchers have been registered as PHITS users, and they apply the code to various research and development fields such as nuclear technology, accelerator design, medical physics, and cosmic-ray research. This paper briefly summarizes the physics models implemented in PHITS, and introduces some important functions useful for specific applications, such as an event generator mode and beam transport functions.

Assessment of RANS and LES Turbulence Modeling for Buoyancy-Aided/Opposed Forced and Mixed Convection

NASA Astrophysics Data System (ADS)

Clifford, Corey; Kimber, Mark

2017-11-01

Over the last 30 years, an industry-wide shift within the nuclear community has led to increased utilization of computational fluid dynamics (CFD) to supplement nuclear reactor safety analyses. One such area that is of particular interest to the nuclear community, specifically to those performing loss-of-flow accident (LOFA) analyses for next-generation very-high temperature reactors (VHTR), is the capacity of current computational models to predict heat transfer across a wide range of buoyancy conditions. In the present investigation, a critical evaluation of Reynolds-averaged Navier-Stokes (RANS) and large-eddy simulation (LES) turbulence modeling techniques is conducted based on CFD validation data collected from the Rotatable Buoyancy Tunnel (RoBuT) at Utah State University. Four different experimental flow conditions are investigated: (1) buoyancy-aided forced convection; (2) buoyancy-opposed forced convection; (3) buoyancy-aided mixed convection; (4) buoyancy-opposed mixed convection. Overall, good agreement is found for both forced convection-dominated scenarios, but an overly-diffusive prediction of the normal Reynolds stress is observed for the RANS-based turbulence models. Low-Reynolds number RANS models perform adequately for mixed convection, while higher-order RANS approaches underestimate the influence of buoyancy on the production of turbulence.

Nuclear safety for the space exploration initiative

NASA Technical Reports Server (NTRS)

Dix, Terry E.

1991-01-01

The results of a study to identify potential hazards arising from nuclear reactor power systems for use on the lunar and Martian surfaces, related safety issues, and resolutions of such issues by system design changes, operating procedures, and other means are presented. All safety aspects of nuclear reactor power systems from prelaunch ground handling to eventual disposal were examined consistent with the level of detail for SP-100 reactor design at the 1988 System Design Review and for launch vehicle and space transport vehicle designs and mission descriptions as defined in the 90-day Space Exploration Initiative (SEI) study. Information from previous aerospace nuclear safety studies was used where appropriate. Safety requirements for the SP-100 space nuclear reactor system were compiled. Mission profiles were defined with emphasis on activities after low earth orbit insertion. Accident scenarios were then qualitatively defined for each mission phase. Safety issues were identified for all mission phases with the aid of simplified event trees. Safety issue resolution approaches of the SP-100 program were compiled. Resolution approaches for those safety issues not covered by the SP-100 program were identified. Additionally, the resolution approaches of the SP-100 program were examined in light of the moon and Mars missions.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

This document is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

Brookhaven highlights for fiscal year 1991, October 1, 1990--September 30, 1991

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rowe, M.S.; Cohen, A.; Greenberg, D.

1991-12-31

This report highlights Brookhaven National Laboratory`s activities for fiscal year 1991. Topics from the four research divisions: Computing and Communications, Instrumentation, Reactors, and Safety and Environmental Protection are presented. The research programs at Brookhaven are diverse, as is reflected by the nine different scientific departments: Accelerator Development, Alternating Gradient Synchrotron, Applied Science, Biology, Chemistry, Medical, National Synchrotron Light Source, Nuclear Energy, and Physics. Administrative and managerial information about Brookhaven are also disclosed. (GHH)

Brookhaven highlights for fiscal year 1991, October 1, 1990--September 30, 1991

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rowe, M.S.; Cohen, A.; Greenberg, D.

1991-01-01

This report highlights Brookhaven National Laboratory's activities for fiscal year 1991. Topics from the four research divisions: Computing and Communications, Instrumentation, Reactors, and Safety and Environmental Protection are presented. The research programs at Brookhaven are diverse, as is reflected by the nine different scientific departments: Accelerator Development, Alternating Gradient Synchrotron, Applied Science, Biology, Chemistry, Medical, National Synchrotron Light Source, Nuclear Energy, and Physics. Administrative and managerial information about Brookhaven are also disclosed. (GHH)

DOE standard 3009 - a reasoned, practical approach to integrating criticality safety into SARs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vessard, S.G.

1995-12-31

In the past there have been efforts by the U.S. Department of Energy (DOE) to provide guidance on those elements that should be included in a facility`s safety analysis report (SAR). In particular, there are two DOE Orders (5480.23, {open_quotes}Nuclear Safety Analysis Reports,{close_quotes} and 5480.24, {open_quotes}Nuclear Criticality Safety{close_quotes}), an interpretive guidance document (NE-70, Interpretive Guidance for DOE Order 5480.24, {open_quotes}Nuclear Criticality Safety{close_quotes}), and DOE Standard DOE-STD-3009-94 {open_quotes}Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports.{close_quotes} Of these, the most practical and useful (pertaining to the application of criticality safety) is DOE-STD-3009-94. This paper is a reviewmore » of Chapters 3, 4, and 6 of this standard and how they provide very clear, helpful, and reasoned criticality safety guidance.« less

10 CFR 72.124 - Criteria for nuclear criticality safety.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Criteria for nuclear criticality safety. 72.124 Section 72.124 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C...

10 CFR 72.124 - Criteria for nuclear criticality safety.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Criteria for nuclear criticality safety. 72.124 Section 72.124 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C...

10 CFR 72.124 - Criteria for nuclear criticality safety.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Criteria for nuclear criticality safety. 72.124 Section 72.124 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C...

10 CFR 72.124 - Criteria for nuclear criticality safety.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Criteria for nuclear criticality safety. 72.124 Section 72.124 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C...

10 CFR 72.124 - Criteria for nuclear criticality safety.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Criteria for nuclear criticality safety. 72.124 Section 72.124 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C...

Providing Nuclear Criticality Safety Analysis Education through Benchmark Experiment Evaluation

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess; J. Blair Briggs; David W. Nigg

2009-11-01

One of the challenges that today's new workforce of nuclear criticality safety engineers face is the opportunity to provide assessment of nuclear systems and establish safety guidelines without having received significant experience or hands-on training prior to graduation. Participation in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and/or the International Reactor Physics Experiment Evaluation Project (IRPhEP) provides students and young professionals the opportunity to gain experience and enhance critical engineering skills.

Passive Safety Features Evaluation of KIPT Neutron Source Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhong, Zhaopeng; Gohar, Yousry

2016-06-01

Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have cooperated on the development, design, and construction of a neutron source facility. The facility was constructed at Kharkov, Ukraine and its commissioning process is underway. It will be used to conduct basic and applied nuclear research, produce medical isotopes, and train young nuclear specialists. The facility has an electron accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100 MeV electrons. Tungsten or natural uranium is the target material for generating neutrons driving the subcritical assembly. The subcritical assemblymore » is composed of WWR-M2 - Russian fuel assemblies with U-235 enrichment of 19.7 wt%, surrounded by beryllium reflector assembles and graphite blocks. The subcritical assembly is seated in a water tank, which is a part of the primary cooling loop. During normal operation, the water coolant operates at room temperature and the total facility power is ~300 KW. The passive safety features of the facility are discussed in in this study. Monte Carlo computer code MCNPX was utilized in the analyses with ENDF/B-VII.0 nuclear data libraries. Negative reactivity temperature feedback was consistently observed, which is important for the facility safety performance. Due to the design of WWR-M2 fuel assemblies, slight water temperature increase and the corresponding water density decrease produce large reactivity drop, which offset the reactivity gain by mistakenly loading an additional fuel assembly. The increase of fuel temperature also causes sufficiently large reactivity decrease. This enhances the facility safety performance because fuel temperature increase provides prompt negative reactivity feedback. The reactivity variation due to an empty fuel position filled by water during the fuel loading process is examined. Also, the loading mistakes of removing beryllium reflector assemblies and replacing them with dummy assemblies were analyzed. In all these circumstances, the reactivity change results do not cause any safety concerns.« less

Parallel computation safety analysis irradiation targets fission product molybdenum in neutronic aspect using the successive over-relaxation algorithm

NASA Astrophysics Data System (ADS)

Susmikanti, Mike; Dewayatna, Winter; Sulistyo, Yos

2014-09-01

One of the research activities in support of commercial radioisotope production program is a safety research on target FPM (Fission Product Molybdenum) irradiation. FPM targets form a tube made of stainless steel which contains nuclear-grade high-enrichment uranium. The FPM irradiation tube is intended to obtain fission products. Fission materials such as Mo99 used widely the form of kits in the medical world. The neutronics problem is solved using first-order perturbation theory derived from the diffusion equation for four groups. In contrast, Mo isotopes have longer half-lives, about 3 days (66 hours), so the delivery of radioisotopes to consumer centers and storage is possible though still limited. The production of this isotope potentially gives significant economic value. The criticality and flux in multigroup diffusion model was calculated for various irradiation positions and uranium contents. This model involves complex computation, with large and sparse matrix system. Several parallel algorithms have been developed for the sparse and large matrix solution. In this paper, a successive over-relaxation (SOR) algorithm was implemented for the calculation of reactivity coefficients which can be done in parallel. Previous works performed reactivity calculations serially with Gauss-Seidel iteratives. The parallel method can be used to solve multigroup diffusion equation system and calculate the criticality and reactivity coefficients. In this research a computer code was developed to exploit parallel processing to perform reactivity calculations which were to be used in safety analysis. The parallel processing in the multicore computer system allows the calculation to be performed more quickly. This code was applied for the safety limits calculation of irradiated FPM targets containing highly enriched uranium. The results of calculations neutron show that for uranium contents of 1.7676 g and 6.1866 g (× 106 cm-1) in a tube, their delta reactivities are the still within safety limits; however, for 7.9542 g and 8.838 g (× 106 cm-1) the limits were exceeded.

Application of Framework for Integrating Safety, Security and Safeguards (3Ss) into the Design Of Used Nuclear Fuel Storage Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Badwan, Faris M.; Demuth, Scott F

Department of Energy’s Office of Nuclear Energy, Fuel Cycle Research and Development develops options to the current commercial fuel cycle management strategy to enable the safe, secure, economic, and sustainable expansion of nuclear energy while minimizing proliferation risks by conducting research and development focused on used nuclear fuel recycling and waste management to meet U.S. needs. Used nuclear fuel is currently stored onsite in either wet pools or in dry storage systems, with disposal envisioned in interim storage facility and, ultimately, in a deep-mined geologic repository. The safe management and disposition of used nuclear fuel and/or nuclear waste is amore » fundamental aspect of any nuclear fuel cycle. Integrating safety, security, and safeguards (3Ss) fully in the early stages of the design process for a new nuclear facility has the potential to effectively minimize safety, proliferation, and security risks. The 3Ss integration framework could become the new national and international norm and the standard process for designing future nuclear facilities. The purpose of this report is to develop a framework for integrating the safety, security and safeguards concept into the design of Used Nuclear Fuel Storage Facility (UNFSF). The primary focus is on integration of safeguards and security into the UNFSF based on the existing Nuclear Regulatory Commission (NRC) approach to addressing the safety/security interface (10 CFR 73.58 and Regulatory Guide 5.73) for nuclear power plants. The methodology used for adaptation of the NRC safety/security interface will be used as the basis for development of the safeguards /security interface and later will be used as the basis for development of safety and safeguards interface. Then this will complete the integration cycle of safety, security, and safeguards. The overall methodology for integration of 3Ss will be proposed, but only the integration of safeguards and security will be applied to the design of the UNFSF. The framework for integration of safeguards and security into the UNFSF will include 1) identification of applicable regulatory requirements, 2) selection of a common system that share dual safeguard and security functions, 3) development of functional design criteria and design requirements for the selected system, 4) identification and integration of the dual safeguards and security design requirements, and 5) assessment of the integration and potential benefit.« less

10 CFR 35.57 - Training for experienced Radiation Safety Officer, teletherapy or medical physicist, authorized...

Code of Federal Regulations, 2011 CFR

2011-01-01

..., teletherapy or medical physicist, authorized medical physicist, authorized user, nuclear pharmacist, and authorized nuclear pharmacist. 35.57 Section 35.57 Energy NUCLEAR REGULATORY COMMISSION MEDICAL USE OF... pharmacist, and authorized nuclear pharmacist. (a)(1) An individual identified as a Radiation Safety Officer...

10 CFR 35.57 - Training for experienced Radiation Safety Officer, teletherapy or medical physicist, authorized...

Code of Federal Regulations, 2014 CFR

2014-01-01

..., teletherapy or medical physicist, authorized medical physicist, authorized user, nuclear pharmacist, and authorized nuclear pharmacist. 35.57 Section 35.57 Energy NUCLEAR REGULATORY COMMISSION MEDICAL USE OF... pharmacist, and authorized nuclear pharmacist. (a)(1) An individual identified as a Radiation Safety Officer...

10 CFR 35.57 - Training for experienced Radiation Safety Officer, teletherapy or medical physicist, authorized...

Code of Federal Regulations, 2012 CFR

2012-01-01

..., teletherapy or medical physicist, authorized medical physicist, authorized user, nuclear pharmacist, and authorized nuclear pharmacist. 35.57 Section 35.57 Energy NUCLEAR REGULATORY COMMISSION MEDICAL USE OF... pharmacist, and authorized nuclear pharmacist. (a)(1) An individual identified as a Radiation Safety Officer...

10 CFR 35.57 - Training for experienced Radiation Safety Officer, teletherapy or medical physicist, authorized...

Code of Federal Regulations, 2013 CFR

2013-01-01

..., teletherapy or medical physicist, authorized medical physicist, authorized user, nuclear pharmacist, and authorized nuclear pharmacist. 35.57 Section 35.57 Energy NUCLEAR REGULATORY COMMISSION MEDICAL USE OF... pharmacist, and authorized nuclear pharmacist. (a)(1) An individual identified as a Radiation Safety Officer...

10 CFR 35.57 - Training for experienced Radiation Safety Officer, teletherapy or medical physicist, authorized...

Code of Federal Regulations, 2010 CFR

2010-01-01

..., teletherapy or medical physicist, authorized medical physicist, authorized user, nuclear pharmacist, and authorized nuclear pharmacist. 35.57 Section 35.57 Energy NUCLEAR REGULATORY COMMISSION MEDICAL USE OF... pharmacist, and authorized nuclear pharmacist. (a)(1) An individual identified as a Radiation Safety Officer...

History and Organizations for Radiological Protection.

PubMed

Kang, Keon Wook

2016-02-01

International Commission on Radiological Protection (ICRP), an independent international organization established in 1925, develops, maintains, and elaborates radiological protection standards, legislation, and guidelines. United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) provides scientific evidence. World Health Organization (WHO) and International Atomic Energy Agency (IAEA) utilise the ICRP recommendations to implement radiation protection in practice. Finally, radiation protection agencies in each country adopt the policies, and adapt them to each situation. In Korea, Nuclear Safety and Security Commission is the governmental body for nuclear safety regulation and Korea Institute of Nuclear Safety is a public organization for technical support and R&D in nuclear safety and radiation protection.

The International Safety Framework for nuclear power source applications in outer space-Useful and substantial guidance

NASA Astrophysics Data System (ADS)

Summerer, L.; Wilcox, R. E.; Bechtel, R.; Harbison, S.

2015-06-01

In 2009, the International Safety Framework for Nuclear Power Source Applications in Outer Space was adopted, following a multi-year process that involved all major space faring nations under the auspices of a partnership between the UN Committee on the Peaceful Uses of Outer Space and the International Atomic Energy Agency. The Safety Framework reflects an international consensus on best practices to achieve safety. Following the 1992 UN Principles Relevant to the Use of Nuclear Power Sources in Outer Space, it is the second attempt by the international community to draft guidance promoting the safety of applications of nuclear power sources in space missions. NPS applications in space have unique safety considerations compared with terrestrial applications. Mission launch and outer space operational requirements impose size, mass and other space environment limitations not present for many terrestrial nuclear facilities. Potential accident conditions could expose nuclear power sources to extreme physical conditions. The Safety Framework is structured to provide guidance for both the programmatic and technical aspects of safety. In addition to sections containing specific guidance for governments and for management, it contains technical guidance pertinent to the design, development and all mission phases of space NPS applications. All sections of the Safety Framework contain elements directly relevant to engineers and space mission designers for missions involving space nuclear power sources. The challenge for organisations and engineers involved in the design and development processes of space nuclear power sources and applications is to implement the guidance provided in the Safety Framework by integrating it into the existing standard space mission infrastructure of design, development and operational requirements, practices and processes. This adds complexity to the standard space mission and launch approval processes. The Safety Framework is deliberately generic to remain relevantly independent of technological progress, of national organisational setups and of space mission types. Implementing its guidance therefore leaves room for interpretation and adaptation. Relying on reported practices, we analyse the guidance particularly relevant to engineers and space mission designers.

Organizational Culture for Safety, Security, and Safeguards in New Nuclear Power Countries

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kovacic, Donald N

2015-01-01

This chapter will contain the following sections: Existing international norms and standards for developing the infrastructure to support new nuclear power programs The role of organizational culture and how it supports the safe, secure, and peaceful application of nuclear power Identifying effective and efficient strategies for implementing safety, security and safeguards in nuclear operations Challenges identified in the implementation of safety, security and safeguards Potential areas for future collaboration between countries in order to support nonproliferation culture

Experimental Criticality Benchmarks for SNAP 10A/2 Reactor Cores

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krass, A.W.

2005-12-19

This report describes computational benchmark models for nuclear criticality derived from descriptions of the Systems for Nuclear Auxiliary Power (SNAP) Critical Assembly (SCA)-4B experimental criticality program conducted by Atomics International during the early 1960's. The selected experimental configurations consist of fueled SNAP 10A/2-type reactor cores subject to varied conditions of water immersion and reflection under experimental control to measure neutron multiplication. SNAP 10A/2-type reactor cores are compact volumes fueled and moderated with the hydride of highly enriched uranium-zirconium alloy. Specifications for the materials and geometry needed to describe a given experimental configuration for a model using MCNP5 are provided. Themore » material and geometry specifications are adequate to permit user development of input for alternative nuclear safety codes, such as KENO. A total of 73 distinct experimental configurations are described.« less

77 FR 36015 - Atomic Safety and Licensing Board; Entergy Nuclear Operations, Inc. (Indian Point Nuclear...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-15

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-247-LR and 50-286-LR; ASLBP No. 07-858-03-LR-BD01] Atomic Safety and Licensing Board; Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3); Notice of Hearing (Application for License Renewal) June 8, 2012. Before Administrative Judges: Lawrence G. McDade, Chairman, Dr....

77 FR 20853 - Entergy Nuclear Operations, Inc.; Establishment of Atomic Safety and Licensing Board

Federal Register 2010, 2011, 2012, 2013, 2014

2012-04-06

... Operations, Inc.; Establishment of Atomic Safety and Licensing Board Pursuant to delegation by the Commission... Atomic Safety and Licensing Board (Board) is being established to preside over the following proceeding... administrative judges: Ann Marshall Young, Chair, Atomic Safety and Licensing Board Panel, U.S. Nuclear...

75 FR 44817 - Notice of Availability of Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-29

... Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services, National... Enrichment Branch, Division of Fuel Cycle Safety and Safeguards, Office of Nuclear Material Safety and... Enrichment Branch, Division of Fuel Cycle Safety and Safeguards, Office of Nuclear Material Safety and...

A COMPARATIVE ANALYSIS BETWEEN FRANCE AND JAPAN ON LOCAL GOVERNMENTS' INVOLVEMENT IN NUCLEAR SAFETY GOVERNANCE

NASA Astrophysics Data System (ADS)

Sugawara, Shin-Etsu; Shiroyama, Hideaki

This paper shows a comparative analysis between France and Japan on the way of the local governments' involvement in nuclear safety governance through some interviews. In France, a law came into force that requires related local governments to establish "Commision Locale d'Information" (CLI), which means the local governments officially involve in nuclear regulatory activity. Meanwhile, in Japan, related local governments substantially involve in the operation of nuclear facilities through the "safety agreements" in spite of the lack of legal authority. As a result of comparative analysis, we can point out some institutional input from French cases as follows: to clarify the local governments' roles in the nuclear regulation system, to establish the official channels of communication among nuclear utilities, national regulatory authorities and local governments, and to stipulate explicitly the transparency as a purpose of safety regulation.

Keeping an Eye on the Prize

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hazi, A U

2007-02-06

Setting performance goals is part of the business plan for almost every company. The same is true in the world of supercomputers. Ten years ago, the Department of Energy (DOE) launched the Accelerated Strategic Computing Initiative (ASCI) to help ensure the safety and reliability of the nation's nuclear weapons stockpile without nuclear testing. ASCI, which is now called the Advanced Simulation and Computing (ASC) Program and is managed by DOE's National Nuclear Security Administration (NNSA), set an initial 10-year goal to obtain computers that could process up to 100 trillion floating-point operations per second (teraflops). Many computer experts thought themore » goal was overly ambitious, but the program's results have proved them wrong. Last November, a Livermore-IBM team received the 2005 Gordon Bell Prize for achieving more than 100 teraflops while modeling the pressure-induced solidification of molten metal. The prestigious prize, which is named for a founding father of supercomputing, is awarded each year at the Supercomputing Conference to innovators who advance high-performance computing. Recipients for the 2005 prize included six Livermore scientists--physicists Fred Streitz, James Glosli, and Mehul Patel and computer scientists Bor Chan, Robert Yates, and Bronis de Supinski--as well as IBM researchers James Sexton and John Gunnels. This team produced the first atomic-scale model of metal solidification from the liquid phase with results that were independent of system size. The record-setting calculation used Livermore's domain decomposition molecular-dynamics (ddcMD) code running on BlueGene/L, a supercomputer developed by IBM in partnership with the ASC Program. BlueGene/L reached 280.6 teraflops on the Linpack benchmark, the industry standard used to measure computing speed. As a result, it ranks first on the list of Top500 Supercomputer Sites released in November 2005. To evaluate the performance of nuclear weapons systems, scientists must understand how materials behave under extreme conditions. Because experiments at high pressures and temperatures are often difficult or impossible to conduct, scientists rely on computer models that have been validated with obtainable data. Of particular interest to weapons scientists is the solidification of metals. ''To predict the performance of aging nuclear weapons, we need detailed information on a material's phase transitions'', says Streitz, who leads the Livermore-IBM team. For example, scientists want to know what happens to a metal as it changes from molten liquid to a solid and how that transition affects the material's characteristics, such as its strength.« less

Nuclear Safety for Space Systems

NASA Astrophysics Data System (ADS)

Offiong, Etim

2010-09-01

It is trite, albeit a truism, to say that nuclear power can provide propulsion thrust needed to launch space vehicles and also, to provide electricity for powering on-board systems, especially for missions to the Moon, Mars and other deep space missions. Nuclear Power Sources(NPSs) are known to provide more capabilities than solar power, fuel cells and conventional chemical means. The worry has always been that of safety. The earliest superpowers(US and former Soviet Union) have designed and launched several nuclear-powered systems, with some failures. Nuclear failures and accidents, however little the number, could be far-reaching geographically, and are catastrophic to humans and the environment. Building on the numerous research works on nuclear power on Earth and in space, this paper seeks to bring to bear, issues relating to safety of space systems - spacecrafts, astronauts, Earth environment and extra terrestrial habitats - in the use and application of nuclear power sources. It also introduces a new formal training course in Space Systems Safety.

Demonstration of emulator-based Bayesian calibration of safety analysis codes: Theory and formulation

DOE PAGES

Yurko, Joseph P.; Buongiorno, Jacopo; Youngblood, Robert

2015-05-28

System codes for simulation of safety performance of nuclear plants may contain parameters whose values are not known very accurately. New information from tests or operating experience is incorporated into safety codes by a process known as calibration, which reduces uncertainty in the output of the code and thereby improves its support for decision-making. The work reported here implements several improvements on classic calibration techniques afforded by modern analysis techniques. The key innovation has come from development of code surrogate model (or code emulator) construction and prediction algorithms. Use of a fast emulator makes the calibration processes used here withmore » Markov Chain Monte Carlo (MCMC) sampling feasible. This study uses Gaussian Process (GP) based emulators, which have been used previously to emulate computer codes in the nuclear field. The present work describes the formulation of an emulator that incorporates GPs into a factor analysis-type or pattern recognition-type model. This “function factorization” Gaussian Process (FFGP) model allows overcoming limitations present in standard GP emulators, thereby improving both accuracy and speed of the emulator-based calibration process. Calibration of a friction-factor example using a Method of Manufactured Solution is performed to illustrate key properties of the FFGP based process.« less

Transactions of the Twenty-First Water Reactor Safety Information Meeting

DOE Office of Scientific and Technical Information (OSTI.GOV)

Monteleone, S.

