Neutron Capture and the Antineutrino Yield from Nuclear Reactors.

PubMed

Huber, Patrick; Jaffke, Patrick

2016-03-25

We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low energies below 3.2 MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach ∼0.9% of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the nonequilibrium correction. For naval reactors the nonlinear correction may reach the 5% level by the end of cycle.

Thermomechanical analysis of fast-burst reactors

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

Miller, J.D.

1994-08-01

Fast-burst reactors are designed to provide intense, short-duration pulses of neutrons. The fission reaction also produces extreme time-dependent heating of the nuclear fuel. An existing transient-dynamic finite element code was modified specifically to compute the time-dependent stresses and displacements due to thermal shock loads of reactors. Thermomechanical analysis was then applied to determine structural feasibility of various concepts for an EDNA-type reactor and to optimize the mechanical design of the new SPR III-M reactor.

On some control problems of dynamic of reactor

NASA Astrophysics Data System (ADS)

Baskakov, A. V.; Volkov, N. P.

2017-12-01

The paper analyzes controllability of the transient processes in some problems of nuclear reactor dynamics. In this case, the mathematical model of nuclear reactor dynamics is described by a system of integro-differential equations consisting of the non-stationary anisotropic multi-velocity kinetic equation of neutron transport and the balance equation of delayed neutrons. The paper defines the formulation of the linear problem on control of transient processes in nuclear reactors with application of spatially distributed actions on internal neutron sources, and the formulation of the nonlinear problems on control of transient processes with application of spatially distributed actions on the neutron absorption coefficient and the neutron scattering indicatrix. The required control actions depend on the spatial and velocity coordinates. The theorems on existence and uniqueness of these control actions are proved in the paper. To do this, the control problems mentioned above are reduced to equivalent systems of integral equations. Existence and uniqueness of the solution for this system of integral equations is proved by the method of successive approximations, which makes it possible to construct an iterative scheme for numerical analyses of transient processes in a given nuclear reactor with application of the developed mathematical model. Sufficient conditions for controllability of transient processes are also obtained. In conclusion, a connection is made between the control problems and the observation problems, which, by to the given information, allow us to reconstruct either the function of internal neutron sources, or the neutron absorption coefficient, or the neutron scattering indicatrix....

Physics-based multiscale coupling for full core nuclear reactor simulation

DOE PAGES

Gaston, Derek R.; Permann, Cody J.; Peterson, John W.; ...

2015-10-01

Numerical simulation of nuclear reactors is a key technology in the quest for improvements in efficiency, safety, and reliability of both existing and future reactor designs. Historically, simulation of an entire reactor was accomplished by linking together multiple existing codes that each simulated a subset of the relevant multiphysics phenomena. Recent advances in the MOOSE (Multiphysics Object Oriented Simulation Environment) framework have enabled a new approach: multiple domain-specific applications, all built on the same software framework, are efficiently linked to create a cohesive application. This is accomplished with a flexible coupling capability that allows for a variety of different datamore » exchanges to occur simultaneously on high performance parallel computational hardware. Examples based on the KAIST-3A benchmark core, as well as a simplified Westinghouse AP-1000 configuration, demonstrate the power of this new framework for tackling—in a coupled, multiscale manner—crucial reactor phenomena such as CRUD-induced power shift and fuel shuffle. 2014 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY-NC-SA license« less

Boronline, a new generation of boron meter

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

Pirat, P.

2011-07-01

Rolls-Royce is a global business providing integrated power systems for use on land, at sea and in the air. The Group has a balanced business portfolio with leading market positions - civil aerospace, defence aerospace, marine and energy Rolls-Royce understands the challenges of design, procurement, manufacture, operation and in-service support of nuclear reactor plants, with over 50 years of experience through the Royal Navy submarine programme. Rolls-Royce can therefore offer full product life-cycle management for new civil nuclear installations, as well as support to existing installations, including plant lifetime extensions. Rolls-Royce produced for 40 years, Instrumentation and Control (I andmore » C) systems of and associated services for nuclear reactors in Europe, including 58 French reactors and others situated in the United States and in others countries, such as China. Rolls-Royce equipped in this domain 200 nuclear reactors in 20 countries. Among all of its nuclear systems, Rolls Royce is presenting to the conference its new generation of on-line boron measurement system, so called Boronline. (authors)« less

The future of nuclear power: A world-wide perspective

NASA Astrophysics Data System (ADS)

Aktar, Ismail

This study analyzes the future of commercial nuclear electric generation worldwide using the Environmental Kuznets Curve (EKC) concept. The Tobit panel data estimation technique is applied to analyze the data between 1980 and 1998 for 105 countries. EKC assumes that low-income countries increase their nuclear reliance in total electric production whereas high-income countries decrease their nuclear reliance. Hence, we expect that high-income countries should shut down existing nuclear reactors and/or not build any new ones. We encounter two reasons for shutdowns: economic or political/environmental concerns. To distinguish these two effects, reasons for shut down are also investigated by using the Hazard Model technique. Hence, the load factor of a reactor is used as an approximation for economic reason to shut down the reactor. If a shut downed reactor had high load factor, this could be attributable to political/environmental concern or else economic concern. The only countries with nuclear power are considered in this model. The two data sets are created. In the first data set, the single entry for each reactor is created as of 1998 whereas in the second data set, the multiple entries are created for each reactor beginning from 1980 to 1998. The dependent variable takes 1 if operational or zero if shut downed. The empirical findings provide strong evidence for EKC relationship for commercial nuclear electric generation. Furthermore, higher natural resources suggest alternative electric generation methods rather than nuclear power. Economic index as an institutional variable suggests higher the economic freedom, lower the nuclear electric generation as expected. This model does not support the idea to cut the carbon dioxide emission via increasing nuclear share. The Hazard Model findings suggest that higher the load factor is, less likely the reactor will shut down. However, if it is still permanently closed downed, then this could be attributable to political hostility against nuclear power. There are also some projections indicating which reactors are most/least likely to be shut downed from the logit model. We also project which countries are most likely to increase/decrease their nuclear reliance from the residuals of EKC model.

Down-selection of candidate alloys for further testing of advanced replacement materials for LWR core internals

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

Was, Gary; Leonard, Keith J.; Tan, Lizhen

Life extension of the existing nuclear reactors imposes irradiation of high fluences to structural materials, resulting in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) Light Water Reactor Sustainability Program to initiate the Advanced Radiation Resistant Materials (ARRM) program, aiming to identify and develop advanced alloys with superiormore » degradation resistance in light water reactor (LWR)-relevant environments by 2024.« less

Nuclear Energy Policy

DTIC Science & Technology

2010-05-27

small modular reactors and extend the lives and improve the operation of existing commercial nuclear power plants. 40 Interdisciplinary MIT Study, The Future of Nuclear Power, Massachusetts Institute of Technology, 2003, p. 79. 41 Gronlund, Lisbeth, David Lochbaum, and Edwin Lyman, Nuclear Power in a Warming World, Union of Concerned Scientists, December 2007. 42 Travis Madsen, Tony Dutzik, and Bernadette Del Chiaro, et al., Generating Failure: How Building Nuclear Power Plants

Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

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

Forsberg, C.W.; Reich, W.J.

1991-09-01

The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactormore » concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs.« 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.

Jet pump-drive system for heat removal

NASA Technical Reports Server (NTRS)

French, James R. (Inventor)

1987-01-01

The invention does away with the necessity of moving parts such as a check valve in a nuclear reactor cooling system. Instead, a jet pump, in combination with a TEMP, is employed to assure safe cooling of a nuclear reactor after shutdown. A main flow exists for a reactor coolant. A point of withdrawal is provided for a secondary flow. A TEMP, responsive to the heat from said coolant in the secondary flow path, automatically pumps said withdrawn coolant to a higher pressure and thus higher velocity compared to the main flow. The high velocity coolant is applied as a driver flow for the jet pump which has a main flow chamber located in the main flow circulation pump. Upon nuclear shutdown and loss of power for the main reactor pumping system, the TEMP/jet pump combination continues to boost the coolant flow in the direction it is already circulating. During the decay time for the nuclear reactor, the jet pump keeps running until the coolant temperature drops to a lower and safe temperature where the heat is no longer a problem. At this lower temperature, the TEMP/jet pump combination ceases its circulation boosting operation. When the nuclear reactor is restarted and the coolant again exceeds the lower temperature setting, the TEMP/jet pump automatically resumes operation. The TEMP/jet pump combination is thus automatic, self-regulating and provides an emergency pumping system free of moving parts.

Reactor Design and Decommissioning - An Overview of International Activities in Post Fukushima Era1 - 12396

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

Devgun, Jas S.; Laraia, Michele; Pescatore, Claudio

Accidents at the Fukushima Dai-ichi reactors as a result of the devastating earthquake and tsunami of March 11, 2011 have not only dampened the nuclear renaissance but have also initiated a re-examination of the design and safety features for the existing and planned nuclear reactors. Even though failures of some of the key site features at Fukushima can be attributed to events that in the past would have been considered as beyond the design basis, the industry as well as the regulatory authorities are analyzing what features, especially passive features, should be designed into the new reactor designs to minimizemore » the potential for catastrophic failures. It is also recognized that since the design of the Fukushima BWR reactors which were commissioned in 1971, many advanced safety features are now a part of the newer reactor designs. As the recovery efforts at the Fukushima site are still underway, decisions with respect to the dismantlement and decommissioning of the damaged reactors and structures have not yet been finalized. As it was with Three Mile Island, it could take several decades for dismantlement, decommissioning and clean up, and the project poses especially tough challenges. Near-term assessments have been issued by several organizations, including the IAEA, the USNRC and others. Results of such investigations will lead to additional improvements in system and site design measures including strengthening of the anti-tsunami defenses, more defense-in-depth features in reactor design, and better response planning and preparation involving reactor sites. The question also arises what would the effect be on the decommissioning scene worldwide, and what would the effect be on the new reactors when they are eventually retired and dismantled. This paper provides an overview of the US and international activities related to recovery and decommissioning including the decommissioning features in the reactor design process and examines these from a new perspective in the post Fukushima -accident era. Accidents at the Fukushima Daiichi reactors in the aftermath of the devastating earthquake and tsunami of March 11, 2011 have slowed down the nuclear renaissance world-wide and may have accelerated decommissioning either because some countries have decided to halt or reduce nuclear, or because the new safety requirements may reduce life-time extensions. Even in countries such as the UK and France that favor nuclear energy production existing nuclear sites are more likely to be chosen as sites for future NPPs. Even as the site recovery efforts continue at Fukushima and any decommissioning decisions are farther into the future, the accidents have focused attention on the reactor designs in general and specifically on the Fukushima type BWRs. The regulatory authorities in many countries have initiated a re-examination of the design of the systems, structures and components and considerations of the capability of the station to cope with beyond-design basis events. Enhancements to SSCs and site features for the existing reactors and the reactors that will be built will also impact the decommissioning phase activities. The newer reactor designs of today not only have enhanced safety features but also take into consideration the features that will facilitate future decommissioning. Lessons learned from past management and operation of reactors as well as the lessons from decommissioning are incorporated into the new designs. However, in the post-Fukushima era, the emphasis on beyond-design-basis capability may lead to significant changes in SSCs, which eventually will also have impact on the decommissioning phase. Additionally, where some countries decide to phase out the nuclear power, many reactors may enter the decommissioning phase in the coming decade. While the formal updating and expanding of existing guidance documents for accident cleanup and decommissioning would benefit by waiting until the Fukushima project has progressed sufficiently for that experience to be reliably interpreted, the development of structured on-line sharing of information and especially the creation of an on-line compendium of methods, tools, and techniques by which damaged fuel and other unique situations have been addressed can be addressed sooner and maintained as new problems and solutions arise and are resolved. The IAEA's new 'WEB 2.0 tool' CONNECT is expected to play a significant role in this and related information-sharing activities. The trend in some countries such as the United States has been to re-license the existing reactors for additional twenty years, beyond the original design life. Given the advances in technology over the past four decades, and considering that the newer designs incorporate significant improvements in safety systems, it may not be economical or technically feasible to retrofit enhancements into some of the older reactors. In such cases, the reactors may be retired from service and decommissioned. Overall, the energy demand in the world continues to rise, with sharp increases in the Asian countries, and nuclear power's role in the world's energy supply is expected to continue. Events at Fukushima have led to a re-examination on many fronts, including reactor design and regulatory requirements. Further changes may occur in these areas in the post-Fukushima era. These changes in turn will also impact the world-wide decommissioning scene and the decommissioning phase of the future reactors. (authors)« less

A Compact Nuclear Fusion Reactor for Space Flights

NASA Astrophysics Data System (ADS)

Nastoyashchiy, Anatoly F.

2006-05-01

A small-scale nuclear fusion reactor is suggested based on the concepts of plasma confinement (with a high pressure gas) which have been patented by the author. The reactor considered can be used as a power setup in space flights. Among the advantages of this reactor is the use of a D3He fuel mixture which at burning gives main reactor products — charged particles. The energy balance considerably improves, as synchrotron radiation turn out "captured" in the plasma volume, and dangerous, in the case of classical magnetic confinement, instabilities in the direct current magnetic field configuration proposed do not exist. As a result, the reactor sizes are quite suitable (of the order of several meters). A possibility of making reactive thrust due to employment of ejection of multiply charged ions formed at injection of pellets from some adequate substance into the hot plasma center is considered.

Improving online risk assessment with equipment prognostics and health monitoring

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

Coble, Jamie B.; Liu, Xiaotong; Briere, Chris

The current approach to evaluating the risk of nuclear power plant (NPP) operation relies on static probabilities of component failure, which are based on industry experience with the existing fleet of nominally similar light water reactors (LWRs). As the nuclear industry looks to advanced reactor designs that feature non-light water coolants (e.g., liquid metal, high temperature gas, molten salt), this operating history is not available. Many advanced reactor designs use advanced components, such as electromagnetic pumps, that have not been used in the US commercial nuclear fleet. Given the lack of rich operating experience, we cannot accurately estimate the evolvingmore » probability of failure for basic components to populate the fault trees and event trees that typically comprise probabilistic risk assessment (PRA) models. Online equipment prognostics and health management (PHM) technologies can bridge this gap to estimate the failure probabilities for components under operation. The enhanced risk monitor (ERM) incorporates equipment condition assessment into the existing PRA and risk monitor framework to provide accurate and timely estimates of operational risk.« less

An Expert Elicitation of the Proliferation Resistance of Using Small Modular Reactors (SMR) for the Expansion of Civilian Nuclear Systems.

PubMed

Siegel, Jonas; Gilmore, Elisabeth A; Gallagher, Nancy; Fetter, Steve

2018-02-01

To facilitate the use of nuclear energy globally, small modular reactors (SMRs) may represent a viable alternative or complement to large reactor designs. One potential benefit is that SMRs could allow for more proliferation resistant designs, manufacturing arrangements, and fuel-cycle practices at widespread deployment. However, there is limited work evaluating the proliferation resistance of SMRs, and existing proliferation assessment approaches are not well suited for these novel arrangements. Here, we conduct an expert elicitation of the relative proliferation resistance of scenarios for future nuclear energy deployment driven by Generation III+ light-water reactors, fast reactors, or SMRs. Specifically, we construct the scenarios to investigate relevant technical and institutional features that are postulated to enhance the proliferation resistance of SMRs. The experts do not consistently judge the scenario with SMRs to have greater overall proliferation resistance than scenarios that rely on conventional nuclear energy generation options. Further, the experts disagreed on whether incorporating a long-lifetime sealed core into an SMR design would strengthen or weaken proliferation resistance. However, regardless of the type of reactor, the experts judged that proliferation resistance would be enhanced by improving international safeguards and operating several multinational fuel-cycle facilities rather than supporting many more national facilities. © 2017 Society for Risk Analysis.

Accelerated development of Zr-containing new generation ferritic steels for advanced nuclear reactors

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

Tan, Lizhen; Yang, Ying; Sridharan, K.

2015-12-01

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 the 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 computationalmore » tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe 2+) irradiation.« less

FMDP reactor alternative summary report. Volume 1 - existing LWR alternative

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

Greene, S.R.; Bevard, B.B.

1996-10-07

Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] are becoming surplus to national defense needs in both the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES&H) consequences if surplus fissile materials are not properly managed. This document summarizes the results of analysis concerned with existing light water reactor plutonium disposition alternatives.

History of Nuclear India

NASA Astrophysics Data System (ADS)

Chaturvedi, Ram

2000-04-01

India emerged as a free and democratic country in 1947, and entered into the nuclear age in 1948 by establishing the Atomic Energy Commission (AEC), with Homi Bhabha as the chairman. Later on the Department of Atomic Energy (DAE) was created under the Office of the Prime Minister Jawahar Lal Nehru. Initially the AEC and DAE received international cooperation, and by 1963 India had two research reactors and four nuclear power reactors. In spite of the humiliating defeat in the border war by China in 1962 and China's nuclear testing in 1964, India continued to adhere to the peaceful uses of nuclear energy. On May 18, 1974 India performed a 15 kt Peaceful Nuclear Explosion (PNE). The western powers considered it nuclear weapons proliferation and cut off all financial and technical help, even for the production of nuclear power. However, India used existing infrastructure to build nuclear power reactors and exploded both fission and fusion devices on May 11 and 13, 1998. The international community viewed the later activity as a serious road block for the Non-Proliferation Treaty and the Comprehensive Test Ban Treaty; both deemed essential to stop the spread of nuclear weapons. India considers these treaties favoring nuclear states and is prepared to sign if genuine nuclear disarmament is included as an integral part of these treaties.

A Nuclear Energy Renaissance in the U.S.?

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

Kessler, Carol E.; Mahy, Heidi A.; Ankrum, Al

2008-01-01

Is it time for a nuclear energy renaissance? Among other things, nuclear power is a carbon neutral source of base load power. With the growth in energy use expected over the next 20 years and the growing negative impacts of global climate changes, the cost of oil and gas, energy security and diversity concerns, and progress on advanced reactor designs, it may be the right time for nuclear power to enter a new age of growth. Asia and Russia are both planning for a nuclear renaissance. In Europe, Finland and France have both taken steps to pursue new nuclear reactors.more » U.S. utilities are preparing for orders of new reactors; one submitted a request to the U.S. Nuclear Regulatory Commission (NRC) to review its request to construct a new reactor on an existing site. What has the industry been doing since nuclear energy was birthed in the 1960s? In those days a bold new industry boasted that nuclear power in the United States was going to be “too cheap to meter”, but as we all know this did not come about for many reasons. Eventually, it became clear that industry had neglected to do its homework. Critiques of the industry were made on safety, security, environment, economic competitiveness (without government support), and nonproliferation. All of these factors need to be effectively addressed to promote the confidence and support of the public – without which a nuclear power program is not feasible.« less

Accelerator Driven Nuclear Energy: The Thorium Option

ScienceCinema

Raja, Rajendran

2018-01-05

Conventional nuclear reactors use enriched Uranium as fuel and produce nuclear waste which needs to be stored away for over 10,000 years.   At the current rate of use, existing sources of Uranium will last for 50-100 years.  We describe a solution to the problem that uses particle accelerators to produce fast neutrons that can be used to burn existing nuclear waste and produce energy.  Such systems, initially proposed by Carlo Rubbia and collaborators in the 1990's, are being seriously considered by many countries as a possible solution to the green energy problem.  Accelerator driven reactors operate in a sub-critical regime and, thus, are safer and can obtain energy from plentiful elements such as Thorium-232 and Uranium-238. What is missing is the high intensity (10MW) accelerator that produces 1 GeV protons. We will describe scenarios which if implemented will make such systems a reality.  

Advanced Instrumentation for Transient Reactor Testing

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

Corradini, Michael L.; Anderson, Mark; Imel, George

Transient testing involves placing fuel or material into the core of specialized materials test reactors that are capable of simulating a range of design basis accidents, including reactivity insertion accidents, that require the reactor produce short bursts of intense highpower neutron flux and gamma radiation. Testing fuel behavior in a prototypic neutron environment under high-power, accident-simulation conditions is a key step in licensing nuclear fuels for use in existing and future nuclear power plants. Transient testing of nuclear fuels is needed to develop and prove the safety basis for advanced reactors and fuels. In addition, modern fuel development and designmore » increasingly relies on modeling and simulation efforts that must be informed and validated using specially designed material performance separate effects studies. These studies will require experimental facilities that are able to support variable scale, highly instrumented tests providing data that have appropriate spatial and temporal resolution. Finally, there are efforts now underway to develop advanced light water reactor (LWR) fuels with enhanced performance and accident tolerance. These advanced reactor designs will also require new fuel types. These new fuels need to be tested in a controlled environment in order to learn how they respond to accident conditions. For these applications, transient reactor testing is needed to help design fuels with improved performance. In order to maximize the value of transient testing, there is a need for in-situ transient realtime imaging technology (e.g., the neutron detection and imaging system like the hodoscope) to see fuel motion during rapid transient excursions with a higher degree of spatial and temporal resolution and accuracy. There also exists a need for new small, compact local sensors and instrumentation that are capable of collecting data during transients (e.g., local displacements, temperatures, thermal conductivity, neutron flux, etc.).« less

A Pilot Study Investigating the Effects of Advanced Nuclear Power Plant Control Room Technologies: Methods and Qualitative Results

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

BLanc, Katya Le; Powers, David; Joe, Jeffrey

2015-08-01

Control room modernization is an important part of life extension for the existing light water reactor fleet. None of the 99 currently operating commercial nuclear power plants in the U.S. has completed a full-scale control room modernization to date. Nuclear power plant main control rooms for the existing commercial reactor fleet remain significantly analog, with only limited digital modernizations. Upgrades in the U.S. do not achieve the full potential of newer technologies that might otherwise enhance plant and operator performance. The goal of the control room upgrade benefits research is to identify previously overlooked benefits of modernization, identify candidate technologiesmore » that may facilitate such benefits, and demonstrate these technologies through human factors research. This report describes a pilot study to test upgrades to the Human Systems Simulation Laboratory at INL.« less

NASA's Kilopower Reactor Development and the Path to Higher Power Missions

NASA Technical Reports Server (NTRS)

Gibson, Marc A.; Oleson, Steven R.; Poston, David I.; McClure, Patrick

2017-01-01

The development of NASAs Kilopower fission reactor is taking large strides toward flight development with several successful tests completed during its technology demonstration trials. The Kilopower reactors are designed to provide 1-10 kW of electrical power to a spacecraft which could be used for additional science instruments as well as the ability to power electric propulsion systems. Power rich nuclear missions have been excluded from NASA proposals because of the lack of radioisotope fuel and the absence of a flight qualified fission system. NASA has partnered with the Department of Energy's National Nuclear Security Administration to develop the Kilopower reactor using existing facilities and infrastructure to determine if the design is ready for flight development. The 3-year Kilopower project started in 2015 with a challenging goal of building and testing a full-scale flight prototypic nuclear reactor by the end of 2017. As the date approaches, the engineering team shares information on the progress of the technology as well as the enabling capabilities it provides for science and human exploration.

NASA's Kilopower Reactor Development and the Path to Higher Power Missions

NASA Technical Reports Server (NTRS)

Gibson, Marc A.; Oleson, Steven R.; Poston, Dave I.; McClure, Patrick

2017-01-01

The development of NASA's Kilopower fission reactor is taking large strides toward flight development with several successful tests completed during its technology demonstration trials. The Kilopower reactors are designed to provide 1-10 kW of electrical power to a spacecraft which could be used for additional science instruments as well as the ability to power electric propulsion systems. Power rich nuclear missions have been excluded from NASA proposals because of the lack of radioisotope fuel and the absence of a flight qualified fission system. NASA has partnered with the Department of Energy's National Nuclear Security Administration to develop the Kilopower reactor using existing facilities and infrastructure to determine if the design is ready for flight development. The 3-year Kilopower project started in 2015 with a challenging goal of building and testing a full-scale flight prototypic nuclear reactor by the end of 2017. As the date approaches, the engineering team shares information on the progress of the technology as well as the enabling capabilities it provides for science and human exploration.

Numerical and Experimental Thermal Responses of Single-cell and Differential Calorimeters: from Out-of-Pile Calibration to Irradiation Campaigns

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

Brun, J.; Reynard-Carette, C.; Carette, M.

2015-07-01

The nuclear radiation energy deposition rate (usually expressed in W.g{sup -1}) is a key parameter for the thermal design of experiments, on materials and nuclear fuel, carried out in experimental channels of irradiation reactors such as the French OSIRIS reactor in Saclay or inside the Polish MARIA reactor. In particular the quantification of the nuclear heating allows to predicting the heat and thermal conditions induced in the irradiation devices or/and structural materials. Various sensors are used to quantify this parameter, in particular radiometric calorimeters also called in-pile calorimeters. Two main kinds of in-pile calorimeter exist with in particular specific designs:more » single-cell calorimeter and differential calorimeter. The present work focuses on these two calorimeter kinds from their out-of-pile calibration step (transient and steady experiments respectively) to comparison between numerical and experimental results obtained from two irradiation campaigns (MARIA reactor and OSIRIS reactor respectively). The main aim of this paper is to propose a steady numerical approach to estimate the single-cell calorimeter response under irradiation conditions. (authors)« less

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

Calculation of the neutron diffusion equation by using Homotopy Perturbation Method

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

Koklu, H., E-mail: koklu@gantep.edu.tr; Ozer, O.; Ersoy, A.

The distribution of the neutrons in a nuclear fuel element in the nuclear reactor core can be calculated by the neutron diffusion theory. It is the basic and the simplest approximation for the neutron flux function in the reactor core. In this study, the neutron flux function is obtained by the Homotopy Perturbation Method (HPM) that is a new and convenient method in recent years. One-group time-independent neutron diffusion equation is examined for the most solved geometrical reactor core of spherical, cubic and cylindrical shapes, in the frame of the HPM. It is observed that the HPM produces excellent resultsmore » consistent with the existing literature.« less

A model for the release, dispersion and environmental impact of a postulated reactor accident from a submerged commercial nuclear power plant

NASA Astrophysics Data System (ADS)

Bertch, Timothy Creston

1998-12-01

Nuclear power plants are inherently suitable for submerged applications and could provide power to the shore power grid or support future underwater applications. The technology exists today and the construction of a submerged commercial nuclear power plant may become desirable. A submerged reactor is safer to humans because the infinite supply of water for heat removal, particulate retention in the water column, sedimentation to the ocean floor and inherent shielding of the aquatic environment would significantly mitigate the effects of a reactor accident. A better understanding of reactor operation in this new environment is required to quantify the radioecological impact and to determine the suitability of this concept. The impact of release to the environment from a severe reactor accident is a new aspect of the field of marine radioecology. Current efforts have been centered on radioecological impacts of nuclear waste disposal, nuclear weapons testing fallout and shore nuclear plant discharges. This dissertation examines the environmental impact of a severe reactor accident in a submerged commercial nuclear power plant, modeling a postulated site on the Atlantic continental shelf adjacent to the United States. This effort models the effects of geography, decay, particle transport/dispersion, bioaccumulation and elimination with associated dose commitment. The use of a source term equivalent to the release from Chernobyl allows comparison between the impacts of that accident and the postulated submerged commercial reactor plant accident. All input parameters are evaluated using sensitivity analysis. The effect of the release on marine biota is determined. Study of the pathways to humans from gaseous radionuclides, consumption of contaminated marine biota and direct exposure as contaminated water reaches the shoreline is conducted. The model developed by this effort predicts a significant mitigation of the radioecological impact of the reactor accident release with a submerged commercial nuclear power plant. The two box models predict the most of the radio-ecological impact occurs during the first eight days after release. The most significant risk to humans is from consumption of biota. The reduction in impact to humans from a large radioactive release makes the concept worthy of further study.

75 FR 16202 - Notice of Issuance of Regulatory Guide

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-31

..., Revision 2, ``An Acceptable Model and Related Statistical Methods for the Analysis of Fuel Densification.... Introduction The U.S. Nuclear Regulatory Commission (NRC) is issuing a revision to an existing guide in the... nuclear power reactors. To meet these objectives, the guide describes statistical methods related to...

Taking Steps to Protect Against the Insider Threat

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

Pope, Noah Gale; Williams, Martha; Lewis, Joel

2015-10-16

Research reactors are required (in accordance with the Safeguards Agreement between the State and the IAEA) to maintain a system of nuclear material accounting and control for reporting quantities of nuclear material received, shipped, and held on inventory. Enhancements to the existing accounting and control system can be made at little additional cost to the facility, and these enhancements can make nuclear material accounting and control useful for nuclear security. In particular, nuclear material accounting and control measures can be useful in protecting against an insider who is intent on unauthorized removal or misuse of nuclear material or misuse ofmore » equipment. An enhanced nuclear material accounting and control system that responds to nuclear security is described in NSS-25G, Use of Nuclear Material Accounting and Control for Nuclear Security Purposes at Facilities, which is scheduled for distribution by the IAEA Department of Nuclear Security later this year. Accounting and control measures that respond to the insider threat are also described in NSS-33, Establishing a System for Control of Nuclear Material for Nuclear Security Purposes at a Facility During Storage, Use and Movement, and in NSS-41, Preventive and Protective Measures against Insider Threats (originally issued as NSS-08), which are available in draft form. This paper describes enhancements to existing material control and accounting systems that are specific to research reactors, and shows how they are important to nuclear security and protecting against an insider.« less

The Nuclear Renaissance in the U.S.

ScienceCinema

Buongiorno, Jacopo

2018-04-23

Nuclear power currently provides 20% of the electricity generation in the U.S. and about 16% worldwide.  As a carbon-free energy source, nuclear is receiving a lot of attention by industry, lawmakers and environmental groups, as they attempt to resolve the issue of man-made climate change.  For the first time in 30 years several U.S. electric utilities have applied for construction and operation licenses of new nuclear power plants.  This talk will review the safety, operational and economic record of the existing U.S. commercial reactor fleet, will provide an overview of the reactor designs considered for the new wave of plant construction, and will discuss several research projects being conducted at the Massachusetts Institute of Technology to support the expansion of nuclear power in the U.S. and overseas.

A survey of Existing V&V, UQ and M&S Data and Knowledge Bases in Support of the Nuclear Energy - Knowledge base for Advanced Modeling and Simulation (NE-KAMS)

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

Hyung Lee; Rich Johnson, Ph.D.; Kimberlyn C. Moussesau

2011-12-01

The Nuclear Energy - Knowledge base for Advanced Modeling and Simulation (NE-KAMS) is being developed at the Idaho National Laboratory in conjunction with Bettis Laboratory, Sandia National Laboratories, Argonne National Laboratory, Oak Ridge National Laboratory, Utah State University and others. The objective of this consortium is to establish a comprehensive knowledge base to provide Verification and Validation (V&V) and Uncertainty Quantification (UQ) and other resources for advanced modeling and simulation (M&S) in nuclear reactor design and analysis. NE-KAMS will become a valuable resource for the nuclear industry, the national laboratories, the U.S. NRC and the public to help ensure themore » safe operation of existing and future nuclear reactors. A survey and evaluation of the state-of-the-art of existing V&V and M&S databases, including the Department of Energy and commercial databases, has been performed to ensure that the NE-KAMS effort will not be duplicating existing resources and capabilities and to assess the scope of the effort required to develop and implement NE-KAMS. The survey and evaluation have indeed highlighted the unique set of value-added functionality and services that NE-KAMS will provide to its users. Additionally, the survey has helped develop a better understanding of the architecture and functionality of these data and knowledge bases that can be used to leverage the development of NE-KAMS.« less

Experiences in utilization of research reactors in Yugoslavia

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

Copic, M.; Gabrovsek, Z.; Pop-Jordanov, J.

1971-06-15

The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied bymore » means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro envisage the role of research reactors in the promotion of nuclear power programs in relation to the size of the program, the competence of domestic industries and the degree of independence where fuel supply is concerned. (author)« less

An experimental investigation of (UF-235)6 fission nuclear-pumped lasers

NASA Technical Reports Server (NTRS)

Miley, G. H.

1979-01-01

A UF6 handling system was designed for use in conjunction with the existing nuclear-pumped laser vacuum system at a nuclear reactor laboratory to perform the experiments described above. A modification to separate the gas fill system from the vacuum system and thus greatly reduce its volume is described as well as operating procedures for the first controlled nuclear pumping experiments with UF6 vapor contained in the laser cell.

MACHINING TEST SPECIMENS FROM HARVESTED ZION RPV SEGMENTS

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

Nanstad, Randy K; Rosseel, Thomas M; Sokolov, Mikhail A

2015-01-01

The decommissioning of the Zion Nuclear Generating Station (NGS) in Zion, Illinois, presents a special and timely opportunity for developing a better understanding of materials degradation and other issues associated with extending the lifetime of existing nuclear power plants (NPPs) beyond 60 years of service. In support of extended service and current operations of the US nuclear reactor fleet, the Oak Ridge National Laboratory (ORNL), through the Department of Energy (DOE), Light Water Reactor Sustainability (LWRS) Program, is coordinating and contracting with Zion Solutions, LLC, a subsidiary of Energy Solutions, an international nuclear services company, the selective procurement of materials,more » structures, components, and other items of interest from the decommissioned reactors. In this paper, we will discuss the acquisition of segments of the Zion Unit 2 Reactor Pressure Vessel (RPV), cutting these segments into blocks from the beltline and upper vertical welds and plate material and machining those blocks into mechanical (Charpy, compact tension, and tensile) test specimens and coupons for microstructural (TEM, SEM, APT, SANS and nano indention) characterization. Access to service-irradiated RPV welds and plate sections will allow through wall attenuation studies to be performed, which will be used to assess current radiation damage models [1].« less

Improved Delayed-Neutron Spectroscopy Using Trapped Ions

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

Norman, Eric B.

The neutrons emitted following the β decay of fission fragments (known as delayed neutrons because they are emitted after fission on a timescale of the β-decay half-lives) play a crucial role in reactor performance and control. Reviews of delayed-neutron properties highlight the need for high-quality data for a wide variety of delayed-neutron emitters to better understand the time dependence and energy spectrum of the neutrons as these properties are essential for a detailed understanding of reactor kinetics needed for reactor safety and to understand the behavior of these reactors under various accident and component-failure scenarios. For fast breeder reactors, criticalitymore » calculations require accurate delayed-neutron energy spectra and approximations that are acceptable for light-water reactors such as assuming the delayed-neutron and fission-neutron energy spectra are identical are not acceptable and improved β-delayed neutron data is needed for safety and accident analyses for these reactors. With improved nuclear data, the delayed neutrons flux and energy spectrum could be calculated from the contributions from individual isotopes and therefore could be accurately modeled for any fuel-cycle concept, actinide mix, or irradiation history. High-quality β-delayed neutron measurements are also critical to constrain modern nuclear-structure calculations and empirical models that predict the decay properties for nuclei for which no data exists and improve the accuracy and flexibility of the existing empirical descriptions of delayed neutrons from fission such as the six-group representation« less

A Comparison Framework for Reactor Anti-Neutrino Detectors in Near-Field Nuclear Safeguards Applications

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

Mendenhall, M.; Bowden, N.; Brodsky, J.

