Towards a supported common NEAMS software stack

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

Cormac Garvey

2012-04-01

The NEAMS IPSC's are developing multidimensional, multiphysics, multiscale simulation codes based on first principles that will be capable of predicting all aspects of current and future nuclear reactor systems. These new breeds of simulation codes will include rigorous verification, validation and uncertainty quantification checks to quantify the accuracy and quality of the simulation results. The resulting NEAMS IPSC simulation codes will be an invaluable tool in designing the next generation of Nuclear Reactors and also contribute to a more speedy process in the acquisition of licenses from the NRC for new Reactor designs. Due to the high resolution of themore » models, the complexity of the physics and the added computational resources to quantify the accuracy/quality of the results, the NEAMS IPSC codes will require large HPC resources to carry out the production simulation runs.« less

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

NASA Astrophysics Data System (ADS)

Piro, Markus Hans Alexander

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

The development of a thermal hydraulic feedback mechanism with a quasi-fixed point iteration scheme for control rod position modeling for the TRIGSIMS-TH application

NASA Astrophysics Data System (ADS)

Karriem, Veronica V.

Nuclear reactor design incorporates the study and application of nuclear physics, nuclear thermal hydraulic and nuclear safety. Theoretical models and numerical methods implemented in computer programs are utilized to analyze and design nuclear reactors. The focus of this PhD study's is the development of an advanced high-fidelity multi-physics code system to perform reactor core analysis for design and safety evaluations of research TRIGA-type reactors. The fuel management and design code system TRIGSIMS was further developed to fulfill the function of a reactor design and analysis code system for the Pennsylvania State Breazeale Reactor (PSBR). TRIGSIMS, which is currently in use at the PSBR, is a fuel management tool, which incorporates the depletion code ORIGEN-S (part of SCALE system) and the Monte Carlo neutronics solver MCNP. The diffusion theory code ADMARC-H is used within TRIGSIMS to accelerate the MCNP calculations. It manages the data and fuel isotopic content and stores it for future burnup calculations. The contribution of this work is the development of an improved version of TRIGSIMS, named TRIGSIMS-TH. TRIGSIMS-TH incorporates a thermal hydraulic module based on the advanced sub-channel code COBRA-TF (CTF). CTF provides the temperature feedback needed in the multi-physics calculations as well as the thermal hydraulics modeling capability of the reactor core. The temperature feedback model is using the CTF-provided local moderator and fuel temperatures for the cross-section modeling for ADMARC-H and MCNP calculations. To perform efficient critical control rod calculations, a methodology for applying a control rod position was implemented in TRIGSIMS-TH, making this code system a modeling and design tool for future core loadings. The new TRIGSIMS-TH is a computer program that interlinks various other functional reactor analysis tools. It consists of the MCNP5, ADMARC-H, ORIGEN-S, and CTF. CTF was coupled with both MCNP and ADMARC-H to provide the heterogeneous temperature distribution throughout the core. Each of these codes is written in its own computer language performing its function and outputs a set of data. TRIGSIMS-TH provides an effective use and data manipulation and transfer between different codes. With the implementation of feedback and control- rod-position modeling methodologies, the TRIGSIMS-TH calculations are more accurate and in a better agreement with measured data. The PSBR is unique in many ways and there are no "off-the-shelf" codes, which can model this design in its entirety. In particular, PSBR has an open core design, which is cooled by natural convection. Combining several codes into a unique system brings many challenges. It also requires substantial knowledge of both operation and core design of the PSBR. This reactor is in operation decades and there is a fair amount of studies and developments in both PSBR thermal hydraulics and neutronics. Measured data is also available for various core loadings and can be used for validation activities. The previous studies and developments in PSBR modeling also aids as a guide to assess the findings of the work herein. In order to incorporate new methods and codes into exiting TRIGSIMS, a re-evaluation of various components of the code was performed to assure the accuracy and efficiency of the existing CTF/MCNP5/ADMARC-H multi-physics coupling. A new set of ADMARC-H diffusion coefficients and cross sections was generated using the SERPENT code. This was needed as the previous data was not generated with thermal hydraulic feedback and the ARO position was used as the critical rod position. The B4C was re-evaluated for this update. The data exchange between ADMARC-H and MCNP5 was modified. The basic core model is given a flexibility to allow for various changes within the core model, and this feature was implemented in TRIGSIMS-TH. The PSBR core in the new code model can be expanded and changed. This allows the new code to be used as a modeling tool for design and analyses of future code loadings.

Evaluation of the DRAGON code for VHTR design analysis.

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

Taiwo, T. A.; Kim, T. K.; Nuclear Engineering Division

2006-01-12

This letter report summarizes three activities that were undertaken in FY 2005 to gather information on the DRAGON code and to perform limited evaluations of the code performance when used in the analysis of the Very High Temperature Reactor (VHTR) designs. These activities include: (1) Use of the code to model the fuel elements of the helium-cooled and liquid-salt-cooled VHTR designs. Results were compared to those from another deterministic lattice code (WIMS8) and a Monte Carlo code (MCNP). (2) The preliminary assessment of the nuclear data library currently used with the code and libraries that have been provided by themore » IAEA WIMS-D4 Library Update Project (WLUP). (3) DRAGON workshop held to discuss the code capabilities for modeling the VHTR.« less

Upgrades to the NESS (Nuclear Engine System Simulation) Code

NASA Technical Reports Server (NTRS)

Fittje, James E.