1993-10-01

This report contains summaries of papers on reactor safety research to be presented at the 21st Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 25--27, 1993. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningfulmore » discussion and information exchange during the course of the meeting and are given in the order of their presentation in each session.« less

Digital Signal Processing Methods for Safety Systems Employed in Nuclear Power Industry

NASA Astrophysics Data System (ADS)

Popescu, George

Some of the major safety concerns in the nuclear power industry focus on the readiness of nuclear power plant safety systems to respond to an abnormal event, the security of special nuclear materials in used nuclear fuels, and the need for physical security to protect personnel and reactor safety systems from an act of terror. Routine maintenance and tests of all nuclear reactor safety systems are performed on a regular basis to confirm the ability of these systems to operate as expected. However, these tests do not determine the reliability of these safety systems and whether the systems will perform for the duration of an accident and whether they will perform their tasks without failure after being engaged. This research has investigated the progression of spindle asynchronous error motion determined from spindle accelerations to predict bearings failure onset. This method could be applied to coolant pumps that are essential components of emergency core cooling systems at all nuclear power plants. Recent security upgrades mandated by the Nuclear Regulatory Commission and the Department of Homeland Security have resulted in implementation of multiple physical security barriers around all of the commercial and research nuclear reactors in the United States. A second part of this research attempts to address an increased concern about illegal trafficking of Special Nuclear Materials (SNM). This research describes a multi element scintillation detector system designed for non - invasive (passive) gamma ray surveillance for concealed SNM that may be within an area or sealed in a package, vehicle or shipping container. Detection capabilities of the system were greatly enhanced through digital signal processing, which allows the combination of two very powerful techniques: 1) Compton Suppression (CS) and 2) Pulse Shape Discrimination (PSD) with less reliance on complicated analog instrumentation.

Annual Report To Congress. Department of Energy Activities Relating to the Defense Nuclear Facilities Safety Board, Calendar Year 2003

DOE Office of Scientific and Technical Information (OSTI.GOV)

None, None

2004-02-28

The Department of Energy (Department) submits an Annual Report to Congress each year detailing the Department’s activities relating to the Defense Nuclear Facilities Safety Board (Board), which provides advice and recommendations to the Secretary of Energy (Secretary) regarding public health and safety issues at the Department’s defense nuclear facilities. In 2003, the Department continued ongoing activities to resolve issues identified by the Board in formal recommendations and correspondence, staff issue reports pertaining to Department facilities, and public meetings and briefings. Additionally, the Department is implementing several key safety initiatives to address and prevent safety issues: safety culture and review ofmore » the Columbia accident investigation; risk reduction through stabilization of excess nuclear materials; the Facility Representative Program; independent oversight and performance assurance; the Federal Technical Capability Program (FTCP); executive safety initiatives; and quality assurance activities. The following summarizes the key activities addressed in this Annual Report.« less

Safety Regulation of Nuclear Power Plant License Renewal

NASA Astrophysics Data System (ADS)

Zhang, Qiaoe; Liu, Ting; Qi, Yuan; Yang, LiLi

2018-01-01

China’s regulations stipulate that a nuclear power plant license is valid for a design life period (generally 30 or 40 years). Whether the nuclear power plant’s license is renewed after the expiration of the license is to be determined based on the safety and economy of the nuclear power plant..

Physical and Mechanical Metallurgy of Zirconium Alloys for Nuclear Applications: A Multi-Scale Computational Study

NASA Astrophysics Data System (ADS)

Glazoff, Michael Vasily

In the post-Fukushima world, thermal and structural stability of materials under extreme conditions is an important issue for the safety of nuclear reactors. Because the nuclear industry will continue using zirconium (Zr) cladding for the foreseeable future, it becomes critical to gain a fundamental understanding of several interconnected problems. First, what are the thermodynamic and kinetic factors affecting oxidation and hydrogen pick-up by these materials at normal, off-normal conditions, and in long-term storage? Secondly, what protective coatings could be used in order to gain valuable time at off-normal conditions (temperature exceeds ~1200°C (2200°F)? Thirdly, the kinetics of the coating's oxidation must be understood. Lastly, one needs automated inspection algorithms allowing identifying cladding's defects. This work attempts to explore the problem from a computational perspective, utilizing first principles atomistic simulations, computational thermodynamics, plasticity theory, and morphological algorithms of image processing for defect identification. It consists of the four parts dealing with these four problem areas preceded by the introduction. In the 1st part, computational thermodynamics and ab initio calculations were used to shed light upon the different stages of zircaloy oxidation and hydrogen pickup, and microstructure optimization to increase thermal stability. The 2 nd part describes the kinetic theory of oxidation of the several materials considered to be perspective coatings for Zr alloys: SiC and ZrSiO4. The 3rd part deals with understanding the respective roles of the two different plasticity mechanisms in Zr nuclear alloys: twinning (at low T) and crystallographic slip (higher T's). For that goal, an advanced plasticity model was proposed. In the 4th part projectional algorithms for defect identification in zircaloy coatings are described. Conclusions and recommendations are presented in the 5th part. This integrative approach's value is in developing multi-faceted understanding of complex processes taking place in nuclear fuel rods. It helped identify several problems pertaining to the safe operations with nuclear fuel: limits of temperature that should be strictly obeyed in storage to retard zircaloy hydriding; understanding the benefits and limitations of coatings; developing in-depth understanding of Zr plasticity; developing original algorithms for defect identification in SiC-braided zircaloy. The obtained results will be useful for the nuclear industry.

Radiation Exposure in X-Ray and CT Examinations

MedlinePlus

... disease. See the X-ray, Interventional Radiology and Nuclear Medicine Radiation Safety page for more information. top of page ... and Radiation Safety X-ray, Interventional Radiology and Nuclear Medicine Radiation Safety Videos related to Radiation Dose in X- ...

75 FR 52046 - Development of U.S. Nuclear Regulatory Commission Safety Culture Policy Statement: Public Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-24

... focuses on the unique aspects of nuclear safety and security and highlights the Commission's expectations... safety culture and (2) high level descriptions or traits of areas important to safety culture. The... NRC headquarters. Please allow time at both locations to register with building security upon entering...

77 FR 32146 - Safety Evaluation Report, International Isotopes Fluorine Products, Inc., Fluorine Extraction...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-05-31

... NUCLEAR REGULATORY COMMISSION [Docket No. 40-9086; NRC-2010-0143] Safety Evaluation Report... Deconversion Plant (FEP/DUP) and will be located in Lea County, New Mexico. The NRC has prepared a Safety..., Deconversion and Enrichment Branch, Division of Fuel Cycle Safety and Safeguards, Office of Nuclear Material...

Linking empowering leadership to safety participation in nuclear power plants: a structural equation model.

PubMed

Martínez-Córcoles, Mario; Schöbel, Markus; Gracia, Francisco J; Tomás, Inés; Peiró, José M

2012-07-01

Safety participation is of paramount importance in guaranteeing the safe running of nuclear power plants. The present study examined the effects of empowering leadership on safety participation. Based on a sample of 495 employees from two Spanish nuclear power plants, structural equation modeling showed that empowering leadership has a significant relationship with safety participation, which is mediated by collaborative team learning. In addition, the results revealed that the relationship between empowering leadership and collaborative learning is partially mediated by the promotion of dialogue and open communication. The implications of these findings for safety research and their practical applications are outlined. An empowering leadership style enhances workers' safety performance, particularly safety participation behaviors. Safety participation is recommended to detect possible rule inconsistencies or misunderstood procedures and make workers aware of critical safety information and issues. Crown Copyright © 2012. Published by Elsevier Ltd. All rights reserved.

Norms Versus Security: What is More Important to Japan’s View of Nuclear Weapons

DTIC Science & Technology

2017-03-01

objectives: “1) prevent the spread of nuclear weapons and weapons technology, 2) promote cooperation in the peaceful uses of nuclear energy , and 3...http://www.world- nuclear.org/information-library/safety-and-security/safety-of-plants/fukushima-accident.aspx. 40 “Japanese Wary of Nuclear Energy ...PewResearchCenter, accessed February 22, 2017. http://www.pewglobal.org/2012/06/05/japanese-wary-of- nuclear - energy / 41 Malcolm Foster, “Thousands

VVER Reactor Safety in Eastern Europe and Former Soviet Union

NASA Astrophysics Data System (ADS)

Papadopoulou, Demetra

2012-02-01

VVER Soviet-designed reactors that operate in Eastern Europe and former Soviet republics have heightened international concern for years due to major safety deficiencies. The governments of countries with VVER reactors have invested millions of dollars toward improving the safety of their nuclear power plants. Most of these reactors will continue to operate for the foreseeable future since they provide urgently-needed electrical power. Given this situation, this paper assesses the radiological consequences of a major nuclear accident in Eastern Europe. The paper also chronicles the efforts launched by the international nuclear community to improve the safety of the reactors and notes the progress made so far through extensive collaborative efforts in Armenia, Bulgaria, the Czech Republic, Hungary, Kazakhstan, Lithuania, Russia, Slovakia, and Ukraine to reduce the risks of nuclear accidents. Western scientific and technical staff collaborated with these countries to improve the safety of their reactor operations by strengthening the ability of the regulator to perform its oversight function, installing safety equipment and technologies, investing time in safety training, and working diligently to establish an enduring safety culture. Still, continued safety improvement efforts are necessary to ensure safe operating practices and achieve timely phase-out of older plants.

Annual report to Congress. Department of Energy activities relating to the Defense Nuclear Facilities Safety Board, calendar year 2000

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

2001-03-01

This Annual Report to the Congress describes the Department of Energy's activities in response to formal recommendations and other interactions with the Defense Nuclear Facilities Safety Board. During 2000, the Department completed its implementation and proposed closure of one Board recommendation and completed all implementation plan milestones associated with two additional Board recommendations. Also in 2000, the Department formally accepted two new Board recommendations and developed implementation plans in response to those recommendations. The Department also made significant progress with a number of broad-based safety initiatives. These include initial implementation of integrated safety management at field sites and within headquartersmore » program offices, issuance of a nuclear safety rule, and continued progress on stabilizing excess nuclear materials to achieve significant risk reduction.« less

Assessment of radiation safety awareness among nuclear medicine nurses: a pilot study

NASA Astrophysics Data System (ADS)

Yunus, N. A.; Abdullah, M. H. R. O.; Said, M. A.; Ch'ng, P. E.

2014-11-01

All nuclear medicine nurses need to have some knowledge and awareness on radiation safety. At present, there is no study to address this issue in Malaysia. The aims of this study were (1) to determine the level of knowledge and awareness on radiation safety among nuclear medicine nurses at Putrajaya Hospital in Malaysia and (2) to assess the effectiveness of a training program provided by the hospital to increase the knowledge and awareness of the nuclear medicine nurses. A total of 27 respondents attending a training program on radiation safety were asked to complete a questionnaire. The questionnaire consists 16 items and were categorized into two main areas, namely general radiation knowledge and radiation safety. Survey data were collected before and after the training and were analyzed using descriptive statistics and paired sample t-test. Respondents were scored out of a total of 16 marks with 8 marks for each area. The findings showed that the range of total scores obtained by the nuclear medicine nurses before and after the training were 6-14 (with a mean score of 11.19) and 13-16 marks (with a mean score of 14.85), respectively. Findings also revealed that the mean score for the area of general radiation knowledge (7.59) was higher than that of the radiation safety (7.26). Currently, the knowledge and awareness on radiation safety among the nuclear medicine nurses are at the moderate level. It is recommended that a national study be conducted to assess and increase the level of knowledge and awareness among all nuclear medicine nurses in Malaysia.

10 CFR 830.200 - Scope.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Scope. 830.200 Section 830.200 Energy DEPARTMENT OF ENERGY NUCLEAR SAFETY MANAGEMENT Safety Basis Requirements § 830.200 Scope. This Subpart establishes safety basis requirements for hazard category 1, 2, and 3 DOE nuclear facilities. ...

10 CFR 830.200 - Scope.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 4 2014-01-01 2014-01-01 false Scope. 830.200 Section 830.200 Energy DEPARTMENT OF ENERGY NUCLEAR SAFETY MANAGEMENT Safety Basis Requirements § 830.200 Scope. This Subpart establishes safety basis requirements for hazard category 1, 2, and 3 DOE nuclear facilities. ...

10 CFR 830.200 - Scope.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 4 2011-01-01 2011-01-01 false Scope. 830.200 Section 830.200 Energy DEPARTMENT OF ENERGY NUCLEAR SAFETY MANAGEMENT Safety Basis Requirements § 830.200 Scope. This Subpart establishes safety basis requirements for hazard category 1, 2, and 3 DOE nuclear facilities. ...

10 CFR 830.200 - Scope.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 4 2012-01-01 2012-01-01 false Scope. 830.200 Section 830.200 Energy DEPARTMENT OF ENERGY NUCLEAR SAFETY MANAGEMENT Safety Basis Requirements § 830.200 Scope. This Subpart establishes safety basis requirements for hazard category 1, 2, and 3 DOE nuclear facilities. ...

76 FR 44623 - Atomic Safety and Licensing Board; In the Matter of Nuclear Innovation North America LLC (South...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-07-26

...] Atomic Safety and Licensing Board; In the Matter of Nuclear Innovation North America LLC (South Texas..., Dr. Gary S. Arnold, Dr. Randall J. Charbeneau. The Atomic Safety and Licensing Board hereby gives... as follows: Mail: Administrative Judge Michael M. Gibson, Atomic Safety and Licensing Board Panel...

10 CFR 1703.105 - Requests for board records not available through the public reading room (FOIA requests).

Code of Federal Regulations, 2014 CFR

2014-01-01

... public reading room (FOIA requests). 1703.105 Section 1703.105 Energy DEFENSE NUCLEAR FACILITIES SAFETY... address for such requests is: Designated FOIA Officer, Defense Nuclear Facilities Safety Board, 625... pertain to an immediate source of risk to the public health and safety or worker safety at a defense...

10 CFR 1703.105 - Requests for board records not available through the public reading room (FOIA requests).

Code of Federal Regulations, 2013 CFR

2013-01-01

... public reading room (FOIA requests). 1703.105 Section 1703.105 Energy DEFENSE NUCLEAR FACILITIES SAFETY... address for such requests is: Designated FOIA Officer, Defense Nuclear Facilities Safety Board, 625... pertain to an immediate source of risk to the public health and safety or worker safety at a defense...

10 CFR 1703.105 - Requests for board records not available through the public reading room (FOIA requests).

Code of Federal Regulations, 2012 CFR

2012-01-01

... public reading room (FOIA requests). 1703.105 Section 1703.105 Energy DEFENSE NUCLEAR FACILITIES SAFETY... address for such requests is: Designated FOIA Officer, Defense Nuclear Facilities Safety Board, 625... pertain to an immediate source of risk to the public health and safety or worker safety at a defense...

10 CFR 1703.105 - Requests for board records not available through the public reading room (FOIA requests).

Code of Federal Regulations, 2010 CFR

2010-01-01

... public reading room (FOIA requests). 1703.105 Section 1703.105 Energy DEFENSE NUCLEAR FACILITIES SAFETY... address for such requests is: Designated FOIA Officer, Defense Nuclear Facilities Safety Board, 625... pertain to an immediate source of risk to the public health and safety or worker safety at a defense...

10 CFR 1703.105 - Requests for board records not available through the public reading room (FOIA requests).

Code of Federal Regulations, 2011 CFR

2011-01-01

... public reading room (FOIA requests). 1703.105 Section 1703.105 Energy DEFENSE NUCLEAR FACILITIES SAFETY... address for such requests is: Designated FOIA Officer, Defense Nuclear Facilities Safety Board, 625... pertain to an immediate source of risk to the public health and safety or worker safety at a defense...

ASME Nuclear Crane Standards for Enhanced Crane Safety and Increased Profit

NASA Astrophysics Data System (ADS)

Parkhurst, Stephen N.

2000-01-01

The ASME NOG-1 standard, 'Rules for Construction of Overhead and Gantry Cranes', covers top running cranes for nuclear facilities; with the ASME NUM-1 standard, 'Rules for Construction of Cranes, Monorails, and Hoists', covering the single girder, underhung, wall and jib cranes, as well as the monorails and hoists. These two ASME nuclear crane standards provide criteria for designing, inspecting and testing overhead handling equipment with enhanced safety to meet the 'defense-in-depth' approach of the United States Nuclear Regulatory Commission (USNRC) documents NUREG 0554 and NUREG 0612. In addition to providing designs for enhanced safety, the ASME nuclear crane standards provide a basis for purchasing overhead handling equipment with standard safety features, based upon accepted engineering principles, and including performance and environmental parameters specific to nuclear facilities. The ASME NOG-1 and ASME NUM-1 standards not only provide enhanced safety for handling a critical load, but also increase profit by minimizing the possibility of load drops, by reducing cumbersome operating restrictions, and by providing the foundation for a sound licensing position. The ASME nuclear crane standards can also increase profit by providing the designs and information to help ensure that the right standard equipment is purchased. Additionally, the ASME nuclear crane standards can increase profit by providing designs and information to help address current issues, such as the qualification of nuclear plant cranes for making 'planned engineered lifts' for steam generator replacement and decommissioning.

75 FR 21605 - Sunshine Act Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-26

... depth federal safety management and oversight policies being developed by DOE and NNSA for defense... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Sunshine Act Notice AGENCY: Defense Nuclear Facilities... in the Sunshine Act'' (5 U.S.C. 552b), notice is hereby given of the Defense Nuclear Facilities...

An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses--Criticality (keff) Predictions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scaglione, John M; Mueller, Don; Wagner, John C

2011-01-01

One of the most significant remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation - in particular, the availability and use of applicable measured data to support validation, especially for fission products. Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. U.S. Nuclear Regulatory Commission (NRC) staff have noted that the rationale for restricting their Interim Staff Guidance on burnup credit (ISG-8) to actinide-only ismore » based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address the issue of validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach (both depletion and criticality) for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the criticality (k{sub eff}) validation approach, and resulting observations and recommendations. Validation of the isotopic composition (depletion) calculations is addressed in a companion paper at this conference. For criticality validation, the approach is to utilize (1) available laboratory critical experiment (LCE) data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the French Haut Taux de Combustion (HTC) program to support validation of the principal actinides and (2) calculated sensitivities, nuclear data uncertainties, and the limited available fission product LCE data to predict and verify individual biases for relevant minor actinides and fission products. This paper (1) provides a detailed description of the approach and its technical bases, (2) describes the application of the approach for representative pressurized water reactor and boiling water reactor safety analysis models to demonstrate its usage and applicability, (3) provides reference bias results based on the prerelease SCALE 6.1 code package and ENDF/B-VII nuclear cross-section data, and (4) provides recommendations for application of the results and methods to other code and data packages.« less

Major safety provisions in nuclear-powered ships

DOE Office of Scientific and Technical Information (OSTI.GOV)

Khlopkin, N.S.; Belyaev, V.M.; Dubrovin, A.M.

1984-12-01

Considerable experience has been accumulated in the Soviet Union on the design, construction and operation of nuclear-powered civilian ships: the icebreakers Lenin, Leonid Brezhnev and Sibir. The nuclear steam plants (NSP) used on these as the main energy source have been found to be highly reliable and safe, and it is desirable to use them in the future not only in icebreakers but also in transport ships for use in ice fields. The Soviet program for building and developing nuclear-powered ships has involved careful attention to safety in ships containing NSP. The experience with the design and operation of nuclearmore » icebreakers in recent years has led to the revision of safety standards for the nuclear ships and correspondingly ship NSP and international guidelines have been developed. If one meets the requirements as set forth in these documents, one has a safe basis for future Soviet nuclear-powered ships. The primary safety provisions for NSP are presented in this paper.« less

Response surface method in geotechnical/structural analysis, phase 1

NASA Astrophysics Data System (ADS)

Wong, F. S.

1981-02-01

In the response surface approach, an approximating function is fit to a long running computer code based on a limited number of code calculations. The approximating function, called the response surface, is then used to replace the code in subsequent repetitive computations required in a statistical analysis. The procedure of the response surface development and feasibility of the method are shown using a sample problem in slop stability which is based on data from centrifuge experiments of model soil slopes and involves five random soil parameters. It is shown that a response surface can be constructed based on as few as four code calculations and that the response surface is computationally extremely efficient compared to the code calculation. Potential applications of this research include probabilistic analysis of dynamic, complex, nonlinear soil/structure systems such as slope stability, liquefaction, and nuclear reactor safety.

The regulatory framework for safe decommissioning of nuclear power plants in Korea

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sangmyeon Ahn; Jungjoon Lee; Chanwoo Jeong

We are having 23 units of nuclear power plants in operation and 5 units of nuclear power plants under construction in Korea as of September 2012. However, we don't have any experience on shutdown permanently and decommissioning of nuclear power plants. There are only two research reactors being decommissioned since 1997. It is realized that improvement of the regulatory framework for decommissioning of nuclear facilities has been emphasized constantly from the point of view of IAEA's safety standards. It is also known that IAEA will prepare the safety requirement on decommissioning of facilities; its title is the Safe Decommissioning ofmore » Facilities, General Safety Requirement Part 6. According to the result of IAEA's Integrated Regulatory Review Service (IRRS) mission to Korea in 2011, it was recommended that the regulatory framework should require decommissioning plans for nuclear installations to be constructed and operated and these plans should be updated periodically. In addition, after the Fukushima nuclear disaster in Japan in March of 2011, preparedness for early decommissioning caused by an unexpected severe accident became important issues and concerns. In this respect, it is acknowledged that the regulatory framework for decommissioning of nuclear facilities in Korea need to be improved. First of all, we focus on identifying the current status and relevant issues of regulatory framework for decommissioning of nuclear power plants compared to the IAEA's safety standards in order to achieve our goal. And then the plan is established for improvement of regulatory framework for decommissioning of nuclear power plants in Korea. It is expected that if the things will go forward as planned, the revised regulatory framework for decommissioning could enhance the safety regime on the decommissioning of nuclear power plants in Korea in light of international standards. (authors)« less

10 CFR 2.106 - Notice of issuance.

Code of Federal Regulations, 2010 CFR

2010-01-01

... previously published. (d) The Director of Nuclear Material Safety and Safeguards will also cause to be... Application-How Initiated § 2.106 Notice of issuance. (a) The Director, Office of New Reactors, Director, Office of Nuclear Reactor Regulation, or Director, Office of Nuclear Material Safety and Safeguards, as...

10 CFR 820.20 - Purpose and scope.

Code of Federal Regulations, 2013 CFR

2013-01-01

... OF ENERGY PROCEDURAL RULES FOR DOE NUCLEAR ACTIVITIES Enforcement Process § 820.20 Purpose and scope... violations of the DOE Nuclear Safety Requirements, for determining, whether a violation has occurred, for... of a violation of: (1) Any DOE Nuclear Safety Requirement set forth in the Code of Federal...

10 CFR 820.20 - Purpose and scope.

Code of Federal Regulations, 2012 CFR

2012-01-01

... OF ENERGY PROCEDURAL RULES FOR DOE NUCLEAR ACTIVITIES Enforcement Process § 820.20 Purpose and scope... violations of the DOE Nuclear Safety Requirements, for determining, whether a violation has occurred, for... of a violation of: (1) Any DOE Nuclear Safety Requirement set forth in the Code of Federal...

10 CFR 820.20 - Purpose and scope.

Code of Federal Regulations, 2011 CFR

2011-01-01

... OF ENERGY PROCEDURAL RULES FOR DOE NUCLEAR ACTIVITIES Enforcement Process § 820.20 Purpose and scope... violations of the DOE Nuclear Safety Requirements, for determining, whether a violation has occurred, for... of a violation of: (1) Any DOE Nuclear Safety Requirement set forth in the Code of Federal...

10 CFR 820.20 - Purpose and scope.

Code of Federal Regulations, 2014 CFR

2014-01-01

... OF ENERGY PROCEDURAL RULES FOR DOE NUCLEAR ACTIVITIES Enforcement Process § 820.20 Purpose and scope... violations of the DOE Nuclear Safety Requirements, for determining, whether a violation has occurred, for... of a violation of: (1) Any DOE Nuclear Safety Requirement set forth in the Code of Federal...

Safety Evaluation Report on Tennessee Valley Authority: Browns Ferry Nuclear Performance Plan: Browns Ferry Unit 2 restart

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1989-04-01

This safety evaluation report (SER) on the information submitted by the Tennessee Valley Authority (TVA) in its Nuclear Performance Plan, through Revision 2, for the Browns Ferry Nuclear Power Station and in supporting documents has been prepared by the US Nuclear Regulatory Commission staff. The plan addresses the plant-specific concerns requiring resolution before startup of Unit 2. The staff will inspect implementation of those programs. Where systems are common to Units 1 and 2 or to Units 2 and 3, the staff safety evaluations of those systems are included herein. 3 refs.

Generation III reactors safety requirements and the design solutions

NASA Astrophysics Data System (ADS)

Felten, P.