Electron anti-neutrino ( e) detectors can support nuclear safeguards, from reactor monitoring to spent fuel characterization. In recent years, the scientific community has developed multiple detector concepts, many of which have been prototyped or deployed for specific measurements by their respective collaborations. However, the diversity of technical approaches, deployment conditions, and analysis techniques complicates direct performance comparison between designs. We have begun development of a simulation framework to compare and evaluate existing and proposed detector designs for nonproliferation applications in a uniform manner. This report demonstrates the intent and capabilities of the framework by evaluating four detector design concepts, calculatingmore » generic reactor antineutrino counting sensitivity, and capabilities in a plutonium disposition application example.« less

Nuclear power: levels of safety.

PubMed

Lidsky, L M

1988-02-01

The rise and fall of the nuclear power industry in the United States is a well-documented story with enough socio-technological conflict to fill dozens of scholarly, and not so scholarly, books. Whatever the reasons for the situation we are now in, and no matter how we apportion the blame, the ultimate choice of whether to use nuclear power in this country is made by the utilities and by the public. Their choices are, finally, based on some form of risk-benefit analysis. Such analysis is done in well-documented and apparently logical form by the utilities and in a rather more inchoate but not necessarily less accurate form by the public. Nuclear power has failed in the United States because both the real and perceived risks outweigh the potential benefits. The national decision not to rely upon nuclear power in its present form is not an irrational one. A wide ranging public balancing of risk and benefit requires a classification of risk which is clear and believable for the public to be able to assess the risks associated with given technological structures. The qualitative four-level safety ladder provides such a framework. Nuclear reactors have been designed which fit clearly and demonstrably into each of the possible qualitative safety levels. Surprisingly, it appears that safer may also mean cheaper. The intellectual and technical prerequisites are in hand for an important national decision. Deployment of a qualitatively different second generation of nuclear reactors can have important benefits for the United States. Surprisingly, it may well be the "nuclear establishment" itself, with enormous investments of money and pride in the existing nuclear systems, that rejects second generation reactors. It may be that we will not have a second generation of reactors until the first generation of nuclear engineers and nuclear power advocates has retired.

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

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

Smith, Cyrus M; Nanstad, Randy K; Clayton, Dwight A

2012-09-01

The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin nextmore » year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.« less

Gaseous-fuel nuclear reactor research for multimegawatt power in space

NASA Technical Reports Server (NTRS)

Thom, K.; Schneider, R. T.; Helmick, H. H.

1977-01-01

In the gaseous-fuel reactor concept, the fissile material is contained in a moderator-reflector cavity and exists in the form of a flowing gas or plasma separated from the cavity walls by means of fluid mechanical forces. Temperatures in excess of structural limitations are possible for low-specific-mass power and high-specific-impulse propulsion in space. Experiments have been conducted with a canister filled with enriched UF6 inserted into a beryllium-reflected cavity. A theoretically predicted critical mass of 6 kg was measured. The UF6 was also circulated through this cavity, demonstrating stable reactor operation with the fuel in motion. Because the flowing gaseous fuel can be continuously processed, the radioactive waste in this type of reactor can be kept small. Another potential of fissioning gases is the possibility of converting the kinetic energy of fission fragments directly into coherent electromagnetic radiation, the nuclear pumping of lasers. Numerous nuclear laser experiments indicate the possibility of transmitting power in space directly from fission energy. The estimated specific mass of a multimegawatt gaseous-fuel reactor power system is from 1 to 5 kg/kW while the companion laser-power receiver station would be much lower in specific mass.

High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems.

PubMed

Mahadevan, Vijay S; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

2014-08-06

An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework.

High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems

PubMed Central

Mahadevan, Vijay S.; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

2014-01-01

An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework. PMID:24982250

A Research Framework for Demonstrating Benefits of Advanced Control Room Technologies

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

Le Blanc, Katya; Boring, Ronald; Joe, Jeffrey

Control Room modernization is an important part of life extension for the existing light water reactor fleet. None of the 99 currently operating commercial nuclear power plants in the U.S. has completed a full-scale control room modernization to date. A full-scale modernization might, for example, entail replacement of all analog panels with digital workstations. Such modernizations have been undertaken successfully in upgrades in Europe and Asia, but the U.S. has yet to undertake a control room upgrade of this magnitude. Instead, nuclear power plant main control rooms for the existing commercial reactor fleet remain significantly analog, with only limited digitalmore » modernizations. Previous research under the U.S. Department of Energy’s Light Water Reactor Sustainability Program has helped establish a systematic process for control room upgrades that support the transition to a hybrid control. While the guidance developed to date helps streamline the process of modernization and reduce costs and uncertainty associated with introducing digital control technologies into an existing control room, these upgrades do not achieve the full potential of newer technologies that might otherwise enhance plant and operator performance. The aim of the control room benefits research presented here is to identify previously overlooked benefits of modernization, identify candidate technologies that may facilitate such benefits, and demonstrate these technologies through human factors research. This report serves as an outline for planned research on the benefits of greater modernization in the main control rooms of nuclear power plants.« less

Reactor safeguards system assessment and design. Volume I

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

Varnado, G.B.; Ericson, D.M. Jr.; Daniel, S.L.

1978-06-01

This report describes the development and application of a methodology for evaluating the effectiveness of nuclear power reactor safeguards systems. Analytic techniques are used to identify the sabotage acts which could lead to release of radioactive material from a nuclear power plant, to determine the areas of a plant which must be protected to assure that significant release does not occur, to model the physical plant layout, and to evaluate the effectiveness of various safeguards systems. The methodology was used to identify those aspects of reactor safeguards systems which have the greatest effect on overall system performance and which, therefore,more » should be emphasized in the licensing process. With further refinements, the methodology can be used by the licensing reviewer to aid in assessing proposed or existing safeguards systems.« less

An Update on Improvements to NiCE Support for PROTEUS

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

Bennett, Andrew; McCaskey, Alexander J.; Billings, Jay Jay

2015-09-01

The Department of Energy Office of Nuclear Energy's Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has supported the development of the NEAMS Integrated Computational Environment (NiCE), a modeling and simulation workflow environment that provides services and plugins to facilitate tasks such as code execution, model input construction, visualization, and data analysis. This report details the development of workflows for the reactor core neutronics application, PROTEUS. This advanced neutronics application (primarily developed at Argonne National Laboratory) aims to improve nuclear reactor design and analysis by providing an extensible and massively parallel, finite-element solver for current and advanced reactor fuel neutronicsmore » modeling. The integration of PROTEUS-specific tools into NiCE is intended to make the advanced capabilities that PROTEUS provides more accessible to the nuclear energy research and development community. This report will detail the work done to improve existing PROTEUS workflow support in NiCE. We will demonstrate and discuss these improvements, including the development of flexible IO services, an improved interface for input generation, and the addition of advanced Fortran development tools natively in the platform.« less

Nuclear Thermal Propulsion: A Joint NASA/DOE/DOD Workshop

NASA Technical Reports Server (NTRS)

Clark, John S. (Editor)

1991-01-01

Papers presented at the joint NASA/DOE/DOD workshop on nuclear thermal propulsion are compiled. The following subject areas are covered: nuclear thermal propulsion programs; Rover/NERVA and NERVA systems; Low Pressure Nuclear Thermal Rocket (LPNTR); particle bed reactor nuclear rocket; hybrid propulsion systems; wire core reactor; pellet bed reactor; foil reactor; Droplet Core Nuclear Rocket (DCNR); open cycle gas core nuclear rockets; vapor core propulsion reactors; nuclear light bulb; Nuclear rocket using Indigenous Martian Fuel (NIMF); mission analysis; propulsion and reactor technology; development plans; and safety issues.

Emulation of reactor irradiation damage using ion beams

DOE PAGES

Was, G. S.; Jiao, Z.; Getto, E.; ...

2014-06-14

The continued operation of existing light water nuclear reactors and the development of advanced nuclear reactor depend heavily on understanding how damage by radiation to levels degrades materials that serve as the structural components in reactor cores. The first high dose ion irradiation experiments on a ferritic-martensitic steel showing that ion irradiation closely emulates the full radiation damage microstructure created in-reactor are described. Ferritic-martensitic alloy HT9 (heat 84425) in the form of a hexagonal fuel bundle duct (ACO-3) accumulated 155 dpa at an average temperature of 443°C in the Fast Flux Test Facility (FFTF). Using invariance theory as a guide,more » irradiation of the same heat was conducted using self-ions (Fe++) at 5 MeV at a temperature of 460°C and to a dose of 188 displacements per atom. The void swelling was nearly identical between the two irradiation and the size and density of precipitates and loops following ion irradiation are within a factor of two of those for neutron irradiation. The level of agreement across all of the principal microstructure changes between ion and reactor irradiation establishes the capability of tailoring ion irradiation to emulate the reactor-irradiated microstructure.« less

Covariance generation and uncertainty propagation for thermal and fast neutron induced fission yields

NASA Astrophysics Data System (ADS)

Terranova, Nicholas; Serot, Olivier; Archier, Pascal; De Saint Jean, Cyrille; Sumini, Marco

2017-09-01

Fission product yields (FY) are fundamental nuclear data for several applications, including decay heat, shielding, dosimetry, burn-up calculations. To be safe and sustainable, modern and future nuclear systems require accurate knowledge on reactor parameters, with reduced margins of uncertainty. Present nuclear data libraries for FY do not provide consistent and complete uncertainty information which are limited, in many cases, to only variances. In the present work we propose a methodology to evaluate covariance matrices for thermal and fast neutron induced fission yields. The semi-empirical models adopted to evaluate the JEFF-3.1.1 FY library have been used in the Generalized Least Square Method available in CONRAD (COde for Nuclear Reaction Analysis and Data assimilation) to generate covariance matrices for several fissioning systems such as the thermal fission of U235, Pu239 and Pu241 and the fast fission of U238, Pu239 and Pu240. The impact of such covariances on nuclear applications has been estimated using deterministic and Monte Carlo uncertainty propagation techniques. We studied the effects on decay heat and reactivity loss uncertainty estimation for simplified test case geometries, such as PWR and SFR pin-cells. The impact on existing nuclear reactors, such as the Jules Horowitz Reactor under construction at CEA-Cadarache, has also been considered.

Testing piezoelectric sensors in a nuclear reactor environment

NASA Astrophysics Data System (ADS)

Reinhardt, Brian T.; Suprock, Andy; Tittmann, Bernhard

2017-02-01

Several Department of Energy Office of Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development (FCRD), Advanced Reactor Concepts (ARC), Light Water Reactor Sustainability, and Next Generation Nuclear Power Plants (NGNP), are investigating new fuels, materials, and inspection paradigms for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials during irradiation. In DOE-NE's FCRD program, ultrasonic based technology was identified as a key approach that should be pursued to obtain the high-fidelity, high-accuracy data required to characterize the behavior and performance of new candidate fuels and structural materials during irradiation testing. The radiation, high temperatures, and pressure can limit the available tools and characterization methods. In this work piezoelectric transducers capable of making these measurements are developed. Specifically, three piezoelectric sensors (Bismuth Titanate, Aluminum Nitride, and Zinc Oxide) are tested in the Massachusetts Institute of Technology Research reactor to a fast neutron fluence of 8.65×1020 nf/cm2. It is demonstrated that Bismuth Titanate is capable of transduction up to 5 × 1020 nf/cm2, Zinc Oxide is capable of transduction up to at least 6.27 × 1020 nf/cm2, and Aluminum Nitride is capable of transduction up to at least 8.65 × 1020 nf/cm2.

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

An Approach for Assessing Development and Deployment Risks in the DOE Fuel Cycle Options Evaluation and Screening Study

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

Gehin, Jess C; Oakley, Brian; Worrall, Andrew

2015-01-01

Abstract One of the key objectives of the U.S. Department of Energy (DOE) Nuclear Energy R&D Roadmap is the development of sustainable nuclear fuel cycles that can improve natural resource utilization and provide solutions to the management of nuclear wastes. Recently, an evaluation and screening (E&S) of fuel cycle systems has been conducted to identify those options that provide the best opportunities for obtaining such improvements and also to identify the required research and development activities that can support the development of advanced fuel cycle options. In order to evaluate and screen the E&S study included nine criteria including Developmentmore » and Deployment Risk (D&DR). More specifically, this criterion was represented by the following metrics: Development time, development cost, deployment cost from prototypic validation to first-of-a-kind commercial, compatibility with the existing infrastructure, existence of regulations for the fuel cycle and familiarity with licensing, and existence of market incentives and/or barriers to commercial implementation of fuel cycle processes. Given the comprehensive nature of the study, a systematic approach was needed to determine metric data for the D&DR criterion, and is presented here. As would be expected, the Evaluation Group representing the once-through use of uranium in thermal reactors is always the highest ranked fuel cycle Evaluation Group for this D&DR criterion. Evaluation Groups that consist of once-through fuel cycles that use existing reactor types are consistently ranked very high. The highest ranked limited and continuous recycle fuel cycle Evaluation Groups are those that recycle Pu in thermal reactors. The lowest ranked fuel cycles are predominately continuous recycle single stage and multi-stage fuel cycles that involve TRU and/or U-233 recycle.« less

Nuclear Reactor Physics

NASA Astrophysics Data System (ADS)

Stacey, Weston M.

2001-02-01

An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

Technology Implementation Plan: Irradiation Testing and Qualification for Nuclear Thermal Propulsion Fuel

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

Harrison, Thomas J.; Howard, Richard H.; Rader, Jordan D.

This document is a notional technology implementation plan (TIP) for the development, testing, and qualification of a prototypic fuel element to support design and construction of a nuclear thermal propulsion (NTP) engine, specifically its pre-flight ground test. This TIP outlines a generic methodology for the progression from non-nuclear out-of-pile (OOP) testing through nuclear in-pile (IP) testing, at operational temperatures, flows, and specific powers, of an NTP fuel element in an existing test reactor. Subsequent post-irradiation examination (PIE) will occur in existing radiological facilities. Further, the methodology is intended to be nonspecific with respect to fuel types and irradiation or examinationmore » facilities. The goals of OOP and IP testing are to provide confidence in the operational performance of fuel system concepts and provide data to program leadership for system optimization and fuel down-selection. The test methodology, parameters, collected data, and analytical results from OOP, IP, and PIE will be documented for reference by the NTP operator and the appropriate regulatory and oversight authorities. Final full-scale integrated testing would be performed separately by the reactor operator as part of the preflight ground test.« less

Optimally moderated nuclear fission reactor and fuel source therefor

DOEpatents

Ougouag, Abderrafi M [Idaho Falls, ID; Terry, William K [Shelley, ID; Gougar, Hans D [Idaho Falls, ID

2008-07-22

An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

Code of Federal Regulations, 2010 CFR

2010-01-01

... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which the...

10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

Code of Federal Regulations, 2011 CFR

2011-01-01

... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which the...

10 CFR 2.103 - Action on applications for byproduct, source, special nuclear material, facility and operator...

Code of Federal Regulations, 2011 CFR

2011-01-01

..., Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Federal and..., Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Nuclear... of this chapter, see § 2.106(d). (b) If the Director, Office of Nuclear Reactor Regulation, Director...

10 CFR 2.103 - Action on applications for byproduct, source, special nuclear material, facility and operator...

Code of Federal Regulations, 2012 CFR

2012-01-01

..., Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Federal and..., Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Nuclear... of this chapter, see § 2.106(d). (b) If the Director, Office of Nuclear Reactor Regulation, Director...

10 CFR 2.101 - Filing of application.

Code of Federal Regulations, 2011 CFR

2011-01-01

... Reactors, the Director, Office of Nuclear Reactor Regulation, the Director, Office of Nuclear Material... Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Federal and State... be requested to: (i) Submit to the Director, Office of Nuclear Reactor Regulation, Director, Office...

10 CFR 2.101 - Filing of application.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Reactors, the Director, Office of Nuclear Reactor Regulation, the Director, Office of Nuclear Material... Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Federal and State... be requested to: (i) Submit to the Director, Office of Nuclear Reactor Regulation, Director, Office...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

10 CFR Appendix A to Part 110 - Illustrative List of Nuclear Reactor Equipment Under NRC Export Licensing Authority

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Illustrative List of Nuclear Reactor Equipment Under NRC... List of Nuclear Reactor Equipment Under NRC Export Licensing Authority Note: A nuclear reactor... core of a nuclear reactor and capable of withstanding the operating pressure of the primary coolant. (2...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems

DOE PAGES

Mahadevan, Vijay S.; Merzari, Elia; Tautges, Timothy; ...

2014-06-30

An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in ordermore » to reduce the overall numerical uncertainty while leveraging available computational resources. Finally, the coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework.« less

Enhancement of NRC station blackout requirements for nuclear power plants

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

McConnell, M. W.

2012-07-01

The U.S. Nuclear Regulatory Commission (NRC) established a Near-Term Task Force (NTTF) in response to Commission direction to conduct a systematic and methodical review of NRC processes and regulations to determine whether the agency should make additional improvements to its regulatory system and to make recommendations to the Commission for its policy direction, in light of the accident at the Fukushima Dai-ichi Nuclear Power Plant. The NTTF's review resulted in a set of recommendations that took a balanced approach to defense-in-depth as applied to low-likelihood, high-consequence events such as prolonged station blackout (SBO) resulting from severe natural phenomena. Part 50,more » Section 63, of Title 10 of the Code of Federal Regulations (CFR), 'Loss of All Alternating Current Power,' currently requires that each nuclear power plant must be able to cool the reactor core and maintain containment integrity for a specified duration of an SBO. The SBO duration and mitigation strategy for each nuclear power plant is site specific and is based on the robustness of the local transmission system and the transmission system operator's capability to restore offsite power to the nuclear power plant. With regard to SBO, the NTTF recommended that the NRC strengthen SBO mitigation capability at all operating and new reactors for design-basis and beyond-design-basis external events. The NTTF also recommended strengthening emergency preparedness for prolonged SBO and multi-unit events. These recommendations, taken together, are intended to clarify and strengthen US nuclear reactor safety regarding protection against and mitigation of the consequences of natural disasters and emergency preparedness during SBO. The focus of this paper is on the existing SBO requirements and NRC initiatives to strengthen SBO capability at all operating and new reactors to address prolonged SBO stemming from design-basis and beyond-design-basis external events. The NRC initiatives are intended to enhance core and spent fuel pool cooling, reactor coolant system integrity, and containment integrity. (authors)« less

Space reactor fuel element testing in upgraded TREAT

NASA Astrophysics Data System (ADS)

Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W. Y.

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc.; a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR); NERVA-derivative; and other concepts are discussed. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggest that full-scale PBR elements could be tested at an average energy deposition of approximately 60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of approximately 100 MW/L may be achievable.

Space reactor fuel element testing in upgraded TREAT

NASA Astrophysics Data System (ADS)

Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

1993-01-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ˜60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ˜100 MW/L may be achievable.

A facility for testing 10 to 100-kWe space power reactors

NASA Astrophysics Data System (ADS)

Carlson, William F.; Bitten, Ernest J.

1993-01-01

This paper describes an existing facility that could be used in a cost-effective manner to test space power reactors in the 10 to 100-kWe range before launch. The facility has been designed to conduct full power tests of 100-kWe SP-100 reactor systems and already has the structural features that would be required for lower power testing. The paper describes a reasonable scenario starting with the acceptance at the test site of the unfueled reactor assembly and the separately shipped nuclear fuel. After fueling the reactor and installing it in the facility, cold critical tests are performed, and the reactor is then shipped to the launch site. The availability of this facility represents a cost-effective means of performing the required prelaunch test program.

Optimize out-of-core thermionic energy conversion for nuclear electric propulsion

NASA Technical Reports Server (NTRS)

Morris, J. F.

1977-01-01

Current designs for out of core thermionic energy conversion (TEC) to power nuclear electric propulsion (NEP) were evaluated. Approaches to improve out of core TEC are emphasized and probabilities for success are indicated. TEC gains are available with higher emitter temperatures and greater power densities. Good potentialities for accommodating external high temperature, high power density TEC with heat pipe cooled reactors exist.

10 CFR Appendix A to Part 110 - Illustrative List of Nuclear Reactor Equipment Under NRC Export Licensing Authority

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Illustrative List of Nuclear Reactor Equipment Under NRC... List of Nuclear Reactor Equipment Under NRC Export Licensing Authority Note—A nuclear reactor basically... nuclear reactor and capable of withstanding the operating pressure of the primary coolant. (2) On-line (e...

10 CFR Appendix A to Part 110 - Illustrative List of Nuclear Reactor Equipment Under NRC Export Licensing Authority

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Illustrative List of Nuclear Reactor Equipment Under NRC... List of Nuclear Reactor Equipment Under NRC Export Licensing Authority Note—A nuclear reactor basically... nuclear reactor and capable of withstanding the operating pressure of the primary coolant. (2) On-line (e...

The benefits of an advanced fast reactor fuel cycle for plutonium management

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

Hannum, W.H.; McFarlane, H.F.; Wade, D.C.

1996-12-31

The United States has no program to investigate advanced nuclear fuel cycles for the large-scale consumption of plutonium from military and civilian sources. The official U.S. position has been to focus on means to bury spent nuclear fuel from civilian reactors and to achieve the spent fuel standard for excess separated plutonium, which is considered by policy makers to be an urgent international priority. Recently, the National Research Council published a long awaited report on its study of potential separation and transmutation technologies (STATS), which concluded that in the nuclear energy phase-out scenario that they evaluated, transmutation of plutonium andmore » long-lived radioisotopes would not be worth the cost. However, at the American Nuclear Society Annual Meeting in June, 1996, the STATS panelists endorsed further study of partitioning to achieve superior waste forms for burial, and suggested that any further consideration of transmutation should be in the context of energy production, not of waste management. 2048 The U.S. Department of Energy (DOE) has an active program for the short-term disposition of excess fissile material and a `focus area` for safe, secure stabilization, storage and disposition of plutonium, but has no current programs for fast reactor development. Nevertheless, sufficient data exist to identify the potential advantages of an advanced fast reactor metallic fuel cycle for the long-term management of plutonium. Advantages are discussed.« less

77 FR 8902 - Draft Regulatory Guide: Issuance, Availability Decommissioning of Nuclear Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-15

... Decommissioning of Nuclear Power Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide... draft regulatory guide (DG) DG-1271 ``Decommissioning of Nuclear Power Reactors.'' This guide describes... Regulatory Guide 1.184, ``Decommissioning of Nuclear Power Reactors,'' dated July 2000. This proposed...

10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear... requirements for immediate notification of the NRC by licensed operating nuclear power reactors are contained...

Online monitoring of the Osiris reactor with the Nucifer neutrino detector

NASA Astrophysics Data System (ADS)

Boireau, G.; Bouvet, L.; Collin, A. P.; Coulloux, G.; Cribier, M.; Deschamp, H.; Durand, V.; Fechner, M.; Fischer, V.; Gaffiot, J.; Gérard Castaing, N.; Granelli, R.; Kato, Y.; Lasserre, T.; Latron, L.; Legou, P.; Letourneau, A.; Lhuillier, D.; Mention, G.; Mueller, Th. A.; Nghiem, T.-A.; Pedrol, N.; Pelzer, J.; Pequignot, M.; Piret, Y.; Prono, G.; Scola, L.; Starzinski, P.; Vivier, M.; Dumonteil, E.; Mancusi, D.; Varignon, C.; Buck, C.; Lindner, M.; Bazoma, J.; Bouvier, S.; Bui, V. M.; Communeau, V.; Cucoanes, A.; Fallot, M.; Gautier, M.; Giot, L.; Guilloux, G.; Lenoir, M.; Martino, J.; Mercier, G.; Milleto, T.; Peuvrel, N.; Porta, A.; Le Quéré, N.; Renard, C.; Rigalleau, L. M.; Roy, D.; Vilajosana, T.; Yermia, F.; Nucifer Collaboration

2016-06-01

Originally designed as a new nuclear reactor monitoring device, the Nucifer detector has successfully detected its first neutrinos. We provide the second-shortest baseline measurement of the reactor neutrino flux. The detection of electron antineutrinos emitted in the decay chains of the fission products, combined with reactor core simulations, provides a new tool to assess both the thermal power and the fissile content of the whole nuclear core and could be used by the International Agency for Atomic Energy to enhance the safeguards of civil nuclear reactors. Deployed at only 7.2 m away from the compact Osiris research reactor core (70 MW) operating at the Saclay research center of the French Alternative Energies and Atomic Energy Commission, the experiment also exhibits a well-suited configuration to search for a new short baseline oscillation. We report the first results of the Nucifer experiment, describing the performances of the ˜0.85 m3 detector remotely operating at a shallow depth equivalent to ˜12 m of water and under intense background radiation conditions. Based on 145 (106) days of data with the reactor on (off), leading to the detection of an estimated 40760 ν¯ e , the mean number of detected antineutrinos is 281 ±7 (stat )±18 (syst )ν¯ e/day , in agreement with the prediction of 277 ±23 ν¯ e/day . Because of the large background, no conclusive results on the existence of light sterile neutrinos could be derived, however. As a first societal application we quantify how antineutrinos could be used for the Plutonium Management and Disposition Agreement.

10 CFR 2.337 - Evidence at a hearing.

Code of Federal Regulations, 2011 CFR

2011-01-01

... chapter by the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director...

10 CFR 2.337 - Evidence at a hearing.

Code of Federal Regulations, 2012 CFR

2012-01-01

... chapter by the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director...

Transportation Shock and Vibration Literature Review

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

Maheras, Steven J.; Lahti, Erik A.; Ross, Steven B.

2013-06-06

This report fulfills the M4 milestone M4FT-13OR08220112, "Report Documenting Experimental Activities." The purpose of this report is to document the results of a literature review conducted of studies related to the vibration and shock associated with the normal conditions of transport for rail shipments of used nuclear fuel from commercial light-water reactors. As discussed in Adkins (2013), the objective of this report is to determine if adequate data exist that would enable the impacts of the shock and vibration associated with the normal conditions of transport on commercial light-water reactor used nuclear fuel shipped in current generation rail transportation casksmore » to be realistically modeled.« less

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 1.43 - Office of Nuclear Reactor Regulation.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 1 2014-01-01 2014-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

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 1.43 - Office of Nuclear Reactor Regulation.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

10 CFR 1.43 - Office of Nuclear Reactor Regulation.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

10 CFR 1.43 - Office of Nuclear Reactor Regulation.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

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...

10 CFR 1.43 - Office of Nuclear Reactor Regulation.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

76 FR 14436 - University of Wisconsin, University of Wisconsin Nuclear Reactor; Notice of Issuance of...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-16

..., University of Wisconsin Nuclear Reactor; Notice of Issuance of Environmental Assessment and Finding of No... operation of the University of Wisconsin Nuclear Reactor. This action is necessary to add supplemental... of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001...

Fail-safe reactivity compensation method for a nuclear reactor

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

Nygaard, Erik T.; Angelo, Peter L.; Aase, Scott B.

The present invention relates generally to the field of compensation methods for nuclear reactors and, in particular to a method for fail-safe reactivity compensation in solution-type nuclear reactors. In one embodiment, the fail-safe reactivity compensation method of the present invention augments other control methods for a nuclear reactor. In still another embodiment, the fail-safe reactivity compensation method of the present invention permits one to control a nuclear reaction in a nuclear reactor through a method that does not rely on moving components into or out of a reactor core, nor does the method of the present invention rely on themore » constant repositioning of control rods within a nuclear reactor in order to maintain a critical state.« less

Structural materials for Gen-IV nuclear reactors: Challenges and opportunities

NASA Astrophysics Data System (ADS)

Murty, K. L.; Charit, I.

2008-12-01

Generation-IV reactor design concepts envisioned thus far cater toward a common goal of providing safer, longer lasting, proliferation-resistant and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-IV reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses and extremely corrosive environment, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This paper presents a summary of various Gen-IV reactor concepts, with emphasis on the structural materials issues depending on the specific application areas. This paper also discusses the challenges involved in using the existing materials under both service and off-normal conditions. Tasks become increasingly complex due to the operation of various fundamental phenomena like radiation-induced segregation, radiation-enhanced diffusion, precipitation, interactions between impurity elements and radiation-produced defects, swelling, helium generation and so forth. Further, high temperature capability (e.g. creep properties) of these materials is a critical, performance-limiting factor. It is demonstrated that novel alloy and microstructural design approaches coupled with new materials processing and fabrication techniques may mitigate the challenges, and the optimum system performance may be achieved under much demanding conditions.

Nuclear reactor cavity floor passive heat removal system

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

Edwards, Tyler A.; Neeley, Gary W.; Inman, James B.

A nuclear reactor includes a reactor core disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor and includes a concrete floor located underneath the nuclear reactor. An ex vessel corium retention system includes flow channels embedded in the concrete floor located underneath the nuclear reactor, an inlet in fluid communication with first ends of the flow channels, and an outlet in fluid communication with second ends of the flow channels. In some embodiments the inlet is in fluid communication with the interior of the radiological containment at a first elevation and the outlet is in fluidmore » communication with the interior of the radiological containment at a second elevation higher than the first elevation. The radiological containment may include a reactor cavity containing a lower portion of the pressure vessel, wherein the concrete floor located underneath the nuclear reactor is the reactor cavity floor.« less

PWR upper/lower internals shield

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

Homyk, W.A.

1995-03-01

During refueling of a nuclear power plant, the reactor upper internals must be removed from the reactor vessel to permit transfer of the fuel. The upper internals are stored in the flooded reactor cavity. Refueling personnel working in containment at a number of nuclear stations typically receive radiation exposure from a portion of the highly contaminated upper intervals package which extends above the normal water level of the refueling pool. This same issue exists with reactor lower internals withdrawn for inservice inspection activities. One solution to this problem is to provide adequate shielding of the unimmersed portion. The use ofmore » lead sheets or blankets for shielding of the protruding components would be time consuming and require more effort for installation since the shielding mass would need to be transported to a support structure over the refueling pool. A preferable approach is to use the existing shielding mass of the refueling pool water. A method of shielding was devised which would use a vacuum pump to draw refueling pool water into an inverted canister suspended over the upper internals to provide shielding from the normally exposed components. During the Spring 1993 refueling of Indian Point 2 (IP2), a prototype shield device was demonstrated. This shield consists of a cylindrical tank open at the bottom that is suspended over the refueling pool with I-beams. The lower lip of the tank is two feet below normal pool level. After installation, the air width of the natural shielding provided by the existing pool water. This paper describes the design, development, testing and demonstration of the prototype device.« less

TOWARD THE DEVELOPMENT OF A CONSENSUS MATERIALS DATABASE FOR PRESSURE TECHNOLGY APPLICATIONS

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

Swindeman, Robert W; Ren, Weiju

The ASME construction code books specify materials and fabrication procedures that are acceptable for pressure technology applications. However, with few exceptions, the materials properties provided in the ASME code books provide no statistics or other information pertaining to material variability. Such information is central to the prediction and prevention of failure events. Many sources of materials data exist that provide variability information but such sources do not necessarily represent a consensus of experts with respect to the reported trends that are represented. Such a need has been identified by the ASME Standards Technology, LLC and initial steps have been takenmore » to address these needs: however, these steps are limited to project-specific applications only, such as the joint DOE-ASME project on materials for Generation IV nuclear reactors. In contrast to light-water reactor technology, the experience base for the Generation IV nuclear reactors is somewhat lacking and heavy reliance must be placed on model development and predictive capability. The database for model development is being assembled and includes existing code alloys such as alloy 800H and 9Cr-1Mo-V steel. Ownership and use rights are potential barriers that must be addressed.« less

10 CFR 52.167 - Issuance of manufacturing license.

Code of Federal Regulations, 2010 CFR

2010-01-01

... proposed reactor(s) can be incorporated into a nuclear power plant and operated at sites having... design and manufacture the proposed nuclear power reactor(s); (5) The proposed inspections, tests... the construction of a nuclear power facility using the manufactured reactor(s). (2) A holder of a...

Five Lectures on Nuclear Reactors Presented at Cal Tech

DOE R&D Accomplishments Database

Weinberg, Alvin M.

1956-02-10

The basic issues involved in the physics and engineering of nuclear reactors are summarized. Topics discussed include theory of reactor design, technical problems in power reactors, physical problems in nuclear power production, and future developments in nuclear power. (C.H.)

Natural Disasters and Safety Risks at Nuclear Power Stations

NASA Astrophysics Data System (ADS)

Tutnova, T.

2012-04-01

In the aftermath of Fukushima natural-technological disaster the global opinion on nuclear energy divided even deeper. While Germany, Italy and the USA are currently reevaluating their previous plans on nuclear growth, many states are committed to expand nuclear energy output. In China and France, where the industry is widely supported by policymakers, there is little talk about abandoning further development of nuclear energy. Moreover, China displays the most remarkable pace of nuclear development in the world: it is responsible for 40% of worldwide reactors under construction, and aims at least to quadruple its nuclear capacity by 2020. In these states the consequences of Fukushima natural-technological accident will probably result in safety checks and advancement of new reactor technologies. Thus, China is buying newer reactor design from the USA which relies on "passive safety systems". It means that emergency power generators, crucial for reactor cooling in case of an accident, won't depend on electricity, so that tsunami won't disable them like it happened in the case of Fukushima. Nuclear energy managed to draw lessons from previous nuclear accidents where technological and human factors played crucial role. But the Fukushima lesson shows that the natural hazards, nevertheless, were undervalued. Though the ongoing technological advancements make it possible to increase the safety of nuclear power plants with consideration of natural risks, it is not just a question of technology improvement. A necessary action that must be taken is the reevaluation of the character and sources of the potential hazards which natural disasters can bring to nuclear industry. One of the examples is a devastating impact of more than one natural disaster happening at the same time. This subject, in fact, was not taken into account before, while it must be a significant point in planning sites for new nuclear power plants. Another important lesson unveiled is that world nuclear industry needs advanced mechanisms of international oversight. The natural-technological disaster that happens in a particular country is a matter of concern of the global community. Hence, the urgent necessity is to develop and adopt a joint mechanism for international consultation in case of serious accident at a nuclear power plant. It is also necessary to work out the list of constraining provisions for building and operating nuclear plants in regions where potential risks of natural-technological catastrophes exist. These provisions should include risk estimate for every particular region, as well as the list of preventive measures to secure the safe operation of nuclear plants located at those sites. As it was stated before, the synergy effects of more than one potential hazard must be taken into account. The main goal of my report is to represent possible methods for mitigating nuclear safety risks associated with natural hazards and technological disasters, review the effectiveness of existing standards and oversight mechanisms, encourage a cooperative discussion of these issues.