2007-01-01

In support of the President's Vision for Space Exploration, the Nuclear Thermal Rocket (NTR) concept is being evaluated as a potential propulsion technology for human expeditions to the moon and Mars. The need for exceptional propulsion system performance in these missions has been documented in numerous studies, and was the primary focus of a considerable effort undertaken during the 1960's and 1970's. The NASA Glenn Research Center is leveraging this past NTR investment in their vehicle concepts and mission analysis studies with the aid of the Nuclear Engine System Simulation (NESS) code. This paper presents the additional capabilities and upgrades made to this code in order to perform higher fidelity NTR propulsion system analysis and design.

Nuclear Energy Knowledge and Validation Center (NEKVaC) Needs Workshop Summary Report

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

Gougar, Hans

2015-02-01

The Department of Energy (DOE) has made significant progress developing simulation tools to predict the behavior of nuclear systems with greater accuracy and of increasing our capability to predict the behavior of these systems outside of the standard range of applications. These analytical tools require a more complex array of validation tests to accurately simulate the physics and multiple length and time scales. Results from modern simulations will allow experiment designers to narrow the range of conditions needed to bound system behavior and to optimize the deployment of instrumentation to limit the breadth and cost of the campaign. Modern validation,more » verification and uncertainty quantification (VVUQ) techniques enable analysts to extract information from experiments in a systematic manner and provide the users with a quantified uncertainty estimate. Unfortunately, the capability to perform experiments that would enable taking full advantage of the formalisms of these modern codes has progressed relatively little (with some notable exceptions in fuels and thermal-hydraulics); the majority of the experimental data available today is the "historic" data accumulated over the last decades of nuclear systems R&D. A validated code-model is a tool for users. An unvalidated code-model is useful for code developers to gain understanding, publish research results, attract funding, etc. As nuclear analysis codes have become more sophisticated, so have the measurement and validation methods and the challenges that confront them. A successful yet cost-effective validation effort requires expertise possessed only by a few, resources possessed only by the well-capitalized (or a willing collective), and a clear, well-defined objective (validating a code that is developed to satisfy the need(s) of an actual user). To that end, the Idaho National Laboratory established the Nuclear Energy Knowledge and Validation Center to address the challenges of modern code validation and to manage the knowledge from past, current, and future experimental campaigns. By pulling together the best minds involved in code development, experiment design, and validation to establish and disseminate best practices and new techniques, the Nuclear Energy Knowledge and Validation Center (NEKVaC or the ‘Center’) will be a resource for industry, DOE Programs, and academia validation efforts.« less

Parameter study of dual-mode space nuclear fission solid core power and propulsion systems, NUROC3A. AMS report No. 1239c

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

Smith, W.W.; Layton, J.P.

1976-09-13

The three-volume report describes a dual-mode nuclear space power and propulsion system concept that employs an advanced solid-core nuclear fission reactor coupled via heat pipes to one of several electric power conversion systems. The NUROC3A systems analysis code was designed to provide the user with performance characteristics of the dual-mode system. Volume 3 describes utilization of the NUROC3A code to produce a detailed parameter study of the system.

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

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

Sartori, E.; Roussin, R.W.

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

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

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

NONE

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

Development of single-copy nuclear intron markers for species-level phylogenetics: Case study with Paullinieae (Sapindaceae).

PubMed

Chery, Joyce G; Sass, Chodon; Specht, Chelsea D

2017-09-01

We developed a bioinformatic pipeline that leverages a publicly available genome and published transcriptomes to design primers in conserved coding sequences flanking targeted introns of single-copy nuclear loci. Paullinieae (Sapindaceae) is used to demonstrate the pipeline. Transcriptome reads phylogenetically closer to the lineage of interest are aligned to the closest genome. Single-nucleotide polymorphisms are called, generating a "pseudoreference" closer to the lineage of interest. Several filters are applied to meet the criteria of single-copy nuclear loci with introns of a desired size. Primers are designed in conserved coding sequences flanking introns. Using this pipeline, we developed nine single-copy nuclear intron markers for Paullinieae. This pipeline is highly flexible and can be used for any group with available genomic and transcriptomic resources. This pipeline led to the development of nine variable markers for phylogenetic study without generating sequence data de novo.

Structural integrity of a confinement vessel for testing nuclear fuels for space propulsion

NASA Astrophysics Data System (ADS)

Bergmann, V. L.

Nuclear propulsion systems for rockets could significantly reduce the travel time to distant destinations in space. However, long before such a concept can become reality, a significant effort must be invested in analysis and ground testing to guide the development of nuclear fuels. Any testing in support of development of nuclear fuels for space propulsion must be safely contained to prevent the release of radioactive materials. This paper describes analyses performed to assess the structural integrity of a test confinement vessel. The confinement structure, a stainless steel pressure vessel with bolted flanges, was designed for operating static pressures in accordance with the ASME Boiler and Pressure Vessel Code. In addition to the static operating pressures, the confinement barrier must withstand static overpressures from off-normal conditions without releasing radioactive material. Results from axisymmetric finite element analyses are used to evaluate the response of the confinement structure under design and accident conditions. For the static design conditions, the stresses computed from the ASME code are compared with the stresses computed by the finite element method.

Interface requirements to couple thermal hydraulics codes to severe accident codes: ICARE/CATHARE

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

Camous, F.; Jacq, F.; Chatelard, P.