2009-03-01

Nuclear energy's public acceptance, and hence its development, depends on its safety. As a reactor designer, we will first briefly remind the basic safety principles of nuclear reactors' design. We will then show how the industry, and in particular Areva with its EPR, made design evolution in the wake of the Three Miles Island accident in 1979. In particular, for this new generation of reactors, severe accidents are taken into account beyond the standard design basis accidents. Today, Areva's EPR meets all so-called "generation III" safety requirements and was licensed by several nuclear safety authorities in the world. Many innovative solutions are integrated in the EPR, some of which will be introduced here.

Post-Cold War Science and Technology at Los Alamos

NASA Astrophysics Data System (ADS)

Browne, John C.

2002-04-01

Los Alamos National Laboratory serves the nation through the development and application of leading-edge science and technology in support of national security. Our mission supports national security by: ensuring the safety, security, and reliability of the U.S. nuclear stockpile; reducing the threat of weapons of mass destruction in support of counter terrorism and homeland defense; and solving national energy, environment, infrastructure, and health security problems. We require crosscutting fundamental and advanced science and technology research to accomplish our mission. The Stockpile Stewardship Program develops and applies, advanced experimental science, computational simulation, and technology to ensure the safety and reliability of U.S. nuclear weapons in the absence of nuclear testing. This effort in itself is a grand challenge. However, the terrorist attack of September 11, 2001, reminded us of the importance of robust and vibrant research and development capabilities to meet new and evolving threats to our national security. Today through rapid prototyping we are applying new, innovative, science and technology for homeland defense, to address the threats of nuclear, chemical, and biological weapons globally. Synergistically, with the capabilities that we require for our core mission, we contribute in many other areas of scientific endeavor. For example, our Laboratory has been part of the NASA effort on mapping water on the moon and NSF/DOE projects studying high-energy astrophysical phenomena, understanding fundamental scaling phenomena of life, exploring high-temperature superconductors, investigating quantum information systems, applying neutrons to condensed-matter and nuclear physics research, developing large-scale modeling and simulations to understand complex phenomena, and exploring nanoscience that bridges the atomic to macroscopic scales. In this presentation, I will highlight some of these post-cold war science and technology advances including our national security contributions, and discuss some of challenges for Los Alamos in the future.

Financial Management: Extending the Financial Statements Audit Requirement of the CFO Act to Additional Federal Agencies

DTIC Science & Technology

2002-05-14

Defense Nuclear Facilities Safety Board has balance-sheet-only audits every 3 to 5 years, most recently for fiscal year 1997. It did not prepare fiscal...associated with the agency’s operations were the most important factors to Have had financial statements audits Defense Nuclear Facilities Safety...audits, the International Trade Commission and the Defense Nuclear Facilities Safety Board, did not have financial statements audits for fiscal year

Ukraine: Current Issues and U.S. Policy

DTIC Science & Technology

2014-02-26

The United States also pledged to continue to cooperate with Ukraine on nuclear safety issues, including the cleanup of the Chernobyl nuclear...Congressional Research Service 13 A significant portion of U.S. aid to Ukraine in the ESF account is dedicated to improving the safety of the Chernobyl nuclear

75 FR 2894 - Withdrawal of Regulatory Guide 1.148

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-19

... Valve Assemblies in Systems Important to Safety in Nuclear Power Plants.'' FOR FURTHER INFORMATION... for Active Valve Assemblies in Systems Important to Safety in Nuclear Power Plants.'' RG 1.148 was... qualifying active mechanical equipment used in nuclear power plants. The NRC is withdrawing RG 1.148 because...

76 FR 44865 - Domestic Licensing of Source Material-Amendments/Integrated Safety Analysis

Federal Register 2010, 2011, 2012, 2013, 2014

2011-07-27

... NUCLEAR REGULATORY COMMISSION 10 CFR Part 40 RIN 3150-AI50 [NRC-2009-0079 and NRC-2011-0080] Domestic Licensing of Source Material--Amendments/Integrated Safety Analysis AGENCY: Nuclear Regulatory Commission. ACTION: Extension of public comment period and public meeting. SUMMARY: The U.S. Nuclear...

Government: Nuclear Safety in Doubt a Year after Accident.

ERIC Educational Resources Information Center

Ember, Lois R.

1980-01-01

A year after the accident at Three Mile Island (TMI), the signals transmitted to the public are still confused. Industry says that nuclear power is safe and that the aftermath of TMI ushers in a new era of safety. Antinuclear activists say TMI sounded nuclear power's death knell. (Author/RE)

10 CFR 35.50 - Training for Radiation Safety Officer.

Code of Federal Regulations, 2011 CFR

2011-01-01

... authorized nuclear pharmacist identified on the licensee's license and has experience with the radiation... Radiation Safety Officer, authorized medical physicist, authorized nuclear pharmacist, or authorized user...

10 CFR 35.50 - Training for Radiation Safety Officer.

Code of Federal Regulations, 2014 CFR

2014-01-01

... authorized nuclear pharmacist identified on the licensee's license and has experience with the radiation... Radiation Safety Officer, authorized medical physicist, authorized nuclear pharmacist, or authorized user...

10 CFR 35.50 - Training for Radiation Safety Officer.

Code of Federal Regulations, 2012 CFR

2012-01-01

... authorized nuclear pharmacist identified on the licensee's license and has experience with the radiation... Radiation Safety Officer, authorized medical physicist, authorized nuclear pharmacist, or authorized user...

10 CFR 35.50 - Training for Radiation Safety Officer.

Code of Federal Regulations, 2013 CFR

2013-01-01

... authorized nuclear pharmacist identified on the licensee's license and has experience with the radiation... Radiation Safety Officer, authorized medical physicist, authorized nuclear pharmacist, or authorized user...

History of nuclear technology development in Japan

NASA Astrophysics Data System (ADS)

Yamashita, Kiyonobu

2015-04-01

Nuclear technology development in Japan has been carried out based on the Atomic Energy Basic Act brought into effect in 1955. The nuclear technology development is limited to peaceful purposes and made in a principle to assure their safety. Now, the technologies for research reactors radiation application and nuclear power plants are delivered to developing countries. First of all, safety measures of nuclear power plants (NPPs) will be enhanced based on lesson learned from TEPCO Fukushima Daiichi NPS accident.

Nuclear Warheads: The Reliable Replacement Warhead Program and the Life Extension Program

DTIC Science & Technology

2006-12-13

Defense Nuclear Facilities Safety Board was created by Congress 1988 "as an independent oversight organization within the Executive Branch charged... nuclear facilities ." U.S. Defense Nuclear Facilities Safety Board. “Who We Are,” at [http://www.dnfsb.gov/about/index.html]. involving CHE and plutonium...approach, if successful, would “reduce or eliminate the need for ESD controls.”42 Kent Fortenberry, Technical Director of the Defense Nuclear Facilities

History of nuclear technology development in Japan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yamashita, Kiyonobu, E-mail: yamashita.kiyonobu@jaea.go.jp; General Advisor Nuclear HRD Centre, Japan Atomic Energy Agency, TOKAI-mura, NAKA-gun, IBARAKI-ken, 319-1195

2015-04-29

Nuclear technology development in Japan has been carried out based on the Atomic Energy Basic Act brought into effect in 1955. The nuclear technology development is limited to peaceful purposes and made in a principle to assure their safety. Now, the technologies for research reactors radiation application and nuclear power plants are delivered to developing countries. First of all, safety measures of nuclear power plants (NPPs) will be enhanced based on lesson learned from TEPCO Fukushima Daiichi NPS accident.

Evaluation of nuclear facility decommissioning projects. Summary report: North Carolina State University Research and Training Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Link, B.W.; Miller, R.L.

1983-08-01

This document summarizes information from the decommissioning of the NCSUR-3 (R-3), a 10 KWt university research and training reactor. The decommissioning data were placed in a computerized information retrieval/manipulation system which permits future utilization of this information in pre-decommissioning activities with other university reactors of similar design. The information is presented both in some detail in its computer output form and also as a manually assembled summarization which highlights the more significant aspects of the decommissioning project. Decommissioning data from a generic study, NUREG/CR 1756, Technology, Safety and Costs of Decommissioning Nuclear Research and Test Reactors, and the decommissioning ofmore » the Ames Laboratory Research Reactor (ALRR), a 5 MWt research reactor, is also included for comparison.« less

HANSF 1.3 Users Manual FAI/98-40-R2 Hanford Spent Nuclear Fuel (SNF) Safety Analysis Model [SEC 1 and 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

DUNCAN, D.R.

The HANSF analysis tool is an integrated model considering phenomena inside a multi-canister overpack (MCO) spent nuclear fuel container such as fuel oxidation, convective and radiative heat transfer, and the potential for fission product release. This manual reflects the HANSF version 1.3.2, a revised version of 1.3.1. HANSF 1.3.2 was written to correct minor errors and to allow modeling of condensate flow on the MCO inner surface. HANSF 1.3.2 is intended for use on personal computers such as IBM-compatible machines with Intel processors running under Lahey TI or digital Visual FORTRAN, Version 6.0, but this does not preclude operation inmore » other environments.« less

HSE's safety assessment principles for criticality safety.

PubMed

Simister, D N; Finnerty, M D; Warburton, S J; Thomas, E A; Macphail, M R

2008-06-01

The Health and Safety Executive (HSE) published its revised Safety Assessment Principles for Nuclear Facilities (SAPs) in December 2006. The SAPs are primarily intended for use by HSE's inspectors when judging the adequacy of safety cases for nuclear facilities. The revised SAPs relate to all aspects of safety in nuclear facilities including the technical discipline of criticality safety. The purpose of this paper is to set out for the benefit of a wider audience some of the thinking behind the final published words and to provide an insight into the development of UK regulatory guidance. The paper notes that it is HSE's intention that the Safety Assessment Principles should be viewed as a reflection of good practice in the context of interpreting primary legislation such as the requirements under site licence conditions for arrangements for producing an adequate safety case and for producing a suitable and sufficient risk assessment under the Ionising Radiations Regulations 1999 (SI1999/3232 www.opsi.gov.uk/si/si1999/uksi_19993232_en.pdf).

Strengthening LLNL Missions through Laboratory Directed Research and Development in High Performance Computing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Willis, D. K.

2016-12-01

High performance computing (HPC) has been a defining strength of Lawrence Livermore National Laboratory (LLNL) since its founding. Livermore scientists have designed and used some of the world’s most powerful computers to drive breakthroughs in nearly every mission area. Today, the Laboratory is recognized as a world leader in the application of HPC to complex science, technology, and engineering challenges. Most importantly, HPC has been integral to the National Nuclear Security Administration’s (NNSA’s) Stockpile Stewardship Program—designed to ensure the safety, security, and reliability of our nuclear deterrent without nuclear testing. A critical factor behind Lawrence Livermore’s preeminence in HPC ismore » the ongoing investments made by the Laboratory Directed Research and Development (LDRD) Program in cutting-edge concepts to enable efficient utilization of these powerful machines. Congress established the LDRD Program in 1991 to maintain the technical vitality of the Department of Energy (DOE) national laboratories. Since then, LDRD has been, and continues to be, an essential tool for exploring anticipated needs that lie beyond the planning horizon of our programs and for attracting the next generation of talented visionaries. Through LDRD, Livermore researchers can examine future challenges, propose and explore innovative solutions, and deliver creative approaches to support our missions. The present scientific and technical strengths of the Laboratory are, in large part, a product of past LDRD investments in HPC. Here, we provide seven examples of LDRD projects from the past decade that have played a critical role in building LLNL’s HPC, computer science, mathematics, and data science research capabilities, and describe how they have impacted LLNL’s mission.« less

Safety aspects of nuclear waste disposal in space

NASA Technical Reports Server (NTRS)

Rice, E. E.; Edgecombe, D. S.; Compton, P. R.

1981-01-01

Safety issues involved in the disposal of nuclear wastes in space as a complement to mined geologic repositories are examined as part of an assessment of the feasibility of nuclear waste disposal in space. General safety guidelines for space disposal developed in the areas of radiation exposure and shielding, containment, accident environments, criticality, post-accident recovery, monitoring systems and isolation are presented for a nuclear waste disposal in space mission employing conventional space technology such as the Space Shuttle. The current reference concept under consideration by NASA and DOE is then examined in detail, with attention given to the waste source and mix, the waste form, waste processing and payload fabrication, shipping casks and ground transport vehicles, launch site operations and facilities, Shuttle-derived launch vehicle, orbit transfer vehicle, orbital operations and space destination, and the system safety aspects of the concept are discussed for each component. It is pointed out that future work remains in the development of an improved basis for the safety guidelines and the determination of the possible benefits and costs of the space disposal option for nuclear wastes.

75 FR 53985 - Southern Nuclear Operating Company Establishment of Atomic Safety And Licensing Board

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-02

...] Southern Nuclear Operating Company Establishment of Atomic Safety And Licensing Board Pursuant to..., notice is hereby given that an Atomic Safety and Licensing Board (Board) is being established to preside...). The Board is comprised of the following administrative judges: G. Paul Bollwerk, III, Chairman, Atomic...

Estimation of concentration of radionuclides in skeletal muscle from blood, based on the data from abandoned animals in Fukushima.

PubMed

Fukuda, Tomokazu

2018-06-01

The damage caused by the earthquake on 11 March, 2011 resulted in a serious nuclear accident in Japan. Due to the damage to the Fukushima Daiichi Nuclear Power Plant (FNPP), large amounts of radioactive substances were released into the environment. In particular, one of the largest safety concerns is radioactive cesium ( 134 Cs and 137 Cs). Due to the FNPP nuclear accident, a 20 km area was restricted from human activity, and various types of domestic animals were left in the zone. We collected the organs and tissues from sacrificed animals to obtain scientific data to evaluate the internal deposition of radioactive compounds. At first, we found there is a strong correlation between blood 137 Cs and organ 137 Cs with data from 44 cattle, indicating that skeletal muscle is the target organ of deposition of radioactive cesium. Second, we analyzed the relationship between blood 137 Cs and muscle 137 Cs within relatively lower radioactive concentration, suggesting that estimation of concentration of 137 Cs is possible from blood concentration of 137 Cs. Finally, we developed computer software to estimate the muscle 137 Cs concentration from blood samples. Our study contributes to the food safety of livestock products. © 2018 The Authors. Animal Science Journal published by John Wiley & Sons Australia, Ltd on behalf of Japanese Society of Animal Science.

Gamma-Ray imaging for nuclear security and safety: Towards 3-D gamma-ray vision

NASA Astrophysics Data System (ADS)

Vetter, Kai; Barnowksi, Ross; Haefner, Andrew; Joshi, Tenzing H. Y.; Pavlovsky, Ryan; Quiter, Brian J.

2018-01-01

The development of portable gamma-ray imaging instruments in combination with the recent advances in sensor and related computer vision technologies enable unprecedented capabilities in the detection, localization, and mapping of radiological and nuclear materials in complex environments relevant for nuclear security and safety. Though multi-modal imaging has been established in medicine and biomedical imaging for some time, the potential of multi-modal data fusion for radiological localization and mapping problems in complex indoor and outdoor environments remains to be explored in detail. In contrast to the well-defined settings in medical or biological imaging associated with small field-of-view and well-constrained extension of the radiation field, in many radiological search and mapping scenarios, the radiation fields are not constrained and objects and sources are not necessarily known prior to the measurement. The ability to fuse radiological with contextual or scene data in three dimensions, in analog to radiological and functional imaging with anatomical fusion in medicine, provides new capabilities enhancing image clarity, context, quantitative estimates, and visualization of the data products. We have developed new means to register and fuse gamma-ray imaging with contextual data from portable or moving platforms. These developments enhance detection and mapping capabilities as well as provide unprecedented visualization of complex radiation fields, moving us one step closer to the realization of gamma-ray vision in three dimensions.

Practical Applications of Cosmic Ray Science: Spacecraft, Aircraft, Ground-Based Computation and Control Systems, Exploration, and Human Health and Safety

NASA Technical Reports Server (NTRS)

Koontz, Steve

2015-01-01

In this presentation a review of galactic cosmic ray (GCR) effects on microelectronic systems and human health and safety is given. The methods used to evaluate and mitigate unwanted cosmic ray effects in ground-based, atmospheric flight, and space flight environments are also reviewed. However not all GCR effects are undesirable. We will also briefly review how observation and analysis of GCR interactions with planetary atmospheres and surfaces and reveal important compositional and geophysical data on earth and elsewhere. About 1000 GCR particles enter every square meter of Earth’s upper atmosphere every second, roughly the same number striking every square meter of the International Space Station (ISS) and every other low- Earth orbit spacecraft. GCR particles are high energy ionized atomic nuclei (90% protons, 9% alpha particles, 1% heavier nuclei) traveling very close to the speed of light. The GCR particle flux is even higher in interplanetary space because the geomagnetic field provides some limited magnetic shielding. Collisions of GCR particles with atomic nuclei in planetary atmospheres and/or regolith as well as spacecraft materials produce nuclear reactions and energetic/highly penetrating secondary particle showers. Three twentieth century technology developments have driven an ongoing evolution of basic cosmic ray science into a set of practical engineering tools needed to design, test, and verify the safety and reliability of modern complex technological systems and assess effects on human health and safety effects. The key technology developments are: 1) high altitude commercial and military aircraft; 2) manned and unmanned spacecraft; and 3) increasingly complex and sensitive solid state micro-electronics systems. Space and geophysical exploration needs drove the development of the instruments and analytical tools needed to recover compositional and structural data from GCR induced nuclear reactions and secondary particle showers. Finally, the possible role of GCR secondary particle showers in addressing an important homeland security problem, finding nuclear contraband and weapons, will be briefly reviewed.

Tank waste remediation system nuclear criticality safety program management review

DOE Office of Scientific and Technical Information (OSTI.GOV)

BRADY RAAP, M.C.

1999-06-24

This document provides the results of an internal management review of the Tank Waste Remediation System (TWRS) criticality safety program, performed in advance of the DOE/RL assessment for closure of the TWRS Nuclear Criticality Safety Issue, March 1994. Resolution of the safety issue was identified as Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) Milestone M-40-12, due September 1999.

New reactor technology: safety improvements in nuclear power systems.

PubMed

Corradini, M L

2007-11-01

Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems.

Perspectives of The Interagency Nuclear Safety Review Panel (INSRP) on future nuclear powered space missions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gray, L.B.; Pyatt, D.W.; Sholtis, J.A.

1993-01-10

The Interagency Nuclear Safety Review Panel (INSRP) has provided reviews of all nuclear powered spacecraft launched by the United States. The two most recent launches were Ulysses in 1990 and Galileo in 1989. One reactor was launched in 1965 (SNAP-10A). All other U.S. space missions have utilized radioisotopic thermoelectric generators (RTGs). There are several missions in the next few years that are to be nuclear powered, including one that would utilize the Topaz II reactor purchased from Russia. INSRP must realign itself to perform parallel safety assessments of a reactor powered space mission, which has not been done in aboutmore » thirty years, and RTG powered missions.« less

Safety evaluation report on Tennessee Valley Authority: Browns Ferry nuclear performance plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1989-10-01

This safety evaluation report (SER) on the information submitted by the Tennessee Valley Authority (TVA) in its Nuclear Performance Plan, through Revision 2, for the Browns Ferry Nuclear Plant and in supporting documents has been prepared by the US Nuclear Regulatory commission staff. The Browns Ferry Nuclear Plant consists of three boiling-water reactors at a site in Limestone County, Alabama. The plan addresses the plant-specific concerns requiring resolution before the startup of Unit 2. The staff will inspect implementation of those TVA programs that address these concerns. Where systems are common to Units 1 and 2 or to Units 2more » and 3, the staff safety evaluations of those systems are included herein. 85 refs.« less

Management of Naturally Occurring Radioactive Materials (NORM) in Canada

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baweja, Anar S.; Tracy, Bliss L.

2008-08-07

In Canada, nuclear and radiological regulatory responsibilities are shared between the provinces/territories and the federal government. The Canadian Nuclear Safety Commission (CNSC) regulates nuclear fuel cycle materials and man-made radionuclides under the Nuclear Safety and Control Act (2000). The provinces and territories regulate NORM arising from industrial activities, not involving the nuclear fuel cycle materials. Present guideline--Canadian Guidelines for the Management of Naturally Occurring Radioactive Materials (NORM)--was published in 2000 in order to bring uniformity to the management of NORM-related procedures to provide adequate radiation protection for workers and the general public. The basic premise of these guidelines is thatmore » the NORM-related activities should not be posing any greater hazard than those activities regulated under the Nuclear Safety and Control Act; these concepts are described in this paper.« less

The COPERNIC3 project: how AREVA is successfully developing an advanced global fuel rod performance code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garnier, Ch.; Mailhe, P.; Sontheimer, F.

2007-07-01

Fuel performance is a key factor for minimizing operating costs in nuclear plants. One of the important aspects of fuel performance is fuel rod design, based upon reliable tools able to verify the safety of current fuel solutions, prevent potential issues in new core managements and guide the invention of tomorrow's fuels. AREVA is developing its future global fuel rod code COPERNIC3, which is able to calculate the thermal-mechanical behavior of advanced fuel rods in nuclear plants. Some of the best practices to achieve this goal are described, by reviewing the three pillars of a fuel rod code: the database,more » the modelling and the computer and numerical aspects. At first, the COPERNIC3 database content is described, accompanied by the tools developed to effectively exploit the data. Then is given an overview of the main modelling aspects, by emphasizing the thermal, fission gas release and mechanical sub-models. In the last part, numerical solutions are detailed in order to increase the computational performance of the code, with a presentation of software configuration management solutions. (authors)« less

Operations Praetorian, Phalanx, Fusileer, and Grenadier, Events HURON LANDING/DIAMOND ACE, MINI JADE, TOMME/MIDNIGHT ZEPHYR, MIDAS MYTH/MILAGRO and MISTY RAIN

DTIC Science & Technology

1993-07-01

General requirements and standards governing safety were based on the FCTC "Safety and Health Compliance Guide for Underground and Nuclear Effects Tests...Defense (DOD)-sponsored underground nuclear tests were conducted from 23 September 1982 through 6 April 1985 to study weapons effects . All six were...weapons- related or effects purposes, and 33 were safety experiments. An additional 24 nuclear experiments were conducted from December 1954 to February

DOE Office of Scientific and Technical Information (OSTI.GOV)

Buchanan, J. R.

The risks of nuclear power and radiation are described to place them in perspective with other potential hazards faced by the public on a day-to-day basis in our complex industrial society. Twenty articles on this general topic that have appeared in Nuclear Safety are reprinted, since they collectively form a valuable reference source. Topics covered include the effects of radiation, riskbenefit concepts, radiation risks relative to other risks, nuclear plant risks relative to fossil plant risks, licensing requirements, nuclear insurance, nuclear industry safety record, and public attitudes. (auth)

Japan-U.S. Relations: Issues for Congress

DTIC Science & Technology

2013-02-15

million Japanese were displaced. Damage to several reactors at the Fukushima Dai-ichi nuclear power plant complex led the government to declare a state of...the March 11, 2011, natural disasters and meltdowns at the Fukushima Daiichi nuclear power plant. Public trust in the safety of nuclear power...aircraft’s safety record may be connected to widespread distrust of the government stemming from the nuclear disaster at the Fukushima Daichi

Challenging Minimum Deterrence: Articulating the Contemporary Relevance of Nuclear Weapons

DTIC Science & Technology

2016-07-13

Command and Control: Nuclear Weapons, the Damascus Accident , and the Illusion of Safety (New York: Penguin Press, 2013), 484. 36. Keir A. Lieber and...Remembrance of Things Past,” 78. 53. Scott D. Sagan, The Limits of Safety: Organizations, Accidents , and Nuclear Weapons (Princeton, NJ: Princeton...16 | Air & Space Power Journal Challenging Minimum Deterrence Articulating the Contemporary Relevance of Nuclear Weapons Maj Joshua D. Wiitala, USAF

Margin of Safety Definition and Examples Used in Safety Basis Documents and the USQ Process

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beaulieu, R. A.

The Nuclear Safety Management final rule, 10 CFR 830, provides an undefined term, margin of safety (MOS). Safe harbors listed in 10 CFR 830, Table 2, such as DOE-STD-3009 use but do not define the term. This lack of definition has created the need for the definition. This paper provides a definition of MOS and documents examples of MOS as applied in a U.S. Department of Energy (DOE) approved safety basis for an existing nuclear facility. If we understand what MOS looks like regarding Technical Safety Requirements (TSR) parameters, then it helps us compare against other parameters that do notmore » involve a MOS. This paper also documents parameters that are not MOS. These criteria could be used to determine if an MOS exists in safety basis documents. This paper helps DOE, including the National Nuclear Security Administration (NNSA) and its contractors responsible for the safety basis improve safety basis documents and the unreviewed safety question (USQ) process with respect to MOS.« less

Nuclear safety considerations in the conceptual design of a fast reactor for space electric power and propulsion

NASA Technical Reports Server (NTRS)

Hsieh, T.-M.; Koenig, D. R.