10 CFR Appendix A to Part 110 - Illustrative List of Nuclear Reactor Equipment Under NRC Export Licensing Authority

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Illustrative List of Nuclear Reactor Equipment Under NRC... List of Nuclear Reactor Equipment Under NRC Export Licensing Authority Note—A nuclear reactor basically includes the items within or attached directly to the reactor vessel, the equipment which controls the...

10 CFR Appendix A to Part 110 - Illustrative List of Nuclear Reactor Equipment Under NRC Export Licensing Authority

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Illustrative List of Nuclear Reactor Equipment Under NRC... List of Nuclear Reactor Equipment Under NRC Export Licensing Authority Note—A nuclear reactor basically includes the items within or attached directly to the reactor vessel, the equipment which controls the...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... behavior of the reactor system during a loss-of-coolant accident. Comparisons to applicable experimental...

Nuclear Data Uncertainties for Typical LWR Fuel Assemblies and a Simple Reactor Core

NASA Astrophysics Data System (ADS)

Rochman, D.; Leray, O.; Hursin, M.; Ferroukhi, H.; Vasiliev, A.; Aures, A.; Bostelmann, F.; Zwermann, W.; Cabellos, O.; Diez, C. J.; Dyrda, J.; Garcia-Herranz, N.; Castro, E.; van der Marck, S.; Sjöstrand, H.; Hernandez, A.; Fleming, M.; Sublet, J.-Ch.; Fiorito, L.

2017-01-01

The impact of the current nuclear data library covariances such as in ENDF/B-VII.1, JEFF-3.2, JENDL-4.0, SCALE and TENDL, for relevant current reactors is presented in this work. The uncertainties due to nuclear data are calculated for existing PWR and BWR fuel assemblies (with burn-up up to 40 GWd/tHM, followed by 10 years of cooling time) and for a simplified PWR full core model (without burn-up) for quantities such as k∞, macroscopic cross sections, pin power or isotope inventory. In this work, the method of propagation of uncertainties is based on random sampling of nuclear data, either from covariance files or directly from basic parameters. Additionally, possible biases on calculated quantities are investigated such as the self-shielding treatment. Different calculation schemes are used, based on CASMO, SCALE, DRAGON, MCNP or FISPACT-II, thus simulating real-life assignments for technical-support organizations. The outcome of such a study is a comparison of uncertainties with two consequences. One: although this study is not expected to lead to similar results between the involved calculation schemes, it provides an insight on what can happen when calculating uncertainties and allows to give some perspectives on the range of validity on these uncertainties. Two: it allows to dress a picture of the state of the knowledge as of today, using existing nuclear data library covariances and current methods.

Reviewing the Nuclear Nonproliferation Treaty

DTIC Science & Technology

2010-05-01

aim was to determine the critical masses of nuclear explosives by tapping two subcritical masses toward each other with a screw- driver, all the...record. Not only has the IAEA failed to find existing covert reactors and fuel-making plants, which are critical to bomb making, the agency still...sort as possible will be critical if the NPT is to re- main effective against further proliferation. Certainly, such a goal informs the present vol

Space reactor fuel element testing in upgraded TREAT

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

Todosow, M.; Bezler, P.; Ludewig, H.

1993-01-14

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. Ifmore » the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.« less

Space reactor fuel element testing in upgraded TREAT

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

Todosow, M.; Bezler, P.; Ludewig, H.

1993-05-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. Ifmore » the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.« less

10 CFR 2.101 - Filing of application.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Nuclear Reactor Regulation, the Director, Office of Nuclear Material Safety and Safeguards, or the... this chapter, see paragraph (g) of this section. (3) If the Director, Office of Nuclear Reactor...) Submit to the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director...

10 CFR 2.101 - Filing of application.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Nuclear Reactor Regulation, the Director, Office of Nuclear Material Safety and Safeguards, or the... this chapter, see paragraph (g) of this section. (3) If the Director, Office of Nuclear Reactor...) Submit to the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director...

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2013 CFR

2013-01-01

... notification of shipment of irradiated reactor fuel and nuclear waste. (a)(1) As specified in paragraphs (b... shipment of irradiated reactor fuel or nuclear waste must contain the following information: (1) The name... nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the...

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2014 CFR

2014-01-01

... notification of shipment of irradiated reactor fuel and nuclear waste. (a)(1) As specified in paragraphs (b... shipment of irradiated reactor fuel or nuclear waste must contain the following information: (1) The name... nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the...

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2011 CFR

2011-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

The alternative strategies of the development of the nuclear power industry in the 21st century

NASA Astrophysics Data System (ADS)

Goverdovskii, A. A.; Kalyakin, S. G.; Rachkov, V. I.

2014-05-01

This paper emphasizes the urgency of scientific-and-technical and sociopolitical problems of the modern nuclear power industry without solving of which the transition from local nuclear power systems now in operation to a large-scale nuclear power industry would be impossible. The existing concepts of the longterm strategy of the development of the nuclear power industry have been analyzed. On the basis of the scenarios having been developed it was shown that the most promising alternative is the orientation towards the closed nuclear fuel cycle with fast neutron reactors (hereinafter referred to as fast reactors) that would meet the requirements on the acceptable safety. It was concluded that the main provisions of "The Strategy of the Development of the Nuclear Power Industry of Russia for the First Half of the 21st Century" approved by the Government of the Russian Federation in the year 2000 remain the same at present as well, although they require to be elaborated with due regard for new realities in the market for fossil fuels, the state of both the Russian and the world economy, as well as tightening of requirements related to safe operation of nuclear power stations (NPSs) (for example, after the severe accident at the Fukushima nuclear power station, Japan) and nonproliferation of nuclear weapons.

Probabilistic Fracture Mechanics of Reactor Pressure Vessels with Populations of Flaws

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

Spencer, Benjamin; Backman, Marie; Williams, Paul

This report documents recent progress in developing a tool that uses the Grizzly and RAVEN codes to perform probabilistic fracture mechanics analyses of reactor pressure vessels in light water reactor nuclear power plants. The Grizzly code is being developed with the goal of creating a general tool that can be applied to study a variety of degradation mechanisms in nuclear power plant components. Because of the central role of the reactor pressure vessel (RPV) in a nuclear power plant, particular emphasis is being placed on developing capabilities to model fracture in embrittled RPVs to aid in the process surrounding decisionmore » making relating to life extension of existing plants. A typical RPV contains a large population of pre-existing flaws introduced during the manufacturing process. The use of probabilistic techniques is necessary to assess the likelihood of crack initiation at one or more of these flaws during a transient event. This report documents development and initial testing of a capability to perform probabilistic fracture mechanics of large populations of flaws in RPVs using reduced order models to compute fracture parameters. The work documented here builds on prior efforts to perform probabilistic analyses of a single flaw with uncertain parameters, as well as earlier work to develop deterministic capabilities to model the thermo-mechanical response of the RPV under transient events, and compute fracture mechanics parameters at locations of pre-defined flaws. The capabilities developed as part of this work provide a foundation for future work, which will develop a platform that provides the flexibility needed to consider scenarios that cannot be addressed with the tools used in current practice.« less

10 CFR 2.108 - Denial of application for failure to supply information.

Code of Federal Regulations, 2012 CFR

2012-01-01

... supply information. (a) The Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and Safeguards, as appropriate, may deny an... of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear...

10 CFR 2.108 - Denial of application for failure to supply information.

Code of Federal Regulations, 2011 CFR

2011-01-01

... supply information. (a) The Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and Safeguards, as appropriate, may deny an... of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear...

Analysis of key safety metrics of thorium utilization in LWRs

DOE PAGES

Ade, Brian J.; Bowman, Stephen M.; Worrall, Andrew; ...

2016-04-08

Here, thorium has great potential to stretch nuclear fuel reserves because of its natural abundance and because it is possible to breed the 232Th isotope into a fissile fuel ( 233U). Various scenarios exist for utilization of thorium in the nuclear fuel cycle, including use in different nuclear reactor types (e.g., light water, high-temperature gas-cooled, fast spectrum sodium, and molten salt reactors), along with use in advanced accelerator-driven systems and even in fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based onmore » concepts that mix thorium with uranium (UO 2 + ThO 2) or that add fertile thorium (ThO 2) fuel pins to typical LWR fuel assemblies. Utilization of mixed fuel assemblies (PuO 2 + ThO 2) is also possible. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts to the nuclear fuel. Thorium and its irradiation products have different nuclear characteristics from those of uranium and its irradiation products. ThO 2, alone or mixed with UO 2 fuel, leads to different chemical and physical properties of the fuel. These key reactor safety–related issues have been studied at Oak Ridge National Laboratory and documented in “Safety and Regulatory Issues of the Thorium Fuel Cycle” (NUREG/CR-7176, U.S. Nuclear Regulatory Commission, 2014). Various reactor analyses were performed using the SCALE code system for comparison of key performance parameters of both ThO 2 + UO 2 and ThO 2 + PuO 2 against those of UO 2 and typical UO 2 + PuO 2 mixed oxide fuels, including reactivity coefficients and power sharing between surrounding UO 2 assemblies and the assembly of interest. The decay heat and radiological source terms for spent fuel after its discharge from the reactor are also presented. Based on this evaluation, potential impacts on safety requirements and identification of knowledge gaps that require additional analysis or research to develop a technical basis for the licensing of thorium fuel are identified.« less

10 CFR 2.1105 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Nuclear Fuel Storage Capacity at Civilian Nuclear Power Reactors § 2.1105 Definitions. As used in this part: (a) Civilian nuclear power reactor means a civilian nuclear power plant required to be licensed... nuclear fuel means fuel that has been withdrawn from a nuclear reactor following irradiation, the...

10 CFR 2.1105 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Nuclear Fuel Storage Capacity at Civilian Nuclear Power Reactors § 2.1105 Definitions. As used in this part: (a) Civilian nuclear power reactor means a civilian nuclear power plant required to be licensed... nuclear fuel means fuel that has been withdrawn from a nuclear reactor following irradiation, the...

Modeling ionization and recombination from low energy nuclear recoils in liquid argon

DOE PAGES

Foxe, M.; Hagmann, C.; Jovanovic, I.; ...

2015-03-27

Coherent elastic neutrino-nucleus scattering (CENNS) is an as-yet undetected, flavor-independent neutrino interaction predicted by the Standard Model. Detection of CENNS could offer benefits for detection of supernova and solar neutrinos in astrophysics, or for detection of antineutrinos for nuclear reactor monitoring and nuclear nonproliferation. One challenge with detecting CENNS is the low energy deposition associated with a typical CENNS nuclear recoil. In addition, nuclear recoils result in lower ionization yields than those produced by electron recoils of the same energy. While a measurement of the nuclear recoil ionization yield in liquid argon in the keV energy range has been recentlymore » reported, a corresponding model for low-energy ionization yield in liquid argon does not exist. For this reason, a Monte Carlo simulation has been developed to predict the ionization yield at sub-10 keV energies. The model consists of two distinct components: (1) simulation of the atomic collision cascade with production of ionization, and (2) the thermalization and drift of ionization electrons in an applied electric field including local recombination. As an application of our results we report updated estimates of detectable ionization in liquid argon from CENNS at a nuclear reactor.« less

Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors

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

M. L. Grossbeck J-P.A. Renier Tim Bigelow

2003-09-30

Burnable poisons are used in nuclear reactors to produce a more level distribution of power in the reactor core and to reduce to necessity for a large control system. An ideal burnable poison would burn at the same rate as the fuel. In this study, separation of neutron-absorbing isotopes was investigated in order to eliminate isotopes that remain as absorbers at the end of fuel life, thus reducing useful fuel life. The isotopes Gd-157, Dy-164, and Er-167 were found to have desirable properties. These isotopes were separated from naturally occurring elements by means of plasma separation to evaluate feasibility andmore » cost. It was found that pure Gd-157 could save approximately $6 million at the end of four years. However, the cost of separation, using the existing facility, made separation cost- ineffective. Using a magnet with three times the field strength is expected to reduce the cost by a factor of ten, making isotopically separated burnable poisons a favorable method of increasing fuel life in commercial reactors, in particular Generation-IV reactors. The project also investigated various burnable poison configurations, and studied incorporation of metallic burnable poisons into fuel cladding.« less

NUCLEAR REACTOR CONTROL SYSTEM

DOEpatents

Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

1959-11-01

A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

10 CFR 140.72 - Indemnity agreements.

Code of Federal Regulations, 2014 CFR

2014-01-01

... (issued pursuant to part 50 of this chapter) authorizing the licensee to operate the nuclear reactor... the licensee to possess and store special nuclear material at the site of the nuclear reactor for use as fuel in operation of the nuclear reactor after issuance of an operating license for the reactor...

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

10 CFR 140.72 - Indemnity agreements.

Code of Federal Regulations, 2010 CFR

2010-01-01

... (issued pursuant to part 50 of this chapter) authorizing the licensee to operate the nuclear reactor... the licensee to possess and store special nuclear material at the site of the nuclear reactor for use as fuel in operation of the nuclear reactor after issuance of an operating license for the reactor...

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

10 CFR 2.621 - Acceptance and docketing of application for early review of site suitability issues in a combined...

Code of Federal Regulations, 2014 CFR

2014-01-01

... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will inform the... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will accept for... New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, that they...

10 CFR 2.621 - Acceptance and docketing of application for early review of site suitability issues in a combined...

Code of Federal Regulations, 2013 CFR

2013-01-01

... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will inform the... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will accept for... New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, that they...

The world's nuclear future - built on material success

NASA Astrophysics Data System (ADS)

Ion, Sue

2010-07-01

In our energy hungry world of the twenty-first century, the future of electricity generation must meet the twin challenges of security of supply and reduced carbon emissions. The expectations for nuclear power programmes to play a part in delivering success on both counts, grows ever higher. The nuclear industry is poised on a renaissance likely to dwarf the heady days of the 1960s and early 1970s. Global supply chain and project management challenges abound, now just as then. The science and engineering of materials will be key to the successful deployment and operation of a new generation of reactor systems and their associated fuel cycles. Understanding and predicting materials performance will be key to achieving life extension of existing assets and underpinning waste disposal options, as well as giving confidence to the designers, their financial backers and governments across the globe, that the next generation of reactors will deliver their full potential.

A cermet fuel reactor for nuclear thermal propulsion

NASA Technical Reports Server (NTRS)

Kruger, Gordon

1991-01-01

Work on the cermet fuel reactor done in the 1960's by General Electric (GE) and the Argonne National Laboratory (ANL) that had as its goal the development of systems that could be used for nuclear rocket propulsion as well as closed cycle propulsion system designs for ship propulsion, space nuclear propulsion, and other propulsion systems is reviewed. It is concluded that the work done in the 1960's has demonstrated that we can have excellent thermal and mechanical performance with cermet fuel. Thousands of hours of testing were performed on the cermet fuel at both GE and AGL, including very rapid transients and some radiation performance history. We conclude that there are no feasibility issues with cermet fuel. What is needed is reactivation of existing technology and qualification testing of a specific fuel form. We believe this can be done with a minimum development risk.

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor [Novel Methodology and Software for Spent Fuel Gross Defect Detection at the Atucha-I Reactor

DOE PAGES

Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek; ...

2017-03-27

Here, fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at themore » site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup.« less

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor [Novel Methodology and Software for Spent Fuel Gross Defect Detection at the Atucha-I Reactor

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

Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek

Here, fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at themore » site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup.« less

The physics design of accelerator-driven transmutation systems

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

Venneri, F.

1995-10-01

Nuclear systems under study in the Los Alamos Accelerator-Driven Transmutation Technology program (ADTT) will allow the destruction of nuclear spent fuel and weapons-return plutonium, as well as the production of nuclear energy from the thorium cycle, without a long-lived radioactive waste stream. The subcritical systems proposed represent a radical departure from traditional nuclear concepts (reactors), yet the actual implementation of ADTT systems is based on modest extrapolations of existing technology. These systems strive to keep the best that the nuclear technology has developed over the years, within a sensible conservative design envelope and eventually manage to offer a safe, lessmore » expensive and more environmentally sound approach to nuclear power.« less

75 FR 12311 - FirstEnergy Nuclear Operating Company; Notice of Consideration of Issuance of Amendment to...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-15

... INFORMATION CONTACT: Michael Mahoney, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission... Licensing Branch III-2, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR... NUCLEAR REGULATORY COMMISSION [Docket No. 50-346] FirstEnergy Nuclear Operating Company; Notice of...

10 CFR 50.30 - Filing of application; oath or affirmation.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and... Director, Office of New Reactors, or the Director, Office of Nuclear Reactor Regulation, or the Director..., operating license, early site permit, combined license, or manufacturing license for a nuclear power reactor...

10 CFR 50.30 - Filing of application; oath or affirmation.

Code of Federal Regulations, 2011 CFR

2011-01-01

... Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and... Director, Office of New Reactors, or the Director, Office of Nuclear Reactor Regulation, or the Director..., operating license, early site permit, combined license, or manufacturing license for a nuclear power reactor...

10 CFR 50.30 - Filing of application; oath or affirmation.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and... Director, Office of New Reactors, or the Director, Office of Nuclear Reactor Regulation, or the Director..., operating license, early site permit, combined license, or manufacturing license for a nuclear power reactor...

10 CFR 50.30 - Filing of application; oath or affirmation.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and... Director, Office of New Reactors, or the Director, Office of Nuclear Reactor Regulation, or the Director..., operating license, early site permit, combined license, or manufacturing license for a nuclear power reactor...

On-line condition monitoring applications in nuclear power plants

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

Hastiemian, H. M.; Feltus, M. A.

2006-07-01

Existing signals from process instruments in nuclear power plants can be sampled while the plant is operating and analyzed to verify the static and dynamic performance of process sensors, identify process-to-sensor problems, detect instrument anomalies such as venturi fouling, measure the vibration of the reactor vessel and its internals, or detect thermal hydraulic anomalies within the reactor coolant system. These applications are important in nuclear plants to satisfy a variety of objectives such as: 1) meeting the plant technical specification requirements; 2) complying with regulatory regulations; 3) guarding against equipment and process degradation; 4) providing a means for incipient failuremore » detection and predictive maintenance; or 5) identifying the root cause of anomalies in equipment and plant processes. The technologies that are used to achieve these objectives are collectively referred to as 'on-line condition monitoring.' This paper presents a review of key elements of these technologies, provides examples of their use in nuclear power plants, and illustrates how they can be integrated into an on-line condition monitoring system for nuclear power plants. (authors)« less

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.

10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors..., including expected occurrences, as low as is reasonably achievable, each licensee of a nuclear power reactor...

10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear power reactor licensee licensed under §§ 50.21(b) or 50.22 holding an operating license under this part...

10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors..., including expected occurrences, as low as is reasonably achievable, each licensee of a nuclear power reactor...

10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear power reactor licensee licensed under §§ 50.21(b) or 50.22 holding an operating license under this part...

10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors..., including expected occurrences, as low as is reasonably achievable, each licensee of a nuclear power reactor...

Tower of Babel: a special report of the nuclear industry

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

Not Available

The southern U.S. region currently maintains 19 operating nuclear reactors, a large number of nuclear-related industries, and numerous radioactive waste storage facilities. To illustrate the greed of nuclear power proponents and the dangers of existing and future nuclear power plant operations, the southern nuclear power industry is surveyed. Detailed are the South's involvement in each phase of the nuclear fuel cycle, from uranium mining to waste disposal; efforts by the region's private electric utility companies to buttress the crumbling supports of the nuclear industry; and the serious threat that nuclear power poses to the region, the nation, and the world.more » The U.S. nuclear power industry can be viewed as a modern Tower of Babel. (4 maps, 20 photos, 2 tables)« less

78 FR 64028 - Decommissioning of Nuclear Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-25

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0035] Decommissioning of Nuclear Power Reactors AGENCY... Commission (NRC) is issuing Revision 1 of regulatory guide (RG) 1.184 ``Decommissioning of Nuclear Power... the NRC's regulations relating to the decommissioning process for nuclear power reactors. The revision...

A Proof of Concept: Grizzly, the LWRS Program Materials Aging and Degradation Pathway Main Simulation Tool

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

Ben Spencer; Jeremey Busby; Richard Martineau

2012-10-01

Nuclear power currently provides a significant fraction of the United States’ non-carbon emitting power generation. In future years, nuclear power must continue to generate a significant portion of the nation’s electricity to meet the growing electricity demand, clean energy goals, and ensure energy independence. New reactors will be an essential part of the expansion of nuclear power. However, given limits on new builds imposed by economics and industrial capacity, the extended service of the existing fleet will also be required.

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

10 CFR 2.603 - Acceptance and docketing of application for early review of site suitability issues in a...

Code of Federal Regulations, 2013 CFR

2013-01-01

... Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will... Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, that they are...

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

10 CFR 2.603 - Acceptance and docketing of application for early review of site suitability issues in a...

Code of Federal Regulations, 2014 CFR

2014-01-01

... Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will... Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, will... Reactors or the Director of the Office of Nuclear Reactor Regulation, as appropriate, that they are...

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

Improvements to Nuclear Data and Its Uncertainties by Theoretical Modeling

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

Danon, Yaron; Nazarewicz, Witold; Talou, Patrick

2013-02-18

This project addresses three important gaps in existing evaluated nuclear data libraries that represent a significant hindrance against highly advanced modeling and simulation capabilities for the Advanced Fuel Cycle Initiative (AFCI). This project will: Develop advanced theoretical tools to compute prompt fission neutrons and gamma-ray characteristics well beyond average spectra and multiplicity, and produce new evaluated files of U and Pu isotopes, along with some minor actinides; Perform state-of-the-art fission cross-section modeling and calculations using global and microscopic model input parameters, leading to truly predictive fission cross-sections capabilities. Consistent calculations for a suite of Pu isotopes will be performed; Implementmore » innovative data assimilation tools, which will reflect the nuclear data evaluation process much more accurately, and lead to a new generation of uncertainty quantification files. New covariance matrices will be obtained for Pu isotopes and compared to existing ones. The deployment of a fleet of safe and efficient advanced reactors that minimize radiotoxic waste and are proliferation-resistant is a clear and ambitious goal of AFCI. While in the past the design, construction and operation of a reactor were supported through empirical trials, this new phase in nuclear energy production is expected to rely heavily on advanced modeling and simulation capabilities. To be truly successful, a program for advanced simulations of innovative reactors will have to develop advanced multi-physics capabilities, to be run on massively parallel super- computers, and to incorporate adequate and precise underlying physics. And all these areas have to be developed simultaneously to achieve those ambitious goals. Of particular interest are reliable fission cross-section uncertainty estimates (including important correlations) and evaluations of prompt fission neutrons and gamma-ray spectra and uncertainties.« less

Investigations on optimization of accident management measures following a station blackout accident in a VVER-1000 pressurized water reactor

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

Tusheva, P.; Schaefer, F.; Kliem, S.

2012-07-01

The reactor safety issues are of primary importance for preserving the health of the population and ensuring no release of radioactivity and fission products into the environment. A part of the nuclear research focuses on improvement of the safety of existing nuclear power plants. Studies, research and efforts are a continuing process at improving the safety and reliability of existing and newly developed nuclear power plants at prevention of a core melt accident. Station blackout (loss of AC power supply) is one of the dominant accidents taken into consideration at performing accident analysis. In case of multiple failures of safetymore » systems it leads to a severe accident. To prevent an accident to turn into a severe one or to mitigate the consequences, accident management measures must be performed. The present paper outlines possibilities for application and optimization of accident management measures following a station blackout accident. Assessed is the behaviour of the nuclear power plant during a station blackout accident without accident management measures and with application of primary/secondary side oriented accident management measures. Discussed are the possibilities for operators ' intervention and the influence of the performed accident management measures on the course of the accident. Special attention has been paid to the effectiveness of the passive feeding and physical phenomena having an influence on the system behaviour. The performed simulations show that the effectiveness of the secondary side feeding procedure can be limited due to an early evaporation or flashing effects in the feed water system. The analyzed cases show that the effectiveness of the accident management measures strongly depends on the initiation criteria applied for depressurization of the reactor coolant system. (authors)« 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

Application of point kinetics equations to the design of a reactivity meter

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

Binney, S.E.; Bakir, A.J.M.

1988-01-01

The time-dependent reactivity of a nuclear reactor is obviously one of the most important reactor parameters that describes the state of the reactor. Although several different types of techniques exist to measure reactivity, only the kinetic method is described here. The paper illustrates the measured reactor power and calculated reactivity for a 70 cents step change in reactivity. These data were taken at 1-s time intervals. It is seen that the reactivity, initially at zero, rises rapidly to a predetermined value (determined by the reactivity change induced in the system) and then returns to zero as the reactor is reestablishedmore » in a critical situation by insertion of another control rod. It is concluded that the method of Tuttle has been adapted to produce a reliable, on-line calculation of reactivity from a time-dependent reactor power signal.« less

76 FR 11521 - Prairie Island Nuclear Generating Plant, Unit 1, Northern States Power Company-Minnesota; Notice...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-02

..., Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001..., Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2011-4557 Filed 3-1... NUCLEAR REGULATORY COMMISSION [Docket No. 50-282; NRC-2011-0040] Prairie Island Nuclear Generating...

10 CFR 140.12 - Amount of financial protection required for other reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000... chapter to operate two or more nuclear reactors at the same location, the total financial protection...

10 CFR 140.12 - Amount of financial protection required for other reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000... chapter to operate two or more nuclear reactors at the same location, the total financial protection...

10 CFR 140.12 - Amount of financial protection required for other reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000... chapter to operate two or more nuclear reactors at the same location, the total financial protection...

10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... domestic non-power reactors. 50.64 Section 50.64 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... Director of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC... Director of the Office of Nuclear Reactor Regulation a written proposal for meeting the requirements of...

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2012 CFR

2012-01-01

... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... of the shipper, carrier, and receiver of the irradiated reactor fuel or nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the shipment, as specified in the...

10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... domestic non-power reactors. 50.64 Section 50.64 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... Director of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC... Director of the Office of Nuclear Reactor Regulation a written proposal for meeting the requirements of...

10 CFR 140.94 - Appendix D-Form of indemnity agreement with Federal agencies.

Code of Federal Regulations, 2011 CFR

2011-01-01

... (hereinafter referred to as the Act). Article I As used in this agreement, 1. Nuclear reactor, byproduct... irradiated or to be irradiated by, the nuclear reactor or reactors subject to the license or licenses... construction of a nuclear reactor with respect to which no operating license has been issued by the Nuclear...

10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... domestic non-power reactors. 50.64 Section 50.64 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... Director of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC... Director of the Office of Nuclear Reactor Regulation a written proposal for meeting the requirements of...

10 CFR 2.102 - Administrative review of application.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of...) The Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office... Energy NUCLEAR REGULATORY COMMISSION AGENCY RULES OF PRACTICE AND PROCEDURE Procedure for Issuance...

10 CFR 2.102 - Administrative review of application.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of...) The Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office... Energy NUCLEAR REGULATORY COMMISSION AGENCY RULES OF PRACTICE AND PROCEDURE Procedure for Issuance...

Cladding and duct materials for advanced nuclear recycle reactors

NASA Astrophysics Data System (ADS)

Allen, T. R.; Busby, J. T.; Klueh, R. L.; Maloy, S. A.; Toloczko, M. B.

2008-01-01

The expanded use of nuclear energy without risk of nuclear weapons proliferation and with safe nuclear waste disposal is a primary goal of the Global Nuclear Energy Partnership (GNEP). To achieve that goal the GNEP is exploring advanced technologies for recycling spent nuclear fuel that do not separate pure plutonium, and advanced reactors that consume transuranic elements from recycled spent fuel. The GNEP’s objectives will place high demands on reactor clad and structural materials. This article discusses the materials requirements of the GNEP’s advanced nuclear recycle reactors program.

Tory II-A: a nuclear ramjet test reactor

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

Hadley, J.W.

Declassified 28 Nov 1973. The first test reactor in the Pluto program, leading to development of a nuclear ramjet engine, is called Tory II-A. While it is not an actual prototype engine, this reactor embodies a core design which is considered feasible for an engine, and operation of the reactor will provide a test of that core type as well as more generalized values in reactor design and testing. The design of Tory II-A and construction of the reactor and of its test facility are described. Operation of the Tory II-A core at a total power of 160 megawatts, withmore » 800 pounds of air per second passing through the core and emerging at a temperature of 2000 deg F, is the central objective of the test program. All other reactor and facility components exist to support operation of the core, and preliminary steps in the test program itself will be directed primarily toward ensuring attalnment of full-power operation and collection of meaningful data on core behavior during that operation. The core, 3 feet in diameter and 41/2 feet long, will be composed of bundled ceramic tubes whose central holes will provide continuous air passages from end to end of the reactor. These tubes are to be composed of a homogeneous mixture of UO/sub 2/ fuel and BeO moderator, compacted and sintered to achieve high strength and density. (30 references) (auth)« less

Laboratory instrumentation modernization at the WPI Nuclear Reactor Facility

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

Not Available

1995-01-01

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Program several laboratory instruments utilized by students and researchers at the WPI Nuclear Reactor Facility have been upgraded or replaced. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduate use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The low power output of the reactor and an ergonomicmore » facility design make it an ideal tool for undergraduate nuclear engineering education and other training. The reactor, its control system, and the associate laboratory equipment are all located in the same room. Over the years, several important milestones have taken place at the WPI reactor. In 1969, the reactor power level was upgraded from 1 kW to 10 kW. The reactor`s Nuclear Regulatory Commission operating license was renewed for 20 years in 1983. In 1988, under DOE Grant No. DE-FG07-86ER75271, the reactor was converted to low-enriched uranium fuel. In 1992, again with partial funding from DOE (Grant No. DE-FG02-90ER12982), the original control console was replaced.« less

Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

PubMed

Hill, R N; Nutt, W M; Laidler, J J

2011-01-01

The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described. Copyright © 2010 Health Physics Society

76 FR 74630 - Making Changes to Emergency Plans for Nuclear Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-01

... NUCLEAR REGULATORY COMMISSION 10 CFR Parts 50 and 52 RIN 3150-AI10 [NRC-2008-0122] Making Changes to Emergency Plans for Nuclear Power Reactors AGENCY: Nuclear Regulatory Commission. ACTION... guide (RG) 1.219, ``Guidance on Making Changes to Emergency Plans for Nuclear Power Reactors.'' This...

10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Immediate notification requirements for operating nuclear power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear...

10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Immediate notification requirements for operating nuclear power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear...

Benefits of Advanced Control Room Technologies: Phase One Upgrades to the HSSL, Research Plan, and Performance Measures

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

Le Blanc, Katya; Joe, Jeffrey; Rice, Brandon

Control Room modernization is an important part of life extension for the existing light water reactor fleet. None of the 99 currently operating commercial nuclear power plants in the U.S. has completed a full-scale control room modernization to date. A full-scale modernization might, for example, entail replacement of all analog panels with digital workstations. Such modernizations have been undertaken successfully in upgrades in Europe and Asia, but the U.S. has yet to undertake a control room upgrade of this magnitude. Instead, nuclear power plant main control rooms for the existing commercial reactor fleet remain significantly analog, with only limited digitalmore » modernizations. Previous research under the U.S. Department of Energy’s Light Water Reactor Sustainability Program has helped establish a systematic process for control room upgrades that support the transition to a hybrid control room. While the guidance developed to date helps streamline the process of modernization and reduce costs and uncertainty associated with introducing digital control technologies into an existing control room, these upgrades do not achieve the full potential of newer technologies that might otherwise enhance plant and operator performance. The aim of the control room benefits research is to identify previously overlooked benefits of modernization, identify candidate technologies that may facilitate such benefits, and demonstrate these technologies through human factors research. This report describes the initial upgrades to the HSSL and outlines the methodology for a pilot test of the HSSL configuration.« less

Basic Nuclear Physics.

ERIC Educational Resources Information Center

Bureau of Naval Personnel, Washington, DC.

Basic concepts of nuclear structures, radiation, nuclear reactions, and health physics are presented in this text, prepared for naval officers. Applications to the area of nuclear power are described in connection with pressurized water reactors, experimental boiling water reactors, homogeneous reactor experiments, and experimental breeder…

Integral isolation valve systems for loss of coolant accident protection

DOEpatents

Kanuch, David J.; DiFilipo, Paul P.

2018-03-20

A nuclear reactor includes a nuclear reactor core comprising fissile material disposed in a reactor pressure vessel having vessel penetrations that exclusively carry flow into the nuclear reactor and at least one vessel penetration that carries flow out of the nuclear reactor. An integral isolation valve (IIV) system includes passive IIVs each comprising a check valve built into a forged flange and not including an actuator, and one or more active IIVs each comprising an active valve built into a forged flange and including an actuator. Each vessel penetration exclusively carrying flow into the nuclear reactor is protected by a passive IIV whose forged flange is directly connected to the vessel penetration. Each vessel penetration carrying flow out of the nuclear reactor is protected by an active IIV whose forged flange is directly connected to the vessel penetration. Each active valve may be a normally closed valve.

Propellant actuated nuclear reactor steam depressurization valve

DOEpatents

Ehrke, Alan C.; Knepp, John B.; Skoda, George I.

1992-01-01

A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

The United States Naval Nuclear Propulsion Program - Over 151 Million Miles Safely Steamed on Nuclear Power

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

None, None

NNSA’s third mission pillar is supporting the U.S. Navy’s ability to protect and defend American interests across the globe. The Naval Reactors Program remains at the forefront of technological developments in naval nuclear propulsion and ensures a commanding edge in warfighting capabilities by advancing new technologies and improvements in naval reactor performance and reliability. In 2015, the Naval Nuclear Propulsion Program pioneered advances in nuclear reactor and warship design – such as increasing reactor lifetimes, improving submarine operational effectiveness, and reducing propulsion plant crewing. The Naval Reactors Program continued its record of operational excellence by providing the technical expertise requiredmore » to resolve emergent issues in the Nation’s nuclear-powered fleet, enabling the Fleet to safely steam more than two million miles. Naval Reactors safely maintains, operates, and oversees the reactors on the Navy’s 82 nuclear-powered warships, constituting more than 45 percent of the Navy’s major combatants.« less

76 FR 64126 - Advisory Committee on Reactor Safeguards; Procedures for Meetings

Federal Register 2010, 2011, 2012, 2013, 2014

2011-10-17

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Procedures for Meetings.... Nuclear Regulatory Commission's (NRC's) Advisory Committee on Reactor Safeguards (ACRS) pursuant to the... specified in the Federal Register Notice, care of the Advisory Committee on Reactor Safeguards, U.S. Nuclear...