1997-07-01

In order to describe with the same code the whole sequence of severe LWR accidents, up to the vessel failure, the Institute of Protection and Nuclear Safety has performed a coupling of the severe accident code ICARE2 to the thermalhydraulics code CATHARE2. The resulting code, ICARE/CATHARE, is designed to be as pertinent as possible in all the phases of the accident. This paper is mainly devoted to the description of the ICARE2-CATHARE2 coupling.

NUclear EVacuation Analysis Code (NUEVAC) : a tool for evaluation of sheltering and evacuation responses following urban nuclear detonations.

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

Yoshimura, Ann S.; Brandt, Larry D.

2009-11-01

The NUclear EVacuation Analysis Code (NUEVAC) has been developed by Sandia National Laboratories to support the analysis of shelter-evacuate (S-E) strategies following an urban nuclear detonation. This tool can model a range of behaviors, including complex evacuation timing and path selection, as well as various sheltering or mixed evacuation and sheltering strategies. The calculations are based on externally generated, high resolution fallout deposition and plume data. Scenario setup and calculation outputs make extensive use of graphics and interactive features. This software is designed primarily to produce quantitative evaluations of nuclear detonation response options. However, the outputs have also proven usefulmore » in the communication of technical insights concerning shelter-evacuate tradeoffs to urban planning or response personnel.« less

Turbopump Design and Analysis Approach for Nuclear Thermal Rockets

NASA Technical Reports Server (NTRS)

Chen, Shu-cheng S.; Veres, Joseph P.; Fittje, James E.

2006-01-01

A rocket propulsion system, whether it is a chemical rocket or a nuclear thermal rocket, is fairly complex in detail but rather simple in principle. Among all the interacting parts, three components stand out: they are pumps and turbines (turbopumps), and the thrust chamber. To obtain an understanding of the overall rocket propulsion system characteristics, one starts from analyzing the interactions among these three components. It is therefore of utmost importance to be able to satisfactorily characterize the turbopump, level by level, at all phases of a vehicle design cycle. Here at NASA Glenn Research Center, as the starting phase of a rocket engine design, specifically a Nuclear Thermal Rocket Engine design, we adopted the approach of using a high level system cycle analysis code (NESS) to obtain an initial analysis of the operational characteristics of a turbopump required in the propulsion system. A set of turbopump design codes (PumpDes and TurbDes) were then executed to obtain sizing and performance characteristics of the turbopump that were consistent with the mission requirements. A set of turbopump analyses codes (PUMPA and TURBA) were applied to obtain the full performance map for each of the turbopump components; a two dimensional layout of the turbopump based on these mean line analyses was also generated. Adequacy of the turbopump conceptual design will later be determined by further analyses and evaluation. In this paper, descriptions and discussions of the aforementioned approach are provided and future outlooks are discussed.

Full 3D visualization tool-kit for Monte Carlo and deterministic transport codes

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

Frambati, S.; Frignani, M.

2012-07-01

We propose a package of tools capable of translating the geometric inputs and outputs of many Monte Carlo and deterministic radiation transport codes into open source file formats. These tools are aimed at bridging the gap between trusted, widely-used radiation analysis codes and very powerful, more recent and commonly used visualization software, thus supporting the design process and helping with shielding optimization. Three main lines of development were followed: mesh-based analysis of Monte Carlo codes, mesh-based analysis of deterministic codes and Monte Carlo surface meshing. The developed kit is considered a powerful and cost-effective tool in the computer-aided design formore » radiation transport code users of the nuclear world, and in particular in the fields of core design and radiation analysis. (authors)« less

10 CFR 50.55a - Codes and standards.

Code of Federal Regulations, 2014 CFR

2014-01-01

..., standard design approval, and standard design certification application under part 52 of this chapter is..., and components must be designed, fabricated, erected, constructed, tested, and inspected to quality... Operation and Maintenance of Nuclear Power Plants; NRC Regulatory Guide (RG) 1.84, Revision 35, “Design...

Nuclear Reaction Models Responsible for Simulation of Neutron-induced Soft Errors in Microelectronics

NASA Astrophysics Data System (ADS)

Watanabe, Y.; Abe, S.

2014-06-01

Terrestrial neutron-induced soft errors in MOSFETs from a 65 nm down to a 25 nm design rule are analyzed by means of multi-scale Monte Carlo simulation using the PHITS-HyENEXSS code system. Nuclear reaction models implemented in PHITS code are validated by comparisons with experimental data. From the analysis of calculated soft error rates, it is clarified that secondary He and H ions provide a major impact on soft errors with decreasing critical charge. It is also found that the high energy component from 10 MeV up to several hundreds of MeV in secondary cosmic-ray neutrons has the most significant source of soft errors regardless of design rule.

Simulation of prompt gamma-ray emission during proton radiotherapy.

PubMed

Verburg, Joost M; Shih, Helen A; Seco, Joao

2012-09-07

The measurement of prompt gamma rays emitted from proton-induced nuclear reactions has been proposed as a method to verify in vivo the range of a clinical proton radiotherapy beam. A good understanding of the prompt gamma-ray emission during proton therapy is key to develop a clinically feasible technique, as it can facilitate accurate simulations and uncertainty analysis of gamma detector designs. Also, the gamma production cross-sections may be incorporated as prior knowledge in the reconstruction of the proton range from the measurements. In this work, we performed simulations of proton-induced nuclear reactions with the main elements of human tissue, carbon-12, oxygen-16 and nitrogen-14, using the nuclear reaction models of the GEANT4 and MCNP6 Monte Carlo codes and the dedicated nuclear reaction codes TALYS and EMPIRE. For each code, we made an effort to optimize the input parameters and model selection. The results of the models were compared to available experimental data of discrete gamma line cross-sections. Overall, the dedicated nuclear reaction codes reproduced the experimental data more consistently, while the Monte Carlo codes showed larger discrepancies for a number of gamma lines. The model differences lead to a variation of the total gamma production near the end of the proton range by a factor of about 2. These results indicate a need for additional theoretical and experimental study of proton-induced gamma emission in human tissue.