1977-01-01

Some nuclear safety aspects of a 3.2 mWt heat pipe cooled fast reactor with out-of-core thermionic converters are discussed. Safety related characteristics of the design including a thin layer of B4C surrounding the core, the use of heat pipes and BeO reflector assembly, the elimination of fuel element bowing, etc., are highlighted. Potential supercriticality hazards and countermeasures are considered. Impacts of some safety guidelines of space transportation system are also briefly discussed, since the currently developing space shuttle would be used as the primary launch vehicle for the nuclear electric propulsion spacecraft.

Annual report to Congress: Department of Energy activities relating to the Defense Nuclear Facilities Safety Board, Calendar Year 1999

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

2000-02-01

This is the tenth Annual Report to the Congress describing Department of Energy activities in response to formal recommendations and other interactions with the Defense Nuclear Facilities Safety Board (Board). The Board, an independent executive-branch agency established in 1988, provides advice and recommendations to the Secretary of Energy regarding public health and safety issues at the Department's defense nuclear facilities. The Board also reviews and evaluates the content and implementation of health and safety standards, as well as other requirements, relating to the design, construction, operation, and decommissioning of the Department's defense nuclear facilities. During 1999, Departmental activities resulted inmore » the closure of nine Board recommendations. In addition, the Department has completed all implementation plan milestones associated with three Board recommendations. One new Board recommendation was received and accepted by the Department in 1999, and a new implementation plan is being developed to address this recommendation. The Department has also made significant progress with a number of broad-based initiatives to improve safety. These include expanded implementation of integrated safety management at field sites, opening of a repository for long-term storage of transuranic wastes, and continued progress on stabilizing excess nuclear materials to achieve significant risk reduction.« less

10 CFR Appendix A to Subpart B of... - General Statement of Safety Basis Policy

Code of Federal Regulations, 2011 CFR

2011-01-01

... for the design and construction of a new DOE nuclear facility or a major modification to an existing... acceptable nuclear safety design criteria for use in preparing a preliminary documented safety analysis. As a... mitigate hazards to workers, the public, or the environment. They include (1) physical, design, structural...

76 FR 37798 - DOE Response to Recommendation 2010-2 of the Defense Nuclear Facilities Safety Board, Pulse Jet...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-28

... Safety Board, Pulse Jet Mixing at the Waste Treatment and Immobilization Plant AGENCY: Department of... their Recommendation 2010-2, concerning Pulse Jet Mixing at the Waste Treatment and Immobilization Plant... Nuclear Facilities Safety Board (Board) Recommendation 2010-2, Pulse Jet Mixing (PJM) at the Waste...

10 CFR 830.201 - Performance of work.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Performance of work. 830.201 Section 830.201 Energy DEPARTMENT OF ENERGY NUCLEAR SAFETY MANAGEMENT Safety Basis Requirements § 830.201 Performance of work. A contractor must perform work in accordance with the safety basis for a hazard category 1, 2, or 3 DOE nuclear...

Root causes and impacts of severe accidents at large nuclear power plants.

PubMed

Högberg, Lars

2013-04-01

The root causes and impacts of three severe accidents at large civilian nuclear power plants are reviewed: the Three Mile Island accident in 1979, the Chernobyl accident in 1986, and the Fukushima Daiichi accident in 2011. Impacts include health effects, evacuation of contaminated areas as well as cost estimates and impacts on energy policies and nuclear safety work in various countries. It is concluded that essential objectives for reactor safety work must be: (1) to prevent accidents from developing into severe core damage, even if they are initiated by very unlikely natural or man-made events, and, recognizing that accidents with severe core damage may nevertheless occur; (2) to prevent large-scale and long-lived ground contamination by limiting releases of radioactive nuclides such as cesium to less than about 100 TBq. To achieve these objectives the importance of maintaining high global standards of safety management and safety culture cannot be emphasized enough. All three severe accidents discussed in this paper had their root causes in system deficiencies indicative of poor safety management and poor safety culture in both the nuclear industry and government authorities.

User Manual for Whisper-1.1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Forrest B.; Rising, Michael Evan; Alwin, Jennifer Louise

2017-01-26

Whisper is a statistical analysis package developed in 2014 to support nuclear criticality safety (NCS) validation [1-3]. It uses the sensitivity profile data for an application as computed by MCNP6 [4-6] along with covariance files [7,8] for the nuclear data to determine a baseline upper-subcritical-limit (USL) for the application. Whisper version 1.0 was first developed and used at LANL in 2014 [3]. During 2015- 2016, Whisper was updated to version 1.1 and is to be included with the upcoming release of MCNP6.2. This document describes the user input and options for running whisper-1.1, including 2 perl utility scripts that simplifymore » ordinary NCS work, whisper_mcnp.pl and whisper_usl.pl. For many detailed references on the theory, applications, nuclear data & covariances, SQA, verification-validation, adjointbased methods for sensitivity-uncertainty analysis, and more – see the Whisper – NCS Validation section of the MCNP Reference Collection at mcnp.lanl.gov. There are currently over 50 Whisper reference documents available.« less

MELCOR computer code manuals

DOE Office of Scientific and Technical Information (OSTI.GOV)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

Safety survey report EBR-II safety survey, ANL-west health protection, industrial safety and fire protection survey

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dunbar, K.A.

1972-01-10

A safety survey covering the disciplines of Reactor Safety, Nuclear Criticality Safety, Health Protection and Industrial Safety and Fire Protection was conducted at the ANL-West EBR-II FEF Complex during the period January 10-18, 1972. In addition, the entire ANL-West site was surveyed for Health Protection and Industrial Safety and Fire Protection. The survey was conducted by members of the AEC Chicago Operations Office, a member of RDT-HQ and a member of the RDT-ID site office. Eighteen recommendations resulted from the survey, eleven in the area of Industrial Safety and Fire Protection, five in the area of Reactor Safety and twomore » in the area of Nuclear Criticality Safety.« less

Scientist vs. Citizen: The Shoreham Controversy

ERIC Educational Resources Information Center

Carl, Ann; Sailor, Vance L.

1972-01-01

Two views on the issue of nuclear reactor safety with reference to the license application of a nuclear power plant on Long Island. Each side presents scientific arguments concerning the safety standards proposed. (AL)

Lessons Learned from the Fukushima Nuclear Accident due to Tohoku Region Pacific Coast Earthquake

NASA Astrophysics Data System (ADS)

Miki, M.; Wada, M.; Takeuchi, N.

2012-01-01

On March 11 2011, Great Eastern Japan Earthquake hit Japan and caused the devastating damage. Fukushima Nuclear Power Station (NPS) also suffered damages and provided the environmental effect with radioactive products. The situation has been settled to some extent about two months after the accidents, and currently, the cooling of reactor is continuing towards settling the situation. Japanese NPSs are designed based on safety requirements and have multiple-folds of hazard controls. However, according to publicly available information, due to the lager-than-anticipated Tsunami, all the power supply were lost, which resulted in loss of hazard controls. Also, although nuclear power plants are equipped with system/procedure in case of loss of all controls, recovery was not made as planned in Fukushima NPSs because assumptions for hazard controls became impractical or found insufficient. In consequence, a state of emergency was declared. Through this accident, many lessons learned have been obtained from the several perspectives. There are many commonality between nuclear safety and space safety. Both industries perform thorough hazard assessments because hazards in both industries can result in loss of life. Therefore, space industry must learn from this accident and reconsider more robust space safety. This paper will introduce lessons learned from Fukushima nuclear accident described in the "Report of the Japanese Government to the IAEA Ministerial Conference on Nuclear Safety" [1], and discuss the considerations to establish more robust safety in the space systems. Detailed information of Fukushima Dai-ichi NPS are referred to this report.

The Regulatory Challenges of Decommissioning Nuclear Power Plants in Korea - 13101

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Jungjoon; Ahn, Sangmyeon; Choi, Kyungwoo

As of 2012, 23 units of nuclear power plants are in operation, but there is no experience of permanent shutdown and decommissioning of nuclear power plant in Korea. It is realized that, since late 1990's, improvement of the regulatory framework for decommissioning has been emphasized constantly from the point of view of International Atomic Energy Agency (IAEA)'s safety standards. And it is known that now IAEA prepare the safety requirement on decommissioning of facilities, its title is the Safe Decommissioning of Facilities, General Safety Requirement Part 6. According to the result of IAEA's Integrated Regulatory Review Service (IRRS) mission tomore » Korea in 2011, it was recommended that the regulatory framework for decommissioning should require decommissioning plans for nuclear installations to be constructed and operated and these plans should be updated periodically. In addition, after the Fukushima nuclear disaster in Japan in March of 2011, preparedness for early decommissioning caused by an unexpected severe accident became also important issues and concerns. In this respect, it is acknowledged that the regulatory framework for decommissioning of nuclear facilities in Korea need to be improved. First of all, we identify the current status and relevant issues of regulatory framework for decommissioning of nuclear power plants compared to the IAEA's safety standards in order to achieve our goal. And then the plan is to be established for improvement of regulatory framework for decommissioning of nuclear power plants in Korea. After dealing with it, it is expected that the revised regulatory framework for decommissioning could enhance the safety regime on the decommissioning of nuclear power plants in Korea in light of international standards. (authors)« less

Modeling and simulation challenges pursued by the Consortium for Advanced Simulation of Light Water Reactors (CASL)

NASA Astrophysics Data System (ADS)

Turinsky, Paul J.; Kothe, Douglas B.

2016-05-01

The Consortium for the Advanced Simulation of Light Water Reactors (CASL), the first Energy Innovation Hub of the Department of Energy, was established in 2010 with the goal of providing modeling and simulation (M&S) capabilities that support and accelerate the improvement of nuclear energy's economic competitiveness and the reduction of spent nuclear fuel volume per unit energy, and all while assuring nuclear safety. To accomplish this requires advances in M&S capabilities in radiation transport, thermal-hydraulics, fuel performance and corrosion chemistry. To focus CASL's R&D, industry challenge problems have been defined, which equate with long standing issues of the nuclear power industry that M&S can assist in addressing. To date CASL has developed a multi-physics ;core simulator; based upon pin-resolved radiation transport and subchannel (within fuel assembly) thermal-hydraulics, capitalizing on the capabilities of high performance computing. CASL's fuel performance M&S capability can also be optionally integrated into the core simulator, yielding a coupled multi-physics capability with untapped predictive potential. Material models have been developed to enhance predictive capabilities of fuel clad creep and growth, along with deeper understanding of zirconium alloy clad oxidation and hydrogen pickup. Understanding of corrosion chemistry (e.g., CRUD formation) has evolved at all scales: micro, meso and macro. CFD R&D has focused on improvement in closure models for subcooled boiling and bubbly flow, and the formulation of robust numerical solution algorithms. For multiphysics integration, several iterative acceleration methods have been assessed, illuminating areas where further research is needed. Finally, uncertainty quantification and data assimilation techniques, based upon sampling approaches, have been made more feasible for practicing nuclear engineers via R&D on dimensional reduction and biased sampling. Industry adoption of CASL's evolving M&S capabilities, which is in progress, will assist in addressing long-standing and future operational and safety challenges of the nuclear industry.

DOE’s Management and Oversight of the Nuclear Weapons Complex

DTIC Science & Technology

1990-03-22

and Economic Development Division Before the Department of Energy Defense Nuclear Facilities Panel Committee on Armed Services House of Representatives...and newly created DOE offices. The Defense Nuclear Facilities Safety Board, whose board members were appointed this past year, was created to provide 6...mandated Defense Nuclear Facilities Safety Board. Continuing dialogue between DOE and the Board can also serve to enhance DOE’s ability to respond more

Decay heat uncertainty for BWR used fuel due to modeling and nuclear data uncertainties

DOE PAGES

Ilas, Germina; Liljenfeldt, Henrik

2017-05-19

Characterization of the energy released from radionuclide decay in nuclear fuel discharged from reactors is essential for the design, safety, and licensing analyses of used nuclear fuel storage, transportation, and repository systems. There are a limited number of decay heat measurements available for commercial used fuel applications. Because decay heat measurements can be expensive or impractical for covering the multitude of existing fuel designs, operating conditions, and specific application purposes, decay heat estimation relies heavily on computer code prediction. Uncertainty evaluation for calculated decay heat is an important aspect when assessing code prediction and a key factor supporting decision makingmore » for used fuel applications. While previous studies have largely focused on uncertainties in code predictions due to nuclear data uncertainties, this study discusses uncertainties in calculated decay heat due to uncertainties in assembly modeling parameters as well as in nuclear data. Capabilities in the SCALE nuclear analysis code system were used to quantify the effect on calculated decay heat of uncertainties in nuclear data and selected manufacturing and operation parameters for a typical boiling water reactor (BWR) fuel assembly. Furthermore, the BWR fuel assembly used as the reference case for this study was selected from a set of assemblies for which high-quality decay heat measurements are available, to assess the significance of the results through comparison with calculated and measured decay heat data.« less

Decay heat uncertainty for BWR used fuel due to modeling and nuclear data uncertainties

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ilas, Germina; Liljenfeldt, Henrik

Characterization of the energy released from radionuclide decay in nuclear fuel discharged from reactors is essential for the design, safety, and licensing analyses of used nuclear fuel storage, transportation, and repository systems. There are a limited number of decay heat measurements available for commercial used fuel applications. Because decay heat measurements can be expensive or impractical for covering the multitude of existing fuel designs, operating conditions, and specific application purposes, decay heat estimation relies heavily on computer code prediction. Uncertainty evaluation for calculated decay heat is an important aspect when assessing code prediction and a key factor supporting decision makingmore » for used fuel applications. While previous studies have largely focused on uncertainties in code predictions due to nuclear data uncertainties, this study discusses uncertainties in calculated decay heat due to uncertainties in assembly modeling parameters as well as in nuclear data. Capabilities in the SCALE nuclear analysis code system were used to quantify the effect on calculated decay heat of uncertainties in nuclear data and selected manufacturing and operation parameters for a typical boiling water reactor (BWR) fuel assembly. Furthermore, the BWR fuel assembly used as the reference case for this study was selected from a set of assemblies for which high-quality decay heat measurements are available, to assess the significance of the results through comparison with calculated and measured decay heat data.« less

Educating Next Generation Nuclear Criticality Safety Engineers at the Idaho National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. D. Bess; J. B. Briggs; A. S. Garcia

2011-09-01

One of the challenges in educating our next generation of nuclear safety engineers is the limitation of opportunities to receive significant experience or hands-on training prior to graduation. Such training is generally restricted to on-the-job-training before this new engineering workforce can adequately provide assessment of nuclear systems and establish safety guidelines. Participation in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) can provide students and young professionals the opportunity to gain experience and enhance critical engineering skills. The ICSBEP and IRPhEP publish annual handbooks that contain evaluations of experiments along withmore » summarized experimental data and peer-reviewed benchmark specifications to support the validation of neutronics codes, nuclear cross-section data, and the validation of reactor designs. Participation in the benchmark process not only benefits those who use these Handbooks within the international community, but provides the individual with opportunities for professional development, networking with an international community of experts, and valuable experience to be used in future employment. Traditionally students have participated in benchmarking activities via internships at national laboratories, universities, or companies involved with the ICSBEP and IRPhEP programs. Additional programs have been developed to facilitate the nuclear education of students while participating in the benchmark projects. These programs include coordination with the Center for Space Nuclear Research (CSNR) Next Degree Program, the Collaboration with the Department of Energy Idaho Operations Office to train nuclear and criticality safety engineers, and student evaluations as the basis for their Master's thesis in nuclear engineering.« less

Security culture for nuclear facilities

NASA Astrophysics Data System (ADS)

Gupta, Deeksha; Bajramovic, Edita

2017-01-01

Natural radioactive elements are part of our environment and radioactivity is a natural phenomenon. There are numerous beneficial applications of radioactive elements (radioisotopes) and radiation, starting from power generation to usages in medical, industrial and agriculture applications. But the risk of radiation exposure is always attached to operational workers, the public and the environment. Hence, this risk has to be assessed and controlled. The main goal of safety and security measures is to protect human life, health, and the environment. Currently, nuclear security considerations became essential along with nuclear safety as nuclear facilities are facing rapidly increase in cybersecurity risks. Therefore, prevention and adequate protection of nuclear facilities from cyberattacks is the major task. Historically, nuclear safety is well defined by IAEA guidelines while nuclear security is just gradually being addressed by some new guidance, especially the IAEA Nuclear Security Series (NSS), IEC 62645 and some national regulations. At the overall level, IAEA NSS 7 describes nuclear security as deterrence and detection of, and response to, theft, sabotage, unauthorized access, illegal transfer or other malicious acts involving nuclear, other radioactive substances and their associated facilities. Nuclear security should be included throughout nuclear facilities. Proper implementation of a nuclear security culture leads to staff vigilance and a high level of security posture. Nuclear security also depends on policy makers, regulators, managers, individual employees and members of public. Therefore, proper education and security awareness are essential in keeping nuclear facilities safe and secure.

Strengthening safety compliance in nuclear power operations: a role-based approach.

PubMed

Martínez-Córcoles, Mario; Gracia, Francisco J; Tomás, Inés; Peiró, José M

2014-07-01

Safety compliance is of paramount importance in guaranteeing the safe running of nuclear power plants. However, it depends mostly on procedures that do not always involve the safest outcomes. This article introduces an empirical model based on the organizational role theory to analyze the influence of legitimate sources of expectations (procedures formalization and leadership) on workers' compliance behaviors. The sample was composed of 495 employees from two Spanish nuclear power plants. Structural equation analysis showed that, in spite of some problematic effects of proceduralization (such as role conflict and role ambiguity), procedure formalization along with an empowering leadership style lead to safety compliance by clarifying a worker's role in safety. Implications of these findings for safety research are outlined, as well as their practical implications. © 2014 Society for Risk Analysis.

Contributions to nuclear safety and radiation technologies in Ukraine by the Science and Technology Center in Ukraine (STCU)

NASA Astrophysics Data System (ADS)

Taranenko, L.; Janouch, F.; Owsiacki, L.

2001-06-01

This paper presents Science and Technology Center in Ukraine (STCU) activities devoted to furthering nuclear and radiation safety, which is a prioritized STCU area. The STCU, an intergovernmental organization with the principle objective of non-proliferation, administers financial support from the USA, Canada, and the EU to Ukrainian projects in various scientific and technological areas; coordinates projects; and promotes the integration of Ukrainian scientists into the international scientific community, including involving western collaborators. The paper focuses on STCU's largest project to date "Program Supporting Y2K Readiness at Ukrainian NPPs" initiated in April 1999 and designed to address possible Y2K readiness problems at 14 Ukrainian nuclear reactors. Other presented projects demonstrate a wide diversity of supported directions in the fields of nuclear and radiation safety, including reactor material improvement ("Improved Zirconium-Based Elements for Nuclear Reactors"), information technologies for nuclear industries ("Ukrainian Nuclear Data Bank in Slavutich"), and radiation health science ("Diagnostics and Treatment of Radiation-Induced Injuries of Human Biopolymers").

RELAP-7 Code Assessment Plan and Requirement Traceability Matrix

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yoo, Junsoo; Choi, Yong-joon; Smith, Curtis L.

2016-10-01

The RELAP-7, a safety analysis code for nuclear reactor system, is under development at Idaho National Laboratory (INL). Overall, the code development is directed towards leveraging the advancements in computer science technology, numerical solution methods and physical models over the last decades. Recently, INL has also been putting an effort to establish the code assessment plan, which aims to ensure an improved final product quality through the RELAP-7 development process. The ultimate goal of this plan is to propose a suitable way to systematically assess the wide range of software requirements for RELAP-7, including the software design, user interface, andmore » technical requirements, etc. To this end, we first survey the literature (i.e., international/domestic reports, research articles) addressing the desirable features generally required for advanced nuclear system safety analysis codes. In addition, the V&V (verification and validation) efforts as well as the legacy issues of several recently-developed codes (e.g., RELAP5-3D, TRACE V5.0) are investigated. Lastly, this paper outlines the Requirement Traceability Matrix (RTM) for RELAP-7 which can be used to systematically evaluate and identify the code development process and its present capability.« less

Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

NASA Astrophysics Data System (ADS)

Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

2018-02-01

Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

75 FR 50009 - Babcock & Wilcox Nuclear Operations Group, Inc.; Establishment of Atomic Safety and Licensing Board

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-16

... Nuclear Operations Group, Inc.; Establishment of Atomic Safety and Licensing Board Pursuant to delegation... 2.106, 2.300, 2.313(a), and 2.318, notice is hereby given that an Atomic Safety and Licensing Board..., 2010. The Board is comprised of the following administrative judges: Alan S. Rosenthal, Chair, Atomic...

78 FR 68102 - Atomic Safety and Licensing Board; In the Matter of Nuclear Innovation North America LLC (South...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-13

...] Atomic Safety and Licensing Board; In the Matter of Nuclear Innovation North America LLC (South Texas... Administrative Judges: Michael M. Gibson, Chairman, Dr. Gary S. Arnold, Dr. Randall J. Charbeneau. The Atomic... follows: Mail: Administrative Judge Michael M. Gibson, Atomic Safety and Licensing Board Panel, Mail Stop...

76 FR 13397 - DOE Response to Recommendation 2010-2 of the Defense Nuclear Facilities Safety Board, Pulse Jet...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-11

... Safety Board, Pulse Jet Mixing at the Waste Treatment and Immobilization Plant AGENCY: Department of..., concerning Pulse Jet Mixing at the Waste Treatment and Immobilization Plant was published in the Federal... Defense Nuclear Facilities Safety Board (Board) Recommendation 2010-2, Pulse Jet Mixing at the Waste...

DOE Office of Scientific and Technical Information (OSTI.GOV)

MITCHELL,GERRY W.; LONGLEY,SUSAN W.; PHILBIN,JEFFREY S.

This Safety Analysis Report (SAR) is prepared in compliance with the requirements of DOE Order 5480.23, Nuclear Safety Analysis Reports, and has been written to the format and content guide of DOE-STD-3009-94 Preparation Guide for U. S. Department of Energy Nonreactor Nuclear Safety Analysis Reports. The Hot Cell Facility is a Hazard Category 2 nonreactor nuclear facility, and is operated by Sandia National Laboratories for the Department of Energy. This SAR provides a description of the HCF and its operations, an assessment of the hazards and potential accidents which may occur in the facility. The potential consequences and likelihood ofmore » these accidents are analyzed and described. Using the process and criteria described in DOE-STD-3009-94, safety-related structures, systems and components are identified, and the important safety functions of each SSC are described. Additionally, information which describes the safety management programs at SNL are described in ancillary chapters of the SAR.« less

Real-time 3D radiation risk assessment supporting simulation of work in nuclear environments.

PubMed

Szőke, I; Louka, M N; Bryntesen, T R; Bratteli, J; Edvardsen, S T; RøEitrheim, K K; Bodor, K

2014-06-01

This paper describes the latest developments at the Institute for Energy Technology (IFE) in Norway, in the field of real-time 3D (three-dimensional) radiation risk assessment for the support of work simulation in nuclear environments. 3D computer simulation can greatly facilitate efficient work planning, briefing, and training of workers. It can also support communication within and between work teams, and with advisors, regulators, the media and public, at all the stages of a nuclear installation's lifecycle. Furthermore, it is also a beneficial tool for reviewing current work practices in order to identify possible gaps in procedures, as well as to support the updating of international recommendations, dissemination of experience, and education of the current and future generation of workers.IFE has been involved in research and development into the application of 3D computer simulation and virtual reality (VR) technology to support work in radiological environments in the nuclear sector since the mid 1990s. During this process, two significant software tools have been developed, the VRdose system and the Halden Planner, and a number of publications have been produced to contribute to improving the safety culture in the nuclear industry.This paper describes the radiation risk assessment techniques applied in earlier versions of the VRdose system and the Halden Planner, for visualising radiation fields and calculating dose, and presents new developments towards implementing a flexible and up-to-date dosimetric package in these 3D software tools, based on new developments in the field of radiation protection. The latest versions of these 3D tools are capable of more accurate risk estimation, permit more flexibility via a range of user choices, and are applicable to a wider range of irradiation situations than their predecessors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tan, Lizhen; Yang, Ying; Tyburska-Puschel, Beata

The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools)more » is an important path to more efficient alloy development and process optimization. Ferritic-martensitic (FM) steels are important structural materials for nuclear reactors due to their advantages over other applicable materials like austenitic stainless steels, notably their resistance to void swelling, low thermal expansion coefficients, and higher thermal conductivity. However, traditional FM steels exhibit a noticeable yield strength reduction at elevated temperatures above ~500°C, which limits their applications in advanced nuclear reactors which target operating temperatures at 650°C or higher. Although oxide-dispersion-strengthened (ODS) ferritic steels have shown excellent high-temperature performance, their extremely high cost, limited size and fabricability of products, as well as the great difficulty with welding and joining, have limited or precluded their commercial applications. Zirconium has shown many benefits to Fe-base alloys such as grain refinement, improved phase stability, and reduced radiation-induced segregation. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of a new generation of Zr-bearing ferritic alloys to be fabricated using conventional steelmaking practices, which have excellent radiation resistance and enhanced high-temperature creep performance greater than Grade 91.« less

Nuclear Regulatory Commission issuances, September 1995. Volume 42, Number 3

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

This book contains an issuance of the Atomic Safety and Licensing Board and a Director`s Decision, both of the US Nuclear Regulatory Commission. The issuance concerns the dismissal of a case by adopting a settlement reached by the Staff of the Nuclear Regulatory Commission and a Radiation Safety Officer of a hospital in which the safety officer pled guilty to deliberate misconduct. The Director`s Decision was to deny a petition to impose a fine on Tennessee Valley Authority concerning alleged harassment of the petitioner and to appoint an independent arbitration board to review all past complaints filed against TVA concerningmore » the Watts Bar Nuclear Plant.« less

Pre-irradiation testing and analysis to support the LWRS Hybrid SiC-CMC-Zircaloy-04 unfueled rodlet irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Isabella J van Rooyen

2012-09-01

Nuclear fuel performance is a significant driver of nuclear power plant operational performance, safety, economics and waste disposal requirements. The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Pathway focuses on improving the scientific knowledge basis to enable the development of high-performance, high burn-up fuels with improved safety and cladding integrity and improved nuclear fuel cycle economics. To achieve significant improvements, fundamental changes are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction.