78 FR 35056 - Effectiveness of the Reactor Oversight Process Baseline Inspection Program

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-11

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0125] Effectiveness of the Reactor Oversight Process... the effectiveness of the reactor oversight process (ROP) baseline inspection program with members of... Nuclear Reactor Regulations, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301...

78 FR 67205 - Advisory Committee on Reactor Safeguards; Procedures for Meetings

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-08

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Procedures for Meetings.... Nuclear Regulatory Commission's (NRC's) Advisory Committee on Reactor Safeguards (ACRS) pursuant to the... specified in the Federal Register Notice, care of the Advisory Committee on Reactor Safeguards, U.S. Nuclear...

5 CFR 5801.102 - Prohibited securities.

Code of Federal Regulations, 2014 CFR

2014-01-01

... licenses for facilities which generate electric energy by means of a nuclear reactor; (2) State or local... reactor or a low-level waste facility; (3) Entities manufacturing or selling nuclear power or test reactors; (4) Architectural-engineering companies providing services relating to a nuclear power reactor...

5 CFR 5801.102 - Prohibited securities.

Code of Federal Regulations, 2010 CFR

2010-01-01

... licenses for facilities which generate electric energy by means of a nuclear reactor; (2) State or local... reactor or a low-level waste facility; (3) Entities manufacturing or selling nuclear power or test reactors; (4) Architectural-engineering companies providing services relating to a nuclear power reactor...

Inside the Black Triangle.

ERIC Educational Resources Information Center

Clamp, Alice

1991-01-01

The energy policies of four emerging democratic governments in Eastern Europe are individually profiled with respect to the challenge of producing more electrical energy, while creating less pollution and cleaning up the existing environmental disorder. Highlighted is the possible change from the burning of lignite coal to use of nuclear reactors.…

Nuclear reactor neutron shielding

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

Speaker, Daniel P; Neeley, Gary W; Inman, James B

A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactormore » cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.« less

Reactor pressure vessel head vents and methods of using the same

DOEpatents

Gels, John L; Keck, David J; Deaver, Gerald A

2014-10-28

Internal head vents are usable in nuclear reactors and include piping inside of the reactor pressure vessel with a vent in the reactor upper head. Piping extends downward from the upper head and passes outside of the reactor to permit the gas to escape or be forcibly vented outside of the reactor without external piping on the upper head. The piping may include upper and lowers section that removably mate where the upper head joins to the reactor pressure vessel. The removable mating may include a compressible bellows and corresponding funnel. The piping is fabricated of nuclear-reactor-safe materials, including carbon steel, stainless steel, and/or a Ni--Cr--Fe alloy. Methods install an internal head vent in a nuclear reactor by securing piping to an internal surface of an upper head of the nuclear reactor and/or securing piping to an internal surface of a reactor pressure vessel.

Flow Induced Vibration Program at Argonne National Laboratory

NASA Astrophysics Data System (ADS)

1984-01-01

The Argonne National Laboratory's Flow Induced Vibration Program, currently residing in the Laboratory's Components Technology Division is discussed. Throughout its existence, the overall objective of the program was to develop and apply new and/or improved methods of analysis and testing for the design evaluation of nuclear reactor plant components and heat exchange equipment from the standpoint of flow induced vibration. Historically, the majority of the program activities were funded by the US Atomic Energy Commission, the Energy Research and Development Administration, and the Department of Energy. Current DOE funding is from the Breeder Mechanical Component Development Division, Office of Breeder Technology Projects; Energy Conversion and Utilization Technology Program, Office of Energy Systems Research; and Division of Engineering, Mathematical and Geosciences, office of Basic Energy Sciences. Testing of Clinch River Breeder Reactor upper plenum components was funded by the Clinch River Breeder Reactor Plant Project Office. Work was also performed under contract with Foster Wheeler, General Electric, Duke Power Company, US Nuclear Regulatory Commission, and Westinghouse.

Non-equilibrium radiation nuclear reactor

NASA Technical Reports Server (NTRS)

Thom, K.; Schneider, R. T. (Inventor)

1978-01-01

An externally moderated thermal nuclear reactor is disclosed which is designed to provide output power in the form of electromagnetic radiation. The reactor is a gaseous fueled nuclear cavity reactor device which can operate over wide ranges of temperature and pressure, and which includes the capability of processing and recycling waste products such as long-lived transuranium actinides. The primary output of the device may be in the form of coherent radiation, so that the reactor may be utilized as a self-critical nuclear pumped laser.

10 CFR 110.26 - General license for the export of nuclear reactor components.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false General license for the export of nuclear reactor components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear reactor...

10 CFR 110.26 - General license for the export of nuclear reactor components.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false General license for the export of nuclear reactor components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear reactor...

10 CFR 110.26 - General license for the export of nuclear reactor components.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false General license for the export of nuclear reactor components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear reactor...

10 CFR 110.26 - General license for the export of nuclear reactor components.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false General license for the export of nuclear reactor components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear reactor...

10 CFR 110.26 - General license for the export of nuclear reactor components.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false General license for the export of nuclear reactor components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear reactor...

10 CFR 2.103 - Action on applications for byproduct, source, special nuclear material, facility and operator...

Code of Federal Regulations, 2013 CFR

2013-01-01

... nuclear material, facility and operator licenses. (a) If the Director, Office of Nuclear Reactor... repository operations area under parts 60 or 63 of this chapter, the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Nuclear Material Safety and Safeguards, or...

10 CFR 2.103 - Action on applications for byproduct, source, special nuclear material, facility and operator...

Code of Federal Regulations, 2014 CFR

2014-01-01

... nuclear material, facility and operator licenses. (a) If the Director, Office of Nuclear Reactor... repository operations area under parts 60 or 63 of this chapter, the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Nuclear Material Safety and Safeguards, or...

76 FR 68514 - Request for a License To Export Reactor Components

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-04

... NUCLEAR REGULATORY COMMISSION Request for a License To Export Reactor Components Pursuant to 10.../docket Number Westinghouse Electric Company Complete reactor 12 Perform seismic China. LLC, August 18... qualification equipment. of AP1000 (design) nuclear reactors. For the Nuclear Regulatory Commission. Dated this...

10 CFR 50.36 - Technical specifications.

Code of Federal Regulations, 2013 CFR

2013-01-01

... nuclear reactors are limits upon important process variables that are found to be necessary to reasonably... Commission terminates the license for the reactor, except for nuclear power reactors licensed under § 50.21(b... for nuclear reactors are settings for automatic protective devices related to those variables having...

10 CFR 50.36 - Technical specifications.

Code of Federal Regulations, 2014 CFR

2014-01-01

... nuclear reactors are limits upon important process variables that are found to be necessary to reasonably... Commission terminates the license for the reactor, except for nuclear power reactors licensed under § 50.21(b... for nuclear reactors are settings for automatic protective devices related to those variables having...

10 CFR 50.36 - Technical specifications.

Code of Federal Regulations, 2012 CFR

2012-01-01

... nuclear reactors are limits upon important process variables that are found to be necessary to reasonably... Commission terminates the license for the reactor, except for nuclear power reactors licensed under § 50.21(b... for nuclear reactors are settings for automatic protective devices related to those variables having...

Technical Application of Nuclear Fission

NASA Astrophysics Data System (ADS)

Denschlag, J. O.

The chapter is devoted to the practical application of the fission process, mainly in nuclear reactors. After a historical discussion covering the natural reactors at Oklo and the first attempts to build artificial reactors, the fundamental principles of chain reactions are discussed. In this context chain reactions with fast and thermal neutrons are covered as well as the process of neutron moderation. Criticality concepts (fission factor η, criticality factor k) are discussed as well as reactor kinetics and the role of delayed neutrons. Examples of specific nuclear reactor types are presented briefly: research reactors (TRIGA and ILL High Flux Reactor), and some reactor types used to drive nuclear power stations (pressurized water reactor [PWR], boiling water reactor [BWR], Reaktor Bolshoi Moshchnosti Kanalny [RBMK], fast breeder reactor [FBR]). The new concept of the accelerator-driven systems (ADS) is presented. The principle of fission weapons is outlined. Finally, the nuclear fuel cycle is briefly covered from mining, chemical isolation of the fuel and preparation of the fuel elements to reprocessing the spent fuel and conditioning for deposit in a final repository.

Machining Test Specimens from Harvested Zion RPV Segments for Through Wall Attenuation Studies

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

Rosseel, Thomas M; Sokolov, Mikhail A; Nanstad, Randy K

2015-01-01

The decommissioning of the Zion Units 1 and 2 Nuclear Generating Station (NGS) in Zion, Illinois presents a special opportunity for developing a better understanding of materials degradation and other issues associated with extending the lifetime of existing Nuclear Power Plants (NPPs) beyond 60 years of service. In support of extended service and current operations of the US nuclear reactor fleet, the Oak Ridge National Laboratory (ORNL), through the Department of Energy (DOE), Light Water Reactor Sustainability (LWRS) Program, is coordinating and contracting with Zion Solutions, LLC, a subsidiary of Energy Solutions, the selective procurement of materials, structures, and componentsmore » from the decommissioned reactors. In this paper, we will discuss the acquisition of segments of the Zion Unit 2 Reactor Pressure Vessel (RPV), the cutting of these segments into sections and blocks from the beltline and upper vertical welds and plate material, the current status of machining those blocks into mechanical (Charpy, compact tension, and tensile) test specimens and coupons for chemical and microstructural (TEM, APT, SANS, and nano indention) characterization, as well as the current test plans and possible collaborative projects. Access to service-irradiated RPV welds and plate sections will allow through wall attenuation studies to be performed, which will be used to assess current radiation damage models (Rosseel et al. (2012) and Rosseel et al. (2015)).« less

10 CFR 2.621 - Acceptance and docketing of application for early review of site suitability issues in a combined...

Code of Federal Regulations, 2012 CFR

2012-01-01

... Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation, as...) The Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor Regulation... of Nuclear Reactor Regulation, as appropriate, that they are complete. (c) If part one of the...

Prospective scenarios of nuclear energy evolution over the 21. century

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

Massara, S.; Tetart, P.; Garzenne, C.

2006-07-01

In this paper, different world scenarios of nuclear energy development over the 21. century are analyzed, by means of the EDF fuel cycle simulation code for nuclear scenario studies, TIRELIRE - STRATEGIE. Three nuclear demand scenarios are considered, and the performance of different nuclear strategies in satisfying these scenarios is analyzed and discussed, focusing on natural uranium consumption and industrial requirements related to the nuclear reactors and the associated fuel cycle facilities. Both thermal-spectrum systems (Pressurized Water Reactor and High Temperature Gas-cooled Reactor) and Fast Reactors are investigated. (authors)

78 FR 57904 - Request for a License To Export; Reactor Components

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-20

... NUCLEAR REGULATORY COMMISSION Request for a License To Export; Reactor Components Pursuant to 10..., systems, related reactors. operation of AP- XR177, 11006121. equipment, and 1000 (design) spare parts. nuclear reactors. Dated this 16th day of September 2013 in Rockville, Maryland. For The Nuclear Regulatory...

10 CFR 2.1115 - Designation of issues for adjudicatory hearing.

Code of Federal Regulations, 2010 CFR

2010-01-01

... at Civilian Nuclear Power Reactors § 2.1115 Designation of issues for adjudicatory hearing. (a) After... reactor already licensed to operate at the site, or any civilian nuclear power reactor for which a... the issuance of a construction permit or operating license for a civilian nuclear power reactor at...

10 CFR 2.108 - Denial of application for failure to supply information.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of... other time as may be specified. (b) The Director, Office of Nuclear Reactor Regulation, Director, Office..., will rule whether an application should be denied by the Director, Office of Nuclear Reactor Regulation...

10 CFR 2.108 - Denial of application for failure to supply information.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of... other time as may be specified. (b) The Director, Office of Nuclear Reactor Regulation, Director, Office..., will rule whether an application should be denied by the Director, Office of Nuclear Reactor Regulation...

Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

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

Farmer, Mitchell T.; Bunt, R.; Corradini, M.

The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affectmore » reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).« less

Environmental assessment of SP-100 ground engineering system test site: Hanford Site, Richland, Washington

NASA Astrophysics Data System (ADS)

1988-12-01

The US Department of Energy (DOE) proposes to modify an existing reactor containment building (decommissioned Plutonium Recycle Test Reactor (PRTR) 309 Building) to provide ground test capability for the prototype SP-100 reactor. The 309 Building (Figure 1.1) is located in the 300 Area on the Hanford Site in Washington State. The National Environmental Policy Act (NEPA) requires that Federal agencies assess the potential impacts that their actions may have on the environment. This Environmental Assessment describes the consideration given to environmental impacts during reactor concept and test site selection, examines the environmental effects of the DOE proposal to ground test the nuclear subsystem, describes alternatives to the proposed action, and examines radiological risks of potential SP-100 use in space.

A Roadmap of Innovative Nuclear Energy System

NASA Astrophysics Data System (ADS)

Sekimoto, Hiroshi

2017-01-01

Nuclear is a dense energy without CO2 emission. It can be used for more than 100,000 years using fast breeder reactors with uranium from the sea. However, it raises difficult problems associated with severe accidents, spent fuel waste and nuclear threats, which should be solved with acceptable costs. Some innovative reactors have attracted interest, and many designs have been proposed for small reactors. These reactors are considered much safer than conventional large reactors and have fewer technical obstructions. Breed-and-burn reactors have high potential to solve all inherent problems for peaceful use of nuclear energy. However, they have some technical problems with materials. A roadmap for innovative reactors is presented herein.

Regulatory Risk Reduction for Advanced Reactor Technologies – FY2016 Status and Work Plan Summary

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

Moe, Wayne Leland

2016-08-01

Millions of public and private sector dollars have been invested over recent decades to realize greater efficiency, reliability, and the inherent and passive safety offered by advanced nuclear reactor technologies. However, a major challenge in experiencing those benefits resides in the existing U.S. regulatory framework. This framework governs all commercial nuclear plant construction, operations, and safety issues and is highly large light water reactor (LWR) technology centric. The framework must be modernized to effectively deal with non-LWR advanced designs if those designs are to become part of the U.S energy supply. The U.S. Department of Energy’s (DOE) Advanced Reactor Technologiesmore » (ART) Regulatory Risk Reduction (RRR) initiative, managed by the Regulatory Affairs Department at the Idaho National Laboratory (INL), is establishing a capability that can systematically retire extraneous licensing risks associated with regulatory framework incompatibilities. This capability proposes to rely heavily on the perspectives of the affected regulated community (i.e., commercial advanced reactor designers/vendors and prospective owner/operators) yet remain tuned to assuring public safety and acceptability by regulators responsible for license issuance. The extent to which broad industry perspectives are being incorporated into the proposed framework makes this initiative unique and of potential benefit to all future domestic non-LWR applicants« less

The United Arab Emirates Nuclear Program and Proposed U.S. Nuclear Cooperation

DTIC Science & Technology

2009-10-28

global efforts to prevent nuclear proliferation” and, “the establishment of reliable sources of nuclear fuel for future civilian light water reactors ...nuclear reactor or on handling spent reactor fuel. (...continued) May 4, 2008; and, Chris...related to the UAE’s proposed nuclear program has already taken place. In August 2008, Virginia’s Thorium Power Ltd. signed two consulting and

The United Arab Emirates Nuclear Program and Proposed U.S. Nuclear Cooperation

DTIC Science & Technology

2009-07-17

global efforts to prevent nuclear proliferation” and, “the establishment of reliable sources of nuclear fuel for future civilian light water reactors ...planned nuclear reactor or on handling spent reactor fuel. (...continued) May 4, 2008...contracting between U.S. firms and the UAE related to the UAE’s proposed nuclear program has already taken place. In August 2008, Virginia’s Thorium Power

Nuclear energy center site survey reactor plant considerations

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

Harty, H.

The Energy Reorganization Act of 1974 required the Nuclear Regulatory Commission (NRC) to make a nuclear energy center site survey (NECSS). Background information for the NECSS report was developed in a series of tasks which include: socioeconomic inpacts; environmental impact (reactor facilities); emergency response capability (reactor facilities); aging of nuclear energy centers; and dry cooled nuclear energy centers.

10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

Code of Federal Regulations, 2012 CFR

2012-01-01

..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

Code of Federal Regulations, 2013 CFR

2013-01-01

..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

Code of Federal Regulations, 2014 CFR

2014-01-01

..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

Navy Nuclear-Powered Surface Ships: Background, Issues, and Options for Congress

DTIC Science & Technology

2010-09-29

to design a smaller scale version of a naval pressurized water reactor , or to design a new reactor type potentially using a thorium liquid salt...integrated nuclear power system capable of use on destroyer- sized vessels either using a pressurized water reactor or a thorium liquid salt reactor ...nuclear reactors for Navy surface ships. The text of Section 246 is as follows: SEC. 246. STUDY ON THORIUM -LIQUID FUELED REACTORS FOR NAVAL FORCES

Nuclear data activities at the n_TOF facility at CERN

NASA Astrophysics Data System (ADS)

Gunsing, F.; Aberle, O.; Andrzejewski, J.; Audouin, L.; Bécares, V.; Bacak, M.; Balibrea-Correa, J.; Barbagallo, M.; Barros, S.; Bečvář, F.; Beinrucker, C.; Belloni, F.; Berthoumieux, E.; Billowes, J.; Bosnar, D.; Brugger, M.; Caamaño, M.; Calviño, F.; Calviani, M.; Cano-Ott, D.; Cardella, R.; Casanovas, A.; Castelluccio, D. M.; Cerutti, F.; Chen, Y. H.; Chiaveri, E.; Colonna, N.; Cortés-Giraldo, M. A.; Cortés, G.; Cosentino, L.; Damone, L. A.; Deo, K.; Diakaki, M.; Domingo-Pardo, C.; Dressler, R.; Dupont, E.; Durán, I.; Fernández-Domínguez, B.; Ferrari, A.; Ferreira, P.; Finocchiaro, P.; Frost, R. J. W.; Furman, V.; Ganesan, S.; García, A. R.; Gawlik, A.; Gheorghe, I.; Glodariu, T.; Gonçalves, I. F.; González, E.; Goverdovski, A.; Griesmayer, E.; Guerrero, C.; Göbel, K.; Harada, H.; Heftrich, T.; Heinitz, S.; Hernández-Prieto, A.; Heyse, J.; Jenkins, D. G.; Jericha, E.; Käppeler, F.; Kadi, Y.; Katabuchi, T.; Kavrigin, P.; Ketlerov, V.; Khryachkov, V.; Kimura, A.; Kivel, N.; Kokkoris, M.; Krtička, M.; Leal-Cidoncha, E.; Lederer, C.; Leeb, H.; Lerendegui, J.; Licata, M.; Lo Meo, S.; Lonsdale, S. J.; Losito, R.; Macina, D.; Marganiec, J.; Martínez, T.; Masi, A.; Massimi, C.; Mastinu, P.; Mastromarco, M.; Matteucci, F.; Maugeri, E. A.; Mazzone, A.; Mendoza, E.; Mengoni, A.; Milazzo, P. M.; Mingrone, F.; Mirea, M.; Montesano, S.; Musumarra, A.; Nolte, R.; Oprea, A.; Palomo-Pinto, F. R.; Paradela, C.; Patronis, N.; Pavlik, A.; Perkowski, J.; Porras, I.; Praena, J.; Quesada, J. M.; Rajeev, K.; Rauscher, T.; Reifarth, R.; Riego-Perez, A.; Robles, M.; Rout, P.; Radeck, D.; Rubbia, C.; Ryan, J. A.; Sabaté-Gilarte, M.; Saxena, A.; Schillebeeckx, P.; Schmidt, S.; Schumann, D.; Sedyshev, P.; Smith, A. G.; Stamatopoulos, A.; Suryanarayana, S. V.; Tagliente, G.; Tain, J. L.; Tarifeño-Saldivia, A.; Tarrío, D.; Tassan-Got, L.; Tsinganis, A.; Valenta, S.; Vannini, G.; Variale, V.; Vaz, P.; Ventura, A.; Vlachoudis, V.; Vlastou, R.; Wallner, A.; Warren, S.; Weigand, M.; Weiss, C.; Wolf, C.; Woods, P. J.; Wright, T.; Žugec, P.

2016-10-01

Nuclear data in general, and neutron-induced reaction cross sections in particular, are important for a wide variety of research fields. They play a key role in the safety and criticality assessment of nuclear technology, not only for existing power reactors but also for radiation dosimetry, medical applications, the transmutation of nuclear waste, accelerator-driven systems, fuel cycle investigations and future reactor systems as in Generation IV. Applications of nuclear data are also related to research fields as the study of nuclear level densities and stellar nucleosynthesis. Simulations and calculations of nuclear technology applications largely rely on evaluated nuclear data libraries. The evaluations in these libraries are based both on experimental data and theoretical models. Experimental nuclear reaction data are compiled on a worldwide basis by the international network of Nuclear Reaction Data Centres (NRDC) in the EXFOR database. The EXFOR database forms an important link between nuclear data measurements and the evaluated data libraries. CERN's neutron time-of-flight facility n_TOF has produced a considerable amount of experimental data since it has become fully operational with the start of the scientific measurement programme in 2001. While for a long period a single measurement station (EAR1) located at 185 m from the neutron production target was available, the construction of a second beam line at 20 m (EAR2) in 2014 has substantially increased the measurement capabilities of the facility. An outline of the experimental nuclear data activities at CERN's neutron time-of-flight facility n_TOF will be presented.

Aircraft and Bases Powered by Compact Nuclear Reactors: Solutions to Projecting Power in Highly Contested Environments and Fossil Fuel Dependence

DTIC Science & Technology

2015-05-01

pushed the depletion date past 2100.21 David Archibald, author of books and papers on climate science and a fellow at the Institute of World...Politics, does not predict explicitly the date of complete exhaustion, but he does note that humans have consumed about half of the world’s supply.22...deuterium, and lithium are plentiful on the earth and in the solar system. As far as fuel for existing and future fission reactors, uranium and

Elastic and inelastic neutron scattering cross sections for fission reactor applications

NASA Astrophysics Data System (ADS)

Hicks, S. F.; Chakraborty, A.; Combs, B.; Crider, B. P.; Downes, L.; Girgis, J.; Kersting, L. J.; Kumar, A.; Lueck, C. J.; McDonough, P. J.; McEllistrem, M. T.; Peters, E. E.; Prados-Estevz, F. M.; Schniederjan, J.; Sidwell, L.; Sigillito, A. J.; Vanhoy, J. R.; Watts, D.; Yates, S. W.

2013-04-01

Nuclear data important for the design and development of the next generation of light-water reactors and future fast reactors include neutron elastic and inelastic scattering cross sections on important structural materials, such as Fe, and on coolant materials, such as Na. These reaction probabilities are needed since neutron reactions impact fuel performance during irradiations and the overall efficiency of reactors. While neutron scattering cross sections from these materials are available for certain incident neutron energies, the fast neutron region, particularly above 2 MeV, has large gaps for which no measurements exist, or the existing uncertainties are large. Measurements have been made at the University of Kentucky Accelerator Laboratory to measure neutron scattering cross sections on both Fe and Na in the region where these gaps occur and to reduce the uncertainties on scattering from the ground state and first excited state of these nuclei. Results from measurements on Fe at incident neutron energies between 2 and 4 MeV will be presented and comparisons will be made to model calculations available from data evaluators.

10 CFR 50.44 - Combustible gas control for nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere with...

10 CFR 50.44 - Combustible gas control for nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere with...

Nuclear Energy and Synthetic Liquid Transportation Fuels

NASA Astrophysics Data System (ADS)

McDonald, Richard

2012-10-01

This talk will propose a plan to combine nuclear reactors with the Fischer-Tropsch (F-T) process to produce synthetic carbon-neutral liquid transportation fuels from sea water. These fuels can be formed from the hydrogen and carbon dioxide in sea water and will burn to water and carbon dioxide in a cycle powered by nuclear reactors. The F-T process was developed nearly 100 years ago as a method of synthesizing liquid fuels from coal. This process presently provides commercial liquid fuels in South Africa, Malaysia, and Qatar, mainly using natural gas as a feedstock. Nuclear energy can be used to separate water into hydrogen and oxygen as well as to extract carbon dioxide from sea water using ion exchange technology. The carbon dioxide and hydrogen react to form synthesis gas, the mixture needed at the beginning of the F-T process. Following further refining, the products, typically diesel and Jet-A, can use existing infrastructure and can power conventional engines with little or no modification. We can then use these carbon-neutral liquid fuels conveniently long into the future with few adverse environmental impacts.

Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

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

Not Available

1992-07-01

Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility.

The WPI reactor-readying for the next generation

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

Bobek, L.M.

1993-01-01

Built in 1959, the 10-kW open-pool nuclear training reactor at Worcester Polytechnic Institute (WPI) was one of the first such facilities in the nation located on a university campus. Since then, the reactor and its related facilities have been used to train two generations of nuclear engineers and scientists for the nuclear industry. With the use of nuclear technology playing an increasing role in many segments of the economy, WPI with its nuclear reactor facility is committed to continuing its mission of training future nuclear engineers and scientists. The WPI reactor includes a 6-in. beam port, graphite thermal column, andmore » in-core sample facility. The reactor, housed in an open 8000-gal tank of water, is designed so that the core is readily accessible. Both the control console and the peripheral counting equipment used for student projects and laboratory exercises are located in the reactor room. This arrangement provides convenience and flexibility in using the reactor for foil activations in neutron flux measurements, diffusion measurements, radioactive decay measurements, and the neutron activation of samples for analysis. In 1988, the reactor was successfully converted to low-enriched uranium fuel.« less

Radiation resistant fiber Bragg grating in random air-line fibers for sensing applications in nuclear reactor cores.

PubMed

Zaghloul, Mohamed A S; Wang, Mohan; Huang, Sheng; Hnatovsky, Cyril; Grobnic, Dan; Mihailov, Stephen; Li, Ming-Jun; Carpenter, David; Hu, Lin-Wen; Daw, Joshua; Laffont, Guillaume; Nehr, Simon; Chen, Kevin P

2018-04-30

This paper reports the testing results of radiation resistant fiber Bragg grating (FBG) in random air-line (RAL) fibers in comparison with FBGs in other radiation-hardened fibers. FBGs in RAL fibers were fabricated by 80 fs ultrafast laser pulse using a phase mask approach. The fiber Bragg gratings tests were carried out in the core region of a 6 MW MIT research reactor (MITR) at a steady temperature above 600°C and an average fast neutron (>1 MeV) flux >1.2 × 10 14 n/cm 2 /s. Fifty five-day tests of FBG sensors showed less than 5 dB reduction in FBG peak strength after over 1 × 10 20 n/cm 2 of accumulated fast neutron dose. The radiation-induced compaction of FBG sensors produced less than 5.5 nm FBG wavelength shift toward shorter wavelength. To test temporal responses of FBG sensors, a number of reactor anomaly events were artificially created to abruptly change reactor power, temperature, and neutron flux over short periods of time. The thermal sensitivity and temporal responses of FBGs were determined at different accumulated doses of neutron flux. Results presented in this paper reveal that temperature-stable Type-II FBGs fabricated in radiation-hardened fibers can survive harsh in-pile conditions. Despite large parameter drift induced by strong nuclear radiation, further engineering and innovation on both optical fibers and fiber devices could lead to useful fiber sensors for various in-pile measurements to improve safety and efficiency of existing and next generation nuclear reactors.

Comparison of evolving photovoltaic and nuclear power systems for earth orbital applications

NASA Technical Reports Server (NTRS)

Rockey, D. E.; Jones, R. M.; Schulman, I.

1982-01-01

Photovoltaic and fission reactor orbital power systems are compared in terms of the end-to-end system power-to-mass ratios. Three PV systems are examined, i.e., a solid substrate with a cell array and a NiCd battery, a modified SEP array and an NiH2 battery, and a 62-micron Si cell array and a fuel cell. All arrays were modeled to be 13.5% efficient and to produce 25 kW dc. The SP-100 reactor consists of the heat source, radiation shield, heat pipes to transfer thermal energy from the reactor to thermoelectric elements, and a waste heat radiator. Consideration is given to system applications in orbits ranging from LEO to GEO, and to mission durations of 1, 5, and 10 yr. PV systems are concluded to be flight-proven, useful out of radiation belts, and best for low to moderate power levels. Limitations exist for operations where atmospheric drag may become a factor and due to the size of a large PV power supply. Space nuclear reactors will continue under development and uses at high power levels and in low altitude orbits are foreseen.

Nuclear reactors built, being built, or planned 1993

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

Not Available

1993-08-01

Nuclear Reactors Built, Being Built, or Planned contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1993. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical datamore » that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: civilian, production, military, export, and critical assembly.« less

Closed fuel cycle with increased fuel burn-up and economy applying of thorium resources

NASA Astrophysics Data System (ADS)

Kulikov, G. G.; Apse, V. A.

2017-01-01

The possible role of existing thorium reserves in the Russian Federation on engaging thorium in being currently closed (U-Pu)-fuel cycle of nuclear power of the country is considered. The application efficiency of thermonuclear neutron sources with thorium blanket for the economical use of existing thorium reserves is demonstrated. The aim of the work is to find solutions of such major tasks as the reduction of both front-end and back-end of nuclear fuel cycle and an enhancing its protection against the uncontrolled proliferation of fissile materials by means of the smallest changes in the fuel cycle. During implementation of the work we analyzed the results obtained earlier by the authors, brought new information on the number of thorium available in the Russian Federation and made further assessments. On the basis of proposal on the inclusion of hybrid reactors with Th-blanket into the future nuclear power for the production of light uranium fraction 232+233+234U, and 231Pa, we obtained the following results: 1. The fuel cycle will shift from fissile 235U to 233U which is more attractive for thermal power reactors. 2. The light uranium fraction is the most "protected" in the uranium component of fuel and mixed with regenerated uranium will in addition become a low enriched uranium fuel, that will weaken the problem of uncontrolled proliferation of fissile materials. 3. 231Pa doping into the fuel stabilizes its multiplying properties that will allow us to implement long-term fuel residence time and eventually to increase the export potential of all nuclear power technologies. 4. The thorium reserves being near city Krasnoufimsk (Russia) are large enough for operation of large-scale nuclear power of the Russian Federation of 70 GWe capacity during more than a quarter century under assumption that thorium is loaded into blankets of hybrid TNS only. The general conclusion: the inclusion of a small number of hybrid reactors with Th-blanket into the future nuclear power will allow us substantially to solve its problems, as well as to increase its export potential.

Summit on Improving the Economics of America's Nuclear Power Plants

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

Collins, John; Mason, Charles

The Summit on Improving the Economics of America’s Nuclear Power Plants was convened May 19, 2016, by Secretary of Energy Ernest Moniz and co-sponsored by Idaho Senator Mike Crapo to stress the importance of existing nuclear reactors in meeting our nation’s energy goals. The summit was also designed to identify and discuss policy options that can be pursued at federal and state levels to address economic challenges, as well as technical options that utilities can use to improve the economic competitiveness of operating nuclear power plants (NPPs) and avoid early plant retirements that are driven by temporary market conditions. Themore » owners of NPPs face difficult economic decisions and are working to improve the performance of existing NPPs. However, it soon became clear that some of the actions taken by states and regional markets have had an impact on the economic viability of existing power plants, including carbon free NPPs. Summit speakers identified concepts and actions that could be taken at state and federal levels to improve the economics of the existing fleet within these regulated and restructured electricity markets. This report summarizes the speeches, concepts, and actions taken.« less

High-intensity power-resolved radiation imaging of an operational nuclear reactor.

PubMed

Beaumont, Jonathan S; Mellor, Matthew P; Villa, Mario; Joyce, Malcolm J

2015-10-09

Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors.

High-intensity power-resolved radiation imaging of an operational nuclear reactor

PubMed Central

Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

2015-01-01

Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

10 CFR 51.107 - Public hearings in proceedings for issuance of combined licenses; limited work authorizations.

Code of Federal Regulations, 2010 CFR

2010-01-01

... the issuance of a combined license for a nuclear power reactor under part 52 of this chapter, the... or Director, Office of Nuclear Reactor Regulation, as appropriate. (b) If a combined license... authorization should be issued as proposed by the Director of New Reactors or the Director of Nuclear Reactor...

10 CFR 51.107 - Public hearings in proceedings for issuance of combined licenses; limited work authorizations.

Code of Federal Regulations, 2012 CFR

2012-01-01

... the issuance of a combined license for a nuclear power reactor under part 52 of this chapter, the... or Director, Office of Nuclear Reactor Regulation, as appropriate. (b) If a combined license... authorization should be issued as proposed by the Director of New Reactors or the Director of Nuclear Reactor...

10 CFR 51.107 - Public hearings in proceedings for issuance of combined licenses; limited work authorizations.

Code of Federal Regulations, 2014 CFR

2014-01-01

... the issuance of a combined license for a nuclear power reactor under part 52 of this chapter, the... or Director, Office of Nuclear Reactor Regulation, as appropriate. (b) If a combined license... authorization should be issued as proposed by the Director of New Reactors or the Director of Nuclear Reactor...

10 CFR 51.107 - Public hearings in proceedings for issuance of combined licenses; limited work authorizations.

Code of Federal Regulations, 2011 CFR

2011-01-01

... the issuance of a combined license for a nuclear power reactor under part 52 of this chapter, the... or Director, Office of Nuclear Reactor Regulation, as appropriate. (b) If a combined license... authorization should be issued as proposed by the Director of New Reactors or the Director of Nuclear Reactor...

10 CFR 51.107 - Public hearings in proceedings for issuance of combined licenses; limited work authorizations.

Code of Federal Regulations, 2013 CFR

2013-01-01

... the issuance of a combined license for a nuclear power reactor under part 52 of this chapter, the... or Director, Office of Nuclear Reactor Regulation, as appropriate. (b) If a combined license... authorization should be issued as proposed by the Director of New Reactors or the Director of Nuclear Reactor...

78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-25

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0237] Cost-Benefit Analysis for Radwaste Systems for Light... (RG) 1.110, ``Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors... components for light water nuclear power reactors. ADDRESSES: Please refer to Docket ID NRC-2013-0237 when...