Potential Effects of Leak-Before-Break on Light Water Reactor Design.

DTIC Science & Technology

1985-08-26

Boiler and Pressure Vessel Code . In fact, section 3 of that code was created for nuclear applications. This... Boiler and Pressure Vessel Code . The only major change which leak-before-break would require in these analyses would be that all piping to be considered...XI of the ASME Boiler and Pressure Vessel Code , and is already required for all Class I piping systems in the plant. Class I systems are those

Monte Carlo Modeling of the Initial Radiation Emitted by a Nuclear Device in the National Capital Region

DTIC Science & Technology

2013-07-01

also simulated in the models. Data was derived from calculations using the three-dimensional Monte Carlo radiation transport code MCNP (Monte Carlo N...32  B.  MCNP PHYSICS OPTIONS ......................................................................................... 33  C.  HAZUS...input deck’) for the MCNP , Monte Carlo N-Particle, radiation transport code. MCNP is a general-purpose code designed to simulate neutron, photon

Convection and thermal radiation analytical models applicable to a nuclear waste repository room

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

Davis, B.W.

1979-01-17

Time-dependent temperature distributions in a deep geologic nuclear waste repository have a direct impact on the physical integrity of the emplaced canisters and on the design of retrievability options. This report (1) identifies the thermodynamic properties and physical parameters of three convection regimes - forced, natural, and mixed; (2) defines the convection correlations applicable to calculating heat flow in a ventilated (forced-air) and in a nonventilated nuclear waste repository room; and (3) delineates a computer code that (a) computes and compares the floor-to-ceiling heat flow by convection and radiation, and (b) determines the nonlinear equivalent conductivity table for a repositorymore » room. (The tables permit the use of the ADINAT code to model surface-to-surface radiation and the TRUMP code to employ two different emissivity properties when modeling radiation exchange between the surface of two different materials.) The analysis shows that thermal radiation dominates heat flow modes in a nuclear waste repository room.« less

Energy spectrum of 208Pb(n,x) reactions

NASA Astrophysics Data System (ADS)

Tel, E.; Kavun, Y.; Özdoǧan, H.; Kaplan, A.

2018-02-01

Fission and fusion reactor technologies have been investigated since 1950's on the world. For reactor technology, fission and fusion reaction investigations are play important role for improve new generation technologies. Especially, neutron reaction studies have an important place in the development of nuclear materials. So neutron effects on materials should study as theoretically and experimentally for improve reactor design. For this reason, Nuclear reaction codes are very useful tools when experimental data are unavailable. For such circumstances scientists created many nuclear reaction codes such as ALICE/ASH, CEM95, PCROSS, TALYS, GEANT, FLUKA. In this study we used ALICE/ASH, PCROSS and CEM95 codes for energy spectrum calculation of outgoing particles from Pb bombardment by neutron. While Weisskopf-Ewing model has been used for the equilibrium process in the calculations, full exciton, hybrid and geometry dependent hybrid nuclear reaction models have been used for the pre-equilibrium process. The calculated results have been discussed and compared with the experimental data taken from EXFOR.

Deuteron nuclear data for the design of accelerator-based neutron sources: Measurement, model analysis, evaluation, and application

NASA Astrophysics Data System (ADS)

Watanabe, Yukinobu; Kin, Tadahiro; Araki, Shouhei; Nakayama, Shinsuke; Iwamoto, Osamu

2017-09-01

A comprehensive research program on deuteron nuclear data motivated by development of accelerator-based neutron sources is being executed. It is composed of measurements of neutron and gamma-ray yields and production cross sections, modelling of deuteron-induced reactions and code development, nuclear data evaluation and benchmark test, and its application to medical radioisotopes production. The goal of this program is to develop a state-of-the-art deuteron nuclear data library up to 200 MeV which will be useful for the design of future (d,xn) neutron sources. The current status and future plan are reviewed.

Nuclear Reaction Models Responsible for Simulation of Neutron-induced Soft Errors in Microelectronics

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

Watanabe, Y., E-mail: watanabe@aees.kyushu-u.ac.jp; Abe, S.

Terrestrial neutron-induced soft errors in MOSFETs from a 65 nm down to a 25 nm design rule are analyzed by means of multi-scale Monte Carlo simulation using the PHITS-HyENEXSS code system. Nuclear reaction models implemented in PHITS code are validated by comparisons with experimental data. From the analysis of calculated soft error rates, it is clarified that secondary He and H ions provide a major impact on soft errors with decreasing critical charge. It is also found that the high energy component from 10 MeV up to several hundreds of MeV in secondary cosmic-ray neutrons has the most significant sourcemore » of soft errors regardless of design rule.« less

Advances in Geologic Disposal System Modeling and Application to Crystalline Rock

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

Mariner, Paul E.; Stein, Emily R.; Frederick, Jennifer M.