Pre-irradiation testing and analysis to support the LWRS Hybrid SiC-CMC-Zircaloy-04 unfueled rodlet irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Isabella J van Rooyen

2013-01-01

Nuclear fuel performance is a significant driver of nuclear power plant operational performance, safety, economics and waste disposal requirements. The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Pathway focuses on improving the scientific knowledge basis to enable the development of high-performance, high burn-up fuels with improved safety and cladding integrity and improved nuclear fuel cycle economics. To achieve significant improvements, fundamental changes are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Winstanley, J. L.

In August 1945, U.S. Navy Captain William Parsons served as the weaponeer aboard the Enola Gay for the mission to Hiroshima (Shelton 1988). In view of the fact that four B-29s had crashed and burned on takeoff from Tinian the night before, Captain Parsons made the decision to arm the gun-type weapon after takeoff for safety reasons (15 kilotons of TNT equivalent). Although he had no control over the success of the takeoff, he could prevent the possibility of a nuclear detonation on Tinian by controlling what we now call the nuclear explosive. As head of the Ordnance Division atmore » Los Alamos and a former gunnery officer, Captain Parsons clearly understood the role of safety in his work. The advent of the pre-assembled implosion weapon where the high explosive and nuclear materials are always in an intimate configuration meant that nuclear explosive safety became a reality at a certain point in development and production not just at the time of delivery by the military. This is the only industry where nuclear materials are intentionally put in contact with high explosives. The agency of the U.S. Government responsible for development and production of U.S. nuclear weapons is the Department of Energy (DOE) (and its predecessor agencies). This paper will be limited to nuclear explosive safety as it is currently practiced within the DOE nuclear weapons« less

Climate change and safety at work with ionizing radiations.

PubMed

Contessa, Gian Marco; Grandi, Carlo; Scognamiglio, Mario; Genovese, Elisabetta; Sandri, Sandro

2016-01-01

The accident at Tokyo Electric Power Company's (TEPCO's) Fukushima Daiichi nuclear power plant (NPP) has been one of the dominant topic in nuclear safety and it has brought new attention on the matter of accidents in NPPs due to external events related to natural causes. Climate change has risen new risks and the growing probability of extreme external events has increased exposure and vulnerability of workers in the nuclear sector. However extreme natural events are a threat not only to NPPs but to all facilities dealing with radioactive material and in an emergency scenario they can affect the effectiveness and implementation of safety devices and procedures and also prevent communications, causing delays in the readiness of response. It is clear that adaptation strategies are necessary to cope with emerging changes in climate and a new nuclear safety culture is growing, that addresses accidents initiated not only by internal but also by external events.

International nuclear safety concerns. Hearing before the Subcommittee on Energy, Nuclear Proliferation, and Government Processes of the Committee on Governmental Affairs, United States Senate, Ninety-Ninth Congress, Second Session, May 8, 1986

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1986-01-01

Allan I. Mendelowitz of the General Accounting Office (GAO) and James R. Shea of the Nuclear Regulatory Commission were the principal witnesses at a hearing on international concerns about reactor safety. The hearing focused on the Soviet accident at Chernobyl as a demonstration that safety matters are a legitimate area of international concern. Among the issues under discussion were safety standards and inspection procedures of the International Atomic Energy Agency (IAEA). Estimates developed by the GAO show that developing countries, which lack a strong technical base or nuclear background to handle emergencies, will have half the reactors in the worldmore » by the year 2000. Mendelowitz and Shea, together with supporting testimony from others in their agencies described international efforts to improve safeguards.« less

Annual report to Congress: Department of Energy activities relating to the Defense Nuclear Facilities Safety Board, calendar year 1998

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1999-02-01

This is the ninth Annual Report to the Congress describing Department of Energy (Department) activities in response to formal recommendations and other interactions with the Defense Nuclear Facilities Safety Board (Board). The Board, an independent executive-branch agency established in 1988, provides advice and recommendations to the Secretary of energy regarding public health and safety issues at the Department`s defense nuclear facilities. The Board also reviews and evaluates the content and implementation of health and safety standards, as well as other requirements, relating to the design, construction, operation, and decommissioning of the Department`s defense nuclear facilities. The locations of the majormore » Department facilities are provided. During 1998, Departmental activities resulted in the proposed closure of one Board recommendation. In addition, the Department has completed all implementation plan milestones associated with four other Board recommendations. Two new Board recommendations were received and accepted by the Department in 1998, and two new implementation plans are being developed to address these recommendations. The Department has also made significant progress with a number of broad-based initiatives to improve safety. These include expanded implementation of integrated safety management at field sites, a renewed effort to increase the technical capabilities of the federal workforce, and a revised plan for stabilizing excess nuclear materials to achieve significant risk reduction.« less

PRELIMINARY NUCLEAR CRITICALITY NUCLEAR SAFETY EVLAUATION FOR THE CONTAINER SURVEILLANCE AND STORAGE CAPABILITY PROJECT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Low, M; Matthew02 Miller, M; Thomas Reilly, T

2007-04-30

Washington Safety Management Solutions (WSMS) provides criticality safety services to Washington Savannah River Company (WSRC) at the Savannah River Site. One activity at SRS is the Container Surveillance and Storage Capability (CSSC) Project, which will perform surveillances on 3013 containers (hereafter referred to as 3013s) to verify that they meet the Department of Energy (DOE) Standard (STD) 3013 for plutonium storage. The project will handle quantities of material that are greater than ANS/ANSI-8.1 single parameter mass limits, and thus required a Nuclear Criticality Safety Evaluation (NCSE). The WSMS methodology for conducting an NCSE is outlined in the WSMS methods manual.more » The WSMS methods manual currently follows the requirements of DOE-O-420.1B, DOE-STD-3007-2007, and the Washington Savannah River Company (WSRC) SCD-3 manual. DOE-STD-3007-2007 describes how a NCSE should be performed, while DOE-O-420.1B outlines the requirements for a Criticality Safety Program (CSP). The WSRC SCD-3 manual implements DOE requirements and ANS standards. NCSEs do not address the Nuclear Criticality Safety (NCS) of non-reactor nuclear facilities that may be affected by overt or covert activities of sabotage, espionage, terrorism or other security malevolence. Events which are beyond the Design Basis Accidents (DBAs) are outside the scope of a double contingency analysis.« less

Integrated deterministic and probabilistic safety analysis for safety assessment of nuclear power plants

DOE PAGES

Di Maio, Francesco; Zio, Enrico; Smith, Curtis; ...

2015-07-06

The present special issue contains an overview of the research in the field of Integrated Deterministic and Probabilistic Safety Assessment (IDPSA) of Nuclear Power Plants (NPPs). Traditionally, safety regulation for NPPs design and operation has been based on Deterministic Safety Assessment (DSA) methods to verify criteria that assure plant safety in a number of postulated Design Basis Accident (DBA) scenarios. Referring to such criteria, it is also possible to identify those plant Structures, Systems, and Components (SSCs) and activities that are most important for safety within those postulated scenarios. Then, the design, operation, and maintenance of these “safety-related” SSCs andmore » activities are controlled through regulatory requirements and supported by Probabilistic Safety Assessment (PSA).« less

INL Director Discusses Lessons Learned from TMI, Fukushima

ScienceCinema

Grossenbacher, John

2017-12-22

Idaho National Laboratory's Director John Grossenbacher explains how the U.S. nuclear industry has boosted its safety procedures as a result of the Three Mile Island (TMI) accident in 1979 and how the industry plans to use current events at Japan's Fukushima nuclear plants to further enhance safety. For more information about INL's nuclear energy research, visit http://www.facebook.com/idahonationallaboratory.

Proceedings of the Nuclear Criticality Technology Safety Workshop

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rene G. Sanchez

1998-04-01

This document contains summaries of most of the papers presented at the 1995 Nuclear Criticality Technology Safety Project (NCTSP) meeting, which was held May 16 and 17 at San Diego, Ca. The meeting was broken up into seven sessions, which covered the following topics: (1) Criticality Safety of Project Sapphire; (2) Relevant Experiments For Criticality Safety; (3) Interactions with the Former Soviet Union; (4) Misapplications and Limitations of Monte Carlo Methods Directed Toward Criticality Safety Analyses; (5) Monte Carlo Vulnerabilities of Execution and Interpretation; (6) Monte Carlo Vulnerabilities of Representation; and (7) Benchmark Comparisons.

Roadmap to an Engineering-Scale Nuclear Fuel Performance & Safety Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Turner, John A; Clarno, Kevin T; Hansen, Glen A

2009-09-01

Developing new fuels and qualifying them for large-scale deployment in power reactors is a lengthy and expensive process, typically spanning a period of two decades from concept to licensing. Nuclear fuel designers serve an indispensable role in the process, at the initial exploratory phase as well as in analysis of the testing results. In recent years fuel performance capabilities based on first principles have been playing more of a role in what has traditionally been an empirically dominated process. Nonetheless, nuclear fuel behavior is based on the interaction of multiple complex phenomena, and recent evolutionary approaches are being applied moremore » on a phenomenon-by-phenomenon basis, targeting localized problems, as opposed to a systematic approach based on a fundamental understanding of all interacting parameters. Advanced nuclear fuels are generally more complex, and less understood, than the traditional fuels used in existing reactors (ceramic UO{sub 2} with burnable poisons and other minor additives). The added challenges are primarily caused by a less complete empirical database and, in the case of recycled fuel, the inherent variability in fuel compositions. It is clear that using the traditional approach to develop and qualify fuels over the entire range of variables pertinent to the U.S. Department of Energy (DOE) Office of Nuclear Energy on a timely basis with available funds would be very challenging, if not impossible. As a result the DOE Office of Nuclear Energy has launched the Nuclear Energy Advanced Modeling and Simulation (NEAMS) approach to revolutionize fuel development. This new approach is predicated upon transferring the recent advances in computational sciences and computer technologies into the fuel development program. The effort will couple computational science with recent advances in the fundamental understanding of physical phenomena through ab initio modeling and targeted phenomenological testing to leapfrog many fuel-development activities. Realizing the full benefits of this approach will likely take some time. However, it is important that the developmental activities for modeling and simulation be tightly coupled with the experimental activities to maximize feedback effects and accelerate both the experimental and analytical elements of the program toward a common objective. The close integration of modeling and simulation and experimental activities is key to developing a useful fuel performance simulation capability, providing a validated design and analysis tool, and understanding the uncertainties within the models and design process. The efforts of this project are integrally connected to the Transmutation Fuels Campaign (TFC), which maintains as a primary objective to formulate, fabricate, and qualify a transuranic-based fuel with added minor actinides for use in future fast reactors. Additional details of the TFC scope can be found in the Transmutation Fuels Campaign Execution Plan. This project is an integral component of the TFC modeling and simulation effort, and this multiyear plan borrowed liberally from the Transmutation Fuels Campaign Modeling and Simulation Roadmap. This document provides the multiyear staged development plan to develop a continuum-level Integrated Performance and Safety Code (IPSC) to predict the behavior of the fuel and cladding during normal reactor operations and anticipated transients up to the point of clad breach.« less

Results of operation and current safety performance of nuclear facilities located in the Russian Federation

NASA Astrophysics Data System (ADS)

Kuznetsov, V. M.; Khvostova, M. S.

2016-12-01

After the NPP radiation accidents in Russia and Japan, a safety statu of Russian nuclear power plants causes concern. A repeated life time extension of power unit reactor plants, designed at the dawn of the nuclear power engineering in the Soviet Union, power augmentation of the plants to 104-109%, operation of power units in a daily power mode in the range of 100-70-100%, the use of untypical for NPP remixed nuclear fuel without a careful study of the results of its application (at least after two operating periods of the research nuclear installations), the aging of operating personnel, and many other management actions of the State Corporation "Rosatom", should attract the attention of the Federal Service for Ecological, Technical and Atomic Supervision (RosTekhNadzor), but this doesn't happen. The paper considers safety issues of nuclear power plants operating in the Russian Federation. The authors collected statistical information on violations in NPP operation over the past 25 years, which shows that even after repeated relaxation over this period of time of safety regulation requirements in nuclear industry and highly expensive NPP modernization, the latter have not become more safe, and the statistics confirms this. At a lower utilization factor high-power pressure-tube reactors RBMK-1000, compared to light water reactors VVER-440 and 1000, have a greater number of violations and that after annual overhauls. A number of direct and root causes of NPP mulfunctions is still high and remains stable for decades. The paper reveals bottlenecks in ensuring nuclear and radiation safety of nuclear facilities. Main outstanding issues on the storage of spent nuclear fuel are defined. Information on emissions and discharges of radioactive substances, as well as fullness of storages of solid and liquid radioactive waste, located at the NPP sites are presented. Russian NPPs stress test results are submitted, as well as data on the coming removal from operation of NPP units is analyzed.

K Basin Hazard Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

PECH, S.H.

This report describes the methodology used in conducting the K Basins Hazard Analysis, which provides the foundation for the K Basins Final Safety Analysis Report. This hazard analysis was performed in accordance with guidance provided by DOE-STD-3009-94, Preparation Guide for U. S. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports and implements the requirements of DOE Order 5480.23, Nuclear Safety Analysis Report.

Computational nuclear quantum many-body problem: The UNEDF project

NASA Astrophysics Data System (ADS)

Bogner, S.; Bulgac, A.; Carlson, J.; Engel, J.; Fann, G.; Furnstahl, R. J.; Gandolfi, S.; Hagen, G.; Horoi, M.; Johnson, C.; Kortelainen, M.; Lusk, E.; Maris, P.; Nam, H.; Navratil, P.; Nazarewicz, W.; Ng, E.; Nobre, G. P. A.; Ormand, E.; Papenbrock, T.; Pei, J.; Pieper, S. C.; Quaglioni, S.; Roche, K. J.; Sarich, J.; Schunck, N.; Sosonkina, M.; Terasaki, J.; Thompson, I.; Vary, J. P.; Wild, S. M.

2013-10-01

The UNEDF project was a large-scale collaborative effort that applied high-performance computing to the nuclear quantum many-body problem. The primary focus of the project was on constructing, validating, and applying an optimized nuclear energy density functional, which entailed a wide range of pioneering developments in microscopic nuclear structure and reactions, algorithms, high-performance computing, and uncertainty quantification. UNEDF demonstrated that close associations among nuclear physicists, mathematicians, and computer scientists can lead to novel physics outcomes built on algorithmic innovations and computational developments. This review showcases a wide range of UNEDF science results to illustrate this interplay.

Nuclear criticality safety staff training and qualifications at Los Alamos National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Monahan, S.P.; McLaughlin, T.P.

1997-05-01

Operations involving significant quantities of fissile material have been conducted at Los Alamos National Laboratory continuously since 1943. Until the advent of the Laboratory`s Nuclear Criticality Safety Committee (NCSC) in 1957, line management had sole responsibility for controlling criticality risks. From 1957 until 1961, the NCSC was the Laboratory body which promulgated policy guidance as well as some technical guidance for specific operations. In 1961 the Laboratory created the position of Nuclear Criticality Safety Office (in addition to the NCSC). In 1980, Laboratory management moved the Criticality Safety Officer (and one other LACEF staff member who, by that time, wasmore » also working nearly full-time on criticality safety issues) into the Health Division office. Later that same year the Criticality Safety Group, H-6 (at that time) was created within H-Division, and staffed by these two individuals. The training and education of these individuals in the art of criticality safety was almost entirely self-regulated, depending heavily on technical interactions between each other, as well as NCSC, LACEF, operations, other facility, and broader criticality safety community personnel. Although the Los Alamos criticality safety group has grown both in size and formality of operations since 1980, the basic philosophy that a criticality specialist must be developed through mentoring and self motivation remains the same. Formally, this philosophy has been captured in an internal policy, document ``Conduct of Business in the Nuclear Criticality Safety Group.`` There are no short cuts or substitutes in the development of a criticality safety specialist. A person must have a self-motivated personality, excellent communications skills, a thorough understanding of the principals of neutron physics, a safety-conscious and helpful attitude, a good perspective of real risk, as well as a detailed understanding of process operations and credible upsets.« less

Software Design Improvements. Part 1; Software Benefits and Limitations

NASA Technical Reports Server (NTRS)

Lalli, Vincent R.; Packard, Michael H.; Ziemianski, Tom

1997-01-01

Computer hardware and associated software have been used for many years to process accounting information, to analyze test data and to perform engineering analysis. Now computers and software also control everything from automobiles to washing machines and the number and type of applications are growing at an exponential rate. The size of individual program has shown similar growth. Furthermore, software and hardware are used to monitor and/or control potentially dangerous products and safety-critical systems. These uses include everything from airplanes and braking systems to medical devices and nuclear plants. The question is: how can this hardware and software be made more reliable? Also, how can software quality be improved? What methodology needs to be provided on large and small software products to improve the design and how can software be verified?

Mathematics and statistics research progress report, period ending June 30, 1983

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beauchamp, J. J.; Denson, M. V.; Heath, M. T.

1983-08-01

This report is the twenty-sixth in the series of progress reports of Mathematics and Statistics Research of the Computer Sciences organization, Union Carbide Corporation Nuclear Division. Part A records research progress in analysis of large data sets, applied analysis, biometrics research, computational statistics, materials science applications, numerical linear algebra, and risk analysis. Collaboration and consulting with others throughout the Oak Ridge Department of Energy complex are recorded in Part B. Included are sections on biological sciences, energy, engineering, environmental sciences, health and safety, and safeguards. Part C summarizes the various educational activities in which the staff was engaged. Part Dmore » lists the presentations of research results, and Part E records the staff's other professional activities during the report period.« less

Nuclear safety, Volume 38, Number 1, January--March 1997

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1997-03-01

This journal contains nine articles which fall under the following categories: (1) general safety considerations; (2) control and instrumentation; (3) design features (4) environmental effects; (5) US Nuclear Regulatory Commission information and analyses; and (6) recent developments.

Safety system augmentation at Russian nuclear power plants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scerbo, J.A.; Satpute, S.N.; Donkin, J.Y.

1996-12-31

This paper describes the design and procurement of a Class IE DC power supply system to upgrade plant safety at the Kola Nuclear Power Plant (NPP). Kola NPP is located above the Arctic circle at Polyarnie Zorie, Murmansk, Russia. Kola NPP consists of four units. Units 1 and 2 have VVER-440/230 type reactors: Units 3 and 4 have VVER-440/213 type reactors. The VVER-440 reactor design is similar to the pressurized water reactor design used in the US. This project provided redundant, Class 1E DC station batteries and DC switchboards for Kola NPP, Units 1 and 2. The new DC powermore » supply system was designed and procured in compliance with current nuclear design practices and requirements. Technical issues that needed to be addressed included reconciling the requirements in both US and Russian codes and satisfying the requirements of the Russian nuclear regulatory authority. Close interface with ATOMENERGOPROEKT (AEP), the Russian design organization, KOLA NPP plant personnel, and GOSATOMNADZOR (GAN), the Russian version of US Nuclear Regulatory Commission, was necessary to develop a design that would assure compliance with current Russian design requirements. Hence, this project was expected to serve as an example for plant upgrades at other similar VVER-440 nuclear plants. In addition to technical issues, the project needed to address language barriers and the logistics of shipping equipment to a remote section of the Former Soviet Union (FSU). This project was executed by Burns and Roe under the sponsorship of the US DOE as part of the International Safety Program (INSP). The INSP is a comprehensive effort, in cooperation with partners in other countries, to improve nuclear safety worldwide. A major element within the INSP is the improvement of the safety of Soviet-designed nuclear reactors.« less

Hidden dangers: Environmental consequences of preparing for war

DOE Office of Scientific and Technical Information (OSTI.GOV)

Birks, J.W.; Ehrlich, A.H.

1990-01-01

This compilation of chapters by some of the world's foremost non-governmental experts, focuses on the military's nuclear mess. Hidden Dangers suggests that in the end, events, not politics, changed operations' in the nuclear complex. After Chernobyl, safety became the pressing issue. Although the continuing stream of revelations of safety and environmental violations within the US nuclear weapons complex may make the 1990 book seem out of date, it remains an indispensable primer for those concerned with the social and environmental consequences of nuclear weapons production.

Space nuclear safety from a user's viewpoint

NASA Technical Reports Server (NTRS)

Campbell, R. W.

1985-01-01

The National Aeronautics and Space Administration (NASA) launched the Jet Propulsion Laboratory's (JPL) two Voyager spacecraft to Jupiter in 1977, each using three radioisotope thermoelectric generators (RTGs) supplied by the Department of Energy (DOE) for onboard electric power. In 1986 NASA will launch JPL's Galileo spacecraft to Jupiter equipped with two DOE supplied RTGs of an improved design. NASA and JPL are also responsible for obtaining a single RTG of this type from DOE and supplying it to the European Space Agency as part of its participation in the International Solar Polar Mission. As a result of these missions, JPL has been deeply involved in space nuclear safety as a user. This paper will give a brief review of the user contributions by JPL - and NASA in general - to the nuclear safety processes and relate them to the overall nuclear safety program necessary for the launch of an RTG. The two major safety areas requiring user support are the ground operations involving RTGs at the launch site and the failure modes and probabilities associated with launch accidents.

Practical Applications of Cosmic Ray Science: Spacecraft, Aircraft, Ground Based Computation and Control Systems and Human Health and Safety

NASA Technical Reports Server (NTRS)

Atwell, William; Koontz, Steve; Normand, Eugene

2012-01-01

In this paper we review the discovery of cosmic ray effects on the performance and reliability of microelectronic systems as well as on human health and safety, as well as the development of the engineering and health science tools used to evaluate and mitigate cosmic ray effects in earth surface, atmospheric flight, and space flight environments. Three twentieth century technological developments, 1) high altitude commercial and military aircraft; 2) manned and unmanned spacecraft; and 3) increasingly complex and sensitive solid state micro-electronics systems, have driven an ongoing evolution of basic cosmic ray science into a set of practical engineering tools (e.g. ground based test methods as well as high energy particle transport and reaction codes) needed to design, test, and verify the safety and reliability of modern complex electronic systems as well as effects on human health and safety. The effects of primary cosmic ray particles, and secondary particle showers produced by nuclear reactions with spacecraft materials, can determine the design and verification processes (as well as the total dollar cost) for manned and unmanned spacecraft avionics systems. Similar considerations apply to commercial and military aircraft operating at high latitudes and altitudes near the atmospheric Pfotzer maximum. Even ground based computational and controls systems can be negatively affected by secondary particle showers at the Earth's surface, especially if the net target area of the sensitive electronic system components is large. Accumulation of both primary cosmic ray and secondary cosmic ray induced particle shower radiation dose is an important health and safety consideration for commercial or military air crews operating at high altitude/latitude and is also one of the most important factors presently limiting manned space flight operations beyond low-Earth orbit (LEO).

Safety and core design of large liquid-metal cooled fast breeder reactors

NASA Astrophysics Data System (ADS)

Qvist, Staffan Alexander

In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

SNTP environmental, safety, and health

NASA Technical Reports Server (NTRS)

Harmon, Charles D.

1993-01-01

Viewgraphs on space nuclear thermal propulsion (SNTP) environmental, safety, and health are presented. Topics covered include: program safety policy; program safety policies; and DEIS public hearing comments.

76 FR 57980 - Senior Executive Service Performance Review Board

Federal Register 2010, 2011, 2012, 2013, 2014

2011-09-19

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Senior Executive Service Performance Review Board AGENCY... the Defense Nuclear Facilities Safety Board (DNFSB) Senior Executive Service (SES) Performance Review... summary rating of the senior executive's performance, the executive's response, and the higher level...

78 FR 55244 - Senior Executive Service Performance Review Board; Membership

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-10

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Senior Executive Service Performance Review Board... the membership of the Defense Nuclear Facilities Safety Board (DNFSB) Senior Executive Service (SES... rating of a senior executive's performance, the executive's response, and the higher level official's...