10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Technical specifications on effluents from nuclear power reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors...

10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Technical specifications on effluents from nuclear power reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors...

10 CFR 140.52 - Indemnity agreements.

Code of Federal Regulations, 2014 CFR

2014-01-01

...) authorizing the licensee to operate the nuclear reactor involved; or (2) The effective date of the license... material at the site of the nuclear reactor for use as fuel in operation of the nuclear reactor after... 10 Energy 2 2014-01-01 2014-01-01 false Indemnity agreements. 140.52 Section 140.52 Energy NUCLEAR...

75 FR 42469 - Firstenergy Nuclear Operating Company; Request for Licensing Action

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-21

... nuclear plant in Ohio, preventing the reactor from restarting until such time that the NRC determines... Commission's regulations. The request has been referred to the Director of the Office of Nuclear Reactor... of Nuclear Reactor Regulation. [FR Doc. 2010-17834 Filed 7-20-10; 8:45 am] BILLING CODE 7590-01-P ...

10 CFR 140.52 - Indemnity agreements.

Code of Federal Regulations, 2010 CFR

2010-01-01

...) authorizing the licensee to operate the nuclear reactor involved; or (2) The effective date of the license... material at the site of the nuclear reactor for use as fuel in operation of the nuclear reactor after... 10 Energy 2 2010-01-01 2010-01-01 false Indemnity agreements. 140.52 Section 140.52 Energy NUCLEAR...

Advanced Test Reactor Tour

ScienceCinema

Miley, Don

2017-12-21

The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored.

The necessity of nuclear reactors for targeted radionuclide therapies.

PubMed

Krijger, Gerard C; Ponsard, Bernard; Harfensteller, Mark; Wolterbeek, Hubert T; Nijsen, Johannes W F

2013-07-01

Nuclear medicine has been contributing towards personalized therapies. Nuclear reactors are required for the working horses of both diagnosis and treatment, i.e., Tc-99m and I-131. In fact, reactors will remain necessary to fulfill the demand for a variety of radionuclides and are essential in the expanding field of targeted radionuclide therapies for cancer. However, the main reactors involved in the global supply are ageing and expected to shut down before 2025. Therefore, the fields of (nuclear) medicine, nuclear industry and politics share a global responsibility, faced with the task to secure future access to suitable nuclear reactors. At the same time, alternative production routes should be industrialized. For this, a coordinating entity should be put into place. Copyright © 2013 Elsevier Ltd. All rights reserved.

An underground nuclear power station using self-regulating heat-pipe controlled reactors

DOEpatents

Hampel, V.E.

1988-05-17

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast- acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor. 5 figs.

Underground nuclear power station using self-regulating heat-pipe controlled reactors

DOEpatents

Hampel, Viktor E.

1989-01-01

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working flud in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor.

Nuclear propulsion apparatus with alternate reactor segments

DOEpatents

Szekely, Thomas

1979-04-03

1. Nuclear propulsion apparatus comprising: A. means for compressing incoming air; B. nuclear fission reactor means for heating said air; C. means for expanding a portion of the heated air to drive said compressing means; D. said nuclear fission reactor means being divided into a plurality of radially extending segments; E. means for directing a portion of the compressed air for heating through alternate segments of said reactor means and another portion of the compressed air for heating through the remaining segments of said reactor means; and F. means for further expanding the heated air from said drive means and the remaining heated air from said reactor means through nozzle means to effect reactive thrust on said apparatus.

Spent Nuclear Fuel Disposition

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

Wagner, John C.

One interdisciplinary field devoted to achieving the end-state of used nuclear fuel (UNF) through reuse and/or permanent disposal. The reuse option aims to make use of the remaining energy content in UNF and reduce the amount of long-lived radioactive materials that require permanent disposal. The planned approach in the U.S., as well as in many other countries worldwide, is direct permanent disposal in a deep geologic repository. Used nuclear fuel is fuel that has been irradiated in a nuclear reactor to the point where it is no longer capable of sustaining operational objectives. The vast majority (by mass) of UNFmore » is from electricity generation in commercial nuclear power reactors. Furthermore, the other main source of UNF in the U.S. is the Department of Energy’s (DOE) and other federal agencies’ operation of reactors in support of federal government missions, such as materials production, nuclear propulsion, research, testing, and training. Upon discharge from a reactor, UNF emits considerable heat from radioactive decay. Some period of active on-site cooling (e.g., 2 or more years) is typically required to facilitate efficient packaging and transportation to a disposition facility. Hence, the field of UNF disposition broadly includes storage, transportation and ultimate disposition. See also: Nuclear Fission (content/nuclear-fission/458400), Nuclear Fuels (/content/nuclear-fuels/458600), Nuclear Fuel Cycle (/content/nuclear-fuel-cycle/458500), Nuclear Fuels Reprocessing (/content/nuclear-fuels-reprocessing/458700), Nuclear Power (/content/nuclear-power/459600), Nuclear Reactor (/content/nuclear-reactor/460100), Radiation (/content/radiation/566300), and Radioactive Waste Management (/content/radioactive-waste-management/568900).« less

Spent Nuclear Fuel Disposition

DOE PAGES

Wagner, John C.

2016-05-22

One interdisciplinary field devoted to achieving the end-state of used nuclear fuel (UNF) through reuse and/or permanent disposal. The reuse option aims to make use of the remaining energy content in UNF and reduce the amount of long-lived radioactive materials that require permanent disposal. The planned approach in the U.S., as well as in many other countries worldwide, is direct permanent disposal in a deep geologic repository. Used nuclear fuel is fuel that has been irradiated in a nuclear reactor to the point where it is no longer capable of sustaining operational objectives. The vast majority (by mass) of UNFmore » is from electricity generation in commercial nuclear power reactors. Furthermore, the other main source of UNF in the U.S. is the Department of Energy’s (DOE) and other federal agencies’ operation of reactors in support of federal government missions, such as materials production, nuclear propulsion, research, testing, and training. Upon discharge from a reactor, UNF emits considerable heat from radioactive decay. Some period of active on-site cooling (e.g., 2 or more years) is typically required to facilitate efficient packaging and transportation to a disposition facility. Hence, the field of UNF disposition broadly includes storage, transportation and ultimate disposition. See also: Nuclear Fission (content/nuclear-fission/458400), Nuclear Fuels (/content/nuclear-fuels/458600), Nuclear Fuel Cycle (/content/nuclear-fuel-cycle/458500), Nuclear Fuels Reprocessing (/content/nuclear-fuels-reprocessing/458700), Nuclear Power (/content/nuclear-power/459600), Nuclear Reactor (/content/nuclear-reactor/460100), Radiation (/content/radiation/566300), and Radioactive Waste Management (/content/radioactive-waste-management/568900).« less

Air Shipment of Highly Enriched Uranium Spent Nuclear Fuel from Romania

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

K. J. Allen; I. Bolshinsky; L. L. Biro

2010-07-01

Romania safely air shipped 23.7 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel from the VVR S research reactor at Magurele, Romania, to the Russian Federation in June 2009. This was the world’s first air shipment of spent nuclear fuel transported in a Type B(U) cask under existing international laws without special exceptions for the air transport licenses. This shipment was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in cooperation with the Romania National Commission for Nuclear Activities Control (CNCAN), the Horiamore » Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH), and the Russian Federation State Corporation Rosatom. The shipment was transported by truck to and from the respective commercial airports in Romania and the Russian Federation and stored at a secure nuclear facility in Russia where it will be converted into low enriched uranium. With this shipment, Romania became the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the work, equipment, and approvals that were required to complete this spent fuel air shipment.« less

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.].

Reactor engineering support of operations at the Davis-Besse nuclear power station

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

Kelley, D.B.

1995-12-31

Reactor engineering functions differ greatly from unit to unit; however, direct support of the reactor operators during reactor startups and operational transients is common to all units. This paper summarizes the support the reactor engineers provide the reactor operators during reactor startups and power changes through the use of automated computer programs at the Davis-Besse nuclear power station.

Progress in space nuclear reactor power systems technology development - The SP-100 program

NASA Technical Reports Server (NTRS)

Davis, H. S.

1984-01-01

Activities related to the development of high-temperature compact nuclear reactors for space applications had reached a comparatively high level in the U.S. during the mid-1950s and 1960s, although only one U.S. nuclear reactor-powered spacecraft was actually launched. After 1973, very little effort was devoted to space nuclear reactor and propulsion systems. In February 1983, significant activities toward the development of the technology for space nuclear reactor power systems were resumed with the SP-100 Program. Specific SP-100 Program objectives are partly related to the determination of the potential performance limits for space nuclear power systems in 100-kWe and 1- to 100-MW electrical classes. Attention is given to potential missions and applications, regimes of possible space power applicability, safety considerations, conceptual system designs, the establishment of technical feasibility, nuclear technology, materials technology, and prospects for the future.

76 FR 69296 - University of Utah, University of Utah TRIGA Nuclear Reactor, Notice of Issuance of Renewed...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-08

... NUCLEAR REGULATORY COMMISSION [Docket No. 50-407, NRC-2011-0153] University of Utah, University of Utah TRIGA Nuclear Reactor, Notice of Issuance of Renewed Facility Operating License No. R-126 AGENCY... University of Utah (UU, the licensee), which authorizes continued operation of the UU TRIGA Nuclear Reactor...

76 FR 35922 - Interim Staff Guidance Regarding the Environmental Report for Applications To Construct and/or...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-20

... Nuclear Reactor Regulation on the information that should be included in the Environmental Report, which...: Mr. Scott Sloan, Project Manager, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory..., Office of Nuclear Reactor Regulation. [FR Doc. 2011-15227 Filed 6-17-11; 8:45 am] BILLING CODE 7590-01-P ...

Space nuclear power systems; Proceedings of the 8th Symposium, Albuquerque, NM, Jan. 6-10, 1991. Pts. 1-3

NASA Technical Reports Server (NTRS)

El-Genk, Mohamed S. (Editor); Hoover, Mark D. (Editor)

1991-01-01

The present conference discusses NASA mission planning for space nuclear power, lunar mission design based on nuclear thermal rockets, inertial-electrostatic confinement fusion for space power, nuclear risk analysis of the Ulysses mission, the role of the interface in refractory metal alloy composites, an advanced thermionic reactor systems design code, and space high power nuclear-pumped lasers. Also discussed are exploration mission enhancements with power-beaming, power requirement estimates for a nuclear-powered manned Mars rover, SP-100 reactor design, safety, and testing, materials compatibility issues for fabric composite radiators, application of the enabler to nuclear electric propulsion, orbit-transfer with TOPAZ-type power sources, the thermoelectric properties of alloys, ruthenium silicide as a promising thermoelectric material, and innovative space-saving device for high-temperature piping systems. The second volume of this conference discusses engine concepts for nuclear electric propulsion, nuclear technologies for human exploration of the solar system, dynamic energy conversion, direct nuclear propulsion, thermionic conversion technology, reactor and power system control, thermal management, thermionic research, effects of radiation on electronics, heat-pipe technology, radioisotope power systems, and nuclear fuels for power reactors. The third volume discusses space power electronics, space nuclear fuels for propulsion reactors, power systems concepts, space power electronics systems, the use of artificial intelligence in space, flight qualifications and testing, microgravity two-phase flow, reactor manufacturing and processing, and space and environmental effects.

10 CFR 2.1105 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Hearing Procedures for Expansion of Spent Nuclear Fuel Storage Capacity at Civilian Nuclear Power Reactors § 2.1105 Definitions. As used in this part: (a) Civilian nuclear power reactor means a civilian... reactor following irradiation, the constituent elements of which have not been separated by reprocessing. ...

75 FR 5357 - In the Matter of Entergy Nuclear Operations, Inc., et al.; Order Extending the Effectiveness of...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-02-02

...). Pilgrim is a boiling water nuclear reactor that is owned by Entergy Nuclear and operated by ENO. The... Generating Unit No. 1 (IP1). IP1 is a pressurized water nuclear reactor that is owned by ENIP2 and maintained... nuclear reactors that are owned by ENIP2 and ENIP3, respectively, and operated by ENO. The facilities are...

High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

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

Mcwilliams, A. J.

2015-09-08

This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniquesmore » through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.« less

Analytical design and performance studies of nuclear furnace tests of small nuclear light bulb models

NASA Technical Reports Server (NTRS)

Latham, T. S.; Rodgers, R. J.

1972-01-01

Analytical studies were continued to identify the design and performance characteristics of a small-scale model of a nuclear light bulb unit cell suitable for testing in a nuclear furnace reactor. Emphasis was placed on calculating performance characteristics based on detailed radiant heat transfer analyses, on designing the test assembly for ease of insertion, connection, and withdrawal at the reactor test cell, and on determining instrumentation and test effluent handling requirements. In addition, a review of candidate test reactors for future nuclear light bulb in-reactor tests was conducted.

Open-Source Integrated Design-Analysis Environment For Nuclear Energy Advanced Modeling & Simulation Final Scientific/Technical Report

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

O'Leary, Patrick

The framework created through the Open-Source Integrated Design-Analysis Environment (IDAE) for Nuclear Energy Advanced Modeling & Simulation grant has simplify and democratize advanced modeling and simulation in the nuclear energy industry that works on a range of nuclear engineering applications. It leverages millions of investment dollars from the Department of Energy's Office of Nuclear Energy for modeling and simulation of light water reactors and the Office of Nuclear Energy's research and development. The IDEA framework enhanced Kitware’s Computational Model Builder (CMB) while leveraging existing open-source toolkits and creating a graphical end-to-end umbrella guiding end-users and developers through the nuclear energymore » advanced modeling and simulation lifecycle. In addition, the work deliver strategic advancements in meshing and visualization for ensembles.« less

Researcher Poses with a Nuclear Rocket Model

NASA Image and Video Library

1961-11-21

A researcher at the NASA Lewis Research Center with slide ruler poses with models of the earth and a nuclear-propelled rocket. The Nuclear Engine for Rocket Vehicle Applications (NERVA) was a joint NASA and Atomic Energy Commission (AEC) endeavor to develop a nuclear-powered rocket for both long-range missions to Mars and as a possible upper-stage for the Apollo Program. The early portion of the program consisted of basic reactor and fuel system research. This was followed by a series of Kiwi reactors built to test nuclear rocket principles in a non-flying nuclear engine. The next phase, NERVA, would create an entire flyable engine. The AEC was responsible for designing the nuclear reactor and overall engine. NASA Lewis was responsible for developing the liquid-hydrogen fuel system. The nuclear rocket model in this photograph includes a reactor at the far right with a hydrogen propellant tank and large radiator below. The payload or crew would be at the far left, distanced from the reactor.

Transmutation of Isotopes --- Ecological and Energy Production Aspects

NASA Astrophysics Data System (ADS)

Gudowski, Waclaw

2000-01-01

This paper describes principles of Accelerator-Driven Transmutation of Nuclear Wastes (ATW) and gives some flavour of the most important topics which are today under investigations in many countries. An assessment of the potential impact of ATW on a future of nuclear energy is also given. Nuclear reactors based on self-sustained fission reactions --- after spectacular development in fifties and sixties, that resulted in deployment of over 400 power reactors --- are wrestling today more with public acceptance than with irresolvable technological problems. In a whole spectrum of reasons which resulted in today's opposition against nuclear power few of them are very relevant for the nuclear physics community and they arose from the fact that development of nuclear power had been handed over to the nuclear engineers and technicians with some generically unresolved problems, which should have been solved properly by nuclear scientists. In a certain degree of simplification one can say, that most of the problems originate from very specific features of a fission phenomenon: self-sustained chain reaction in fissile materials and very strong radioactivity of fission products and very long half-life of some of the fission and activation products. And just this enormous concentration of radioactive fission products in the reactor core is the main problem of managing nuclear reactors: it requires unconditional guarantee for the reactor core integrity in order to avoid radioactive contamination of the environment; it creates problems to handle decay heat in the reactor core and finally it makes handling and/or disposal of spent fuel almost a philosophical issue, due to unimaginable long time scales of radioactive decay of some isotopes. A lot can be done to improve the design of conventional nuclear reactors (like Light Water Reactors); new, better reactors can be designed but it seems today very improbable to expect any radical change in the public perception of conventional nuclear power. In this context a lot of hopes and expectations have been expressed for novel systems called Accelerator-Driven Systems, Accelerator-Driven Transmutation of Waste or just Hybrid Reactors. All these names are used for description of the same nuclear system combining a powerful particle accelerator with a subcritical reactor. A careful analysis of possible environmental impact of ATW together with limitation of this technology is presented also in this paper.

10 CFR 2.340 - Initial decision in certain contested proceedings; immediate effectiveness of initial decisions...

Code of Federal Regulations, 2014 CFR

2014-01-01

... no significant hazards consideration, the Commission, the Director, Office of Nuclear Reactor... determination of no significant hazards consideration, the Commission, the Director, Office of Nuclear Reactor... no significant hazards consideration, the Commission, the Director, Office of Nuclear Reactor...

10 CFR 2.340 - Initial decision in certain contested proceedings; immediate effectiveness of initial decisions...

Code of Federal Regulations, 2013 CFR

2013-01-01

... no significant hazards consideration, the Commission, the Director, Office of Nuclear Reactor... determination of no significant hazards consideration, the Commission, the Director, Office of Nuclear Reactor... no significant hazards consideration, the Commission, the Director, Office of Nuclear Reactor...

Liquid metal magnetohydrodynamics

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

Lielpeteris, J.; Moreau, R.

1989-01-01

Liquid metal MHD is the subject of this book. It is of central importance in fields like metals processing, energy conversion, nuclear engineering (fast breeders or fusion reactors), geomagnetism and astrophysics. In some circumstances fluid flow phenomena are controlled by an existing magnetic field; the melts in induction furnaces or the liquid metal blanket around future tokamak fusion reactors being significant examples. In other cases the application of an external magnetic field (or of an electric current) may generate drastic modifications in the fluid motion and in the transfer rates; such effects may be used to develop new technologies (electromagneticmore » shaping) or to improve existing techniques (electromagnetic stirring in continuous casting). In the core of the Earth, fluid motion and magnetic fields are both present and their interaction governs important phenomena.« less

78 FR 73898 - Operator Licensing Examination Standards for Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-09

... Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft NUREG; request for comment. SUMMARY: The U.S..., Revision 10, ``Operator Licensing Examination Standards for Power Reactors.'' DATES: Submit comments [email protected] . Both of the Office of New Reactors; or Timothy Kolb, Office of Nuclear Reactor Regulation, U...

A Programmable Liquid Collimator for Both Coded Aperture Adaptive Imaging and Multiplexed Compton Scatter Tomography

DTIC Science & Technology

2012-03-01

environments where a source is either weak or shielded. A vehicle of this type could survey large areas after a nuclear attack or a nuclear reactor accident...to prevent its detection by γ-rays. The best application for unmanned vehicles is the detection of radioactive material after a nuclear reactor ...accident or a nuclear weapon detonation [70]. Whether by a nuclear detonation or a nuclear reactor accident, highly radioactive substances could be dis

Nuclear Energy Enabling Technologies (NEET) Reactor Materials: News for the Reactor Materials Crosscut, May 2016

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

Maloy, Stuart Andrew

In this newsletter for Nuclear Energy Enabling Technologies (NEET) Reactor Materials, pages 1-3 cover highlights from the DOE-NE (Nuclear Energy) programs, pages 4-6 cover determining the stress-strain response of ion-irradiated metallic materials via spherical nanoindentation, and pages 7-8 cover theoretical approaches to understanding long-term materials behavior in light water reactors.

10 CFR 2.603 - Acceptance and docketing of application for early review of site suitability issues in a...

Code of Federal Regulations, 2012 CFR

2012-01-01

... processing, the Director of the Office of New Reactors or the Director of the Office of Nuclear Reactor... days. (b)(1) The Director of the Office of New Reactors or the Director of the Office of Nuclear... Nuclear Reactor Regulation, as appropriate, that they are complete. (c) If part one of the application is...

Solution of heat removal from nuclear reactors by natural convection

NASA Astrophysics Data System (ADS)

Zitek, Pavel; Valenta, Vaclav

2014-03-01

This paper summarizes the basis for the solution of heat removal by natural convection from both conventional nuclear reactors and reactors with fuel flowing coolant (such as reactors with molten fluoride salts MSR).The possibility of intensification of heat removal through gas lift is focused on. It might be used in an MSR (Molten Salt Reactor) for cleaning the salt mixture of degassed fission products and therefore eliminating problems with iodine pitting. Heat removal by natural convection and its intensification increases significantly the safety of nuclear reactors. Simultaneously the heat removal also solves problems with lifetime of pumps in the primary circuit of high-temperature reactors.

10 CFR Appendix N to Part 52 - Standardization of Nuclear Power Plant Designs: Combined Licenses To Construct and Operate...

Code of Federal Regulations, 2011 CFR

2011-01-01

... Licenses To Construct and Operate Nuclear Power Reactors of Identical Design at Multiple Sites N Appendix N... Designs: Combined Licenses To Construct and Operate Nuclear Power Reactors of Identical Design at Multiple... construct and operate nuclear power reactors of identical design (“common design”) to be located at multiple...

AIR SHIPMENT OF HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL FROM ROMANIA AND LIBYA

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

Christopher Landers; Igor Bolshinsky; Ken Allen

2010-07-01

In June 2009 Romania successfully completed the world’s first air shipment of highly enriched uranium (HEU) spent nuclear fuel transported in Type B(U) casks under existing international laws and without special exceptions for the air transport licenses. Special 20-foot ISO shipping containers and cask tiedown supports were designed to transport Russian TUK 19 shipping casks for the Romanian air shipment and the equipment was certified for all modes of transport, including road, rail, water, and air. In December 2009 Libya successfully used this same equipment for a second air shipment of HEU spent nuclear fuel. Both spent fuel shipments weremore » transported by truck from the originating nuclear facilities to nearby commercial airports, were flown by commercial cargo aircraft to a commercial airport in Yekaterinburg, Russia, and then transported by truck to their final destinations at the Production Association Mayak facility in Chelyabinsk, Russia. Both air shipments were performed under the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI). The Romania air shipment of 23.7 kg of HEU spent fuel from the VVR S research reactor was the last of three HEU fresh and spent fuel shipments under RRRFR that resulted in Romania becoming the 3rd RRRFR participating country to remove all HEU. Libya had previously completed two RRRFR shipments of HEU fresh fuel so the 5.2 kg of HEU spent fuel air shipped from the IRT 1 research reactor in December made Libya the 4th RRRFR participating country to remove all HEU. This paper describes the equipment, preparations, and license approvals required to safely and securely complete these two air shipments of spent nuclear fuel.« less

Air Shipment of Spent Nuclear Fuel from Romania to Russia

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

Igor Bolshinsky; Ken Allen; Lucian Biro

Romania successfully completed the world’s first air shipment of spent nuclear fuel transported in Type B(U) casks under existing international laws and without shipment license special exceptions when the last Romanian highly enriched uranium (HEU) spent nuclear fuel was transported to the Russian Federation in June 2009. This air shipment required the design, fabrication, and licensing of special 20 foot freight containers and cask tiedown supports to transport the eighteen TUK 19 shipping casks on a Russian commercial cargo aircraft. The new equipment was certified for transport by road, rail, water, and air to provide multi modal transport capabilities formore » shipping research reactor spent fuel. The equipment design, safety analyses, and fabrication were performed in the Russian Federation and transport licenses were issued by both the Russian and Romanian regulatory authorities. The spent fuel was transported by truck from the VVR S research reactor to the Bucharest airport, flown by commercial cargo aircraft to the airport at Yekaterinburg, Russia, and then transported by truck to the final destination in a secure nuclear facility at Chelyabinsk, Russia. This shipment of 23.7 kg of HEU was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in close cooperation with the Rosatom State Atomic Energy Corporation and the International Atomic Energy Agency, and was managed in Romania by the National Commission for Nuclear Activities Control (CNCAN). This paper describes the planning, shipment preparations, equipment design, and license approvals that resulted in the safe and secure air shipment of this spent nuclear fuel.« less

Total absorption spectroscopy of fission fragments relevant for reactor antineutrino spectra

DOE PAGES

Fallot, M.; Porta, A.; Meur, L. Le; ...

2017-09-13

Here, the accurate determination of reactor antineutrino spectra remains a very active research topic for which new methods of study have emerged in recent years. Indeed, following the long-recognized reactor anomaly (measured antineutrino deficit in short baseline reactor experiments when compared with spectral predictions), the three international reactor neutrino experiments Double Chooz, Daya Bay and Reno have recently demonstrated the existence of spectral distortions in their measurements with respect to the same predictions. These spectral predictions were obtained through the conversion of integral beta-energy spectra obtained at the ILL research reactor. Several studies have shown that the underlying nuclear physicsmore » required for the conversion of these spectra into antineutrino spectra is not totally understood. An alternative to such converted spectra is a complementary approach that consists of determining the antineutrino spectrum by means of the measurement and processing of nuclear data. The beta properties of some key fission products suffer from the pandemonium effect which can be circumvented by the use of the Total Absorption Gamma-ray Spectroscopy technique (TAGS). The two main contributors to the Pressurized Water Reactor antineutrino spectrum in the region where the spectral distortion has been observed are 92Rb and 142Cs, which have been measured at the radioactive beam facility of the University of Jyvaskyla in two TAGS experiments. We present the results of the analysis of the TAGS measurements of the β-decay properties of 92Rb along with preliminary results on 142Cs and report on the measurements already performed.« less

Total absorption spectroscopy of fission fragments relevant for reactor antineutrino spectra

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

Fallot, M.; Porta, A.; Meur, L. Le

Here, the accurate determination of reactor antineutrino spectra remains a very active research topic for which new methods of study have emerged in recent years. Indeed, following the long-recognized reactor anomaly (measured antineutrino deficit in short baseline reactor experiments when compared with spectral predictions), the three international reactor neutrino experiments Double Chooz, Daya Bay and Reno have recently demonstrated the existence of spectral distortions in their measurements with respect to the same predictions. These spectral predictions were obtained through the conversion of integral beta-energy spectra obtained at the ILL research reactor. Several studies have shown that the underlying nuclear physicsmore » required for the conversion of these spectra into antineutrino spectra is not totally understood. An alternative to such converted spectra is a complementary approach that consists of determining the antineutrino spectrum by means of the measurement and processing of nuclear data. The beta properties of some key fission products suffer from the pandemonium effect which can be circumvented by the use of the Total Absorption Gamma-ray Spectroscopy technique (TAGS). The two main contributors to the Pressurized Water Reactor antineutrino spectrum in the region where the spectral distortion has been observed are 92Rb and 142Cs, which have been measured at the radioactive beam facility of the University of Jyvaskyla in two TAGS experiments. We present the results of the analysis of the TAGS measurements of the β-decay properties of 92Rb along with preliminary results on 142Cs and report on the measurements already performed.« less

Advantages of liquid fluoride thorium reactor in comparison with light water reactor

NASA Astrophysics Data System (ADS)

Bahri, Che Nor Aniza Che Zainul; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

2015-04-01

Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

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

Energy from nuclear fission()

NASA Astrophysics Data System (ADS)

Ripani, M.

2015-08-01

The main features of nuclear fission as physical phenomenon will be revisited, emphasizing its peculiarities with respect to other nuclear reactions. Some basic concepts underlying the operation of nuclear reactors and the main types of reactors will be illustrated, including fast reactors, showing the most important differences among them. The nuclear cycle and radioactive-nuclear-waste production will be also discussed, along with the perspectives offered by next generation nuclear assemblies being proposed. The current situation of nuclear power in the world, its role in reducing carbon emission and the available resources will be briefly illustrated.

Assessment of environmental public exposure from a hypothetical nuclear accident for Unit-1 Bushehr nuclear power plant.

PubMed

Sohrabi, M; Ghasemi, M; Amrollahi, R; Khamooshi, C; Parsouzi, Z

2013-05-01

Unit-1 of the Bushehr nuclear power plant (BNPP-1) is a VVER-type reactor with 1,000-MWe power constructed near Bushehr city at the coast of the Persian Gulf, Iran. The reactor has been recently operational to near its full power. The radiological impact of nuclear power plant (NPP) accidents is of public concern, and the assessment of radiological consequences of any hypothetical nuclear accident on public exposure is vital. The hypothetical accident scenario considered in this paper is a design-basis accident, that is, a primary coolant leakage to the secondary circuit. This scenario was selected in order to compare and verify the results obtained in the present paper with those reported in the Final Safety Analysis Report (FSAR 2007) of the BNPP-1 and to develop a well-proven methodology that can be used to study other and more severe hypothetical accident scenarios for this reactor. In the present study, the version 2.01 of the PC COSYMA code was applied. In the early phase of the accidental releases, effective doses (from external and internal exposures) as well as individual and collective doses (due to the late phase of accidental releases) were evaluated. The surrounding area of the BNPP-1 within a radius of 80 km was subdivided into seven concentric rings and 16 sectors, and distribution of population and agricultural products was calculated for this grid. The results show that during the first year following the modeled hypothetical accident, the effective doses do not exceed the limit of 5 mSv, for the considered distances from the BNPP-1. The results obtained in this study are in good agreement with those in the FSAR-2007 report. The agreement obtained is in light of many inherent uncertainties and variables existing in the two modeling procedures applied and proves that the methodology applied here can also be used to model other severe hypothetical accident scenarios of the BNPP-1 such as a small and large break in the reactor coolant system as well as beyond design-basis accidents. Such scenarios are planned to be studied in the near future, for this reactor.

Space nuclear power systems; Proceedings of the 8th Symposium, Albuquerque, NM, Jan. 6-10, 1991. Pts. 1-3

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.; Hoover, Mark D.

1991-07-01

The present conference discusses NASA mission planning for space nuclear power, lunar mission design based on nuclear thermal rockets, inertial-electrostatic confinement fusion for space power, nuclear risk analysis of the Ulysses mission, the role of the interface in refractory metal alloy composites, an advanced thermionic reactor systems design code, and space high power nuclear-pumped lasers. Also discussed are exploration mission enhancements with power-beaming, power requirement estimates for a nuclear-powered manned Mars rover, SP-100 reactor design, safety, and testing, materials compatibility issues for fabric composite radiators, application of the enabler to nuclear electric propulsion, orbit-transfer with TOPAZ-type power sources, the thermoelectric properties of alloys, ruthenium silicide as a promising thermoelectric material, and innovative space-saving device for high-temperature piping systems. The second volume of this conference discusses engine concepts for nuclear electric propulsion, nuclear technologies for human exploration of the solar system, dynamic energy conversion, direct nuclear propulsion, thermionic conversion technology, reactor and power system control, thermal management, thermionic research, effects of radiation on electronics, heat-pipe technology, radioisotope power systems, and nuclear fuels for power reactors. The third volume discusses space power electronics, space nuclear fuels for propulsion reactors, power systems concepts, space power electronics systems, the use of artificial intelligence in space, flight qualifications and testing, microgravity two-phase flow, reactor manufacturing and processing, and space and environmental effects. (For individual items see A93-13752 to A93-13937)

10 CFR 140.12 - Amount of financial protection required for other reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Amount of financial protection required for other reactors... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000...

10 CFR 140.12 - Amount of financial protection required for other reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Amount of financial protection required for other reactors... reactors. (a) Each licensee is required to have and maintain financial protection for each nuclear reactor... of financial protection required for any nuclear reactor under this section be less than $4,500,000...

World Energy Data System (WENDS). Volume XI. Nuclear fission program summaries

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

Not Available

1979-06-01

Brief management and technical summaries of nuclear fission power programs are presented for nineteen countries. The programs include the following: fuel supply, resource recovery, enrichment, fuel fabrication, light water reactors, heavy water reactors, gas cooled reactors, breeder reactors, research and test reactors, spent fuel processing, waste management, and safety and environment. (JWR)

Fast Breeder Reactors in Sweden: Vision and Reality.

PubMed

Fjaestad, Maja

2015-01-01

The fast breeder is a type of nuclear reactor that aroused much attention in the 1950s and '60s. Its ability to produce more nuclear fuel than it consumes offered promises of cheap and reliable energy. Sweden had advanced plans for a nuclear breeder program, but canceled them in the middle of the 1970s with the rise of nuclear skepticism. The article investigates the nuclear breeder as a technological vision. The nuclear breeder reactor is an example of a technological future that did not meet its industrial expectations. But that does not change the fact that the breeder was an influential technology. Decisions about the contemporary reactors were taken with the idea that in a foreseeable future they would be replaced with the efficient breeder. The article argues that general themes in the history of the breeder reactor can deepen our understanding of the mechanisms behind technological change.

75 FR 62610 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Plant...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-10-12

...: Draft Regulatory Guide DG-1237, ``Guidance on Making Changes to Emergency Plans for Nuclear Power Reactors,'' Interim Staff Guidance (ISG) NSIR/DPR-ISG-01, ``Emergency Planning for Nuclear Power Plants... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS...

Nuclear waste disposal utilizing a gaseous core reactor

NASA Technical Reports Server (NTRS)

Paternoster, R. R.

1975-01-01

The feasibility of a gaseous core nuclear reactor designed to produce power to also reduce the national inventories of long-lived reactor waste products through nuclear transmutation was examined. Neutron-induced transmutation of radioactive wastes is shown to be an effective means of shortening the apparent half life.

77 FR 39521 - Application for a License To Export Nuclear Reactor Major Components and Equipment

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-03

... LLC reactor coolant equipment for four constructing four plant May 14, 2012 pumps with motors, APR1400... Emirates. XR176 monitoring and plant in Braka. 110060011 control equipment, auxiliary equipment and... NUCLEAR REGULATORY COMMISSION Application for a License To Export Nuclear Reactor Major Components...

10 CFR 725.15 - Requirements for approval of applications.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Secret Restricted Data in C-91, Nuclear Reactors for Rocket Propulsion, will be approved only if the... capable of making a contribution to research and development in the field of nuclear reactors for rocket... the field of nuclear reactors for rocket propulsion preparatory to the submission of a research and...

10 CFR 725.15 - Requirements for approval of applications.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Secret Restricted Data in C-91, Nuclear Reactors for Rocket Propulsion, will be approved only if the... capable of making a contribution to research and development in the field of nuclear reactors for rocket... the field of nuclear reactors for rocket propulsion preparatory to the submission of a research and...

10 CFR 725.15 - Requirements for approval of applications.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Secret Restricted Data in C-91, Nuclear Reactors for Rocket Propulsion, will be approved only if the... capable of making a contribution to research and development in the field of nuclear reactors for rocket... the field of nuclear reactors for rocket propulsion preparatory to the submission of a research and...