The Used Fuel Disposition Campaign (UFDC) of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Fuel Cycle Technology (OFCT) is conducting research and development (R&D) on geologic disposal of used nuclear fuel (UNF) and high-level nuclear waste (HLW). Two of the high priorities for UFDC disposal R&D are design concept development and disposal system modeling (DOE 2011). These priorities are directly addressed in the UFDC Generic Disposal Systems Analysis (GDSA) work package, which is charged with developing a disposal system modeling and analysis capability for evaluating disposal system performance for nuclear waste in geologic mediamore » (e.g., salt, granite, clay, and deep borehole disposal). This report describes specific GDSA activities in fiscal year 2016 (FY 2016) toward the development of the enhanced disposal system modeling and analysis capability for geologic disposal of nuclear waste. The GDSA framework employs the PFLOTRAN thermal-hydrologic-chemical multi-physics code and the Dakota uncertainty sampling and propagation code. Each code is designed for massively-parallel processing in a high-performance computing (HPC) environment. Multi-physics representations in PFLOTRAN are used to simulate various coupled processes including heat flow, fluid flow, waste dissolution, radionuclide release, radionuclide decay and ingrowth, precipitation and dissolution of secondary phases, and radionuclide transport through engineered barriers and natural geologic barriers to the biosphere. Dakota is used to generate sets of representative realizations and to analyze parameter sensitivity.« less

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

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

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

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

Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2

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

Coryell, E.W.; Siefken, L.J.; Harvego, E.A.

1997-07-01

The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures.more » The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds.« less

Integrated system for production of neutronics and photonics calculational constants. Program SIGMA1 (Version 77-1): Doppler broaden evaluated cross sections in the Evaluated Nuclear Data File/Version B (ENDF/B) format

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

Cullen, D.E.

1977-01-12

A code, SIGMA1, has been designed to Doppler broaden evaluated cross sections in the ENDF/B format. The code can only be applied to tabulated data that vary linearly in energy and cross section between tabulated points. This report describes the methods used in the code and serves as a user's guide to the code.

SCALE Code System

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

Rearden, Bradley T.; Jessee, Matthew Anderson

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministicmore » and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.« less

SCALE Code System 6.2.1

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

Rearden, Bradley T.; Jessee, Matthew Anderson

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministicmore » and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.« less

Development and preliminary verification of the 3D core neutronic code: COCO

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

Lu, H.; Mo, K.; Li, W.

As the recent blooming economic growth and following environmental concerns (China)) is proactively pushing forward nuclear power development and encouraging the tapping of clean energy. Under this situation, CGNPC, as one of the largest energy enterprises in China, is planning to develop its own nuclear related technology in order to support more and more nuclear plants either under construction or being operation. This paper introduces the recent progress in software development for CGNPC. The focus is placed on the physical models and preliminary verification results during the recent development of the 3D Core Neutronic Code: COCO. In the COCO code,more » the non-linear Green's function method is employed to calculate the neutron flux. In order to use the discontinuity factor, the Neumann (second kind) boundary condition is utilized in the Green's function nodal method. Additionally, the COCO code also includes the necessary physical models, e.g. single-channel thermal-hydraulic module, burnup module, pin power reconstruction module and cross-section interpolation module. The preliminary verification result shows that the COCO code is sufficient for reactor core design and analysis for pressurized water reactor (PWR). (authors)« less

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

DOE PAGES

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

2017-08-02

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

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

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

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

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

Common radiation analysis model for 75,000 pound thrust NERVA engine (1137400E)

NASA Technical Reports Server (NTRS)

Warman, E. A.; Lindsey, B. A.

1972-01-01

The mathematical model and sources of radiation used for the radiation analysis and shielding activities in support of the design of the 1137400E version of the 75,000 lbs thrust NERVA engine are presented. The nuclear subsystem (NSS) and non-nuclear components are discussed. The geometrical model for the NSS is two dimensional as required for the DOT discrete ordinates computer code or for an azimuthally symetrical three dimensional Point Kernel or Monte Carlo code. The geometrical model for the non-nuclear components is three dimensional in the FASTER geometry format. This geometry routine is inherent in the ANSC versions of the QAD and GGG Point Kernal programs and the COHORT Monte Carlo program. Data are included pertaining to a pressure vessel surface radiation source data tape which has been used as the basis for starting ANSC analyses with the DASH code to bridge into the COHORT Monte Carlo code using the WANL supplied DOT angular flux leakage data. In addition to the model descriptions and sources of radiation, the methods of analyses are briefly described.

Neutron transport analysis for nuclear reactor design

DOEpatents

Vujic, Jasmina L.

1993-01-01

Replacing regular mesh-dependent ray tracing modules in a collision/transfer probability (CTP) code with a ray tracing module based upon combinatorial geometry of a modified geometrical module (GMC) provides a general geometry transfer theory code in two dimensions (2D) for analyzing nuclear reactor design and control. The primary modification of the GMC module involves generation of a fixed inner frame and a rotating outer frame, where the inner frame contains all reactor regions of interest, e.g., part of a reactor assembly, an assembly, or several assemblies, and the outer frame, with a set of parallel equidistant rays (lines) attached to it, rotates around the inner frame. The modified GMC module allows for determining for each parallel ray (line), the intersections with zone boundaries, the path length between the intersections, the total number of zones on a track, the zone and medium numbers, and the intersections with the outer surface, which parameters may be used in the CTP code to calculate collision/transfer probability and cross-section values.