77 FR 54570 - Senior Executive Service Performance Review Board

Federal Register 2010, 2011, 2012, 2013, 2014

2012-09-05

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Senior Executive Service Performance Review Board AGENCY... the Defense Nuclear Facilities Safety Board (DNFSB) Senior Executive Service (SES) Performance Review.... The PRB shall review and evaluate the initial summary rating of the senior executive's performance...

10 CFR 830.1 - Scope.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 4 2013-01-01 2013-01-01 false Scope. 830.1 Section 830.1 Energy DEPARTMENT OF ENERGY NUCLEAR SAFETY MANAGEMENT § 830.1 Scope. This part governs the conduct of DOE contractors, DOE personnel... affect, the safety of DOE nuclear facilities. ...

10 CFR 830.1 - Scope.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Scope. 830.1 Section 830.1 Energy DEPARTMENT OF ENERGY NUCLEAR SAFETY MANAGEMENT § 830.1 Scope. This part governs the conduct of DOE contractors, DOE personnel... affect, the safety of DOE nuclear facilities. ...

10 CFR 830.1 - Scope.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 4 2011-01-01 2011-01-01 false Scope. 830.1 Section 830.1 Energy DEPARTMENT OF ENERGY NUCLEAR SAFETY MANAGEMENT § 830.1 Scope. This part governs the conduct of DOE contractors, DOE personnel... affect, the safety of DOE nuclear facilities. ...

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jose Reyes

In recent years it has been recognized that the application of passive safety systems (i.e., those whose operation takes advantage of natural forces such as convection and gravity), can contribute to simplification and potentially to improved economics of new nuclear power plant designs. In 1991 the IAEA Conference on ''The Safety of Nuclear Power: Strategy for the Future'' noted that for new plants the use of passive safety features is a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions, and should be used wherever appropriate''.

NNSA Administrator Tom D'Agostino's speech at the 2009 ISM Conference

ScienceCinema

Thomas D'Agostino

2017-12-09

National Nuclear Security Administration Administrator Thomas DAgostino addressed the 2009 Department of Energy Integrated Safety Management (ISM) Conference in Knoxville, Tenn., on Wednesday, Aug. 26. In his remarks, Administrator DAgostino highlighted the NNSAs track record of developing innovative approaches to workplace safety. And, while he noted the improvements in NNSAs safety record over the years, the Administrator highlighted the need to ensure that workers across the nuclear security enterprise never become complacent in their approach to safety.

Safety features of subcritical fluid fueled systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bell, C.R.

1995-10-01

Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitativemore » in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved is very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible.« less

Radioactive release during nuclear accidents in Chernobyl and Fukushima

NASA Astrophysics Data System (ADS)

Nur Ain Sulaiman, Siti; Mohamed, Faizal; Rahim, Ahmad Nabil Ab

2018-01-01

Nuclear accidents that occurred in Chernobyl and Fukushima have initiated many research interests to understand the cause and mechanism of radioactive release within reactor compound and to the environment. Common types of radionuclide release are the fission products from the irradiated fuel rod itself. In case of nuclear accident, the focus of monitoring will be mostly on the release of noble gases, I-131 and Cs-137. As these are the only accidents have been rated within International Nuclear Events Scale (INES) Level 7, the radioactive release to the environment was one of the critical insights to be monitored. It was estimated that the release of radioactive material to the atmosphere due to Fukushima accident was approximately 10% of the Chernobyl accident. By referring to the previous reports using computational code systems to model the release rate, the release activity of I-131 and Cs-137 in Chernobyl was significantly higher compare to Fukushima. The simulation code also showed that Chernobyl had higher release rate of both radionuclides on the day of accident. Other factors affecting the radioactive release for Fukushima and Chernobyl accidents such as the current reactor technology and safety measures are also compared for discussion.

Performance Assessment of the Commercial CFD Software for the Prediction of the Reactor Internal Flow

NASA Astrophysics Data System (ADS)

Lee, Gong Hee; Bang, Young Seok; Woo, Sweng Woong; Kim, Do Hyeong; Kang, Min Ku

2014-06-01

As the computer hardware technology develops the license applicants for nuclear power plant use the commercial CFD software with the aim of reducing the excessive conservatism associated with using simplified and conservative analysis tools. Even if some of CFD software developer and its user think that a state of the art CFD software can be used to solve reasonably at least the single-phase nuclear reactor problems, there is still limitation and uncertainty in the calculation result. From a regulatory perspective, Korea Institute of Nuclear Safety (KINS) is presently conducting the performance assessment of the commercial CFD software for nuclear reactor problems. In this study, in order to examine the validity of the results of 1/5 scaled APR+ (Advanced Power Reactor Plus) flow distribution tests and the applicability of CFD in the analysis of reactor internal flow, the simulation was conducted with the two commercial CFD software (ANSYS CFX V.14 and FLUENT V.14) among the numerous commercial CFD software and was compared with the measurement. In addition, what needs to be improved in CFD for the accurate simulation of reactor core inlet flow was discussed.

Interdisciplinary Team-Teaching Experience for a Computer and Nuclear Energy Course for Electrical and Computer Engineering Students

ERIC Educational Resources Information Center

Kim, Charles; Jackson, Deborah; Keiller, Peter

2016-01-01

A new, interdisciplinary, team-taught course has been designed to educate students in Electrical and Computer Engineering (ECE) so that they can respond to global and urgent issues concerning computer control systems in nuclear power plants. This paper discusses our experience and assessment of the interdisciplinary computer and nuclear energy…

Nuclear Data Needs for Generation IV Nuclear Energy Systems

NASA Astrophysics Data System (ADS)

Rullhusen, Peter

2006-04-01

Nuclear data needs for generation IV systems. Future of nuclear energy and the role of nuclear data / P. Finck. Nuclear data needs for generation IV nuclear energy systems-summary of U.S. workshop / T. A. Taiwo, H. S. Khalil. Nuclear data needs for the assessment of gen. IV systems / G. Rimpault. Nuclear data needs for generation IV-lessons from benchmarks / S. C. van der Marck, A. Hogenbirk, M. C. Duijvestijn. Core design issues of the supercritical water fast reactor / M. Mori ... [et al.]. GFR core neutronics studies at CEA / J. C. Bosq ... [et al]. Comparative study on different phonon frequency spectra of graphite in GCR / Young-Sik Cho ... [et al.]. Innovative fuel types for minor actinides transmutation / D. Haas, A. Fernandez, J. Somers. The importance of nuclear data in modeling and designing generation IV fast reactors / K. D. Weaver. The GIF and Mexico-"everything is possible" / C. Arrenondo Sánchez -- Benmarks, sensitivity calculations, uncertainties. Sensitivity of advanced reactor and fuel cycle performance parameters to nuclear data uncertainties / G. Aliberti ... [et al.]. Sensitivity and uncertainty study for thermal molten salt reactors / A. Biduad ... [et al.]. Integral reactor physics benchmarks- The International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPHEP) / J. B. Briggs, D. W. Nigg, E. Sartori. Computer model of an error propagation through micro-campaign of fast neutron gas cooled nuclear reactor / E. Ivanov. Combining differential and integral experiments on [symbol] for reducing uncertainties in nuclear data applications / T. Kawano ... [et al.]. Sensitivity of activation cross sections of the Hafnium, Tanatalum and Tungsten stable isotopes to nuclear reaction mechanisms / V. Avrigeanu ... [et al.]. Generating covariance data with nuclear models / A. J. Koning. Sensitivity of Candu-SCWR reactors physics calculations to nuclear data files / K. S. Kozier, G. R. Dyck. The lead cooled fast reactor benchmark BREST-300: analysis with sensitivity method / V. Smirnov ... [et al.]. Sensitivity analysis of neutron cross-sections considered for design and safety studies of LFR and SFR generation IV systems / K. Tucek, J. Carlsson, H. Wider -- Experiments. INL capabilities for nuclear data measurements using the Argonne intense pulsed neutron source facility / J. D. Cole ... [et al.]. Cross-section measurements in the fast neutron energy range / A. Plompen. Recent measurements of neutron capture cross sections for minor actinides by a JNC and Kyoto University Group / H. Harada ... [et al.]. Determination of minor actinides fission cross sections by means of transfer reactions / M. Aiche ... [et al.] -- Evaluated data libraries. Nuclear data services from the NEA / H. Henriksson, Y. Rugama. Nuclear databases for energy applications: an IAEA perspective / R. Capote Noy, A. L. Nichols, A. Trkov. Nuclear data evaluation for generation IV / G. Noguère ... [et al.]. Improved evaluations of neutron-induced reactions on americium isotopes / P. Talou ... [et al.]. Using improved ENDF-based nuclear data for candu reactor calculations / J. Prodea. A comparative study on the graphite-moderated reactors using different evaluated nuclear data / Do Heon Kim ... [et al.].

Comparative study between single core model and detail core model of CFD modelling on reactor core cooling behaviour

NASA Astrophysics Data System (ADS)

Darmawan, R.

2018-01-01

Nuclear power industry is facing uncertainties since the occurrence of the unfortunate accident at Fukushima Daiichi Nuclear Power Plant. The issue of nuclear power plant safety becomes the major hindrance in the planning of nuclear power program for new build countries. Thus, the understanding of the behaviour of reactor system is very important to ensure the continuous development and improvement on reactor safety. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last four decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. Recently, 2D and 3D system level codes such as CFD are being explored. This paper discusses a comparative study on two different approaches of CFD modelling on reactor core cooling behaviour.

Nuclear Regulatory Commission Issuances. Opinions and decisions of the Nuclear Regulatory Commission with selected orders: July 1, 1992--December 31, 1992, Volume 36, Pages 1--396

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1992-12-31

This is the thirty-sixth volume of issuances (1-396) of the Nuclear Regulatory Commission and its Atomic Safety and Licensing Boards, Administrative Law Judges, and Office Directors. It covers the period from July 1, 1992-December 31, 1992. Atomic Safety and Licensing Boards are authorized by Section 191 of the Atomic Energy Act of 1954. These Boards, comprised of three members conduct adjudicatory hearings on applications to construct and operate nuclear power plants and related facilities and issue initial decisions which, subject to internal review and appellate procedures, become the final Commission action with respect to those applications. Boards are drawn frommore » the Atomic Safety and Licensing Board Panel, comprised of lawyers, nuclear physicists and engineers, environmentalists, chemists, and economists. The Atomic Energy Commission first established Licensing Boards in 1962 and the Panel in 1967.« less

The Italian experience on T/H best estimate codes: Achievements and perspectives

DOE Office of Scientific and Technical Information (OSTI.GOV)

Alemberti, A.; D`Auria, F.; Fiorino, E.

1997-07-01

Themalhydraulic system codes are complex tools developed to simulate the power plants behavior during off-normal conditions. Among the objectives of the code calculations the evaluation of safety margins, the operator training, the optimization of the plant design and of the emergency operating procedures, are mostly considered in the field of the nuclear safety. The first generation of codes was developed in the United States at the end of `60s. Since that time, different research groups all over the world started the development of their own codes. At the beginning of the `80s, the second generation codes were proposed; these differmore » from the first generation codes owing to the number of balance equations solved (six instead of three), the sophistication of the constitutive models and of the adopted numerics. The capabilities of available computers have been fully exploited during the years. The authors then summarize some of the major steps in the process of developing, modifying, and advancing the capabilities of the codes. They touch on the fact that Italian, and for that matter non-American, researchers have not been intimately involved in much of this work. They then describe the application of these codes in Italy, even though there are no operating or under construction nuclear power plants at this time. Much of this effort is directed at the general question of plant safety in the face of transient type events.« less

Safety evaluation report on Tennessee Valley Authority: Browns Ferry Nuclear Performance Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1991-01-01

This safety evaluation report (SER) was prepared by the US Nuclear Regulatory Commission (NRC) staff and represents the second and last supplement (SSER 2) to the staff's original SER published as Volume 3 of NUREG-1232 in April 1989. Supplement 1 of Volume 3 of NUREG-1232 (SSER 1) was published in October 1989. Like its predecessors, SSER 2 is composed of numerous safety evaluations by the staff regarding specific elements contained in the Browns Ferry Nuclear Performance Plan (BFNPP), Volume 3 (up to and including Revision 2), submitted by the Tennessee Valley Authority (TVA) for the Browns Ferry Nuclear Plant (BFN).more » The Browns Ferry Nuclear Plant consists of three boiling-water reactors (BWRs) at a site in Limestone County, Alabama. The BFNPP describes the corrective action plans and commitments made by TVA to resolve deficiencies with its nuclear programs before the startup of Unit 2. The staff has inspected and will continue to inspect TVA's implementation of these BFNPP corrective action plans that address staff concerns about TVA's nuclear program. SSER 2 documents the NRC staff's safety evaluations and conclusions for those elements of the BFNPP that were not previously addressed by the staff or that remained open as a result of unresolved issues identified by the staff in previous SERs and inspections.« less

75 FR 56999 - Senior Executive Service Performance Review Board

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-17

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Senior Executive Service Performance Review Board AGENCY... the Defense Nuclear Facilities Safety Board (DNFSB) Senior Executive Service (SES) Performance Review... performance review boards. The PRB shall review and evaluate the initial summary rating of the senior...

Safety engineering: KTA code of practice. Lifting mechanisms in nuclear plant

NASA Astrophysics Data System (ADS)

Lifting mechanisms safety requirements are discussed in accordance with the present state of development of science and engineering for the protection of life, health, and assets against the dangers of nuclear energy and the ill effects of ionizing radiation.

75 FR 43495 - Sunshine Act Notice

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-26

... DEFENSE NUCLEAR FACILITIES SAFETY BOARD Sunshine Act Notice AGENCY: Defense Nuclear Facilities..., structures, and components, and (5) safety-related design aspects of new facilities or modifications of existing facilities needed to deliver high-level waste feed. The Board will be prepared to accept any other...

The in-depth safety assessment (ISA) pilot projects in Ukraine.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kot, C. A.

1998-02-10

Ukraine operates pressurized water reactors of the Soviet-designed type, VVER. All Ukrainian plants are currently operating with annually renewable permits until they update their safety analysis reports (SARs). After approval of the SARS by the Ukrainian Nuclear Regulatory Authority, the plants will be granted longer-term operating licenses. In September 1995, the Nuclear Regulatory Authority and the Government Nuclear Power Coordinating Committee of Ukraine issued a new contents requirement for the safety analysis reports of VVERs in Ukraine. It contains requirements in three major areas: design basis accident (DBA) analysis, probabilistic risk assessment (PRA), and beyond design-basis accident (BDBA) analysis. Themore » DBA requirements are an expanded version of the older SAR requirements. The last two requirements, on PRA and BDBA, are new. The US Department of Energy (USDOE), through the International Nuclear Safety Program (INSP), has initiated an assistance and technology transfer program to Ukraine to assist their nuclear power stations in developing a Western-type technical basis for the new SARS. USDOE sponsored In-Depth Safety Assessments (ISAs) have been initiated at three pilot nuclear reactor units in Ukraine, South Ukraine Unit 1, Zaporizhzhya Unit 5, and Rivne Unit 1. USDOE/INSP have structured the ISA program in such a way as to provide maximum assistance and technology transfer to Ukraine while encouraging and supporting the Ukrainian plants to take the responsibility and initiative and to perform the required assessments.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

LaSalle, F.R.; Golbeg, P.R.; Chenault, D.M.

For reactor and nuclear facilities, both Title 10, Code of Federal Regulations, Part 50, and US Department of Energy Order 6430.1A require assessments of the interaction of non-Safety Class 1 piping and equipment with Safety Class 1 piping and equipment during a seismic event to maintain the safety function. The safety class systems of nuclear reactors or nuclear facilities are designed to the applicable American Society of Mechanical Engineers standards and Seismic Category 1 criteria that require rigorous analysis, construction, and quality assurance. Because non-safety class systems are generally designed to lesser standards and seismic criteria, they may become missilesmore » during a safe shutdown earthquake. The resistance of piping, tubing, and equipment to seismically generated missiles is addressed in the paper. Gross plastic and local penetration failures are considered with applicable test verification. Missile types and seismic zones of influence are discussed. Field qualification data are also developed for missile evaluation.« less

Proceedings of the American Power Conference. Volume 58-II

DOE Office of Scientific and Technical Information (OSTI.GOV)

McBride, A.E.

1996-11-01

This book is part 2 of the proceedings of the American Power Conference, Technology for Competition and Globalization, 1996. The topics of the papers include structural plant design; challenges of the global marketplace; thermal hydraulic methods for nuclear power plant safety and operation; decontamination and decommissioning; competitive operations and maintenance; fuel opportunities; cooling; competitive power pricing; operations; transformers; relays; plant controls; training to meet the competitive future; burning technologies; ash and byproducts utilization; advanced systems; computer tools for plant design; globalization of power; power system protection and power quality; life extension; grounding; and transmission line equipment.

The role of PRA in the safety assessment of VVER Nuclear Power Plants in Ukraine.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kot, C.

1999-05-10

Ukraine operates thirteen (13) Soviet-designed pressurized water reactors, VVERS. All Ukrainian plants are currently operating with annually renewable permits until they update their safety analysis reports (SARs), in accordance with new SAR content requirements issued in September 1995, by the Nuclear Regulatory Authority and the Government Nuclear Power Coordinating Committee of Ukraine. The requirements are in three major areas: design basis accident (DBA) analysis, probabilistic risk assessment (PRA), and beyond design-basis accident (BDBA) analysis. The last two requirements, on PRA and BDBA, are new, and the DBA requirements are an expanded version of the older SAR requirements. The US Departmentmore » of Energy (USDOE), as part of its Soviet-Designed Reactor Safety activities, is providing assistance and technology transfer to Ukraine to support their nuclear power plants (NPPs) in developing a Western-type technical basis for the new SARs. USDOE sponsored In-Depth Safety Assessments (ISAs) are in progress at three pilot nuclear reactor units in Ukraine, South Ukraine Unit 1, Zaporizhzhya Unit 5, and Rivne Unit 1, and a follow-on study has been initiated at Khmenytskyy Unit 1. The ISA projects encompass most areas of plant safety evaluation, but the initial emphasis is on performing a detailed, plant-specific Level 1 Internal Events PRA. This allows the early definition of the plant risk profile, the identification of risk significant accident sequences and plant vulnerabilities and provides guidance for the remainder of the safety assessments.« less

Quarterly report on Defense Nuclear Facilities Safety Board Recommendation 90-7 for the period ending December 31, 1992

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cash, R.J.; Dukelow, G.T.; Forbes, C.J.

1993-03-01

This is the seventh quarterly report on the progress of activities addressing safety issues associated with Hanford Site high-level radioactive waste tanks that contain ferrocyanide compounds. In the presence of oxidizing materials, such as nitrates or nitrites, ferrocyanide can be made to explode in the laboratory by heating it to high temperatures [above 285{degrees}C (545{degrees}F)]. In the mid 1950s approximately 140 metric tons of ferrocyanide were added to 24 underground high-level radioactive waste tanks. An implementation plan (Cash 1991) responding to the Defense Nuclear Facilities Safety Board Recommendation 90-7 (FR 1990) was issued in March 1991 describing the activities thatmore » were planned and underway to address each of the six parts of Recommendation 90-7. A revision to the original plan was transmitted to US Department of Energy by Westinghouse Hanford Company in December 1992. Milestones completed this quarter are described in this report. Contents of this report include: Introduction; Defense Nuclear Facilities Safety Board Implementation Plan Task Activities (Defense Nuclear Facilities Safety Board Recommendation for enhanced temperature measurement, Recommendation for continuous temperature monitoring, Recommendation for cover gas monitoring, Recommendation for ferrocyanide waste characterization, Recommendation for chemical reaction studies, and Recommendation for emergency response planning); Schedules; and References. All actions recommended by the Defense Nuclear Facilities Safety Board for emergency planning by Hanford Site emergency preparedness organizations have been completed.« less

Analysis of radiation safety for Small Modular Reactor (SMR) on PWR-100 MWe type

NASA Astrophysics Data System (ADS)

Udiyani, P. M.; Husnayani, I.; Deswandri; Sunaryo, G. R.

2018-02-01

Indonesia as an archipelago country, including big, medium and small islands is suitable to construction of Small Medium/Modular reactors. Preliminary technology assessment on various SMR has been started, indeed the SMR is grouped into Light Water Reactor, Gas Cooled Reactor, and Solid Cooled Reactor and from its site it is group into Land Based reactor and Water Based Reactor. Fukushima accident made people doubt about the safety of Nuclear Power Plant (NPP), which impact on the public perception of the safety of nuclear power plants. The paper will describe the assessment of safety and radiation consequences on site for normal operation and Design Basis Accident postulation of SMR based on PWR-100 MWe in Bangka Island. Consequences of radiation for normal operation simulated for 3 units SMR. The source term was generated from an inventory by using ORIGEN-2 software and the consequence of routine calculated by PC-Cream and accident by PC Cosyma. The adopted methodology used was based on site-specific meteorological and spatial data. According to calculation by PC-CREAM 08 computer code, the highest individual dose in site area for adults is 5.34E-02 mSv/y in ESE direction within 1 km distance from stack. The result of calculation is that doses on public for normal operation below 1mSv/y. The calculation result from PC Cosyma, the highest individual dose is 1.92.E+00 mSv in ESE direction within 1km distance from stack. The total collective dose (all pathway) is 3.39E-01 manSv, with dominant supporting from cloud pathway. Results show that there are no evacuation countermeasure will be taken based on the regulation of emergency.

Real-time graphic display utility for nuclear safety applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yang, S.; Huang, X.; Taylor, J.

2006-07-01

With the increasing interests in the nuclear energy, new nuclear power plants will be constructed and licensed, and older generation ones will be upgraded for assuring continuing operation. The tendency of adopting the latest proven technology and the fact of older parts becoming obsolete have made the upgrades imperative. One of the areas for upgrades is the older CRT display being replaced by the latest graphics displays running under modern real time operating system (RTOS) with safety graded modern computer. HFC has developed a graphic display utility (GDU) under the QNX RTOS. A standard off-the-shelf software with a long historymore » of performance in industrial applications, QNX RTOS used for safety applications has been examined via a commercial dedication process that is consistent with the regulatory guidelines. Through a commercial survey, a design life cycle and an operating history evaluation, and necessary tests dictated by the dedication plan, it is reasonably confirmed that the QNX RTOS was essentially equivalent to what would be expected in the nuclear industry. The developed GDU operates and communicates with the existing equipment through a dedicated serial channel of a flat panel controller (FPC) module. The FPC module drives a flat panel display (FPD) monitor. A touch screen mounted on the FPD serves as the normal operator interface with the FPC/FPD monitor system. The GDU can be used not only for replacing older CRTs but also in new applications. The replacement of the older CRT does not disturb the function of the existing equipment. It not only provides modern proven technology upgrade but also improves human ergonomics. The FPC, which can be used as a standalone controller running with the GDU, is an integrated hardware and software module. It operates as a single board computer within a control system, and applies primarily to the graphics display, targeting, keyboard and mouse. During normal system operation, the GDU has two sources of data input: a serial interface with field equipment and a serial input from the FPD touch screen. The mechanism for data collection from the field equipment consists of the regular exchange of the data update request messages and target commands sent to the equipment and the update messages returned to the FPC. The data updates from field equipment control displays presented on the graphic pages. Touch screen contacts are decoded to identify physical position that was contacted. If that position corresponds with one of the buttons on the graphic page, the software uses that input to initiate the function defined for the particular button contacted. In this paper, the FPC will be illustrated as a standalone system as well as a module in a dedicated control system. The GDU design concepts and its design flow will be demonstrated. The dedication process of the QNX RTOS needed for the GDU will be highlighted. Finally, the GDU with a specific application example used in one of the nuclear power plants will be presented. (authors)« less

The Fundamentals and Status of Nuclear Power

NASA Astrophysics Data System (ADS)

Matzie, Regis A.

2011-11-01

Nuclear power has enormous potential to provide clean, safe base-load electricity to the world's growing population. Harnessing this potential in an economic and responsible manner is not without challenges. Safety remains the principal tenet of our operating fleet, which currently provides ˜20% of U.S. electricity generated. The performance of this fleet from economic and safety standpoints has improved dramatically over the past several decades. This nuclear generation also represents greater than 70% of the emission free electricity with hydroelectric power providing the majority of the remainder. There have been many lessons learned from the more than 50 years of experience with nuclear power and these have been factored into the new designs now being constructed worldwide. These new designs, which have enhanced safety compared to the operating fleet, have been simplified by employing passive safety systems and modular construction. There are applications for licenses of more than 20 new reactors under review by the U.S. Nuclear Regulatory Commission; the first of these licenses will be completed in early 2012, and the first new U.S. reactor will start operating in 2016. Yet there are still more improvements that can be made and these are being pursued to achieve an even greater deployment of nuclear power technology.