Small Reactor for Deep Space Exploration

ScienceCinema

none,

2018-06-06

This is the first demonstration of a space nuclear reactor system to produce electricity in the United States since 1965, and an experiment demonstrated the first use of a heat pipe to cool a small nuclear reactor and then harvest the heat to power a Stirling engine at the Nevada National Security Site's Device Assembly Facility confirms basic nuclear reactor physics and heat transfer for a simple, reliable space power system.

10 CFR Appendix N to Part 50 - Standardization of Nuclear Power Plant Designs: Permits To Construct and Licenses To Operate...

Code of Federal Regulations, 2011 CFR

2011-01-01

... Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites N Appendix N... Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites Section 101... nuclear power reactors of essentially the same design to be located at different sites. 1 1 If the design...

10 CFR Appendix N to Part 50 - Standardization of Nuclear Power Plant Designs: Permits To Construct and Licenses To Operate...

Code of Federal Regulations, 2012 CFR

2012-01-01

... Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites N Appendix N... Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites Section 101... nuclear power reactors of essentially the same design to be located at different sites. 1 1 If the design...

10 CFR Appendix N to Part 50 - Standardization of Nuclear Power Plant Designs: Permits To Construct and Licenses To Operate...

Code of Federal Regulations, 2013 CFR

2013-01-01

... Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites N Appendix N... Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites Section 101... nuclear power reactors of essentially the same design to be located at different sites. 1 1 If the design...

The United Arab Emirates Nuclear Program and Proposed U.S. Nuclear Cooperation

DTIC Science & Technology

2009-12-23

reactors deployed” in the UAE. Some Members of Congress had welcomed the UAE government’s stated commitments not to pursue proliferation-sensitive...for the planned nuclear reactor or on handling spent reactor fuel. (...continued) May...firms and the UAE related to the UAE’s proposed nuclear program has already taken place. In August 2008, Virginia’s Thorium Power Ltd. signed two

A brief history of design studies on innovative nuclear reactors

NASA Astrophysics Data System (ADS)

Sekimoto, Hiroshi

2014-09-01

In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970's the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980's the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

Innovations in Nuclear Infrastructure and Education

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

John Bernard

The decision to implement the Innovation in Nuclear Infrastructure and Engineering Program (INIE) was an important first step towards ensuring that the United States preserves its worldwide leadership role in the field of nuclear science and engineering. Prior to INIE, university nuclear science and engineering programs were waning, undergraduate student enrollment was down, university research reactors were being shut down, while others faced the real possibility of closure. For too long, cutting edge research in the areas of nuclear medicine, neutron scattering, radiochemistry, and advanced materials was undervalued and therefore underfunded. The INIE program corrected this lapse in focus andmore » direction and started the process of drawing a new blueprint with positive goals and objectives that supports existing as well the next generation of educators, students and researchers.« less

Nuclear power in the 21st century: Challenges and possibilities.

PubMed

Horvath, Akos; Rachlew, Elisabeth

2016-01-01

The current situation and possible future developments for nuclear power--including fission and fusion processes--is presented. The fission nuclear power continues to be an essential part of the low-carbon electricity generation in the world for decades to come. There are breakthrough possibilities in the development of new generation nuclear reactors where the life-time of the nuclear waste can be reduced to some hundreds of years instead of the present time-scales of hundred thousand of years. Research on the fourth generation reactors is needed for the realisation of this development. For the fast nuclear reactors, a substantial research and development effort is required in many fields--from material sciences to safety demonstration--to attain the envisaged goals. Fusion provides a long-term vision for an efficient energy production. The fusion option for a nuclear reactor for efficient production of electricity has been set out in a focussed European programme including the international project of ITER after which a fusion electricity DEMO reactor is envisaged.

76 FR 65541 - Assuring the Availability of Funds for Decommissioning Nuclear Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2011-10-21

... NUCLEAR REGULATORY COMMISSION [NRC-2009-0263] Assuring the Availability of Funds for Decommissioning Nuclear Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Regulatory guide; issuance. SUMMARY: The U.S. Nuclear Regulatory Commission (NRC or Commission) is issuing a revision to Regulatory...

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

Grabaskas, David; Bucknor, Matthew; Jerden, James

A mechanistic source term (MST) calculation attempts to realistically assess the transport and release of radionuclides from a reactor system to the environment during a specific accident sequence. The U.S. Nuclear Regulatory Commission (NRC) has repeatedly stated its expectation that advanced reactor vendors will utilize an MST during the U.S. reactor licensing process. As part of a project to examine possible impediments to sodium fast reactor (SFR) licensing in the U.S., an analysis was conducted regarding the current capabilities to perform an MST for a metal fuel SFR. The purpose of the project was to identify and prioritize any gapsmore » in current computational tools, and the associated database, for the accurate assessment of an MST. The results of the study demonstrate that an SFR MST is possible with current tools and data, but several gaps exist that may lead to possibly unacceptable levels of uncertainty, depending on the goals of the MST analysis.« less

Space Nuclear Power Plant Pre-Conceptual Design Report, For Information

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

B. Levine

2006-01-27

This letter transmits, for information, the Project Prometheus Space Nuclear Power Plant (SNPP) Pre-Conceptual Design Report completed by the Naval Reactors Prime Contractor Team (NRPCT). This report documents the work pertaining to the Reactor Module, which includes integration of the space nuclear reactor with the reactor radiation shield, energy conversion, and instrumentation and control segments. This document also describes integration of the Reactor Module with the Heat Rejection segment, the Power Conditioning and Distribution subsystem (which comprise the SNPP), and the remainder of the Prometheus spaceship.

76 FR 16842 - Request for a License To Export Reactor Components

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-25

... NUCLEAR REGULATORY COMMISSION Request for a License To Export Reactor Components Pursuant to 10.... Mechanical Corporation. coolant pump 1000 (design) maintenance, and systems, related reactors. operation of AP- equipment, and 1000 (design) spare parts. nuclear reactors. February 10, 2011 February 23, 2011...

Reduce, reuse and recycle: a green solution to Canada's medical isotope shortage.

PubMed

Galea, R; Ross, C; Wells, R G

2014-05-01

Due to the unforeseen maintenance issues at the National Research Universal (NRU) reactor at Chalk River and coincidental shutdowns of other international reactors, a global shortage of medical isotopes (in particular technetium-99m, Tc-99m) occurred in 2009. The operation of these research reactors is expensive, their age creates concerns about their continued maintenance and the process results in a large amount of long-lived nuclear waste, whose storage cost has been subsidized by governments. While the NRU has since revived its operations, it is scheduled to cease isotope production in 2016. The Canadian government created the Non-reactor based medical Isotope Supply Program (NISP) to promote research into alternative methods for producing medical isotopes. The NRC was a member of a collaboration looking into the use of electron linear accelerators (LINAC) to produce molybdenum-99 (Mo-99), the parent isotope of Tc-99m. This paper outlines NRC's involvement in every step of this process, from the production, chemical processing, recycling and preliminary animal studies to demonstrate the equivalence of LINAC Tc-99m with the existing supply. This process stems from reusing an old idea, reduces the nuclear waste to virtually zero and recycles material to create a green solution to Canada's medical isotope shortage. © 2013 Published by Elsevier Ltd.

Heat Pipe Powered Stirling Conversion for the Demonstration Using Flattop Fission (DUFF) Test

NASA Technical Reports Server (NTRS)

Gibson, Marc A.; Briggs, Maxwell H.; Sanzi, James L.; Brace, Michael H.

2013-01-01

Design concepts for small Fission Power Systems (FPS) have shown that heat pipe cooled reactors provide a passive, redundant, and lower mass option to transfer heat from the fuel to the power conversion system, as opposed to pumped loop designs typically associated with larger FPS. Although many systems have been conceptually designed and a few making it to electrically heated testing, none have been coupled to a real nuclear reactor. A demonstration test named DUFF Demonstration Using Flattop Fission, was planned by the Los Alamos National Lab (LANL) to use an existing criticality experiment named Flattop to provide the nuclear heat source. A team from the NASA Glenn Research Center designed, built, and tested a heat pipe and power conversion system to couple to Flattop with the end goal of making electrical power. This paper will focus on the design and testing performed in preparation for the DUFF test.

Effect of temperature on the fracture toughness in the nuclear reactor pressure vessel steel (SA508-3)

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

Oh, S.W.; Lim, M.B.; Yoon, H.K.

1994-12-31

The elastic-plastic fracture toughness J{sub IC} of the Nuclear Reactor Pressure Vessel Steel (SA508-3) which has high toughness was obtained at three temperatures (room temperature, {minus}20 C, 200 C) using a 1/2 CT specimen. Especially the two methods recommended in ASTM and JSME were compared. It was found that difficulty exists in obtaining J{sub IC} by ASTM R-curve method, while JSME R-curve method yielded good results. The stretched zone width method gave slightly larger J{sub IC} values than those by the R-curve method for SA508-3 steel and the blunting line was not affected by the test temperatures. The relation betweenmore » SZW and J, SZW and J/E and SZW and J/{sigma}{sub ys} before initiation of a stable crack growth in the fracture toughness test at three temperatures is described.« less

Quantity and management of spent fuel from prototype and research reactors in Germany

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

Dorr, Sabine; Bollingerfehr, Wilhelm; Filbert, Wolfgang

Within the scope of an R and D project (project identification number FKZ 02 S 8679) sponsored by BMBF (Federal Ministry of Education and Research), the current state of storage and management of fuel elements from prototype and research reactors was established, and an approach for their future storage/management was developed. The spent fuels from prototype and research reactors in Germany that require disposal were specified and were described in regard to their repository-relevant characteristics. As there are currently no casks licensed for disposal in Germany, descriptions of casks that were considered to be suitable were provided. Based on themore » information provided on the spent fuel from prototype and research reactors and the potential casks, a technical disposal concept was developed. In this context, concepts to integrate the spent fuel from prototype and research reactors into existing disposal concepts for spent fuel from German nuclear power plants and for waste from reprocessing were developed for salt and clay formations. (authors)« less

75 FR 51025 - Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle Technology Subcommittee

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-18

... DEPARTMENT OF ENERGY Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle... meeting. SUMMARY: This notice announces an open meeting of the Reactor and Fuel Cycle Technology (RFCT... back end of the nuclear fuel cycle. The Commission will provide advice and make recommendations on...

Early Program Development

NASA Image and Video Library

1963-01-01

This artist's concept from 1963 shows a proposed NERVA (Nuclear Engine for Rocket Vehicle Application) incorporating the NRX-A1, the first NERVA-type cold flow reactor. The NERVA engine, based on Kiwi nuclear reactor technology, was intended to power a RIFT (Reactor-In-Flight-Test) nuclear stage, for which Marshall Space Flight Center had development responsibility.

75 FR 36648 - Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle Technologies Subcommittee

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-28

... DEPARTMENT OF ENERGY Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle Technologies Subcommittee AGENCY: Office of Nuclear Energy, DOE. ACTION: Notice of open meeting correction. On June 21, 2010, the Department of Energy published a notice announcing an open meeting of the Reactor...

10 CFR 2.340 - Initial decision in certain contested proceedings; immediate effectiveness of initial decisions...

Code of Federal Regulations, 2010 CFR

2010-01-01

... those matters, the Commission, the Director, Office of Nuclear Reactor Regulation or Director, Office of... Nuclear Reactor Regulation, as appropriate, after making the requisite findings, will issue, deny or... acceptance criteria in nuclear power reactor combined licenses. In any initial decision under § 52.103(g) of...

10 CFR 50.63 - Loss of all alternating current power.

Code of Federal Regulations, 2014 CFR

2014-01-01

... information defined below to the Director of the Office of Nuclear Reactor Regulation by April 17, 1989. For... Office of Nuclear Reactor Regulation, by 270 days after the date of license issuance. For each light... accordance with paragraph (c)(3) of this section, submit to the Director of the Office of Nuclear Reactor...

10 CFR 2.337 - Evidence at a hearing.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of... the proceeding under subpart A of part 51 of this chapter by the Director, Office of Nuclear Reactor... 10 Energy 1 2014-01-01 2014-01-01 false Evidence at a hearing. 2.337 Section 2.337 Energy NUCLEAR...

10 CFR 2.340 - Initial decision in certain contested proceedings; immediate effectiveness of initial decisions...

Code of Federal Regulations, 2011 CFR

2011-01-01

... those matters, the Commission, the Director, Office of Nuclear Reactor Regulation or Director, Office of... Nuclear Reactor Regulation, as appropriate, after making the requisite findings, will issue, deny or... acceptance criteria in nuclear power reactor combined licenses. In any initial decision under § 52.103(g) of...

10 CFR 50.63 - Loss of all alternating current power.

Code of Federal Regulations, 2012 CFR

2012-01-01

... information defined below to the Director of the Office of Nuclear Reactor Regulation by April 17, 1989. For... Office of Nuclear Reactor Regulation, by 270 days after the date of license issuance. For each light... accordance with paragraph (c)(3) of this section, submit to the Director of the Office of Nuclear Reactor...

10 CFR 50.63 - Loss of all alternating current power.

Code of Federal Regulations, 2013 CFR

2013-01-01

... information defined below to the Director of the Office of Nuclear Reactor Regulation by April 17, 1989. For... Office of Nuclear Reactor Regulation, by 270 days after the date of license issuance. For each light... accordance with paragraph (c)(3) of this section, submit to the Director of the Office of Nuclear Reactor...

10 CFR 2.340 - Initial decision in certain contested proceedings; immediate effectiveness of initial decisions...

Code of Federal Regulations, 2012 CFR

2012-01-01

... those matters, the Commission, the Director, Office of Nuclear Reactor Regulation or Director, Office of... Nuclear Reactor Regulation, as appropriate, after making the requisite findings, will issue, deny or... acceptance criteria in nuclear power reactor combined licenses. In any initial decision under § 52.103(g) of...

10 CFR 2.337 - Evidence at a hearing.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of... the proceeding under subpart A of part 51 of this chapter by the Director, Office of Nuclear Reactor... 10 Energy 1 2013-01-01 2013-01-01 false Evidence at a hearing. 2.337 Section 2.337 Energy NUCLEAR...

10 CFR 50.63 - Loss of all alternating current power.

Code of Federal Regulations, 2011 CFR

2011-01-01

... information defined below to the Director of the Office of Nuclear Reactor Regulation by April 17, 1989. For... Office of Nuclear Reactor Regulation, by 270 days after the date of license issuance. For each light... accordance with paragraph (c)(3) of this section, submit to the Director of the Office of Nuclear Reactor...

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)

The rate of decay of fresh fission products from a nuclear reactor

NASA Astrophysics Data System (ADS)

Dolan, David J.

Determining the rate of decay of fresh fission products from a nuclear reactor is complex because of the number of isotopes involved, different types of decay, half-lives of the isotopes, and some isotopes decay into other radioactive isotopes. Traditionally, a simplified rule of 7s and 10s is used to determine the dose rate from nuclear weapons and can be to estimate the dose rate from fresh fission products of a nuclear reactor. An experiment was designed to determine the dose rate with respect to time from fresh fission products of a nuclear reactor. The experiment exposed 0.5 grams of unenriched Uranium to a fast and thermal neutron flux from a TRIGA Research Reactor (Lakewood, CO) for ten minutes. The dose rate from the fission products was measured by four Mirion DMC 2000XB electronic personal dosimeters over a period of six days. The resulting dose rate following a rule of 10s: the dose rate of fresh fission products from a nuclear reactor decreases by a factor of 10 for every 10 units of time.

Radiation chemistry for modern nuclear energy development

NASA Astrophysics Data System (ADS)

Chmielewski, Andrzej G.; Szołucha, Monika M.

2016-07-01

Radiation chemistry plays a significant role in modern nuclear energy development. Pioneering research in nuclear science, for example the development of generation IV nuclear reactors, cannot be pursued without chemical solutions. Present issues related to light water reactors concern radiolysis of water in the primary circuit; long-term storage of spent nuclear fuel; radiation effects on cables and wire insulation, and on ion exchangers used for water purification; as well as the procedures of radioactive waste reprocessing and storage. Radiation effects on materials and enhanced corrosion are crucial in current (II/III/III+) and future (IV) generation reactors, and in waste management, deep geological disposal and spent fuel reprocessing. The new generation of reactors (III+ and IV) impose new challenges for radiation chemists due to their new conditions of operation and the usage of new types of coolant. In the case of the supercritical water-cooled reactor (SCWR), water chemistry control may be the key factor in preventing corrosion of reactor structural materials. This paper mainly focuses on radiation effects on long-term performance and safety in the development of nuclear power plants.

Supplying the nuclear arsenal: Production reactor technology, management, and policy, 1942--1992

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

Carlisle, R.P.; Zenzen, J.M.

1994-01-01

This book focuses on the lineage of America`s production reactors, those three at Hanford and their descendants, the reactors behind America`s nuclear weapons. The work will take only occasional sideways glances at the collateral lines of descent, the reactor cousins designed for experimental purposes, ship propulsion, and electric power generation. Over the decades from 1942 through 1992, fourteen American production reactors made enough plutonium to fuel a formidable arsenal of more than twenty thousand weapons. In the last years of that period, planners, nuclear engineers, and managers struggled over designs for the next generation of production reactors. The story ofmore » fourteen individual machines and of the planning effort to replace them might appear relatively narrow. Yet these machines lay at the heart of the nation`s nuclear weapons complex. The story of these machines is the story of arming the winning weapon, supplying the nuclear arms race. This book is intended to capture the history of the first fourteen production reactors, and associated design work, in the face of the end of the Cold War.« less

Nuclear reactors built, being built, or planned, 1994

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

NONE

1995-07-01

This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1994. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: a commercial reactor locator map and tables of the characteristicmore » and statistical data that follow; a table of abbreviations; tables of data for reactors operating, being built, or planned; and tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company -- working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).« less

Nuclear reactors built, being built, or planned: 1995

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

NONE

1996-08-01

This report contains unclassified information about facilities built, being built, or planned in the US for domestic use or export as of December 31, 1995. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristicmore » and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company--working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).« less

Reactor experiments to study luminescence of He-Ne and He-Kr gaseous mixtures, excited by the products of 6Li (n, α) 3H nuclear reaction

NASA Astrophysics Data System (ADS)

Batyrbekov, E. G.; Gordienko, Yu. N.; Barsukov, N. I.; Ponkratov, Yu. V.; Kulsartov, T. V.; Khassenov, M. U.; Zaurbekova, Zh. A.; Tulubayev, Ye. Y.; Samarkhanov, K. K.

2018-04-01

The spectral studies of optical radiation of gaseous mixtures are of interest for solving problems associated with finding gaseous media with high energy conversion efficiency of nuclear reactions into the energy of laser or spontaneous emission [1, 2]. Such media can be used to extract energy from nuclear and fusion reactors in the form of optical radiation, and also to control and adjust the nuclear reactors parameters. This paper presents the preliminary results of the reactor experiments to study the spectral-luminescent properties of gas mixtures (based on He, Ne and Kr noble gases) excited by the products of 6Li(n,α)3H nuclear reaction at different levels of the stationary power of the IVG.1M reactor.

Control console replacement at the WPI Reactor

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

Not Available

1992-01-01

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« less

Advanced Demonstration and Test Reactor Options Study

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

Petti, David Andrew; Hill, R.; Gehin, J.

Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power’s share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercializationmore » of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy’s (DOE’s) broader commitment to pursuing an “all of the above” clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate “advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy”. Advanced reactors are defined in this study as reactors that use coolants other than water. Advanced reactor technologies have the potential to expand the energy applications, enhance the competitiveness, and improve the sustainability of nuclear energy.« less

76 FR 79229 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-21

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Notice of Meeting In... Advisory Committee on Reactor Safeguards (ACRS) will hold a meeting on January 19-20, 2012, 11545 Rockville... Cooling Systems for Light- Water Nuclear Power Reactors'' (Open)--The Committee will hear presentations by...

77 FR 16270 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-20

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0070] Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft..., ``Updated Aging Management Criteria for PWR Reactor Vessel Internal Components.'' This draft LR-ISG revises...

77 FR 23513 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-04-19

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0070] Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft...-ISG), LR-ISG-2011-04, ``Updated Aging Management Criteria for PWR Reactor Vessel Internal Components...

NUCLEAR REACTOR

DOEpatents

Treshow, M.

1961-09-01

A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

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.

Analysis of the TREAT LEU Conceptual Design

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

Connaway, H. M.; Kontogeorgakos, D. C.; Papadias, D. D.

2016-03-01

Analyses were performed to evaluate the performance of the low enriched uranium (LEU) conceptual design fuel for the conversion of the Transient Reactor Test Facility (TREAT) from its current highly enriched uranium (HEU) fuel. TREAT is an experimental nuclear reactor designed to produce high neutron flux transients for the testing of reactor fuels and other materials. TREAT is currently in non-operational standby, but is being restarted under the U.S. Department of Energy’s Resumption of Transient Testing Program. The conversion of TREAT is being pursued in keeping with the mission of the Department of Energy National Nuclear Security Administration’s Material Managementmore » and Minimization (M3) Reactor Conversion Program. The focus of this study was to demonstrate that the converted LEU core is capable of maintaining the performance of the existing HEU core, while continuing to operate safely. Neutronic and thermal hydraulic simulations have been performed to evaluate the performance of the LEU conceptual-design core under both steady-state and transient conditions, for both normal operation and reactivity insertion accident scenarios. In addition, ancillary safety analyses which were performed for previous LEU design concepts have been reviewed and updated as-needed, in order to evaluate if the converted LEU core will function safely with all existing facility systems. Simulations were also performed to evaluate the detailed behavior of the UO 2-graphite fuel, to support future fuel manufacturing decisions regarding particle size specifications. The results of these analyses will be used in conjunction with work being performed at Idaho National Laboratory and Los Alamos National Laboratory, in order to develop the Conceptual Design Report project deliverable.« less

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.

Measurement instruments for automatically monitoring the water chemistry of reactor coolant at nuclear power stations equipped with VVER reactors. Selection of measurement instruments and experience gained from their operation at Russian and foreign NPSs

NASA Astrophysics Data System (ADS)

Ivanov, Yu. A.

2007-12-01

An analytical review is given of Russian and foreign measurement instruments employed in a system for automatically monitoring the water chemistry of the reactor coolant circuit and used in the development of projects of nuclear power stations equipped with VVER-1000 reactors and the nuclear station project AES 2006. The results of experience gained from the use of such measurement instruments at nuclear power stations operating in Russia and abroad are presented.

CIRFT Data Update and Data Analyses for Spent Nuclear Fuel Vibration Reliability Study

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

Wang, Jy-An John; Wang, Hong

The objective of this research is to collect experimental data on spent nuclear fuel (SNF) from pressurized water reactors (PWRs), including the H. B. Robinson Nuclear Power Station (HBR), Catawba Nuclear Station, North Anna Nuclear Power Station (NA), and the Limerick Nuclear Power Station (LMK) boiling water reactor (BWR).

78 FR 37591 - Entergy Nuclear Operations, Inc., Entergy Nuclear Indian Point Unit 2, LLC, Issuance of Director...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-21

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0284; Docket No. 50-247; License No. DPR-26] Entergy Nuclear Operations, Inc., Entergy Nuclear Indian Point Unit 2, LLC, Issuance of Director's Decision Notice is hereby given that the Deputy Director, Reactor Safety Programs, Office of Nuclear Reactor...

Space nuclear reactors — A post-operational disposal strategy

NASA Astrophysics Data System (ADS)

Angelo, Joseph A.; Buden, David

If 100-kWe and multimegawatt-electric class space nuclear reactors are to play a significant role in humanity's push into cislunar and heliocentric space in the next millennium, the obvious advantages of space nuclear power plants should not be denied to space mission planners due to a failure to develop internationally-acceptable post-operational disposal strategies for spent reactor cores. This is true whether the space reactor has shut down at the end of its normal mission lifetime or in response to an onboard system failure/emergency which causes a premature mission termination. Up until now the great majority of aerospace nuclear safety efforts have concentrated on prelaunch, launch and reactor startup activities. In fact, with the exception of the development of the "nuclear safe orbit" (NSO) concept, little technical attention has yet been given to the post-operational disposal of future space reactors. This paper describes the technical alternatives available for the safe, acceptable disposal of space reactors that could be used in a wide variety of space applications in the 21st Century. Post-operational core radioactivity levels for typical advanced design (hundred kWe-class) space reactors are presented as a function of decay time and contrasted to the spent core radionuclide inventory of the SNAP-10A system, the only nuclear reactor operated in space by the United States. The role of a permanent space station, smart robotic systems, and an operating lunar base in support of spent core disposal strategies is also presented, including use of a selected portion of the lunar surface as an internationally-designated spent reactor core repository.

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24

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

Block, R.C.; Feiner, F.

1995-09-01

Technical papers accepted for presentation at the Seventh International Topical Meeting on Nuclear Reactor Thermal-Hydraulics are included in the present Proceedings. Except for the invited papers in the plenary session, all other papers are contributed papers. The topics of the meeting encompass all major areas of nuclear thermal-hydraulics, including analytical and experimental works on the fundamental mechanisms of fluid flow and heat transfer, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Because of the complex nature of nuclear reactors and power plants, several papers dealmore » with the combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. The participation in the conference by the authors from several countries and four continents makes the Proceedings a comprehensive review of the recent progress in the field of nuclear reactor thermal-hydraulics worldwide. Individual papers have been cataloged separately.« less

Fuel supply of nuclear power industry with the introduction of fast reactors

NASA Astrophysics Data System (ADS)

Muraviev, E. V.

2014-12-01

The results of studies conducted for the validation of the updated development strategy for nuclear power industry in Russia in the 21st century are presented. Scenarios with different options for the reprocessing of spent fuel of thermal reactors and large-scale growth of nuclear power industry based on fast reactors of inherent safety with a breeding ratio of ˜1 in a closed nuclear fuel cycle are considered. The possibility of enhanced fuel breeding in fast reactors is also taken into account in the analysis. The potential to establish a large-scale nuclear power industry that covers 100% of the increase in electric power requirements in Russia is demonstrated. This power industry may be built by the end of the century through the introduction of fast reactors (replacing thermal ones) with a gross uranium consumption of up to ˜1 million t and the termination of uranium mining even if the reprocessing of spent fuel of thermal reactors is stopped or suffers a long-term delay.

10 CFR 1.32 - Office of the Executive Director for Operations.

Code of Federal Regulations, 2014 CFR

2014-01-01

... of Nuclear Reactor Regulation, the Office of New Reactors, the Office of Nuclear Material Safety and... Section 1.32 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION... Nuclear Regulatory Research, the Office of Nuclear Security and Incident Response, and the NRC Regional...

10 CFR 1.32 - Office of the Executive Director for Operations.

Code of Federal Regulations, 2012 CFR

2012-01-01

... of Nuclear Reactor Regulation, the Office of New Reactors, the Office of Nuclear Material Safety and... Section 1.32 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION... Nuclear Regulatory Research, the Office of Nuclear Security and Incident Response, and the NRC Regional...

10 CFR 1.32 - Office of the Executive Director for Operations.

Code of Federal Regulations, 2011 CFR

2011-01-01

... of Nuclear Reactor Regulation, the Office of New Reactors, the Office of Nuclear Material Safety and... Section 1.32 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION... Nuclear Regulatory Research, the Office of Nuclear Security and Incident Response, and the NRC Regional...

10 CFR 1.32 - Office of the Executive Director for Operations.

Code of Federal Regulations, 2013 CFR

2013-01-01

... of Nuclear Reactor Regulation, the Office of New Reactors, the Office of Nuclear Material Safety and... Section 1.32 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION... Nuclear Regulatory Research, the Office of Nuclear Security and Incident Response, and the NRC Regional...

A brief history of design studies on innovative nuclear reactors

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

Sekimoto, Hiroshi, E-mail: hsekimot@gmail.com

2014-09-30

In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970’s the TMI accident occurred and many nuclear reactor contracts were cancelled in USAmore » and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980’s the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.« less

75 FR 70042 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-16

... NUCLEAR REGULATORY COMMISSION [Docket Nos (Redacted), License Nos (Redacted), EA (Redacted); NRC- 2010-0351] In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order Modifying License (Effective Immediately) I. The licensees identified in...

75 FR 79423 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-20

... NUCLEAR REGULATORY COMMISSION [Docket Nos. (Redacted), License Nos.: (Redacted), EA (Redacted); NRC- 2010-0351] In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order Modifying License (Effective Immediately) I The licensees identified in...

Determination of parameters of a nuclear reactor through noise measurements

DOEpatents

Cohn, C.E.

1975-07-15

A method of measuring parameters of a nuclear reactor by noise measurements is described. Noise signals are developed by the detectors placed in the reactor core. The polarity coincidence between the noise signals is used to develop quantities from which various parameters of the reactor can be calculated. (auth)

Liquid level, void fraction, and superheated steam sensor for nuclear-reactor cores. [PWR; BWR

DOEpatents

Tokarz, R.D.

1981-10-27

This disclosure relates to an apparatus for monitoring the presence of coolant in liquid or mixed liquid and vapor, and superheated gaseous phases at one or more locations within an operating nuclear reactor core, such as pressurized water reactor or a boiling water reactor.

78 FR 71675 - Update of the Office of Nuclear Reactor Regulation's Electronic Operating Reactor Correspondence

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-29

... correspondence to addressees and subscribers through a computer-based email distribution system. Since then, the... Electronic Operating Reactor Correspondence The U.S. Nuclear Regulatory Commission (NRC) is issuing this... available operating reactor licensing correspondence, effective December 9, 2013. Official agency records...

Effects of an Advanced Reactor’s Design, Use of Automation, and Mission on Human Operators

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

Jeffrey C. Joe; Johanna H. Oxstrand

The roles, functions, and tasks of the human operator in existing light water nuclear power plants (NPPs) are based on sound nuclear and human factors engineering (HFE) principles, are well defined by the plant’s conduct of operations, and have been validated by years of operating experience. However, advanced NPPs whose engineering designs differ from existing light-water reactors (LWRs) will impose changes on the roles, functions, and tasks of the human operators. The plans to increase the use of automation, reduce staffing levels, and add to the mission of these advanced NPPs will also affect the operator’s roles, functions, and tasks.more » We assert that these factors, which do not appear to have received a lot of attention by the design engineers of advanced NPPs relative to the attention given to conceptual design of these reactors, can have significant risk implications for the operators and overall plant safety if not mitigated appropriately. This paper presents a high-level analysis of a specific advanced NPP and how its engineered design, its plan to use greater levels of automation, and its expanded mission have risk significant implications on operator performance and overall plant safety.« less

Understanding the reaction of nuclear graphite with molecular oxygen: Kinetics, transport, and structural evolution

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

Kane, Joshua J.; Contescu, Cristian I.; Smith, Rebecca E.

A thorough understanding of oxidation is important when considering the health and integrity of graphite components in graphite reactors. For the next generation of graphite reactors, HTGRs specifically, an unlikely air ingress has been deemed significant enough to have made its way into the licensing applications of many international licensing bodies. While a substantial body of literature exists on nuclear graphite oxidation in the presence of molecular oxygen and significant efforts have been made to characterize oxidation kinetics of various grades, the value of existing information is somewhat limited. Often, multiple competing processes, including reaction kinetics, mass transfer, and microstructuralmore » evolution, are lumped together into a single rate expression that limits the ability to translate this information to different conditions. This article reviews the reaction of graphite with molecular oxygen in terms of the reaction kinetics, gas transport, and microstructural evolution of graphite. It also presents the foundations of a model for the graphite-molecular oxygen reaction system that is kinetically independent of graphite grade, and is capable of describing both the bulk and local oxidation rates under a wide range of conditions applicable to air-ingress.« less

Understanding the reaction of nuclear graphite with molecular oxygen: Kinetics, transport, and structural evolution

DOE PAGES

Kane, Joshua J.; Contescu, Cristian I.; Smith, Rebecca E.; ...

2017-06-08

A thorough understanding of oxidation is important when considering the health and integrity of graphite components in graphite reactors. For the next generation of graphite reactors, HTGRs specifically, an unlikely air ingress has been deemed significant enough to have made its way into the licensing applications of many international licensing bodies. While a substantial body of literature exists on nuclear graphite oxidation in the presence of molecular oxygen and significant efforts have been made to characterize oxidation kinetics of various grades, the value of existing information is somewhat limited. Often, multiple competing processes, including reaction kinetics, mass transfer, and microstructuralmore » evolution, are lumped together into a single rate expression that limits the ability to translate this information to different conditions. This article reviews the reaction of graphite with molecular oxygen in terms of the reaction kinetics, gas transport, and microstructural evolution of graphite. It also presents the foundations of a model for the graphite-molecular oxygen reaction system that is kinetically independent of graphite grade, and is capable of describing both the bulk and local oxidation rates under a wide range of conditions applicable to air-ingress.« less

Method for passive cooling liquid metal cooled nuclear reactors, and system thereof

DOEpatents

Hunsbedt, Anstein; Busboom, Herbert J.

1991-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel.

Search for eV sterile neutrinos at a nuclear reactor — the Stereo project

NASA Astrophysics Data System (ADS)

Haser, J.; Stereo Collaboration

2016-05-01

The re-analyses of the reference spectra of reactor antineutrinos together with a revised neutrino interaction cross section enlarged the absolute normalization of the predicted neutrino flux. The tension between previous reactor measurements and the new prediction is significant at 2.7 σ and is known as “reactor antineutrino anomaly”. In combination with other anomalies encountered in neutrino oscillation measurements, this observation revived speculations about the existence of a sterile neutrino in the eV mass range. Mixing of this light sterile neutrino with the active flavours would lead to a modification of the detected antineutrino flux. An oscillation pattern in energy and space could be resolved by a detector at a distance of few meters from a reactor core: the neutrino detector of the Stereo project will be located at about 10 m distance from the ILL research reactor in Grenoble, France. Lengthwise separated in six target cells filled with 2 m3 Gd-loaded liquid scintillator in total, the experiment will search for a position-dependent distortion in the energy spectrum.

REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

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

Nichols, T.; Beals, D.; Sternat, M.

2011-07-18

Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Manymore » research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical analysis. The fuel assembly was modeled using MONTEBURNS(MCNP5/ ORIGEN2.2) and MCNPX/CINDER90. The results from the models have been compared to each other and to the measured data.« less

Coupled IVPs to Investigate a Nuclear Reactor Poison Burn Up

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

Faghihi, F.