Neutron transport analysis for nuclear reactor design

DOEpatents

Vujic, J.L.

1993-11-30

Replacing regular mesh-dependent ray tracing modules in a collision/transfer probability (CTP) code with a ray tracing module based upon combinatorial geometry of a modified geometrical module (GMC) provides a general geometry transfer theory code in two dimensions (2D) for analyzing nuclear reactor design and control. The primary modification of the GMC module involves generation of a fixed inner frame and a rotating outer frame, where the inner frame contains all reactor regions of interest, e.g., part of a reactor assembly, an assembly, or several assemblies, and the outer frame, with a set of parallel equidistant rays (lines) attached to it, rotates around the inner frame. The modified GMC module allows for determining for each parallel ray (line), the intersections with zone boundaries, the path length between the intersections, the total number of zones on a track, the zone and medium numbers, and the intersections with the outer surface, which parameters may be used in the CTP code to calculate collision/transfer probability and cross-section values. 28 figures.

Study of a new design of p-N semiconductor detector array for nuclear medicine imaging by monte carlo simulation codes.

PubMed

Hajizadeh-Safar, M; Ghorbani, M; Khoshkharam, S; Ashrafi, Z

2014-07-01

Gamma camera is an important apparatus in nuclear medicine imaging. Its detection part is consists of a scintillation detector with a heavy collimator. Substitution of semiconductor detectors instead of scintillator in these cameras has been effectively studied. In this study, it is aimed to introduce a new design of P-N semiconductor detector array for nuclear medicine imaging. A P-N semiconductor detector composed of N-SnO2 :F, and P-NiO:Li, has been introduced through simulating with MCNPX monte carlo codes. Its sensitivity with different factors such as thickness, dimension, and direction of emission photons were investigated. It is then used to configure a new design of an array in one-dimension and study its spatial resolution for nuclear medicine imaging. One-dimension array with 39 detectors was simulated to measure a predefined linear distribution of Tc(99_m) activity and its spatial resolution. The activity distribution was calculated from detector responses through mathematical linear optimization using LINPROG code on MATLAB software. Three different configurations of one-dimension detector array, horizontal, vertical one sided, and vertical double-sided were simulated. In all of these configurations, the energy windows of the photopeak were ± 1%. The results show that the detector response increases with an increase of dimension and thickness of the detector with the highest sensitivity for emission photons 15-30° above the surface. Horizontal configuration array of detectors is not suitable for imaging of line activity sources. The measured activity distribution with vertical configuration array, double-side detectors, has no similarity with emission sources and hence is not suitable for imaging purposes. Measured activity distribution using vertical configuration array, single side detectors has a good similarity with sources. Therefore, it could be introduced as a suitable configuration for nuclear medicine imaging. It has been shown that using semiconductor P-N detectors such as P-NiO:Li, N-SnO2 :F for gamma detection could be possibly applicable for design of a one dimension array configuration with suitable spatial resolution of 2.7 mm for nuclear medicine imaging.

ENEL overall PWR plant models and neutronic integrated computing systems

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

Pedroni, G.; Pollachini, L.; Vimercati, G.

1987-01-01

To support the design activity of the Italian nuclear energy program for the construction of pressurized water reactors, the Italian Electricity Board (ENEL) needs to verify the design as a whole (that is, the nuclear steam supply system and balance of plant) both in steady-state operation and in transient. The ENEL has therefore developed two computer models to analyze both operational and incidental transients. The models, named STRIP and SFINCS, perform the analysis of the nuclear as well as the conventional part of the plant (the control system being properly taken into account). The STRIP model has been developed bymore » means of the French (Electricite de France) modular code SICLE, while SFINCS is based on the Italian (ENEL) modular code LEGO. STRIP validation was performed with respect to Fessenheim French power plant experimental data. Two significant transients were chosen: load step and total load rejection. SFINCS validation was performed with respect to Saint-Laurent French power plant experimental data and also by comparing the SFINCS-STRIP responses.« less

RELAP-7 Development Updates

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

Zhang, Hongbin; Zhao, Haihua; Gleicher, Frederick Nathan

RELAP-7 is a nuclear systems safety analysis code being developed at the Idaho National Laboratory, and is the next generation tool in the RELAP reactor safety/systems analysis application series. RELAP-7 development began in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway of the Light Water Reactor Sustainability (LWRS) program. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support nuclear power safety analysis. The code is beingmore » developed based on Idaho National Laboratory’s modern scientific software development framework – MOOSE (the Multi-Physics Object-Oriented Simulation Environment). The initial development goal of the RELAP-7 approach focused primarily on the development of an implicit algorithm capable of strong (nonlinear) coupling of the dependent hydrodynamic variables contained in the 1-D/2-D flow models with the various 0-D system reactor components that compose various boiling water reactor (BWR) and pressurized water reactor nuclear power plants (NPPs). During Fiscal Year (FY) 2015, the RELAP-7 code has been further improved with expanded capability to support boiling water reactor (BWR) and pressurized water reactor NPPs analysis. The accumulator model has been developed. The code has also been coupled with other MOOSE-based applications such as neutronics code RattleSnake and fuel performance code BISON to perform multiphysics analysis. A major design requirement for the implicit algorithm in RELAP-7 is that it is capable of second-order discretization accuracy in both space and time, which eliminates the traditional first-order approximation errors. The second-order temporal is achieved by a second-order backward temporal difference, and the one-dimensional second-order accurate spatial discretization is achieved with the Galerkin approximation of Lagrange finite elements. During FY-2015, we have done numerical verification work to verify that the RELAP-7 code indeed achieves 2nd-order accuracy in both time and space for single phase models at the system level.« less

NEAMS Update. Quarterly Report for October - December 2011.