Oak Ridge National Laboratory Health and Safety Long-Range Plan: Fiscal years 1989--1995

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1989-06-01

The health and safety of its personnel is the first concern of ORNL and its management. The ORNL Health and Safety Program has the responsibility for ensuring the health and safety of all individuals assigned to ORNL activities. This document outlines the principal aspects of the ORNL Health and Safety Long-Range Plan and provides a framework for management use in the future development of the health and safety program. Each section of this document is dedicated to one of the health and safety functions (i.e., health physics, industrial hygiene, occupational medicine, industrial safety, nuclear criticality safety, nuclear facility safety, transportationmore » safety, fire protection, and emergency preparedness). Each section includes functional mission and objectives, program requirements and status, a summary of program needs, and program data and funding summary. Highlights of FY 1988 are included.« less

MELCOR computer code manuals: Primer and user`s guides, Version 1.8.3 September 1994. Volume 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users` Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

78 FR 15351 - International Trade Administration

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-11

.... \\1\\ ASEAN Member countries include, Brunei, Burma, Cambodia, Indonesia, Laos, Malaysia, the... generation safety standards. The revised nuclear capacity target for 2020 is now 58 GWe. During the same timeframe, the State Council approved the 12th Five- Year Plan for Nuclear Safety and Radioactive Pollution...

10 CFR 1707.103 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 4 2013-01-01 2013-01-01 false Definitions. 1707.103 Section 1707.103 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD TESTIMONY BY DNFSB EMPLOYEES AND PRODUCTION OF OFFICIAL RECORDS IN LEGAL PROCEEDINGS General Provisions § 1707.103 Definitions. DNFSB means the Defense Nuclear Facilities Safety Board...

10 CFR 1707.103 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Definitions. 1707.103 Section 1707.103 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD TESTIMONY BY DNFSB EMPLOYEES AND PRODUCTION OF OFFICIAL RECORDS IN LEGAL PROCEEDINGS General Provisions § 1707.103 Definitions. DNFSB means the Defense Nuclear Facilities Safety Board...

10 CFR 1707.103 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 4 2012-01-01 2012-01-01 false Definitions. 1707.103 Section 1707.103 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD TESTIMONY BY DNFSB EMPLOYEES AND PRODUCTION OF OFFICIAL RECORDS IN LEGAL PROCEEDINGS General Provisions § 1707.103 Definitions. DNFSB means the Defense Nuclear Facilities Safety Board...

10 CFR 1707.103 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 4 2011-01-01 2011-01-01 false Definitions. 1707.103 Section 1707.103 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD TESTIMONY BY DNFSB EMPLOYEES AND PRODUCTION OF OFFICIAL RECORDS IN LEGAL PROCEEDINGS General Provisions § 1707.103 Definitions. DNFSB means the Defense Nuclear Facilities Safety Board...

10 CFR 1707.103 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 4 2014-01-01 2014-01-01 false Definitions. 1707.103 Section 1707.103 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD TESTIMONY BY DNFSB EMPLOYEES AND PRODUCTION OF OFFICIAL RECORDS IN LEGAL PROCEEDINGS General Provisions § 1707.103 Definitions. DNFSB means the Defense Nuclear Facilities Safety Board...

Thermal-hydraulic interfacing code modules for CANDU reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liu, W.S.; Gold, M.; Sills, H.

1997-07-01

The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

Robust Online Monitoring for Calibration Assessment of Transmitters and Instrumentation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ramuhalli, Pradeep; Coble, Jamie B.; Shumaker, Brent

Robust online monitoring (OLM) technologies are expected to enable the extension or elimination of periodic sensor calibration intervals in operating and new reactors. These advances in OLM technologies will improve the safety and reliability of current and planned nuclear power systems through improved accuracy and increased reliability of sensors used to monitor key parameters. In this article, we discuss an overview of research being performed within the Nuclear Energy Enabling Technologies (NEET)/Advanced Sensors and Instrumentation (ASI) program, for the development of OLM algorithms to use sensor outputs and, in combination with other available information, 1) determine whether one or moremore » sensors are out of calibration or failing and 2) replace a failing sensor with reliable, accurate sensor outputs. Algorithm development is focused on the following OLM functions: • Signal validation • Virtual sensing • Sensor response-time assessment These algorithms incorporate, at their base, a Gaussian Process-based uncertainty quantification (UQ) method. Various plant models (using kernel regression, GP, or hierarchical models) may be used to predict sensor responses under various plant conditions. These predicted responses can then be applied in fault detection (sensor output and response time) and in computing the correct value (virtual sensing) of a failing physical sensor. The methods being evaluated in this work can compute confidence levels along with the predicted sensor responses, and as a result, may have the potential for compensating for sensor drift in real-time (online recalibration). Evaluation was conducted using data from multiple sources (laboratory flow loops and plant data). Ongoing research in this project is focused on further evaluation of the algorithms, optimization for accuracy and computational efficiency, and integration into a suite of tools for robust OLM that are applicable to monitoring sensor calibration state in nuclear power plants.« less

Lessons learned from the Galileo and Ulysses flight safety review experience

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bennett, Gary L.

In preparation for the launches of the Galileo and Ulysses spacecraft, a very comprehensive aerospace nuclear safety program and flight safety review were conducted. A review of this work has highlighted a number of important lessons which should be considered in the safety analysis and review of future space nuclear systems. These lessons have been grouped into six general categories: (1) establishment of the purpose, objectives and scope of the safety process; (2) establishment of charters defining the roles of the various participants; (3) provision of adequate resources; (4) provision of timely peer-reviewed information to support the safety program; (5)more » establishment of general ground rules for the safety review; and (6) agreement on the kinds of information to be provided from the safety review process.« less

Publications of LASL research, 1972--1976

DOE Office of Scientific and Technical Information (OSTI.GOV)

Petersen, L.

1977-04-01

This bibliography is a compilation of unclassified work done at the Los Alamos Scientific Laboratory and published during the years 1972 to 1976. Publications too late for inclusion in earlier compilations are also listed. Declassification of previously classified reports is considered to constitute publication. The bibliography includes LASL reports, journal articles, books, conference papers, papers published in congressional hearings, theses, patents, etc. The following subject areas are included: aerospace studies; analytical technology; astrophysics; atomic and molecular physics, equation of state, opacity; biology and medicine; chemical dynamics and kinetics; chemistry; cryogenics; crystallography; CTR and plasma physics; earth science and engineering; energymore » (nonnuclear); engineering and equipment; EPR, ESR, NMR studies; explosives and detonations; fission physics; health and safety; hydrodynamics and radiation transport; instruments; lasers; mathematics and computers; medium-energy physics; metallurgy and ceramics technology; neutronics and criticality studies; nuclear physics; nuclear safeguards; physics; reactor technology; solid state science; and miscellaneous (including Project Rover). (RWR)« less

The European ASAMPSA_E project : towards guidance to model the impact of high amplitude natural hazards in the probabilistic safety assessment of nuclear power plants. Information on the project progress and needs from the geosciences.

NASA Astrophysics Data System (ADS)

Raimond, Emmanuel; Decker, Kurt; Guigueno, Yves; Klug, Joakim; Loeffler, Horst

2015-04-01

The Fukushima nuclear accident in Japan resulted from the combination of two correlated extreme external events (earthquake and tsunami). The consequences, in particular flooding, went beyond what was considered in the initial engineering design design of nuclear power plants (NPPs). Such situations can in theory be identified using probabilistic safety assessment (PSA) methodology. PSA results may then lead industry (system suppliers and utilities) or Safety Authorities to take appropriate decisions to reinforce the defence-in-depth of the NPP for low probability event but high amplitude consequences. In reality, the development of such PSA remains a challenging task. Definitions of the design basis of NPPs, for example, require data on events with occurrence probabilities not higher than 10-4 per year. Today, even lower probabilities, down to 10-8, are expected and typically used for probabilistic safety analyses (PSA) of NPPs and the examination of so-called design extension conditions. Modelling the combinations of natural or man-made hazards that can affect a NPP and affecting some meaningful probability of occurrence seems to be difficult. The European project ASAMPSAE (www.asampsa.eu) gathers more than 30 organizations (industry, research, safety control) from Europe, US and Japan and aims at identifying some meaningful practices to extend the scope and the quality of the existing probabilistic safety analysis developed for nuclear power plants. It offers a framework to discuss, at a technical level, how "extended PSA" can be developed efficiently and be used to verify if the robustness of Nuclear Power Plants (NPPs) in their environment is sufficient. The paper will present the objectives of this project, some first lessons and introduce which type of guidance is being developed. It will explain the need of expertise from geosciences to support the nuclear safety assessment in the different area (seismotectonic, hydrological, meteorological and biological hazards, …).

Using Machine Learning to Predict MCNP Bias

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grechanuk, Pavel Aleksandrovi

For many real-world applications in radiation transport where simulations are compared to experimental measurements, like in nuclear criticality safety, the bias (simulated - experimental k eff) in the calculation is an extremely important quantity used for code validation. The objective of this project is to accurately predict the bias of MCNP6 [1] criticality calculations using machine learning (ML) algorithms, with the intention of creating a tool that can complement the current nuclear criticality safety methods. In the latest release of MCNP6, the Whisper tool is available for criticality safety analysts and includes a large catalogue of experimental benchmarks, sensitivity profiles,more » and nuclear data covariance matrices. This data, coming from 1100+ benchmark cases, is used in this study of ML algorithms for criticality safety bias predictions.« less

The near boiling reactor: Conceptual design of a small inherently safe nuclear reactor to extend the operational capability of the Victoria Class submarine

NASA Astrophysics Data System (ADS)

Cole, Christopher J. P.

Nuclear power has several unique advantages over other air independent energy sources for nuclear combat submarines. An inherently safe, small nuclear reactor, capable of supply the hotel load of the Victoria Class submarines, has been conceptually developed. The reactor is designed to complement the existing diesel electric power generation plant presently onboard the submarine. The reactor, rated at greater than 1 MW thermal, will supply electricity to the submarine's batteries through an organic Rankine cycle energy conversion plant at 200 kW. This load will increase the operational envelope of the submarine by providing up to 28 continuous days submerged, allowing for an enhanced indiscretion ratio (ratio of time spent on the surface versus time submerged) and a limited under ice capability. The power plant can be fitted into the existing submarine by inserting a 6 m hull plug. With its simplistic design and inherent safety features, the reactor plant will require a minimal addition to the crew. The reactor employs TRISO fuel particles for increased safety. The light water coolant remains at atmospheric pressure, exiting the core at 96°C. Burn-up control and limiting excess reactivity is achieved through movable reflector plates. Shut down and regulatory control is achieved through the thirteen hafnium control rods. Inherent safety is achieved through the negative prompt and delayed temperature coefficients, as well as the negative void coefficient. During a transient, the boiling of the moderator results in a sudden drop in reactivity, essentially shutting down the reactor. It is this characteristic after which the reactor has been named. The design of the reactor was achieved through modelling using computer codes such as MCNP5, WIMS-AECL, FEMLAB, and MicroShield5, in addition to specially written software for kinetics, heat transfer and fission product poisoning calculations. The work has covered a broad area of research and has highlighted additional areas that should be investigated. These include developing a detailed point nodel kinetic model coupled with a finite element heat transfer model, undertaking radiation protection shielding calculations in accordance with international and national regulations, and exploring the effects of advanced fuels.

Numerical Tests for the Problem of U-Pu Fuel Burnup in Fuel Rod and Polycell Models Using the MCNP Code

NASA Astrophysics Data System (ADS)

Muratov, V. G.; Lopatkin, A. V.

An important aspect in the verification of the engineering techniques used in the safety analysis of MOX-fuelled reactors, is the preparation of test calculations to determine nuclide composition variations under irradiation and analysis of burnup problem errors resulting from various factors, such as, for instance, the effect of nuclear data uncertainties on nuclide concentration calculations. So far, no universally recognized tests have been devised. A calculation technique has been developed for solving the problem using the up-to-date calculation tools and the latest versions of nuclear libraries. Initially, in 1997, a code was drawn up in an effort under ISTC Project No. 116 to calculate the burnup in one VVER-1000 fuel rod, using the MCNP Code. Later on, the authors developed a computation technique which allows calculating fuel burnup in models of a fuel rod, or a fuel assembly, or the whole reactor. It became possible to apply it to fuel burnup in all types of nuclear reactors and subcritical blankets.

The ORNL Chemical Technology Division, 1950-1994

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jolley, R.L.; Genung, R.K.; McNeese, L.E.

1994-10-01

This document attempts to reconstruct the role played by the Chemical Technology Division (Chem Tech) of the Oak Ridge National Laboratory (ORNL) in the atomic era since the 1940`s related to the development and production of nuclear weapons and power reactors. Chem Tech`s early contributions were landmark pioneering studies. Unknown and dimly perceived problems like chemical hazards, radioactivity, and criticality had to be dealt with. New chemical concepts and processes had to be developed to test the new theories being developed by physicists. New engineering concepts had to be developed and demonstrated in order to build facilities and equipment thatmore » had never before been attempted. Chem Tech`s role was chemical separations, especially uranium and plutonium, and nuclear fuel reprocessing. With diversification of national and ORNL missions, Chem Tech undertook R&D studies in many areas including biotechnology; clinical and environmental chemistry; nuclear reactors; safety regulations; effective and safe waste management and disposal; computer modeling and informational databases; isotope production; and environmental control. The changing mission of Chem Tech are encapsulated in the evolving activities.« less

Publications of LASL research, 1974

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kerr, A.K.

1975-05-01

This bibliography includes Los Alamos Scientific Laboratory reports, papers released as non-Los Alamos reports, journal articles, books, chapters of books, conference papers (whether published separately or as part of conference proceedings issued as books or reports), papers published in congressional hearings, theses, and U. S. patents. Publications by LASL authors which are not records of Laboratory-sponsored work are included when the Library becomes aware of them. The entries are arranged in sections by broad subject categories; within each section they are alphabetical by title. The following subject categories are included: aerospace studies; analytical technology; astrophysics; atomic and molecular physics, equationmore » of state, opacity; biology and medicine; chemical dynamics and kinetics; chemistry; cryogenics; crystallography; CTR and plasma studies; earth science and engineering; energy (non-nuclear); engineering and equipment; EPR, ESR, NMR studies; explosives and detonations; fission physics; health and safety; hydrodynamics and radiation transport; instruments; lasers; mathematics and computers; medium-energy physics; metallurgy and ceramics technology; neutronic and criticality studies; nuclear physics; nuclear safeguards; physics; reactor technology; solid state science; and miscellaneous (including Project Rover). Author, numerical and KWIC indexes are included. (RWR)« less

RELAP-7 Development Updates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, Hongbin; Zhao, Haihua; Gleicher, Frederick Nathan

RELAP-7 is a nuclear systems safety analysis code being developed at the Idaho National Laboratory, and is the next generation tool in the RELAP reactor safety/systems analysis application series. RELAP-7 development began in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway of the Light Water Reactor Sustainability (LWRS) program. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support nuclear power safety analysis. The code is beingmore » developed based on Idaho National Laboratory’s modern scientific software development framework – MOOSE (the Multi-Physics Object-Oriented Simulation Environment). The initial development goal of the RELAP-7 approach focused primarily on the development of an implicit algorithm capable of strong (nonlinear) coupling of the dependent hydrodynamic variables contained in the 1-D/2-D flow models with the various 0-D system reactor components that compose various boiling water reactor (BWR) and pressurized water reactor nuclear power plants (NPPs). During Fiscal Year (FY) 2015, the RELAP-7 code has been further improved with expanded capability to support boiling water reactor (BWR) and pressurized water reactor NPPs analysis. The accumulator model has been developed. The code has also been coupled with other MOOSE-based applications such as neutronics code RattleSnake and fuel performance code BISON to perform multiphysics analysis. A major design requirement for the implicit algorithm in RELAP-7 is that it is capable of second-order discretization accuracy in both space and time, which eliminates the traditional first-order approximation errors. The second-order temporal is achieved by a second-order backward temporal difference, and the one-dimensional second-order accurate spatial discretization is achieved with the Galerkin approximation of Lagrange finite elements. During FY-2015, we have done numerical verification work to verify that the RELAP-7 code indeed achieves 2nd-order accuracy in both time and space for single phase models at the system level.« less

10 CFR 72.70 - Safety analysis report updating.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Safety analysis report updating. 72.70 Section 72.70 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Records...

10 CFR 1707.301 - Fees.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 4 2011-01-01 2011-01-01 false Fees. 1707.301 Section 1707.301 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD TESTIMONY BY DNFSB EMPLOYEES AND PRODUCTION OF OFFICIAL RECORDS IN LEGAL PROCEEDINGS... (authentication) of copies of records. The Defense Nuclear Facilities Safety Board may certify that records are...

10 CFR 1707.301 - Fees.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Fees. 1707.301 Section 1707.301 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD TESTIMONY BY DNFSB EMPLOYEES AND PRODUCTION OF OFFICIAL RECORDS IN LEGAL PROCEEDINGS... (authentication) of copies of records. The Defense Nuclear Facilities Safety Board may certify that records are...

10 CFR 1707.301 - Fees.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 4 2014-01-01 2014-01-01 false Fees. 1707.301 Section 1707.301 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD TESTIMONY BY DNFSB EMPLOYEES AND PRODUCTION OF OFFICIAL RECORDS IN LEGAL PROCEEDINGS... (authentication) of copies of records. The Defense Nuclear Facilities Safety Board may certify that records are...

10 CFR 1707.301 - Fees.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 4 2013-01-01 2013-01-01 false Fees. 1707.301 Section 1707.301 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD TESTIMONY BY DNFSB EMPLOYEES AND PRODUCTION OF OFFICIAL RECORDS IN LEGAL PROCEEDINGS... (authentication) of copies of records. The Defense Nuclear Facilities Safety Board may certify that records are...

10 CFR 1707.301 - Fees.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 4 2012-01-01 2012-01-01 false Fees. 1707.301 Section 1707.301 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD TESTIMONY BY DNFSB EMPLOYEES AND PRODUCTION OF OFFICIAL RECORDS IN LEGAL PROCEEDINGS... (authentication) of copies of records. The Defense Nuclear Facilities Safety Board may certify that records are...

10 CFR 830.5 - Enforcement.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 4 2012-01-01 2012-01-01 false Enforcement. 830.5 Section 830.5 Energy DEPARTMENT OF ENERGY NUCLEAR SAFETY MANAGEMENT § 830.5 Enforcement. The requirements in this part are DOE Nuclear Safety Requirements and are subject to enforcement by all appropriate means, including the imposition of...

10 CFR 830.5 - Enforcement.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 4 2013-01-01 2013-01-01 false Enforcement. 830.5 Section 830.5 Energy DEPARTMENT OF ENERGY NUCLEAR SAFETY MANAGEMENT § 830.5 Enforcement. The requirements in this part are DOE Nuclear Safety Requirements and are subject to enforcement by all appropriate means, including the imposition of...

10 CFR 830.5 - Enforcement.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Enforcement. 830.5 Section 830.5 Energy DEPARTMENT OF ENERGY NUCLEAR SAFETY MANAGEMENT § 830.5 Enforcement. The requirements in this part are DOE Nuclear Safety Requirements and are subject to enforcement by all appropriate means, including the imposition of...

10 CFR 830.5 - Enforcement.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 4 2011-01-01 2011-01-01 false Enforcement. 830.5 Section 830.5 Energy DEPARTMENT OF ENERGY NUCLEAR SAFETY MANAGEMENT § 830.5 Enforcement. The requirements in this part are DOE Nuclear Safety Requirements and are subject to enforcement by all appropriate means, including the imposition of...

A Figure of Merit: Quantifying the Probability of a Nuclear Reactor Accident.

PubMed

Wellock, Thomas R

In recent decades, probabilistic risk assessment (PRA) has become an essential tool in risk analysis and management in many industries and government agencies. The origins of PRA date to the 1975 publication of the U.S. Nuclear Regulatory Commission's (NRC) Reactor Safety Study led by MIT professor Norman Rasmussen. The "Rasmussen Report" inspired considerable political and scholarly disputes over the motives behind it and the value of its methods and numerical estimates of risk. The Report's controversies have overshadowed the deeper technical origins of risk assessment. Nuclear experts had long sought to express risk in a "figure of merit" to verify the safety of weapons and, later, civilian reactors. By the 1970s, technical advances in PRA gave the methodology the potential to serve political ends, too. The Report, it was hoped, would prove nuclear power's safety to a growing chorus of critics. Subsequent attacks on the Report's methods and numerical estimates damaged the NRC's credibility. PRA's fortunes revived when the 1979 Three Mile Island accident demonstrated PRA's potential for improving the safety of nuclear power and other technical systems. Nevertheless, the Report's controversies endure in mistrust of PRA and its experts.

IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Strydom, Gerhard; Bostelmann, F.

The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained).more » SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on the HTGR Uncertainty Analysis in Modelling (UAM) be implemented. This CRP is a continuation of the previous IAEA and Organization for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) international activities on Verification and Validation (V&V) of available analytical capabilities for HTGR simulation for design and safety evaluations. Within the framework of these activities different numerical and experimental benchmark problems were performed and insight was gained about specific physics phenomena and the adequacy of analysis methods.« less

Nuclear Technology Series. Course 24: Nuclear Systems and Safety.

ERIC Educational Resources Information Center

Center for Occupational Research and Development, Inc., Waco, TX.

This technical specialty course is one of thirty-five courses designed for use by two-year postsecondary institutions in five nuclear technician curriculum areas: (1) radiation protection technician, (2) nuclear instrumentation and control technician, (3) nuclear materials processing technician, (4) nuclear quality-assurance/quality-control…

Moving forward with safety culture.

PubMed

Weber, Michael

2012-04-01

Radiation safety and protection of people are shared goals of the Health Physics Society (HPS) and the U.S. Nuclear Regulatory Commission (NRC). A positive safety culture contributes to achieving radiation safety and protection of people, which are important to both the HPS and the NRC. Through unprecedented collaboration and engagement with diverse stakeholders, the NRC and the stakeholders developed a Safety Culture Policy Statement. The policy statement defines safety culture and describes the traits of a positive safety culture. Consideration of both safety and security issues and the interface of safety and security are underlying principles that support the policy. Examination of significant events, both within the nuclear industry and in society at large, illustrates how weaknesses in these traits can contribute to the occurrence and consequences of safety incidents, including serious injury and loss of life. With the policy statement in place, the NRC is moving forward with outreach and education about safety culture. Health physicists and other radiation safety specialists play an essential role in enhancing safety culture.

The Interagency Nuclear Safety Review Panel's Galileo safety evaluation report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nelson, R.C.; Gray, L.B.; Huff, D.A.

The safety evaluation report (SER) for Galileo was prepared by the Interagency Nuclear Safety Review Panel (INSRP) coordinators in accordance with Presidential directive/National Security Council memorandum 25. The INSRP consists of three coordinators appointed by their respective agencies, the Department of Defense, the Department of Energy (DOE), and the National Aeronautics and Space Administration (NASA). These individuals are independent of the program being evaluated and depend on independent experts drawn from the national technical community to serve on the five INSRP subpanels. The Galileo SER is based on input provided by the NASA Galileo Program Office, review and assessment ofmore » the final safety analysis report prepared by the Office of Special Applications of the DOE under a memorandum of understanding between NASA and the DOE, as well as other related data and analyses. The SER was prepared for use by the agencies and the Office of Science and Technology Policy, Executive Office of the Present for use in their launch decision-making process. Although more than 20 nuclear-powered space missions have been previously reviewed via the INSRP process, the Galileo review constituted the first review of a nuclear power source associated with launch aboard the Space Transportation System.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soubies, B.; Henry, J.Y.; Le Meur, M.

1300 MWe pressurised water reactors (PWRs), like the 1400 MWe reactors, operate with microprocessor-based safety systems. This is particularly the case for the Digital Integrated Protection System (SPIN), which trips the reactor in an emergency and sets in action the safeguard functions. The softwares used in these systems must therefore be highly dependable in the execution of their functions. In the case of SPIN, three players are working at different levels to achieve this goal: the protection system manufacturer, Merlin Gerin; the designer of the nuclear steam supply system, Framatome; the operator of the nuclear power plants, Electricite de Francemore » (EDF), which is also responsible for the safety of its installations. Regulatory licenses are issued by the French safety authority, the Nuclear Installations Safety Directorate (French abbreviation DSIN), subsequent to a successful examination of the technical provisions adopted by the operator. This examination is carried out by the IPSN and the standing group on nuclear reactors. This communication sets out: the methods used by the manufacturer to develop SPIN software for the 1400 MWe PWRs (N4 series); the approach adopted by the IPSN to evaluate the safety software of the protection system for the N4 series of reactors.« less

Light Water Reactor Sustainability Program Reactor Safety Technologies Pathway Technical Program Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Corradini, M. L.; Peko, D.; Farmer, M.

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safetymore » initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.« less

Light Water Reactor Sustainability Program: Reactor Safety Technologies Pathway Technical Program Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Corradini, M. L.

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safetymore » initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.« less

Mathematics and Statistics Research Department progress report, period ending June 30, 1982

DOE Office of Scientific and Technical Information (OSTI.GOV)

Denson, M.V.; Funderlic, R.E.; Gosslee, D.G.