2009-09-09

A set of coupled IVPs that describe the change rate of an important poison, in a nuclear reactor, has been written herein. Specifically, in this article, we have focused on the samarium-149 (as a poison) burnup in a desired pressurized water nuclear reactor and its concentration are given using our MATLAB-linked 'solver'.

Developing the European Center of Competence on VVER-Type Nuclear Power Reactors

ERIC Educational Resources Information Center

Geraskin, Nikolay; Pironkov, Lyubomir; Kulikov, Evgeny; Glebov, Vasily

2017-01-01

This paper presents the results of the European educational projects CORONA and CORONA-II which are dedicated to preserving and further developing nuclear knowledge and competencies in the area of VVER-type nuclear power reactors technologies (Water-Water Energetic Reactor, WWER or VVER). The development of the European Center of Competence for…

Coupled IVPs to Investigate a Nuclear Reactor Poison Burn Up

NASA Astrophysics Data System (ADS)

Faghihi, F.; Saidi-Nezhad, M.

2009-09-01

A set of coupled IVPs that describe the change rate of an important poison, in a nuclear reactor, has been written herein. Specifically, in this article, we have focused on the samarium-149 (as a poison) burnup in a desired pressurized water nuclear reactor and its concentration are given using our MATLAB-linked "solver."

10 CFR 2.110 - Filing and administrative action on submittals for standard design approval or early review of...

Code of Federal Regulations, 2012 CFR

2012-01-01

..., Office of New Reactors, or Director, Office of Nuclear Reactor Regulation, as appropriate shall publish... Nuclear Reactor Regulation, as appropriate shall publish in the Federal Register a determination as to... standard design approval or early review of site suitability issues. 2.110 Section 2.110 Energy NUCLEAR...

10 CFR 2.110 - Filing and administrative action on submittals for standard design approval or early review of...

Code of Federal Regulations, 2013 CFR

2013-01-01

..., Office of Nuclear Reactor Regulation, as appropriate shall publish in the Federal Register a notice of... Director, Office of New Reactors or Director, Office of Nuclear Reactor Regulation, as appropriate shall... standard design approval or early review of site suitability issues. 2.110 Section 2.110 Energy NUCLEAR...

10 CFR 2.110 - Filing and administrative action on submittals for standard design approval or early review of...

Code of Federal Regulations, 2014 CFR

2014-01-01

..., Office of Nuclear Reactor Regulation, as appropriate shall publish in the Federal Register a notice of... Director, Office of New Reactors or Director, Office of Nuclear Reactor Regulation, as appropriate shall... standard design approval or early review of site suitability issues. 2.110 Section 2.110 Energy NUCLEAR...

Light Water Reactor Sustainability Program: Integrated Program Plan

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

None, None

Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 60%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to grow by about 24% from 2013 to 2040 . At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license, for a total of 60 years of operation (the oldest commercial plants in the Unitedmore » States reached their 40th anniversary in 2009). Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity for 40- and 60-year license periods. If current operating nuclear power plants do not operate beyond 60 years (and new nuclear plants are not built quickly enough to replace them), the total fraction of generated electrical energy from nuclear power will rapidly decline. That decline will be accelerated if plants are shut down before 60 years of operation. Decisions on extended operation ultimately rely on economic factors; however, economics can often be improved through technical advancements. The U.S. Department of Energy Office of Nuclear Energy’s 2010 Research and Development Roadmap (2010 Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: 1. Develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; 2. Develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; 3. Develop sustainable nuclear fuel cycles; and 4. Understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer-than-initially-licensed lifetime. It has two facets with respect to long-term operations: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the industry to implement technology to exceed the performance of the current labor-intensive business model.« less

75 FR 76498 - Firstenergy Nuclear Operating Company, Davis-Besse Nuclear Power Station; Environmental...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-08

... Company, Davis-Besse Nuclear Power Station; Environmental Assessment And Finding of No Significant Impact... operation of the Davis-Besse Nuclear Power Station, Unit 1 (DBNPS), located in Ottawa County, Ohio. In... the reactor coolant pressure boundary of light-water nuclear power reactors provide adequate margins...

78 FR 24438 - Evaluations of Explosions Postulated To Occur at Nearby Facilities and on Transportation Routes...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-25

... Nearby Facilities and on Transportation Routes Near Nuclear Power Plants AGENCY: Nuclear Regulatory... Nearby Facilities and on Transportation Routes Near Nuclear Power Plants.'' This regulatory guide describes for applicants seeking nuclear power reactor licenses and licensees of nuclear power reactors...

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

Mulder, R. U.; Benneche, P. E.; Hosticka, B.

The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these users institutions is enhanced by the use of the nuclear facilities.

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

Function of university reactors in operator licensing training for nuclear utilities

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

Wicks, F.

1985-11-01

The director of the Division of the US Nuclear Regulatory Commission in generic letter 84-10, dated April 26, 1984, spoke the requirement that applicants for senior reactor operator licenses for power reactors shall have performed then reactor startups. Simulator startups were not acknowledged. Startups performed on a university reactor are acceptable. The content and results of a five-day program combining instruction and experiments with the Rensselaer reactor are summarized.

Reactor operation environmental information document

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

Haselow, J.S.; Price, V.; Stephenson, D.E.

1989-12-01

The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimalmore » impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.« less

Exploratory study of several advanced nuclear-MHD power plant systems.

NASA Technical Reports Server (NTRS)

Williams, J. R.; Clement, J. D.; Rosa, R. J.; Yang, Y. Y.

1973-01-01

In order for efficient multimegawatt closed cycle nuclear-MHD systems to become practical, long-life gas cooled reactors with exit temperatures of about 2500 K or higher must be developed. Four types of nuclear reactors which have the potential of achieving this goal are the NERVA-type solid core reactor, the colloid core (rotating fluidized bed) reactor, the 'light bulb' gas core reactor, and the 'coaxial flow' gas core reactor. Research programs aimed at developing these reactors have progressed rapidly in recent years so that prototype power reactors could be operating by 1980. Three types of power plant systems which use these reactors have been analyzed to determine the operating characteristics, critical parameters and performance of these power plants. Overall thermal efficiencies as high as 80% are projected, using an MHD turbine-compressor cycle with steam bottoming, and slightly lower efficiencies are projected for an MHD motor-compressor cycle.

The scheme for evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Saldikov, I. S.; Ternovykh, M. Yu; Fomichenko, P. A.; Gerasimov, A. S.

2017-01-01

The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of power. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. To solve the closed nuclear fuel modeling tasks REPRORYV code was developed. It simulates the mass flow for nuclides in the closed fuel cycle. This paper presents the results of modeling of a closed nuclear fuel cycle, nuclide flows considering the influence of the uncertainty on the outcome of neutron-physical characteristics of the reactor.

Progress and challenges of nuclear science development in Vietnam - an outlook on the occassion of the 10-th anniversary of the Dalat Nuclear Research Reactor

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

Hien, P.D.

1994-12-31

Over ten years since the commissioning of the Dalat nuclear research reactor a number of nuclear techniques have been developed and applied in Vietnam Manufacturing of radioisotopes and nuclear instruments, development of isotope tracer and nuclear analytical techniques for environmental studies, exploitation of filtered neutron beams, ... have been major activities of reactor utilizations. Efforts made during ten years of reactor operation have resulted also in establishing and sustaining the applications of nuclear techniques in medicine, industry, agriculture, etc. The successes achieved and lessons teamed over the past ten years are discussed illustrating the approaches taken for developing the nuclearmore » science in the conditions of a country having a very low national income and experiencing a transition from a centrally planned to a market-oriented economic system.« less

The Euratom Seventh Framework Programme FP7 (2007-2011)

NASA Astrophysics Data System (ADS)

Garbil, R.

2010-10-01

The objective of the Seventh Euratom Framework Program in the area of nuclear fission and radiation protection is to establish a sound scientific and technical basis to accelerate practical developments of nuclear energy related to resource efficiency, enhancing safety performance, cost-effectiveness and safer management of long-lived radioactive waste. Key cross-cutting topics such as the nuclear fuel cycle, actinide chemistry, risk analysis, safety assessment, even societal and governance issues are linked to the individual technical areas. Research need to explore new scientific and techno- logical opportunities and to respond in a flexible way to new policy needs that arise. The following activities are to be pursued. (a) Management of radioactive waste, research on partitioning and transmutation and/or other concepts aimed at reducing the amount and/or hazard of the waste for disposal; (b) Reactor systems research to underpin the con- tinued safe operation of all relevant types of existing reactor systems (including fuel cycle facilities), life-time extension, development of new advanced safety assessment methodologies and waste-management aspects of future reactor systems; (c) Radiation protection research in particular on the risks from low doses on medical uses and on the management of accidents; (d) Infrastructures and support given to the availability of, and cooperation between, research infrastructures necessary to maintain high standards of technical achievement, innovation and safety in the European nuclear sector and Research Area. (e) Human resources, mobility and training support to be provided for the retention and further development of scientific competence, human capacity through joint training activities in order to guarantee the availability of suitably qualified researchers, engineers and employees in the nuclear sector over the longer term.

THE ECONOMICS OF REPROCESSING vs DIRECT DISPOSAL OF SPENT NUCLEAR FUEL

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

Matthew Bunn; Steve Fetter; John P. Holdren

This report assesses the economics of reprocessing versus direct disposal of spent nuclear fuel. The breakeven uranium price at which reprocessing spent nuclear fuel from existing light-water reactors (LWRs) and recycling the resulting plutonium and uranium in LWRs would become economic is assessed, using central estimates of the costs of different elements of the nuclear fuel cycle (and other fuel cycle input parameters), for a wide range of range of potential reprocessing prices. Sensitivity analysis is performed, showing that the conclusions reached are robust across a wide range of input parameters. The contribution of direct disposal or reprocessing and recyclingmore » to electricity cost is also assessed. The choice of particular central estimates and ranges for the input parameters of the fuel cycle model is justified through a review of the relevant literature. The impact of different fuel cycle approaches on the volume needed for geologic repositories is briefly discussed, as are the issues surrounding the possibility of performing separations and transmutation on spent nuclear fuel to reduce the need for additional repositories. A similar analysis is then performed of the breakeven uranium price at which deploying fast neutron breeder reactors would become competitive compared with a once-through fuel cycle in LWRs, for a range of possible differences in capital cost between LWRs and fast neutron reactors. Sensitivity analysis is again provided, as are an analysis of the contribution to electricity cost, and a justification of the choices of central estimates and ranges for the input parameters. The equations used in the economic model are derived and explained in an appendix. Another appendix assesses the quantities of uranium likely to be recoverable worldwide in the future at a range of different possible future prices.« less

On fundamental quality of fission chain reaction to oppose rapid runaways of nuclear reactors

NASA Astrophysics Data System (ADS)

Kulikov, G. G.; Shmelev, A. N.; Apse, V. A.; Kulikov, E. G.

2017-01-01

It has been shown that the in-hour equation characterizes the barriers and resistibility of fission chain reaction (FCR) against rapid runaways in nuclear reactors. Traditionally, nuclear reactors are characterized by the presence of barriers based on delayed and prompt neutrons. A new barrier based on the reflector neutrons that can occur when the fast reactor core is surrounded by a weakly absorbing neutron reflector with heavy atomic weight was proposed. It has been shown that the safety of this fast reactor is substantially improved, and considerable elongation of prompt neutron lifetime "devalues" the role of delayed neutron fraction as the maximum permissible reactivity for the reactor safety.

Tritium release during nuclear power operation in China.

PubMed

Yang, D J; Chen, X Q; Li, B

2012-06-01

Overviews were evaluated of tritium releases and related doses to the public from airborne and liquid effluents from nuclear power plants on the mainland of China before 2009. The differences between tritium releases from various nuclear power plants were also evaluated. The tritium releases are mainly from liquid pathways for pressurised water reactors, but tritium releases between airborne and liquid effluents are comparable for heavy water reactors. The airborne release from a heavy water reactor is obviously higher than that from a pressurised water reactor.

Nuclear Security: Action May Be Needed to Reassess the Security of NRC-Licensed Research Reactors. Report to the Ranking Member, Subcommittee on National Security and Foreign Affairs, Committee on Oversight and Government Reform, House of Representatives. GAO-08-403

ERIC Educational Resources Information Center

Aloise, Gene

2008-01-01

There are 37 research reactors in the United States, mostly located on college campuses. Of these, 33 reactors are licensed and regulated by the Nuclear Regulatory Commission (NRC). Four are operated by the Department of Energy (DOE) and are located at three national laboratories. Although less powerful than commercial nuclear power reactors,…

Nuclear reactor apparatus

DOEpatents

Wade, Elman E.

1978-01-01

A lifting, rotating and sealing apparatus for nuclear reactors utilizing rotating plugs above the nuclear reactor core. This apparatus permits rotation of the plugs to provide under the plug refueling of a nuclear core. It also provides a means by which positive top core holddown can be utilized. Both of these operations are accomplished by means of the apparatus lifting the top core holddown structure off the nuclear core while stationary, and maintaining this structure in its elevated position during plug rotation. During both of these operations, the interface between the rotating member and its supporting member is sealingly maintained.

Isotopic signature of atmospheric xenon released from light water reactors.

PubMed

Kalinowski, Martin B; Pistner, Christoph

2006-01-01

A global monitoring system for atmospheric xenon radioactivity is being established as part of the International Monitoring System to verify compliance with the Comprehensive Nuclear-Test-Ban Treaty (CTBT). The isotopic activity ratios of (135)Xe, (133m)Xe, (133)Xe and (131m)Xe are of interest for distinguishing nuclear explosion sources from civilian releases. Simulations of light water reactor (LWR) fuel burn-up through three operational reactor power cycles are conducted to explore the possible xenon isotopic signature of nuclear reactor releases under different operational conditions. It is studied how ratio changes are related to various parameters including the neutron flux, uranium enrichment and fuel burn-up. Further, the impact of diffusion and mixing on the isotopic activity ratio variability are explored. The simulations are validated with reported reactor emissions. In addition, activity ratios are calculated for xenon isotopes released from nuclear explosions and these are compared to the reactor ratios in order to determine whether the discrimination of explosion releases from reactor effluents is possible based on isotopic activity ratios.

KERENA safety concept in the context of the Fukushima accident

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

Zacharias, T.; Novotny, C.; Bielor, E.

Within the last three years AREVA NP and E.On KK finalized the basic design of KERENA which is a medium sized innovative boiling water reactor, based on the operational experience of German BWR nuclear power plants (NPPs). It is a generation III reactor design with a net electrical output of about 1250 MW. It combines active safety equipment of service-proven designs with new passive safety components, both safety classified. The passive systems utilize basic laws of physics, such as gravity and natural convection, enabling them to function without electric power. Even actuation of these systems is performed thanks to basicmore » physic laws. The degree of diversity in component and system design, achieved by combining active and passive equipment, results in a very low core damage frequency. The Fukushima accident enhanced the world wide discussion about the safety of operating nuclear power plants. World wide stress tests for operating nuclear power plants are being performed embracing both natural and man made hazards. Beside the assessment of existing power plants, also new designs are analyzed regarding the system response to beyond design base accidents. KERENA's optimal combination of diversified cooling systems (active and passive) allows passing efficiently such tests, with a high level of confidence. This paper describes the passive safety components and the KERENA reactor behavior after a Fukushima like accident. (authors)« less

Handbook explaining the fundamentals of nuclear and atomic physics

NASA Technical Reports Server (NTRS)

Hanlen, D. F.; Morse, W. J.

1969-01-01

Indoctrination document presents nuclear, reactor, and atomic physics in an easy, straightforward manner. The entire subject of nuclear physics including atomic structure ionization, isotopes, radioactivity, and reactor dynamics is discussed.

Reactivity control assembly for nuclear reactor

DOEpatents

Bollinger, Lawrence R.

1984-01-01

Reactivity control assembly for nuclear reactor comprises supports stacked above reactor core for holding control rods. Couplers associated with the supports and a vertically movable drive shaft have lugs at their lower ends for engagement with the supports.

Changing concepts of geologic structure and the problem of siting nuclear reactors: Examples from Washington State

NASA Astrophysics Data System (ADS)

Tabor, R. W.

1986-09-01

The conflict between regulation and healthy evolution of geological science has contributed to the difficulties of siting nuclear reactors. On the Columbia Plateau in Washington, but for conservative design of the Hanford reactor facility, the recognition of the little-understood Olympic-Wallowa lineament as a major, possibly still active structural alinement might have jeopardized the acceptability of the site for nuclear reactors. On the Olympic Peninsula, evolving concepts of compressive structures and their possible recent activity and the current recognition of a subducting Juan de Fuca plate and its potential for generating great earthquakes—both concepts little-considered during initial site selection—may delay final acceptance of the Satsop site. Conflicts of this sort are inevitable but can be accommodated if they are anticipated in the reactor-licensing process. More important, society should be increasing its store of geologic knowledge now, during the current recess in nuclear reactor siting.

Nuclear reactors built, being built, or planned 1996

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

NONE

1997-08-01

This publication contains unclassified information about facilities, built, being built, or planned in the United States for domestic use or export as of December 31, 1996. The Office of Scientific and Technical Information, U.S. Department of Energy, gathers this information annually from Washington headquarters, and field offices of DOE; from the U.S. Nuclear Regulatory Commission (NRC); from the U. S. reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from U.S. and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables ofmore » the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled.« less

Double-clad nuclear fuel safety rod

DOEpatents

McCarthy, William H.; Atcheson, Donald B.; Vaidyanathan, Swaminathan

1984-01-01

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

Double-clad nuclear-fuel safety rod

DOEpatents

McCarthy, W.H.; Atcheson, D.B.

1981-12-30

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

Status of Fuel Development and Manufacturing for Space Nuclear Reactors at BWX Technologies

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

Carmack, W.J.; Husser, D.L.; Mohr, T.C.

2004-02-04

New advanced nuclear space propulsion systems will soon seek a high temperature, stable fuel form. BWX Technologies Inc (BWXT) has a long history of fuel manufacturing. UO2, UCO, and UCx have been fabricated at BWXT for various US and international programs. Recent efforts at BWXT have focused on establishing the manufacturing techniques and analysis capabilities needed to provide a high quality, high power, compact nuclear reactor for use in space nuclear powered missions. To support the production of a space nuclear reactor, uranium nitride has recently been manufactured by BWXT. In addition, analytical chemistry and analysis techniques have been developedmore » to provide verification and qualification of the uranium nitride production process. The fabrication of a space nuclear reactor will require the ability to place an unclad fuel form into a clad structure for assembly into a reactor core configuration. To this end, BWX Technologies has reestablished its capability for machining, GTA welding, and EB welding of refractory metals. Specifically, BWX Technologies has demonstrated GTA welding of niobium flat plate and EB welding of niobium and Nb-1Zr tubing. In performing these demonstration activities, BWX Technologies has established the necessary infrastructure to manufacture UO2, UCx, or UNx fuel, components, and complete reactor assemblies in support of space nuclear programs.« less

2011.2 Revision of the Evaluated Nuclear Data Library (ENDL2011.2)

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

Beck, B.; Descalles, M. A.; Mattoon, C.

LLNL's Computational Nuclear Physics Group and Nuclear Theory and Modeling Group have col- laborated to create the 2011.2 revised release of the Evaluated Nuclear Data Library (ENDL2011.2). ENDL2011.2 is designed to support LLNL's current and future nuclear data needs and will be em- ployed in nuclear reactor, nuclear security and stockpile stewardship simulations with ASC codes. This database is currently the most complete nuclear database for Monte Carlo and deterministic transport of neutrons and charged particles. This library was assembled with strong support from the ASC PEM and Attribution programs, leveraged with support from Campaign 4 and the DOE/O cemore » of Science's US Nuclear Data Program. This document lists the revisions made in ENDL2011.2 compared with the data existing in the original ENDL2011.0 release and the ENDL2011.1-rc4 re- lease candidate of April 2015. These changes are made in parallel with some similar revisions for ENDL2009.2.« less

77 FR 60479 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-03

... Pressurized Water Reactor Spent Fuel in Transportation and Storage Casks AGENCY: Nuclear Regulatory Commission... 3, entitled, ``Burnup Credit in the Criticality Safety Analyses of PWR [Pressurized Water Reactor... water reactor spent nuclear fuel (SNF) in transportation packages and storage casks. SFST-ISG-8...

Nuclear reactor fuel containment safety structure

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

Rosewell, M.P.

A nuclear reactor fuel containment safety structure is disclosed and is shown to include an atomic reactor fuel shield with a fuel containment chamber and exhaust passage means, and a deactivating containment base attached beneath the fuel reactor shield and having exhaust passages, manifold, and fluxing and control material and vessels. 1 claim, 8 figures.

10 CFR 72.210 - General license issued.

Code of Federal Regulations, 2010 CFR

2010-01-01

... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.210 General license issued. A general license is... reactor sites to persons authorized to possess or operate nuclear power reactors under 10 CFR part 50 or...

10 CFR 1.44 - Office of New Reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR... Office of New Reactors. The Office of New Reactors— (a) Develops, promulgates and implements regulations... safeguarding of nuclear reactor facilities licensed under part 52 of this chapter prior to initial commencement...

10 CFR 72.210 - General license issued.

Code of Federal Regulations, 2011 CFR

2011-01-01

... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.210 General license issued. A general license is... reactor sites to persons authorized to possess or operate nuclear power reactors under 10 CFR part 50 or...

10 CFR 1.44 - Office of New Reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR... Office of New Reactors. The Office of New Reactors— (a) Develops, promulgates and implements regulations... safeguarding of nuclear reactor facilities licensed under part 52 of this chapter prior to initial commencement...

Application of Molten Salt Reactor Technology to Nuclear Electric Propulsion Mission

NASA Technical Reports Server (NTRS)

Patton, Bruce; Sorensen, Kirk; Rodgers, Stephen L. (Technical Monitor)

2002-01-01

Nuclear electric propulsion (NEP) and planetary surface power missions require reactors that are lightweight, operationally robust, and scalable in power for widely varying scientific mission objectives. Molten salt reactor technology meets all of these requirements and offers an interesting alternative to traditional gas cooled, liquid metal, and heat pipe space reactors.

10 CFR 1.44 - Office of New Reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR... safeguarding of nuclear reactor facilities licensed under part 52 of this chapter prior to initial commencement... Office of New Reactors. The Office of New Reactors— (a) Develops, promulgates and implements regulations...

10 CFR 1.44 - Office of New Reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 1 2014-01-01 2014-01-01 false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR... safeguarding of nuclear reactor facilities licensed under part 52 of this chapter prior to initial commencement... Office of New Reactors. The Office of New Reactors— (a) Develops, promulgates and implements regulations...

10 CFR 1.44 - Office of New Reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR... safeguarding of nuclear reactor facilities licensed under part 52 of this chapter prior to initial commencement... Office of New Reactors. The Office of New Reactors— (a) Develops, promulgates and implements regulations...

Reactor pressure vessel embrittlement: Insights from neural network modelling

NASA Astrophysics Data System (ADS)

Mathew, J.; Parfitt, D.; Wilford, K.; Riddle, N.; Alamaniotis, M.; Chroneos, A.; Fitzpatrick, M. E.

2018-04-01

Irradiation embrittlement of steel pressure vessels is an important consideration for the operation of current and future light water nuclear reactors. In this study we employ an ensemble of artificial neural networks in order to provide predictions of the embrittlement using two literature datasets, one based on US surveillance data and the second from the IVAR experiment. We use these networks to examine trends with input variables and to assess various literature models including compositional effects and the role of flux and temperature. Overall, the networks agree with the existing literature models and we comment on their more general use in predicting irradiation embrittlement.

Current and anticipated uses of thermal-hydraulic codes in Germany

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

Teschendorff, V.; Sommer, F.; Depisch, F.

1997-07-01

In Germany, one third of the electrical power is generated by nuclear plants. ATHLET and S-RELAP5 are successfully applied for safety analyses of the existing PWR and BWR reactors and possible future reactors, e.g. EPR. Continuous development and assessment of thermal-hydraulic codes are necessary in order to meet present and future needs of licensing organizations, utilities, and vendors. Desired improvements include thermal-hydraulic models, multi-dimensional simulation, computational speed, interfaces to coupled codes, and code architecture. Real-time capability will be essential for application in full-scope simulators. Comprehensive code validation and quantification of uncertainties are prerequisites for future best-estimate analyses.

The Birth of Nuclear-Generated Electricity

DOE R&D Accomplishments Database

1999-09-01

The Experimental Breeder Reactor-I (EBR-I), built in Idaho in 1949, generated the first usable electricity from nuclear power on December 20, 1951. More importantly, the reactor was used to prove that it was possible to create more nuclear fuel in the reactor than it consumed during operation -- fuel breeding. The EBR-I facility is now a National Historic Landmark open to the public.

Turbulence coefficients and stability studies for the coaxial flow or dissimiliar fluids. [gaseous core nuclear reactors

NASA Technical Reports Server (NTRS)

Weinstein, H.; Lavan, Z.

1975-01-01

Analytical investigations of fluid dynamics problems of relevance to the gaseous core nuclear reactor program are presented. The vortex type flow which appears in the nuclear light bulb concept is analyzed along with the fluid flow in the fuel inlet region for the coaxial flow gaseous core nuclear reactor concept. The development of numerical methods for the solution of the Navier-Stokes equations for appropriate geometries is extended to the case of rotating flows and almost completes the gas core program requirements in this area. The investigations demonstrate that the conceptual design of the coaxial flow reactor needs further development.

Evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Tikhomirov, Georgy; Ternovykh, Mikhail; Saldikov, Ivan; Fomichenko, Peter; Gerasimov, Alexander

2017-09-01

The strategy of the development of nuclear power in Russia provides for use of fast power reactors in closed nuclear fuel cycle. The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of energy. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. The closed nuclear fuel cycle concept of the PRORYV assumes self-supplied mode of operation with fuel regeneration by neutron capture reaction in non-enriched uranium, which is used as a raw material. Operating modes of reactors and its characteristics should be chosen so as to provide the self-sufficient mode by using of fissile isotopes while refueling by depleted uranium and to support this state during the entire period of reactor operation. Thus, the actual issue is modeling fuel handling processes. To solve these problems, the code REPRORYV (Recycle for PRORYV) has been developed. It simulates nuclide streams in non-reactor stages of the closed fuel cycle. At the same time various verified codes can be used to evaluate in-core characteristics of a reactor. By using this approach various options for nuclide streams and assess the impact of different plutonium content in the fuel, fuel processing conditions, losses during fuel processing, as well as the impact of initial uncertainties on neutron-physical characteristics of reactor are considered in this study.

76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-16

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0055] Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of Final Design Approval The U.S. Nuclear Regulatory Commission has issued a final design approval (FDA) to GE Hitachi Nuclear Energy (GEH) for the economic...

75 FR 16517 - FirstEnergy Nuclear Operating Company; Environmental Assessment and Finding of No Significant Impact

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-01

... Perry. FOR FURTHER INFORMATION CONTACT: Michael Mahoney, Office of Nuclear Reactor Regulation, U.S... Nuclear Reactor Regulation. [FR Doc. 2010-7331 Filed 3-31-10; 8:45 am] BILLING CODE 7590-01-P ... NUCLEAR REGULATORY COMMISSION [Docket No. 50-440; NRC-2010-0124] FirstEnergy Nuclear Operating...

Developments and Tendencies in Fission Reactor Concepts

NASA Astrophysics Data System (ADS)

Adamov, E. O.; Fuji-Ie, Y.

This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC) - as an advanced and promising reactor system that offers solutions to the above problems. The difference (not confrontation) between the approaches to nuclear power development based on the principles of “inherent safety” and “natural safety” is demonstrated.

Energy Innovation Hubs: A Home for Scientific Collaboration

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

Chu, Steven

Secretary Chu will host a live, streaming Q&A session with the directors of the Energy Innovation Hubs on Tuesday, March 6, at 2:15 p.m. EST. The directors will be available for questions regarding their teams' work and the future of American energy. Ask your questions in the comments below, or submit them on Facebook, Twitter (@energy), or send an e-mail to newmedia@hq.doe.gov, prior or during the live event. Dr. Hank Foley is the director of the Greater Philadelphia Innovation Cluster for Energy-Efficient Buildings, which is pioneering new data intensive techniques for designing and operating energy efficient buildings, including advanced computermore » modeling. Dr. Douglas Kothe is the director of the Consortium for Advanced Simulation of Light Water Reactors, which uses powerful supercomputers to create "virtual" reactors that will help improve the safety and performance of both existing and new nuclear reactors. Dr. Nathan Lewis is the director of the Joint Center for Artificial Photosynthesis, which focuses on how to produce fuels from sunlight, water, and carbon dioxide. The Energy Innovation Hubs are major integrated research centers, with researchers from many different institutions and technical backgrounds. Each hub is focused on a specific high priority goal, rapidly accelerating scientific discoveries and shortening the path from laboratory innovation to technological development and commercial deployment of critical energy technologies. Ask your questions in the comments below, or submit them on Facebook, Twitter (@energy), or send an e-mail to newmedia@energy.gov, prior or during the live event. The Energy Innovation Hubs are major integrated research centers, with researchers from many different institutions and technical backgrounds. Each Hub is focused on a specific high priority goal, rapidly accelerating scientific discoveries and shortening the path from laboratory innovation to technological development and commercial deployment of critical energy technologies. Dr. Hank Holey is the director of the Greater Philadelphia Innovation Cluster for Energy-Efficient Buildings, which is pioneering new data intensive techniques for designing and operating energy efficient buildings, including advanced computer modeling. Dr. Douglas Kothe is the director of the Modeling and Simulation for Nuclear Reactors Hub, which uses powerful supercomputers to create "virtual" reactors that will help improve the safety and performance of both existing and new nuclear reactors. Dr. Nathan Lewis is the director of the Joint Center for Artificial Photosynthesis Hub, which focuses on how to produce biofuels from sunlight, water, and carbon dioxide.« less

Energy Innovation Hubs: A Home for Scientific Collaboration

ScienceCinema

Chu, Steven

2017-12-11

Secretary Chu will host a live, streaming Q&A session with the directors of the Energy Innovation Hubs on Tuesday, March 6, at 2:15 p.m. EST. The directors will be available for questions regarding their teams' work and the future of American energy. Ask your questions in the comments below, or submit them on Facebook, Twitter (@energy), or send an e-mail to newmedia@hq.doe.gov, prior or during the live event. Dr. Hank Foley is the director of the Greater Philadelphia Innovation Cluster for Energy-Efficient Buildings, which is pioneering new data intensive techniques for designing and operating energy efficient buildings, including advanced computer modeling. Dr. Douglas Kothe is the director of the Consortium for Advanced Simulation of Light Water Reactors, which uses powerful supercomputers to create "virtual" reactors that will help improve the safety and performance of both existing and new nuclear reactors. Dr. Nathan Lewis is the director of the Joint Center for Artificial Photosynthesis, which focuses on how to produce fuels from sunlight, water, and carbon dioxide. The Energy Innovation Hubs are major integrated research centers, with researchers from many different institutions and technical backgrounds. Each hub is focused on a specific high priority goal, rapidly accelerating scientific discoveries and shortening the path from laboratory innovation to technological development and commercial deployment of critical energy technologies. Ask your questions in the comments below, or submit them on Facebook, Twitter (@energy), or send an e-mail to newmedia@energy.gov, prior or during the live event. The Energy Innovation Hubs are major integrated research centers, with researchers from many different institutions and technical backgrounds. Each Hub is focused on a specific high priority goal, rapidly accelerating scientific discoveries and shortening the path from laboratory innovation to technological development and commercial deployment of critical energy technologies. Dr. Hank Holey is the director of the Greater Philadelphia Innovation Cluster for Energy-Efficient Buildings, which is pioneering new data intensive techniques for designing and operating energy efficient buildings, including advanced computer modeling. Dr. Douglas Kothe is the director of the Modeling and Simulation for Nuclear Reactors Hub, which uses powerful supercomputers to create "virtual" reactors that will help improve the safety and performance of both existing and new nuclear reactors. Dr. Nathan Lewis is the director of the Joint Center for Artificial Photosynthesis Hub, which focuses on how to produce biofuels from sunlight, water, and carbon dioxide.

Control console replacement at the WPI Reactor. [Final report

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

Not Available

1992-12-31

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« less

Nuclear power: the bargain we can't afford

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

Morgan, R.

1977-01-01

This is a handbook for citizens who wish to raise questions about the costs of atomic energy. It explains, step-by-step, why nuclear reactors have failed to produce low-cost electricity, and it tells citizens how they can use economic arguments to challenge nuclear expansion. Part One, The Costs of Nuclear Energy, contains 7 chapters--The Price of Power (electricity is big business); Mushrooming Capital Costs (nuclear construction costs are skyrocketing); Nuclear Lemons (reactors spend much of their time closed for repairs); The Faulty Fuel Cycle (turning uranium into electricity is not as simple as the utilities say); Hidden Costs (goverment subsidies obscuremore » the true costs of atomic energy); Ratepayer Roulette (nuclear problems translate into higher electric rates); and Alternatives to the Atom (coal-fired power and energy conservation can meet future energy needs more cheaply than nuclear energy). Part Two, Challenging Nuclear Power, contains 3 chapters--Regulators and Reactors (state utility commissions can eliminate the power companies' bias toward nuclear energy); Legislation, Licensing, and Lawsuits (nuclear critics can challenge reactor construction in numerous forums); and Winning the Battle (building an organization is a crucial step in fighting nuclear power). (MCW)« less

Navy Nuclear-Powered Surface Ships: Background, Issues, and Options for Congress

DTIC Science & Technology

2010-06-10

scale pressurized water reactors suitable for destroyer-sized vessels or for alternative nuclear power systems using thorium liquid salt technology...or to design a new reactor type potentially using a thorium liquid salt reactor developed for maritime use. The committee recommends an increase of...either using a pressurized water reactor or a thorium liquid salt reactor . (Page 158) Senate The Senate Armed Services Committee, in its report

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

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

Su'ud, Zaki; Anshari, Rio

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environmentmore » such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.« less

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

NASA Astrophysics Data System (ADS)

Su'ud, Zaki; Anshari, Rio

2012-06-01

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

Nuclear reactors built, being built, or planned, 1991

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

Simpson, B.

1992-07-01

This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1991. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor ismore » an American company -- working either independently or in cooperation with a foreign company (Part 4, in each section). Critical assembly refers to an assembly of fuel and assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).« less

10 CFR 2.1105 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-01-01

... Hearing Procedures for Expansion of Spent Nuclear Fuel Storage Capacity at Civilian Nuclear Power Reactors § 2.1105 Definitions. As used in this part: (a) Civilian nuclear power reactor means a civilian... 10 Energy 1 2011-01-01 2011-01-01 false Definitions. 2.1105 Section 2.1105 Energy NUCLEAR...