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

Bradley, K.

2012-02-16

The Advanced Modeling and Simulation Office within the DOE Office of Nuclear Energy (NE) has been charged with revolutionizing the design tools used to build nuclear power plants during the next 10 years. To accomplish this, the DOE has brought together the national laboratories, U.S. universities, and the nuclear energy industry to establish the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program. The mission of NEAMS is to modernize computer modeling of nuclear energy systems and improve the fidelity and validity of modeling results using contemporary software environments and high-performance computers. NEAMS will create a set of engineering-level codes aimedmore » at designing and analyzing the performance and safety of nuclear power plants and reactor fuels. The truly predictive nature of these codes will be achieved by modeling the governing phenomena at the spatial and temporal scales that dominate the behavior. These codes will be executed within a simulation environment that orchestrates code integration with respect to spatial meshing, computational resources, and execution to give the user a common 'look and feel' for setting up problems and displaying results. NEAMS is building upon a suite of existing simulation tools, including those developed by the federal Scientific Discovery through Advanced Computing and Advanced Simulation and Computing programs. NEAMS also draws upon existing simulation tools for materials and nuclear systems, although many of these are limited in terms of scale, applicability, and portability (their ability to be integrated into contemporary software and hardware architectures). NEAMS investments have directly and indirectly supported additional NE research and development programs, including those devoted to waste repositories, safeguarded separations systems, and long-term storage of used nuclear fuel. NEAMS is organized into two broad efforts, each comprising four elements. The quarterly highlights October-December 2011 are: (1) Version 1.0 of AMP, the fuel assembly performance code, was tested on the JAGUAR supercomputer and released on November 1, 2011, a detailed discussion of this new simulation tool is given; (2) A coolant sub-channel model and a preliminary UO{sub 2} smeared-cracking model were implemented in BISON, the single-pin fuel code, more information on how these models were developed and benchmarked is given; (3) The Object Kinetic Monte Carlo model was implemented to account for nucleation events in meso-scale simulations and a discussion of the significance of this advance is given; (4) The SHARP neutronics module, PROTEUS, was expanded to be applicable to all types of reactors, and a discussion of the importance of PROTEUS is given; (5) A plan has been finalized for integrating the high-fidelity, three-dimensional reactor code SHARP with both the systems-level code RELAP7 and the fuel assembly code AMP. This is a new initiative; (6) Work began to evaluate the applicability of AMP to the problem of dry storage of used fuel and to define a relevant problem to test the applicability; (7) A code to obtain phonon spectra from the force-constant matrix for a crystalline lattice has been completed. This important bridge between subcontinuum and continuum phenomena is discussed; (8) Benchmarking was begun on the meso-scale, finite-element fuels code MARMOT to validate its new variable splitting algorithm; (9) A very computationally demanding simulation of diffusion-driven nucleation of new microstructural features has been completed. An explanation of the difficulty of this simulation is given; (10) Experiments were conducted with deformed steel to validate a crystal plasticity finite-element code for bodycentered cubic iron; (11) The Capability Transfer Roadmap was completed and published as an internal laboratory technical report; (12) The AMP fuel assembly code input generator was integrated into the NEAMS Integrated Computational Environment (NiCE). More details on the planned NEAMS computing environment is given; and (13) The NEAMS program website (neams.energy.gov) is nearly ready to launch.« less

Nuclear Data Uncertainty Propagation to Reactivity Coefficients of a Sodium Fast Reactor

NASA Astrophysics Data System (ADS)

Herrero, J. J.; Ochoa, R.; Martínez, J. S.; Díez, C. J.; García-Herranz, N.; Cabellos, O.

2014-04-01

The assessment of the uncertainty levels on the design and safety parameters for the innovative European Sodium Fast Reactor (ESFR) is mandatory. Some of these relevant safety quantities are the Doppler and void reactivity coefficients, whose uncertainties are quantified. Besides, the nuclear reaction data where an improvement will certainly benefit the design accuracy are identified. This work has been performed with the SCALE 6.1 codes suite and its multigroups cross sections library based on ENDF/B-VII.0 evaluation.

Preliminary Design of a Manned Nuclear Electric Propulsion Vehicle Using Genetic Algorithms

NASA Technical Reports Server (NTRS)

Irwin, Ryan W.; Tinker, Michael L.

2005-01-01

Nuclear electric propulsion (NEP) vehicles will be needed for future manned missions to Mars and beyond. Candidate designs must be identified for further detailed design from a large array of possibilities. Genetic algorithms have proven their utility in conceptual design studies by effectively searching a large design space to pinpoint unique optimal designs. This research combined analysis codes for NEP subsystems with a genetic algorithm. The use of penalty functions with scaling ratios was investigated to increase computational efficiency. Also, the selection of design variables for optimization was considered to reduce computation time without losing beneficial design search space. Finally, trend analysis of a reference mission to the asteroids yielded a group of candidate designs for further analysis.

Risk Informed Design and Analysis Criteria for Nuclear Structures

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

Salmon, Michael W.