1982-08-01

This report is the twenty-fifth in the series of progress reports of the Mathematics and Statistics Research Department of the Computer Sciences Division, Union Carbide Corporation Nuclear Division (UCC-ND). Part A records research progress in analysis of large data sets, biometrics research, computational statistics, materials science applications, moving boundary problems, numerical linear algebra, and risk analysis. Collaboration and consulting with others throughout the UCC-ND complex are recorded in Part B. Included are sections on biology, chemistry, energy, engineering, environmental sciences, health and safety, materials science, safeguards, surveys, and the waste storage program. Part C summarizes the various educational activities inmore » which the staff was engaged. Part D lists the presentations of research results, and Part E records the staff's other professional activities during the report period.« less

Progress in Decommissioning of Ignalina NPP Unit 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ancius, Darius; Krenevicius, Rimantas; Kutas, Saulius

2002-07-01

The aim of the paper is to present the Lithuanian legal framework regarding the nuclear safety in Decommissioning and Waste Management, and the progress in the Decommissioning Programme of the unit 1 of Ignalina Nuclear Power Plant (INPP). INPP is the only nuclear plant in Lithuania. It comprises two RBMK-1500 reactors. After Lithuania has restored its independence, responsibility for Ignalina NPP was transferred to the Republic of Lithuania. To ensure the control of the Nuclear Safety in Lithuania, The State Nuclear Power Safety Inspectorate (VATESI) was created on 18 October 1991, by a resolution of the Lithuanian Government. Significant workmore » has been performed over the last decade, aiming at upgrading the safety level of the Ignalina NPP with reference to the International standards. On 5 October 1999 the Seimas (Parliament) adopted the National Energy Strategy: It has been decided that unit 1 of Ignalina NPP will be closed down before 2005, The conditions and precise final date of the decommissioning of Unit 2 will be stated in the updated National Energy strategy in 2004. On 20-21 June 2000, the International Donors' Conference for the Decommissioning of Ignalina NPP took place in Vilnius. More than 200 Millions Euro were pledged of which 165 M funded directly from the European Union's budget, as financial support to the Decommissioning projects. The Decommissioning Program encompasses legal, organizational, financial and technical means including the social and economical impacts in the region of Ignalina. The Program is financed from International Support Fund, State budget, National Decommissioning Fund of Ignalina NPP and other funds. Decommissioning of Ignalina NPP is subject to VATESI license according to the Law on Nuclear Energy. The Government established the licensing procedure in the so-called 'Procedure for licensing of Nuclear Activities'; and the document 'General Requirements for Decommissioning of the Ignalina NPP' has been issued by VATESI. A very important issue is the technical support to VATESI and the Lithuanian TSO's (Technical Support Organisations) in their activities within the licensing process related to the Decommissioning of INPP. This includes regulatory assistance in the preparation of decommissioning and radioactive waste management regulatory documents, and technical assistance in the review of the safety case presented by the operator. The Institute for Radioprotection and Nuclear Safety (IRSN, France) and the French Nuclear Safety Authority (DSIN) as well as Swedish International Project (SIP) are providing their support to VATESI in these areas. (authors)« less

Initial empirical analysis of nuclear power plant organization and its effect on safety performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Olson, J.; McLaughlin, S.D.; Osborn, R.N.

This report contains an analysis of the relationship between selected aspects of organizational structure and the safety-related performance of nuclear power plants. The report starts by identifying and operationalizing certain key dimensions of organizational structure that may be expected to be related to plant safety performance. Next, indicators of plant safety performance are created by combining existing performance measures into more reliable indicators. Finally, the indicators of plant safety performance using correlational and discriminant analysis. The overall results show that plants with better developed coordination mechanisms, shorter vertical hierarchies, and a greater number of departments tend to perform more safely.

NEAMS Update. Quarterly Report for October - December 2011.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bradley, K.

2012-02-16

The Advanced Modeling and Simulation Office within the DOE Office of Nuclear Energy (NE) has been charged with revolutionizing the design tools used to build nuclear power plants during the next 10 years. To accomplish this, the DOE has brought together the national laboratories, U.S. universities, and the nuclear energy industry to establish the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program. The mission of NEAMS is to modernize computer modeling of nuclear energy systems and improve the fidelity and validity of modeling results using contemporary software environments and high-performance computers. NEAMS will create a set of engineering-level codes aimedmore » at designing and analyzing the performance and safety of nuclear power plants and reactor fuels. The truly predictive nature of these codes will be achieved by modeling the governing phenomena at the spatial and temporal scales that dominate the behavior. These codes will be executed within a simulation environment that orchestrates code integration with respect to spatial meshing, computational resources, and execution to give the user a common 'look and feel' for setting up problems and displaying results. NEAMS is building upon a suite of existing simulation tools, including those developed by the federal Scientific Discovery through Advanced Computing and Advanced Simulation and Computing programs. NEAMS also draws upon existing simulation tools for materials and nuclear systems, although many of these are limited in terms of scale, applicability, and portability (their ability to be integrated into contemporary software and hardware architectures). NEAMS investments have directly and indirectly supported additional NE research and development programs, including those devoted to waste repositories, safeguarded separations systems, and long-term storage of used nuclear fuel. NEAMS is organized into two broad efforts, each comprising four elements. The quarterly highlights October-December 2011 are: (1) Version 1.0 of AMP, the fuel assembly performance code, was tested on the JAGUAR supercomputer and released on November 1, 2011, a detailed discussion of this new simulation tool is given; (2) A coolant sub-channel model and a preliminary UO{sub 2} smeared-cracking model were implemented in BISON, the single-pin fuel code, more information on how these models were developed and benchmarked is given; (3) The Object Kinetic Monte Carlo model was implemented to account for nucleation events in meso-scale simulations and a discussion of the significance of this advance is given; (4) The SHARP neutronics module, PROTEUS, was expanded to be applicable to all types of reactors, and a discussion of the importance of PROTEUS is given; (5) A plan has been finalized for integrating the high-fidelity, three-dimensional reactor code SHARP with both the systems-level code RELAP7 and the fuel assembly code AMP. This is a new initiative; (6) Work began to evaluate the applicability of AMP to the problem of dry storage of used fuel and to define a relevant problem to test the applicability; (7) A code to obtain phonon spectra from the force-constant matrix for a crystalline lattice has been completed. This important bridge between subcontinuum and continuum phenomena is discussed; (8) Benchmarking was begun on the meso-scale, finite-element fuels code MARMOT to validate its new variable splitting algorithm; (9) A very computationally demanding simulation of diffusion-driven nucleation of new microstructural features has been completed. An explanation of the difficulty of this simulation is given; (10) Experiments were conducted with deformed steel to validate a crystal plasticity finite-element code for bodycentered cubic iron; (11) The Capability Transfer Roadmap was completed and published as an internal laboratory technical report; (12) The AMP fuel assembly code input generator was integrated into the NEAMS Integrated Computational Environment (NiCE). More details on the planned NEAMS computing environment is given; and (13) The NEAMS program website (neams.energy.gov) is nearly ready to launch.« less

Automating Nuclear-Safety-Related SQA Procedures with Custom Applications

DOE PAGES

Freels, James D.

2016-01-01

Nuclear safety-related procedures are rigorous for good reason. Small design mistakes can quickly turn into unwanted failures. Researchers at Oak Ridge National Laboratory worked with COMSOL to define a simulation app that automates the software quality assurance (SQA) verification process and provides results in less than 24 hours.

78 FR 15054 - Agency Information Collection Activities: Proposed Collection; Comment Request

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-08

... nuclear materials to ensure adequate protection of public health and safety, to promote the common defense and security, and to protect the environment. One way to support this mission is through the... assess the safety performance of commercial nuclear power plants and to respond to any decline in...

10 CFR 1706.1 - Scope; statement of policy.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 4 2011-01-01 2011-01-01 false Scope; statement of policy. 1706.1 Section 1706.1 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD ORGANIZATIONAL AND CONSULTANT CONFLICTS OF INTERESTS § 1706.1... the Defense Nuclear Facilities Safety Board will follow in determining whether a contractor or offeror...

10 CFR 1707.205 - Processing demands or requests.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 4 2011-01-01 2011-01-01 false Processing demands or requests. 1707.205 Section 1707.205 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD TESTIMONY BY DNFSB EMPLOYEES AND PRODUCTION OF OFFICIAL... and/or produce official records and information. (b) The Defense Nuclear Facilities Safety Board will...

10 CFR 1707.205 - Processing demands or requests.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 4 2013-01-01 2013-01-01 false Processing demands or requests. 1707.205 Section 1707.205 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD TESTIMONY BY DNFSB EMPLOYEES AND PRODUCTION OF OFFICIAL... and/or produce official records and information. (b) The Defense Nuclear Facilities Safety Board will...

10 CFR 1707.205 - Processing demands or requests.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Processing demands or requests. 1707.205 Section 1707.205 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD TESTIMONY BY DNFSB EMPLOYEES AND PRODUCTION OF OFFICIAL... and/or produce official records and information. (b) The Defense Nuclear Facilities Safety Board will...

10 CFR 1707.204 - Service of subpoenas or requests.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 4 2013-01-01 2013-01-01 false Service of subpoenas or requests. 1707.204 Section 1707.204 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD TESTIMONY BY DNFSB EMPLOYEES AND PRODUCTION OF... be served on the General Counsel, Defense Nuclear Facilities Safety Board, 625 Indiana Avenue, NW...

10 CFR 1705.03 - Systems of records notification.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 4 2014-01-01 2014-01-01 false Systems of records notification. 1705.03 Section 1705.03 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PRIVACY ACT § 1705.03 Systems of records notification. (a... writing. Written requests should be directed to: Privacy Act Officer, Defense Nuclear Facilities Safety...

10 CFR 1707.205 - Processing demands or requests.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 4 2014-01-01 2014-01-01 false Processing demands or requests. 1707.205 Section 1707.205 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD TESTIMONY BY DNFSB EMPLOYEES AND PRODUCTION OF OFFICIAL... and/or produce official records and information. (b) The Defense Nuclear Facilities Safety Board will...

10 CFR 1707.205 - Processing demands or requests.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 4 2012-01-01 2012-01-01 false Processing demands or requests. 1707.205 Section 1707.205 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD TESTIMONY BY DNFSB EMPLOYEES AND PRODUCTION OF OFFICIAL... and/or produce official records and information. (b) The Defense Nuclear Facilities Safety Board will...

10 CFR 1707.204 - Service of subpoenas or requests.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Service of subpoenas or requests. 1707.204 Section 1707.204 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD TESTIMONY BY DNFSB EMPLOYEES AND PRODUCTION OF... be served on the General Counsel, Defense Nuclear Facilities Safety Board, 625 Indiana Avenue, NW...

10 CFR 1706.1 - Scope; statement of policy.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Scope; statement of policy. 1706.1 Section 1706.1 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD ORGANIZATIONAL AND CONSULTANT CONFLICTS OF INTERESTS § 1706.1... the Defense Nuclear Facilities Safety Board will follow in determining whether a contractor or offeror...

10 CFR 1707.204 - Service of subpoenas or requests.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 4 2014-01-01 2014-01-01 false Service of subpoenas or requests. 1707.204 Section 1707.204 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD TESTIMONY BY DNFSB EMPLOYEES AND PRODUCTION OF... be served on the General Counsel, Defense Nuclear Facilities Safety Board, 625 Indiana Avenue, NW...

10 CFR 1706.1 - Scope; statement of policy.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 4 2013-01-01 2013-01-01 false Scope; statement of policy. 1706.1 Section 1706.1 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD ORGANIZATIONAL AND CONSULTANT CONFLICTS OF INTERESTS § 1706.1... the Defense Nuclear Facilities Safety Board will follow in determining whether a contractor or offeror...

10 CFR 1707.204 - Service of subpoenas or requests.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 4 2011-01-01 2011-01-01 false Service of subpoenas or requests. 1707.204 Section 1707.204 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD TESTIMONY BY DNFSB EMPLOYEES AND PRODUCTION OF... be served on the General Counsel, Defense Nuclear Facilities Safety Board, 625 Indiana Avenue, NW...

10 CFR 1706.1 - Scope; statement of policy.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 4 2012-01-01 2012-01-01 false Scope; statement of policy. 1706.1 Section 1706.1 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD ORGANIZATIONAL AND CONSULTANT CONFLICTS OF INTERESTS § 1706.1... the Defense Nuclear Facilities Safety Board will follow in determining whether a contractor or offeror...

10 CFR 1707.204 - Service of subpoenas or requests.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 4 2012-01-01 2012-01-01 false Service of subpoenas or requests. 1707.204 Section 1707.204 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD TESTIMONY BY DNFSB EMPLOYEES AND PRODUCTION OF... be served on the General Counsel, Defense Nuclear Facilities Safety Board, 625 Indiana Avenue, NW...

10 CFR 1706.1 - Scope; statement of policy.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 4 2014-01-01 2014-01-01 false Scope; statement of policy. 1706.1 Section 1706.1 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD ORGANIZATIONAL AND CONSULTANT CONFLICTS OF INTERESTS § 1706.1... the Defense Nuclear Facilities Safety Board will follow in determining whether a contractor or offeror...

10 CFR 1705.08 - Appeals from correction denials.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 4 2014-01-01 2014-01-01 false Appeals from correction denials. 1705.08 Section 1705.08 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PRIVACY ACT § 1705.08 Appeals from correction denials. (a... in writing. This appeal should be directed to The Chairman, Defense Nuclear Facilities Safety Board...

10 CFR 1705.06 - Appeals from access denials.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 4 2012-01-01 2012-01-01 false Appeals from access denials. 1705.06 Section 1705.06 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PRIVACY ACT § 1705.06 Appeals from access denials. When.... This appeal should be directed to The Chairman, Defense Nuclear Facilities Safety Board, 625 Indiana...

10 CFR 1705.03 - Systems of records notification.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 4 2013-01-01 2013-01-01 false Systems of records notification. 1705.03 Section 1705.03 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PRIVACY ACT § 1705.03 Systems of records notification. (a... writing. Written requests should be directed to: Privacy Act Officer, Defense Nuclear Facilities Safety...

10 CFR 1705.08 - Appeals from correction denials.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 4 2012-01-01 2012-01-01 false Appeals from correction denials. 1705.08 Section 1705.08 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PRIVACY ACT § 1705.08 Appeals from correction denials. (a... in writing. This appeal should be directed to The Chairman, Defense Nuclear Facilities Safety Board...

10 CFR 1705.03 - Systems of records notification.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 4 2012-01-01 2012-01-01 false Systems of records notification. 1705.03 Section 1705.03 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PRIVACY ACT § 1705.03 Systems of records notification. (a... writing. Written requests should be directed to: Privacy Act Officer, Defense Nuclear Facilities Safety...

10 CFR 1705.07 - Requests for correction of records.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 4 2013-01-01 2013-01-01 false Requests for correction of records. 1705.07 Section 1705.07 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PRIVACY ACT § 1705.07 Requests for correction of..., Defense Nuclear Facilities Safety Board, 625 Indiana Avenue, NW., Suite 700, Washington, DC 20004. The...

10 CFR 1705.08 - Appeals from correction denials.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 4 2013-01-01 2013-01-01 false Appeals from correction denials. 1705.08 Section 1705.08 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PRIVACY ACT § 1705.08 Appeals from correction denials. (a... in writing. This appeal should be directed to The Chairman, Defense Nuclear Facilities Safety Board...

10 CFR 1705.07 - Requests for correction of records.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 4 2012-01-01 2012-01-01 false Requests for correction of records. 1705.07 Section 1705.07 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PRIVACY ACT § 1705.07 Requests for correction of..., Defense Nuclear Facilities Safety Board, 625 Indiana Avenue, NW., Suite 700, Washington, DC 20004. The...

10 CFR 1705.06 - Appeals from access denials.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 4 2013-01-01 2013-01-01 false Appeals from access denials. 1705.06 Section 1705.06 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PRIVACY ACT § 1705.06 Appeals from access denials. When.... This appeal should be directed to The Chairman, Defense Nuclear Facilities Safety Board, 625 Indiana...

10 CFR 1705.07 - Requests for correction of records.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 4 2014-01-01 2014-01-01 false Requests for correction of records. 1705.07 Section 1705.07 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PRIVACY ACT § 1705.07 Requests for correction of..., Defense Nuclear Facilities Safety Board, 625 Indiana Avenue, NW., Suite 700, Washington, DC 20004. The...

10 CFR 1705.06 - Appeals from access denials.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 4 2014-01-01 2014-01-01 false Appeals from access denials. 1705.06 Section 1705.06 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PRIVACY ACT § 1705.06 Appeals from access denials. When.... This appeal should be directed to The Chairman, Defense Nuclear Facilities Safety Board, 625 Indiana...

75 FR 45678 - Notice of Availability of Interim Staff Guidance Document for Fuel Cycle Facilities

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-03

... Document for Fuel Cycle Facilities AGENCY: Nuclear Regulatory Commission. ACTION: Notice of availability..., Division of Fuel Cycle Safety and Safeguards, Office of Nuclear Material Safety and Safeguards, U.S... Commission (NRC) prepares and issues Interim Staff Guidance (ISG) documents for fuel cycle facilities. These...

78 FR 23312 - Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-18

... NUCLEAR REGULATORY COMMISSION [Docket No. 70-3103; NRC-2010-0264] Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National Enrichment Facility, Eunice, New Mexico..., Division of Fuel Cycle Safety, and Safeguards Office of Nuclear Material Safety, and Safeguards. [FR Doc...

10 CFR 1705.06 - Appeals from access denials.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Appeals from access denials. 1705.06 Section 1705.06 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PRIVACY ACT § 1705.06 Appeals from access denials. When.... This appeal should be directed to The Chairman, Defense Nuclear Facilities Safety Board, 625 Indiana...

10 CFR 1705.03 - Systems of records notification.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Systems of records notification. 1705.03 Section 1705.03 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PRIVACY ACT § 1705.03 Systems of records notification. (a... writing. Written requests should be directed to: Privacy Act Officer, Defense Nuclear Facilities Safety...

10 CFR 1705.08 - Appeals from correction denials.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Appeals from correction denials. 1705.08 Section 1705.08 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PRIVACY ACT § 1705.08 Appeals from correction denials. (a... in writing. This appeal should be directed to The Chairman, Defense Nuclear Facilities Safety Board...

10 CFR 1705.07 - Requests for correction of records.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Requests for correction of records. 1705.07 Section 1705.07 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PRIVACY ACT § 1705.07 Requests for correction of..., Defense Nuclear Facilities Safety Board, 625 Indiana Avenue, NW., Suite 700, Washington, DC 20004. The...

10 CFR Appendix A to Subpart B of... - General Statement of Safety Basis Policy

Code of Federal Regulations, 2010 CFR

2010-01-01

... Analysis Reports for Nuclear Power Plants, or successor document. (2) A DOE nonreactor nuclear facility... with DOE Policy 450.2A, “Identifying, Implementing and Complying with Environment, Safety and Health..., the public and the environment from adverse consequences. These analyses and hazard controls...

10 CFR 1705.07 - Requests for correction of records.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 4 2011-01-01 2011-01-01 false Requests for correction of records. 1705.07 Section 1705.07 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PRIVACY ACT § 1705.07 Requests for correction of..., Defense Nuclear Facilities Safety Board, 625 Indiana Avenue, NW., Suite 700, Washington, DC 20004. The...

10 CFR 1705.03 - Systems of records notification.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 4 2011-01-01 2011-01-01 false Systems of records notification. 1705.03 Section 1705.03 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PRIVACY ACT § 1705.03 Systems of records notification. (a... writing. Written requests should be directed to: Privacy Act Officer, Defense Nuclear Facilities Safety...

10 CFR 1705.06 - Appeals from access denials.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 4 2011-01-01 2011-01-01 false Appeals from access denials. 1705.06 Section 1705.06 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PRIVACY ACT § 1705.06 Appeals from access denials. When.... This appeal should be directed to The Chairman, Defense Nuclear Facilities Safety Board, 625 Indiana...

10 CFR 1705.08 - Appeals from correction denials.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 4 2011-01-01 2011-01-01 false Appeals from correction denials. 1705.08 Section 1705.08 Energy DEFENSE NUCLEAR FACILITIES SAFETY BOARD PRIVACY ACT § 1705.08 Appeals from correction denials. (a... in writing. This appeal should be directed to The Chairman, Defense Nuclear Facilities Safety Board...

Graduate Research Assistant Program for Professional Development at Oak Ridge National Laboratory (ORNL) Global Nuclear Security Technology Division (GNSTD)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eipeldauer, Mary D; Shelander Jr, Bruce R

2012-01-01

The southeast is a highly suitable environment for establishing a series of nuclear safety, security and safeguards 'professional development' courses. Oak Ridge National Laboratory (ORNL) provides expertise in the research component of these subjects while the Y-12 Nuclear Security Complex handles safeguards/security and safety applications. Several universities (i.e., University of Tennessee, Knoxville (UTK), North Carolina State University, University of Michigan, and Georgia Technology Institute) in the region, which offer nuclear engineering and public policy administration programs, and the Howard Baker Center for Public Policy make this an ideal environment for learning. More recently, the Institute for Nuclear Security (INS) wasmore » established between ORNL, Y-12, UTK and Oak Ridge Associate Universities (ORAU), with a focus on five principal areas. These areas include policy, law, and diplomacy; education and training; science and technology; operational and intelligence capability building; and real-world missions and applications. This is a new approach that includes professional development within the graduate research assistant program addressing global needs in nuclear security, safety and safeguards.« less

LWRS ATR Irradiation Testing Readiness Status

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kristine Barrett

2012-09-01

The Light Water Reactor Sustainability (LWRS) Program was established by the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors. The LWRS Program is divided into four R&D Pathways: (1) Materials Aging and Degradation; (2) Advanced Light Water Reactor Nuclear Fuels; (3) Advanced Instrumentation, Information and Control Systems; and (4) Risk-Informed Safety Margin Characterization. This report describes an irradiation testing readiness analysis in preparation of LWRS experiments for irradiation testing at the Idaho National Laboratory (INL) Advanced Testmore » Reactor (ATR) under Pathway (2). The focus of the Advanced LWR Nuclear Fuels Pathway is to improve the scientific knowledge basis for understanding and predicting fundamental performance of advanced nuclear fuel and cladding in nuclear power plants during both nominal and off-nominal conditions. This information will be applied in the design and development of high-performance, high burn-up fuels with improved safety, cladding integrity, and improved nuclear fuel cycle economics« less

77 FR 59994 - Nuclear Fuel Services, Inc., Erwin, TN; Issuance of License Renewal

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-01

... NUCLEAR REGULATORY COMMISSION [Docket No. 70-143; NRC-2009-0435] Nuclear Fuel Services, Inc., Erwin, TN; Issuance of License Renewal AGENCY: U.S. Nuclear Regulatory Commission. ACTION: Notice of... Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001...

Nuclear Energy and the Environment.

ERIC Educational Resources Information Center

International Atomic Energy Agency, Vienna (Austria).

"Nuclear Energy and the Environment" is a pocket folder of removable leaflets concerned with two major topics: Nuclear energy and Nuclear Techniques. Under Nuclear Energy, leaflets concerning the topics of "Radiation--A Fact of Life,""The Impact of a Fact: 1963 Test Ban Treaty,""Energy Needs and Nuclear Power,""Power Reactor Safety,""Transport,"…

Hand held data collection and monitoring system for nuclear facilities

DOEpatents

Brayton, D.D.; Scharold, P.G.; Thornton, M.W.; Marquez, D.L.

1999-01-26

Apparatus and method is disclosed for a data collection and monitoring system that utilizes a pen based hand held computer unit which has contained therein interaction software that allows the user to review maintenance procedures, collect data, compare data with historical trends and safety limits, and input new information at various collection sites. The system has a means to allow automatic transfer of the collected data to a main computer data base for further review, reporting, and distribution purposes and uploading updated collection and maintenance procedures. The hand held computer has a running to-do list so sample collection and other general tasks, such as housekeeping are automatically scheduled for timely completion. A done list helps users to keep track of all completed tasks. The built-in check list assures that work process will meet the applicable processes and procedures. Users can hand write comments or drawings with an electronic pen that allows the users to directly interface information on the screen. 15 figs.

Hand held data collection and monitoring system for nuclear facilities

DOEpatents

Brayton, Darryl D.; Scharold, Paul G.; Thornton, Michael W.; Marquez, Diana L.

1999-01-01

Apparatus and method is disclosed for a data collection and monitoring system that utilizes a pen based hand held computer unit which has contained therein interaction software that allows the user to review maintenance procedures, collect data, compare data with historical trends and safety limits, and input new information at various collection sites. The system has a means to allow automatic transfer of the collected data to a main computer data base for further review, reporting, and distribution purposes and uploading updated collection and maintenance procedures. The hand held computer has a running to-do list so sample collection and other general tasks, such as housekeeping are automatically scheduled for timely completion. A done list helps users to keep track of all completed tasks. The built-in check list assures that work process will meet the applicable processes and procedures. Users can hand write comments or drawings with an electronic pen that allows the users to directly interface information on the screen.