10 CFR 2.1105 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Hearing Procedures for Expansion of Spent Nuclear Fuel Storage Capacity at Civilian Nuclear Power Reactors § 2.1105 Definitions. As used in this part: (a) Civilian nuclear power reactor means a civilian... 10 Energy 1 2012-01-01 2012-01-01 false Definitions. 2.1105 Section 2.1105 Energy NUCLEAR...

Measuring Human Performance in Simulated Nuclear Power Plant Control Rooms Using Eye Tracking

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

Kovesdi, Casey Robert; Rice, Brandon Charles; Bower, Gordon Ross

Control room modernization will be an important part of life extension for the existing light water reactor fleet. As part of modernization efforts, personnel will need to gain a full understanding of how control room technologies affect performance of human operators. Recent advances in technology enables the use of eye tracking technology to continuously measure an operator’s eye movement, which correlates with a variety of human performance constructs such as situation awareness and workload. This report describes eye tracking metrics in the context of how they will be used in nuclear power plant control room simulator studies.

High Power MPD Nuclear Electric Propulsion (NEP) for Artificial Gravity HOPE Missions to Callisto

NASA Technical Reports Server (NTRS)

McGuire, Melissa L.; Borowski, Stanley K.; Mason, Lee M.; Gilland, James

2003-01-01

This documents the results of a one-year multi-center NASA study on the prospect of sending humans to Jupiter's moon, Callisto, using an all Nuclear Electric Propulsion (NEP) space transportation system architecture with magnetoplasmadynamic (MPD) thrusters. The fission reactor system utilizes high temperature uranium dioxide (UO2) in tungsten (W) metal matrix cermet fuel and electricity is generated using advanced dynamic Brayton power conversion technology. The mission timeframe assumes on-going human Moon and Mars missions and existing space infrastructure to support launch of cargo and crewed spacecraft to Jupiter in 2041 and 2045, respectively.

Weld monitor and failure detector for nuclear reactor system

DOEpatents

Sutton, Jr., Harry G.

1987-01-01

Critical but inaccessible welds in a nuclear reactor system are monitored throughout the life of the reactor by providing small aperture means projecting completely through the reactor vessel wall and also through the weld or welds to be monitored. The aperture means is normally sealed from the atmosphere within the reactor. Any incipient failure or cracking of the weld will cause the environment contained within the reactor to pass into the aperture means and thence to the outer surface of the reactor vessel where its presence is readily detected.

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Advance notification of shipment of irradiated reactor... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... required under this section for shipments of irradiated reactor fuel in quantities less than that subject...

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Advance notification of shipment of irradiated reactor... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... required under this section for shipments of irradiated reactor fuel in quantities less than that subject...

Reactivity control assembly for nuclear reactor. [LMFBR

DOEpatents

Bollinger, L.R.

1982-03-17

This invention, which resulted from a contact with the United States Department of Energy, relates to a control mechanism for a nuclear reactor and, more particularly, to an assembly for selectively shifting different numbers of reactivity modifying rods into and out of the core of a nuclear reactor. It has been proposed heretofore to control the reactivity of a breeder reactor by varying the depth of insertion of control rods (e.g., rods containing a fertile material such as ThO/sub 2/) in the core of the reactor, thereby varying the amount of neutron-thermalizing coolant and the amount of neutron-capturing material in the core. This invention relates to a mechanism which can advantageously be used in this type of reactor control system.

Reactor design and integration into a nuclear electric spacecraft

NASA Technical Reports Server (NTRS)

Phillips, W. M.; Koenig, D. R.

1978-01-01

One of the well-defined applications for nuclear power in space is nuclear electric propulsion (NEP). Mission studies have identified the optimum power level (400 kWe). A single Shuttle launch requirement and science-package integration have added additional constraints to the design. A reactor design which will meet these constraints has been studied. The reactor employs 90 fuel elements, each heat pipe cooled. Reactor control is obtained with BeO/B4C drums in a BeO reflector. The balance of the spacecraft is shielded from the reactor with LiH. Power conditioning and reactor control drum drives are located behind the LiH with the power conditioning. Launch safety, mechanical design and integration with the power conversion subsystem are discussed.

10 CFR 2.501 - Notice of hearing on application under subpart F of 10 CFR part 52 for a license to manufacture...

Code of Federal Regulations, 2011 CFR

2011-01-01

... part 52 for a license to manufacture nuclear power reactors. 2.501 Section 2.501 Energy NUCLEAR... Procedures Applicable to Proceedings for the Issuance of Licenses To Manufacture Nuclear Power Reactors To Be... power reactors. (a) In the case of an application under subpart F of part 52 of this chapter for a...

10 CFR 2.501 - Notice of hearing on application under subpart F of 10 CFR part 52 for a license to manufacture...

Code of Federal Regulations, 2010 CFR

2010-01-01

... part 52 for a license to manufacture nuclear power reactors. 2.501 Section 2.501 Energy NUCLEAR... Procedures Applicable to Proceedings for the Issuance of Licenses To Manufacture Nuclear Power Reactors To Be... power reactors. (a) In the case of an application under subpart F of part 52 of this chapter for a...

Nuclear Energy Policy

DTIC Science & Technology

2009-12-10

Small Modular Reactors Rising cost estimates for large conventional nuclear power plants—widely projected to be $6 billion or more—have contributed to growing interest in proposals for smaller, modular reactors. Ranging from about 40 to 350 megawatts of electrical capacity, such reactors would be only a fraction of the size of current commercial reactors. Several modular reactors would be installed together to make up a power block with a single control room, under most concepts. Modular reactor concepts would use a variety of technologies,

2009.1 Revision of the Evaluated Nuclear Data Library (ENDL2009.1)

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

Thompson, I. J.; Beck, B.; Descalles, M. A.

LLNL’s Computational Nuclear Data and Theory Group have created a 2009.1 revised release of the Evaluated Nuclear Data Library (ENDL2009.1). This library is designed to support LLNL’s current and future nuclear data needs and will be employed in nuclear reactor, nuclear security and stockpile stewardship simulations with ASC codes. The ENDL2009 database was the most complete nuclear database for Monte Carlo and deterministic transport of neutrons and charged particles. It was assembled with strong support from the ASC PEM and Attribution programs, leveraged with support from Campaign 4 and the DOE/Office of Science’s US Nuclear Data Program. This document listsmore » the revisions and fixes made in a new release called ENDL2009.1, by comparing with the existing data in the original release which is now called ENDL2009.0. These changes are made in conjunction with the revisions for ENDL2011.1, so that both the .1 releases are as free as possible of known defects.« less

2009.3 Revision of the Evaluated Nuclear Data Library (ENDL2009.3)

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

Thompson, I. J.; Beck, B.; Descalle, M. A.

LLNL's Computational Nuclear Data and Theory Group have created a 2009.3 revised release of the Evaluated Nuclear Data Library (ENDL2009.3). This library is designed to support LLNL's current and future nuclear data needs and will be employed in nuclear reactor, nuclear security and stockpile stewardship simulations with ASC codes. The ENDL2009 database was the most complete nuclear database for Monte Carlo and deterministic transport of neutrons and charged particles. It was assembled with strong support from the ASC PEM and Attribution programs, leveraged with support from Campaign 4 and the DOE/Office of Science's US Nuclear Data Program. This document listsmore » the revisions and fixes made in a new release called ENDL2009.3, by com- paring with the existing data in the previous release ENDL2009.2. These changes are made in conjunction with the revisions for ENDL2011.3, so that both the .3 releases are as free as possible of known defects.« less

Non-Nuclear Testing of Compact Reactor Technologies at NASA MSFC

NASA Technical Reports Server (NTRS)

Houts, Michael G.; Pearson, J. Boise; Godfroy, Thomas J.

2011-01-01

Safe, reliable, compact, autonomous, long-life fission systems have numerous potential applications, both terrestrially and in space. Technologies and facilities developed in support of these systems could be useful to a variety of concepts. At moderate power levels, fission systems can be designed to operate for decades without the need for refueling. In addition, fast neutron damage to cladding and structural materials can be maintained at an acceptable level. Nuclear design codes have advanced to the stage where high confidence in the behavior and performance of a system can be achieved prior to initial testing. To help ensure reactor affordability, an optimal strategy must be devised for development and qualification. That strategy typically involves a combination of non-nuclear and nuclear testing. Non-nuclear testing is particularly useful for concepts in which nuclear operating characteristics are well understood and nuclear effects such as burnup and radiation damage are not likely to be significant. To be mass efficient, a SFPS must operate at higher coolant temperatures and use different types of power conversion than typical terrestrial reactors. The primary reason is the difficulty in rejecting excess heat to space. Although many options exist, NASA s current reference SFPS uses a fast spectrum, pumped-NaK cooled reactor coupled to a Stirling power conversion subsystem. The reference system uses technology with significant terrestrial heritage while still providing excellent performance. In addition, technologies from the SFPS system could be applicable to compact terrestrial systems. Recent non-nuclear testing at NASA s Early Flight Fission Test Facility (EFF-TF) has helped assess the viability of the reference SFPS and evaluate methods for system integration. In July, 2011 an Annular Linear Induction Pump (ALIP) provided by Idaho National Laboratory was tested at the EFF-TF to assess performance and verify suitability for use in a10 kWe technology demonstration unit (TDU). In November, 2011 testing of a 37-pin core simulator (designed in conjunction with Los Alamos National Laboratory) for use with the TDU will occur. Previous testing at the EFFTF has included the thermal and mechanical coupling of a pumped NaK loop to Stirling engines (provided by GRC). Testing related to heat pipe cooled systems, gas cooled systems, heat exchangers, and other technologies has also been performed. Integrated TDU testing will begin at GRC in 2013. Thermal simulators developed at the EFF-TF are capable of operating over the temperature and power range typically of interest to compact reactors. Small and large diameter simulators have been developed, and simulators (coupled with the facility) are able to closely match the axial and radial power profile of all potential systems of interest. A photograph of the TDU core simulator during assembly is provided in Figure 2.

Developing the European Center of Competence on VVER-type nuclear power reactors

NASA Astrophysics Data System (ADS)

Geraskin, Nikolay; Pironkov, Lyubomir; Kulikov, Evgeny; Glebov, Vasily

2017-09-01

This paper presents the results of the European educational projects CORONA and CORONA-II which are dedicated to preserving and further developing nuclear knowledge and competencies in the area of VVER-type nuclear power reactors technologies (Water-Water Energetic Reactor, WWER or VVER). The development of the European Center of Competence for VVER-technology is focused on master's degree programmes. The specifics of a systematic approach to training in the area of VVER-type nuclear power reactors technologies are analysed. This paper discusses enhancement of the training opportunities of the European Center that have arisen from advances in methodology and distance education. With a special attention paid to the European Nuclear Education Network (ENEN), the possibilities of further development of the international cooperation between European countries and educational institutions are examined.

Preliminary results of calculations for heavy-water nuclear-power-plant reactors employing 235U, 233U, and 232Th as a fuel and meeting requirements of a nonproliferation of nuclear weapons

NASA Astrophysics Data System (ADS)

Ioffe, B. L.; Kochurov, B. P.

2012-02-01

A physical design is developed for a gas-cooled heavy-water nuclear reactor intended for a project of a nuclear power plant. As a fuel, the reactor would employ thorium with a small admixture of enriched uranium that contains not more than 20% of 235U. It operates in the open-cycle mode involving 233U production from thorium and its subsequent burnup. The reactor meets the conditions of a nonproliferation of nuclear weapons: the content of fissionable isotopes in uranium at all stages of the process, including the final one, is below the threshold for constructing an atomic bomb, the amount of product plutonium being extremely small.

78 FR 17450 - Entergy Nuclear Operations, Inc.; Entergy Operations, Inc.; Biweekly Notice; Notice of Issuance...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-21

... INFORMATION CONTACT: Nageswaran Kalyanam, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory..., Office of Nuclear Reactor Regulation. [FR Doc. 2013-06510 Filed 3-20-13; 8:45 am] BILLING CODE 7590-01-P ... NUCLEAR REGULATORY COMMISSION [NRC-2013-0012] [Docket Nos. 50-458, 50-155, 72-043, 50-003, 50-247...

Environmental Cracking and Irradiation Resistant Stainless Steels by Additive Manufacturing

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

Rebak, Raul B.; Lou, Xiaoyuan

Metal additive manufacturing (AM), or metal 3D printing is an emergent advanced manufacturing method that can create near net shape geometries directly from computer models. This technology can provide the capability to rapidly fabricate complex parts that may be required to enhance the integrity of reactor internals components. Such opportunities may be observed during a plant refueling outage and AM parts can be rapidly custom designed, manufactured and deployed within the outage interval. Additive manufacturing of stainless steel (SS) components can add business benefits on fast delivery on repair hardware, installation tooling, new design prototypes tests, etc. For the nuclearmore » industry, the supply chain is always an issue for reactor service. AM can provide through-life supply chain (40-60 years) for high-value low-volume components. In the meantime, the capability of generating complex geometries and functional gradient materials will improve the performance, reduce the overall component cost, plant asset management cost and increase the plant reliability by the improvement in materials performance in nuclear environments. While extensive work has been conducted regarding additively manufacturing of austenitic SS parts, most efforts focused only on basic attributes such as porosity, residual stress, basic tensile properties, along with components yield and process monitoring. Little work has been done to define and evaluate the material requirements for nuclear applications. Technical gaps exist, which limit this technology adoption in the nuclear industry, which includes high manufacturing cost, unknown risks, limited nuclear related data, lack of specification and qualification methods, and no prior business experience. The main objective of this program was to generate research data to address all these technical gaps and establish a commercial practice to use AM technology in the nuclear power industry. The detailed objectives are listed as follows: (1) Evaluate nuclear related properties of AM 316L SS, including microstructure, tensile properties, impact toughness, stress corrosion cracking (SCC), corrosion fatigue (CF), irradiation effects, and irradiation assisted stress corrosion cracking (IASCC). (2) Understand the correlations among laser processing, heat treatment, microstructure and SCC/irradiation properties; (3) Optimize and improve the manufacturing process to achieve enhanced nuclear application properties; (4) Fabricate, evaluate, qualify and test a prototype reactor component to demonstrate the commercial viability and cost benefit; (5) Create regulatory approval path and commercialization plans for the production of a commercial reactor component.« less

Fissioning uranium plasmas and nuclear-pumped lasers

NASA Technical Reports Server (NTRS)

Schneider, R. T.; Thom, K.

1975-01-01

Current research into uranium plasmas, gaseous-core (cavity) reactors, and nuclear-pumped lasers is discussed. Basic properties of fissioning uranium plasmas are summarized together with potential space and terrestrial applications of gaseous-core reactors and nuclear-pumped lasers. Conditions for criticality of a uranium plasma are outlined, and it is shown that the nonequilibrium state and the optical thinness of a fissioning plasma can be exploited for the direct conversion of fission fragment energy into coherent light (i.e., for nuclear-pumped lasers). Successful demonstrations of nuclear-pumped lasers are described together with gaseous-fuel reactor experiments using uranium hexafluoride.

Mission design considerations for nuclear risk mitigation

NASA Technical Reports Server (NTRS)

Stancati, Mike; Collins, John

1993-01-01

Strategies for the mitigation of the nuclear risk associated with two specific mission operations are discussed. These operations are the safe return of nuclear thermal propulsion reactors to earth orbit and the disposal of lunar/Mars spacecraft reactors.

A Potential NASA Research Reactor to Support NTR Development

NASA Technical Reports Server (NTRS)

Eades, Michael; Gerrish, Harold; Hardin, Leroy

2013-01-01

In support of efforts for research into the design and development of a man rated Nuclear Thermal Rocket (NTR) engine, the National Aeronautics and Space Administration (NASA), Marshall Space Flight Center (MSFC), is evaluating the potential for building a Nuclear Regulatory Commission (NRC) licensed research reactor. The proposed reactor would be licensed by NASA and operated jointly by NASA and university partners. The purpose of this reactor would be to perform further research into the technologies and systems needed for a successful NTR project and promote nuclear training and education.

A Simple Global View of Fuel Burnup

NASA Astrophysics Data System (ADS)

Sekimoto, Hiroshi

2017-01-01

Reactor physics and fuel burnup are discussed in order to obtain a simple global view of the effects of nuclear reactor characteristics to fuel cycle system performance. It may provide some idea of free thinking and overall vision, though it is still a small part of nuclear energy system. At the beginning of this lecture, governing equations for nuclear reactors are presented. Since the set of these equations is so big and complicated, it is simplified by imposing some extreme conditions and the nuclear equilibrium equation is derived. Some features of future nuclear equilibrium state are obtained by solving this equation. The contribution of a nucleus charged into reactor core to the system performance indexes such as criticality is worth for understanding the importance of each nuclide. It is called nuclide importance and can be evaluated by using the equations adjoint to the nuclear equilibrium equation. Examples of some importances and their application to criticalily search problem are presented.

Advantages of liquid fluoride thorium reactor in comparison with light water reactor

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

Bahri, Che Nor Aniza Che Zainul, E-mail: anizazainul@gmail.com; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

2015-04-29

Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclearmore » waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.« less

77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-20

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0010] Knowledge and Abilities Catalog for Nuclear Power... comment a draft NUREG, NUREG-2104, Revision 0, ``Knowledge and Abilities Catalog for Nuclear Power Plant... developed using this Catalog along with the Operator Licensing Examination Standards for Power Reactors...

U.S. Nuclear Cooperation with India: Issues for Congress

DTIC Science & Technology

2010-04-08

Kazakhstan might start uranium exports to India in 2009,” Panorama , February 6, 2009. “Chennai Daily Report: India, Kazakhstan Set To Sign Nuclear Reactor...reactors producing more than 5 MW thermal or special nuclear material connected therewith. U.S. Nuclear Cooperation with India: Issues for Congress

78 FR 26812 - University of California, Irvine; License Renewal for University of California, Irvine Nuclear...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-08

... NUCLEAR REGULATORY COMMISSION [Docket No. 50-326; NRC-2010-0217] University of California, Irvine; License Renewal for University of California, Irvine Nuclear Reactor Facility; Supplemental Information... Renewal for University of California, Irvine Nuclear Reactor Facility,'' to inform the public that the NRC...

10 CFR 52.35 - Use of site for other purposes.

Code of Federal Regulations, 2013 CFR

2013-01-01

..., Office of New Reactors or Director, Office of Nuclear Reactor Regulation, as appropriate, (Director) of... NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER PLANTS... informs the Director that the holder no longer intends to use the site for a nuclear power plant, the...

10 CFR 52.35 - Use of site for other purposes.

Code of Federal Regulations, 2014 CFR

2014-01-01

..., Office of New Reactors or Director, Office of Nuclear Reactor Regulation, as appropriate, (Director) of... NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER PLANTS... informs the Director that the holder no longer intends to use the site for a nuclear power plant, the...

10 CFR 52.35 - Use of site for other purposes.

Code of Federal Regulations, 2012 CFR

2012-01-01

..., Office of New Reactors or Director, Office of Nuclear Reactor Regulation, as appropriate, (Director) of... NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER PLANTS... informs the Director that the holder no longer intends to use the site for a nuclear power plant, the...

10 CFR 52.35 - Use of site for other purposes.

Code of Federal Regulations, 2010 CFR

2010-01-01

..., Office of New Reactors or Director, Office of Nuclear Reactor Regulation, as appropriate, (Director) of... NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER PLANTS... informs the Director that the holder no longer intends to use the site for a nuclear power plant, the...

10 CFR 52.35 - Use of site for other purposes.

Code of Federal Regulations, 2011 CFR

2011-01-01

..., Office of New Reactors or Director, Office of Nuclear Reactor Regulation, as appropriate, (Director) of... NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER PLANTS... informs the Director that the holder no longer intends to use the site for a nuclear power plant, the...

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

Swift, Alicia L.

There is no better time than now to close the loophole in Article IV of the Nuclear Non-proliferation Treaty (NPT) that excludes military uses of fissile material from nuclear safeguards. Several countries have declared their intention to pursue and develop naval reactor technology, including Argentina, Brazil, Iran, and Pakistan, while other countries such as China, India, Russia, and the United States are expanding their capabilities. With only a minority of countries using low enriched uranium (LEU) fuel in their naval reactors, it is possible that a state could produce highly enriched uranium (HEU) under the guise of a nuclear navymore » while actually stockpiling the material for a nuclear weapon program. This paper examines the likelihood that non-nuclear weapon states exploit the loophole to break out from the NPT and also the regional ramifications of deterrence and regional stability of expanding naval forces. Possible solutions to close the loophole are discussed, including expanding the scope of the Fissile Material Cut-off Treaty, employing LEU fuel instead of HEU fuel in naval reactors, amending the NPT, creating an export control regime for naval nuclear reactors, and forming individual naval reactor safeguards agreements.« less

75 FR 13 - Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-04

...The Nuclear Regulatory Commission (NRC) is amending its regulations to provide alternate fracture toughness requirements for protection against pressurized thermal shock (PTS) events for pressurized water reactor (PWR) pressure vessels. This final rule provides alternate PTS requirements based on updated analysis methods. This action is desirable because the existing requirements are based on unnecessarily conservative probabilistic fracture mechanics analyses. This action reduces regulatory burden for those PWR licensees who expect to exceed the existing requirements before the expiration of their licenses, while maintaining adequate safety, and may choose to comply with the final rule as an alternative to complying with the existing requirements.

Innovations for In-Pile Measurements in the Framework of the CEA-SCK•CEN Joint Instrumentation Laboratory

NASA Astrophysics Data System (ADS)

Villard, Jean-Francois; Schyns, Marc

2010-12-01

Optimizing the life cycle of nuclear systems under safety constraints requires high-performance experimental programs to reduce uncertainties on margins and limits. In addition to improvement in modeling and simulation, innovation in instrumentation is crucial for analytical and integral experiments conducted in research reactors. The quality of nuclear research programs relies obviously on an excellent knowledge of their experimental environment which constantly calls for better online determination of neutron and gamma flux. But the combination of continuously increasing scientific requirements and new experimental domains -brought for example by Generation IV programsnecessitates also major innovations for in-pile measurements of temperature, dimensions, pressure or chemical analysis in innovative mediums. At the same time, the recent arising of a European platform around the building of the Jules Horowitz Reactor offers new opportunities for research institutes and organizations to pool their resources in order to face these technical challenges. In this situation, CEA (French Nuclear Energy Commission) and SCK'CEN (Belgian Nuclear Research Centre) have combined their efforts and now share common developments through a Joint Instrumentation Laboratory. Significant progresses have thus been obtained recently in the field of in-pile measurements, on one hand by improvement of existing measurement methods, and on the other hand by introduction in research reactors of original measurement techniques. This paper highlights the state-of-the-art and the main requirements regarding in-pile measurements, particularly for the needs of current and future irradiation programs performed in material testing reactors. Some of the main on-going developments performed in the framework of the Joint Instrumentation Laboratory are also described, such as: - a unique fast neutron flux measurement system using fission chambers with 242Pu deposit and a specific online data processing, - an optical system designed to perform in-pile dimensional measurements of material samples under irradiation, - an acoustical instrumentation allowing the online characterization of fission gas release in Pressurized Water Reactor fuel rods. For each example, the obtained results, expected impacts and development status are detailed.

10 CFR 100.4 - Communications.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Communications. 100.4 Section 100.4 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) REACTOR SITE CRITERIA § 100.4 Communications. Except where otherwise... Reactor Regulation or Director, Office of New Reactors, as appropriate, U.S. Nuclear Regulatory Commission...

10 CFR 100.4 - Communications.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Communications. 100.4 Section 100.4 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) REACTOR SITE CRITERIA § 100.4 Communications. Except where otherwise... Reactor Regulation or Director, Office of New Reactors, as appropriate, U.S. Nuclear Regulatory Commission...

Comparative evaluation of solar, fission, fusion, and fossil energy resources. Part 2: Power from nuclear fission

NASA Technical Reports Server (NTRS)

Clement, J. D.

1973-01-01

Different types of nuclear fission reactors and fissionable materials are compared. Special emphasis is placed upon the environmental impact of such reactors. Graphs and charts comparing reactor facilities in the U. S. are presented.

OVERVIEW OF NUCLEAR PHYSICS LABORATORY (IMMEDIATELY EAST OF SPSE REACTOR ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

OVERVIEW OF NUCLEAR PHYSICS LABORATORY (IMMEDIATELY EAST OF SP-SE REACTOR ROOM), LEVEL -15’, LOOKING SOUTHWEST. NOTE SLIDING STEEL PLATE DOOR BETWEEN LABORATORY AND REACTOR ROOM - Physics Assembly Laboratory, Area A/M, Savannah River Site, Aiken, Aiken County, SC

10 CFR 55.5 - Communications.

Code of Federal Regulations, 2012 CFR

2012-01-01

.... Nuclear Regulatory Commission, Washington, DC 20555-0001. The guidance discusses, among other topics, the.... (b)(1) Except for test and research reactor facilities, the Director, Office of Nuclear Reactor... involving a test and research reactor facility licensed under 10 CFR part 50 and any related inquiry...

10 CFR 55.5 - Communications.

Code of Federal Regulations, 2013 CFR

2013-01-01

.... Nuclear Regulatory Commission, Washington, DC 20555-0001. The guidance discusses, among other topics, the.... (b)(1) Except for test and research reactor facilities, the Director, Office of Nuclear Reactor... involving a test and research reactor facility licensed under 10 CFR part 50 and any related inquiry...

Space Nuclear Reactor Engineering

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

Poston, David Irvin

We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

10 CFR 100.4 - Communications.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Reactor Regulation or Director, Office of New Reactors, as appropriate, U.S. Nuclear Regulatory Commission... 10 Energy 2 2012-01-01 2012-01-01 false Communications. 100.4 Section 100.4 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) REACTOR SITE CRITERIA § 100.4 Communications. Except where otherwise...

Liquid metal cooled nuclear reactor plant system

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1993-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

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.

SCW Pressure-Channel Nuclear Reactor Some Design Features

NASA Astrophysics Data System (ADS)

Pioro, Igor L.; Khan, Mosin; Hopps, Victory; Jacobs, Chris; Patkunam, Ruban; Gopaul, Sandeep; Bakan, Kurtulus

Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30-35% to about 45-48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (˜1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. Some design features of the Canadian concept related to fuel channels are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

NASA Astrophysics Data System (ADS)

Stacey, W. M.

2009-09-01

The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation's energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.

Nuclear design of a vapor core reactor for space nuclear propulsion

NASA Astrophysics Data System (ADS)

Dugan, Edward T.; Watanabe, Yoichi; Kuras, Stephen A.; Maya, Isaac; Diaz, Nils J.

1993-01-01

Neutronic analysis methodology and results are presented for the nuclear design of a vapor core reactor for space nuclear propulsion. The Nuclear Vapor Thermal Reactor (NVTR) Rocket Engine uses modified NERVA geometry and systems which the solid fuel replaced by uranium tetrafluoride vapor. The NVTR is an intermediate term gas core thermal rocket engine with specific impulse in the range of 1000-1200 seconds; a thrust of 75,000 lbs for a hydrogen flow rate of 30 kg/s; average core exit temperatures of 3100 K to 3400 K; and reactor thermal powers of 1400 to 1800 MW. Initial calculations were performed on epithermal NVTRs using ZrC fuel elements. Studies are now directed at thermal NVTRs that use fuel elements made of C-C composite. The large ZrC-moderated reactors resulted in thrust-to-weight ratios of only 1 to 2; the compact C-C composite systems yield thrust-to-weight ratios of 3 to 5.

Reactor power system deployment and startup

NASA Technical Reports Server (NTRS)

Wetch, J. R.; Nelin, C. J.; Britt, E. J.; Klein, G.

1985-01-01

This paper addresses issues that should receive further examination in the near-term as concept selection for development of a U.S. space reactor power system is approached. The issues include: the economics, practicality and system reliability associated with transfer of nuclear spacecraft from low earth shuttle orbits to operational orbits, via chemical propulsion versus nuclear electric propulsion; possible astronaut supervised reactor and nuclear electric propulsion startup in low altitude Shuttle orbit; potential deployment methods for nuclear powered spacecraft from Shuttle; the general public safety of low altitude startup and nuclear safe and disposal orbits; the question of preferred reactor power level; and the question of frozen versus molten alkali metal coolant during launch and deployment. These issues must be considered now because they impact the SP-100 concept selection, power level selection, weight and size limits, use of deployable radiators, reliability requirements, and economics, as well as the degree of need for and the urgency of developing space reactor power systems.

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

International academic program in technologies of light-water nuclear reactors. Phases of development and implementation

NASA Astrophysics Data System (ADS)

Geraskin, N. I.; Glebov, V. B.

2017-01-01

The results of implementation of European educational projects CORONA and CORONA II dedicated to preserving and further developing nuclear knowledge and competencies in the area of technologies of light-water nuclear reactors are analyzed. Present article addresses issues of design and implementation of the program for specialized training in the branch of technologies of light-water nuclear reactors. The systematic approach has been used to construct the program for students of nuclear specialties, which corresponding to IAEA standards and commonly accepted nuclear principles recognized in the European Union. Possibilities of further development of the international cooperation between countries and educational institutions are analyzed. Special attention is paid to e-learning/distance training, nuclear knowledge preservation and interaction with European Nuclear Education Network.

EMERGENCY SHUTDOWN FOR NUCLEAR REACTORS

DOEpatents

Paget, J.A.; Koutz, S.L.; Stone, R.S.; Stewart, H.B.

1963-12-24

An emergency shutdown or scram apparatus for use in a nuclear reactor that includes a neutron absorber suspended from a temperature responsive substance that is selected to fail at a preselected temperature in excess of the normal reactor operating temperature, whereby the neutron absorber is released and allowed to fall under gravity to a preselected position within the reactor core is presented. (AEC)

A survey of nuclear-related agreements and possibilities for nuclear cooperation in South Asia: Cooperative Monitoring Center Occasional Paper/15

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

RAJEN,GAURAV

2000-04-01

Several existing nuclear-related agreements already require India and Pakistan, as members, to share information. The agreements are bilateral, regional, and international. Greater nuclear transparency between India and Pakistan could be promoted by first understanding the information flows required by existing agreements. This understanding is an essential step for developing projects that can incrementally advance the sensitivity of the information being shared. This paper provides a survey of existing nuclear-related agreements involving India and Pakistan, and suggests future confidence-building projects using the frameworks provided by these agreements. The Bilateral Agreement on the Prohibition of Attack against Nuclear Reactors and Nuclear Facilitiesmore » is discussed as a basis for creating further agreements on restricting the use and deployment of nuclear weapons. The author suggests options for enhancing the value of the list of nuclear facilities exchanged annually as a part of this agreement. The International Atomic Energy Agency's regional cooperation agreement among countries in the Asia-Pacific region is an opportunity for greater subregional nuclear cooperation in South Asia. Linking the regional agreement with South Asian environmental cooperation and marine pollution protection efforts could provide a framework for projects involving Indian and Pakistani coastal nuclear facilities. Programs of the Food and Agriculture Organization of the United Nations that use nuclear techniques to increase food and crop production and optimize water management in arid areas also provide similar opportunities for nuclear cooperation. Other frameworks for nuclear cooperation originate from international conventions related to nuclear safety, transportation of nuclear wastes, worker protection against ionizing radiation, and the nondeployment of nuclear weapons in certain areas. The information shared by existing frameworks includes: laws and regulations (including internal inspection procedures that enforce compliance); lists of nuclear facilities; emergency response procedures and available resources; information related to the transportation of nuclear wastes (particularly via shipping); understanding and notification of accidental releases; and radionuclide release data from select coastal facilities. Incremental increases in the sensitivity of the information being shared could strengthen norms for Indian and Pakistani nuclear transparency. This paper suggests seven technology-based Indian and Pakistani nuclear transparency projects for consideration. Existing nuclear-related agreements provide an information-sharing framework within which the projects could occur. Eventually, as confidence increases and new agreements are negotiated, future projects could begin to deal with the accounting of fissile materials and nuclear weapons disposition and control.« less

The role of mass spectrometry to study the Oklo-Bangombé natural reactors.

PubMed

De Laeter, J R; Hidaka, H

2007-01-01

The discovery of the existence of chain reactions at the Oklo natural reactors in Gabon, Central Africa in 1972 was a triumph for the accuracy of mass spectrometric measurements, in that a 0.5% anomaly in the (235)U/(238)U ratio of certain U ore samples indicated a depletion in (235)U. Mass spectrometric techniques thereafter played a dominant role in determining the nuclear parameters of the reactor zones themselves, and in deciphering the geochemical characteristics of various elements in the U-rich ore and in the surrounding rock strata. The variations in the isotopic composition of a large number of elements, caused by a combination of nuclear fission, neutron capture and radioactive decay, provide a powerful tool for investigating this unique geological environment. Mass spectrometry can be used to measure the present-day elemental and isotopic abundances of numerous elements, so as to decipher the past history of the reactors and examine the retentivity/mobility of these elements. Many of the fission products have a radioactive decay history that have been used to date the age and duration of the reactor zones, and to provide insight into their nuclear and geochemical behavior as a function of time. The Oklo fission reactors and their near neighbor at Bangombé, some 30 km to the south-east of Oklo, are unique in that not only did they become critical some 2 x 10(9) years ago, but also the deposits have been preserved over this period of geological time. The long-term geochemical behavior of actinides and fission products have been extensively studied by a variety of mass spectrometric techniques over the past 30 years to provide us with significant information on the mobility/retentivity of this material in a natural geological repository. The Oklo-Bangombé natural reactors are therefore geological analogs that can be evaluated in terms of possible radioactive waste containment sites. As more reactor zones were discovered, it was realized that they could be classified into two groups according to their burial depth in the Oklo mine-site. Reactor Zones 10, 13, and 16 were buried more deeply, and were therefore less weathered than the other zones. The less-weathered zones are of great importance in mobility/retentivity studies and therefore to the question of radioactive waste containment. Isotopic studies of these natural reactors are also of value in physics to examine possible variations in fundamental constants over the past 2 billion years.

10 CFR 55.5 - Communications.

Code of Federal Regulations, 2011 CFR

2011-01-01

.... Nuclear Regulatory Commission, Washington, DC 20555-0001. The guidance discusses, among other topics, the.... (b)(1) Except for test and research reactor facilities, the Director, Office of Nuclear Reactor... this part involving a test and research reactor facility licensed under 10 CFR part 50 and any related...