2015-06-17

Target performance can be achieved by defining design basis ground motion from results of a probabilistic seismic hazards assessment, and introducing known levels of conservatism in the design above the DBE. ASCE 4, 43, DOE-STD-1020 defined the DBE at 4x10-4 and introduce only slight levels of conservatism in response. ASCE 4, 43, DOE-STD-1020 assume code capacities shoot for about 98% NEP. There is a need to have a uniform target (98% NEP) for code developers (ACI, AISC, etc.) to aim for. In considering strengthening options, one must also consider cost/risk reduction achieved.

Overview of the Graphical User Interface for the GERM Code (GCR Event-Based Risk Model

NASA Technical Reports Server (NTRS)

Kim, Myung-Hee; Cucinotta, Francis A.

2010-01-01

The descriptions of biophysical events from heavy ions are of interest in radiobiology, cancer therapy, and space exploration. The biophysical description of the passage of heavy ions in tissue and shielding materials is best described by a stochastic approach that includes both ion track structure and nuclear interactions. A new computer model called the GCR Event-based Risk Model (GERM) code was developed for the description of biophysical events from heavy ion beams at the NASA Space Radiation Laboratory (NSRL). The GERM code calculates basic physical and biophysical quantities of high-energy protons and heavy ions that have been studied at NSRL for the purpose of simulating space radiobiological effects. For mono-energetic beams, the code evaluates the linear-energy transfer (LET), range (R), and absorption in tissue equivalent material for a given Charge (Z), Mass Number (A) and kinetic energy (E) of an ion. In addition, a set of biophysical properties are evaluated such as the Poisson distribution of ion or delta-ray hits for a specified cellular area, cell survival curves, and mutation and tumor probabilities. The GERM code also calculates the radiation transport of the beam line for either a fixed number of user-specified depths or at multiple positions along the Bragg curve of the particle. The contributions from primary ion and nuclear secondaries are evaluated. The GERM code accounts for the major nuclear interaction processes of importance for describing heavy ion beams, including nuclear fragmentation, elastic scattering, and knockout-cascade processes by using the quantum multiple scattering fragmentation (QMSFRG) model. The QMSFRG model has been shown to be in excellent agreement with available experimental data for nuclear fragmentation cross sections, and has been used by the GERM code for application to thick target experiments. The GERM code provides scientists participating in NSRL experiments with the data needed for the interpretation of their experiments, including the ability to model the beam line, the shielding of samples and sample holders, and the estimates of basic physical and biological outputs of the designed experiments. We present an overview of the GERM code GUI, as well as providing training applications.

The virtual digital nuclear power plant: A modern tool for supporting the lifecycle of VVER-based nuclear power units

NASA Astrophysics Data System (ADS)

Arkadov, G. V.; Zhukavin, A. P.; Kroshilin, A. E.; Parshikov, I. A.; Solov'ev, S. L.; Shishov, A. V.

2014-10-01

The article describes the "Virtual Digital VVER-Based Nuclear Power Plant" computerized system comprising a totality of verified initial data (sets of input data for a model intended for describing the behavior of nuclear power plant (NPP) systems in design and emergency modes of their operation) and a unified system of new-generation computation codes intended for carrying out coordinated computation of the variety of physical processes in the reactor core and NPP equipment. Experiments with the demonstration version of the "Virtual Digital VVER-Based NPP" computerized system has shown that it is in principle possible to set up a unified system of computation codes in a common software environment for carrying out interconnected calculations of various physical phenomena at NPPs constructed according to the standard AES-2006 project. With the full-scale version of the "Virtual Digital VVER-Based NPP" computerized system put in operation, the concerned engineering, design, construction, and operating organizations will have access to all necessary information relating to the NPP power unit project throughout its entire lifecycle. The domestically developed commercial-grade software product set to operate as an independently operating application to the project will bring about additional competitive advantages in the modern market of nuclear power technologies.

Design and implementation of a nuclear weapons management system submodule: Shipboard security force system. Master's thesis

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

Settlemyer, S.R.

1991-09-01

The Nuclear Weapons Management System combines the strengths of an expert system with the flexibility of a database management system to assist the Weapons Officer, Security Officer, and the Personnel Reliability Program Officer in the performance of administrative duties associated with the nuclear weapons programs in the United States Navy. This thesis examines the need for, and ultimately the design of, a system that will assist the Security Officer in administrative duties associated with the Shipboard Self Defense Force. This system, designed and coded utilizing dBASE IV, can be implemented as a stand alone system. Furthermore, it interfaces with themore » expert system submodule that handles the PRP screening process.« less

ASME code considerations for the compact heat exchanger

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

Nestell, James; Sham, Sam

2015-08-31

The mission of the U.S. Department of Energy (DOE), Office of Nuclear Energy is to advance nuclear power in order to meet the nation's energy, environmental, and energy security needs. Advanced high temperature reactor systems such as sodium fast reactors and high and very high temperature gas-cooled reactors are being considered for the next generation of nuclear reactor plant designs. The coolants for these high temperature reactor systems include liquid sodium and helium gas. Supercritical carbon dioxide (sCO₂), a fluid at a temperature and pressure above the supercritical point of CO₂, is currently being investigated by DOE as a workingmore » fluid for a nuclear or fossil-heated recompression closed Brayton cycle energy conversion system that operates at 550°C (1022°F) at 200 bar (2900 psi). Higher operating temperatures are envisioned in future developments. All of these design concepts require a highly effective heat exchanger that transfers heat from the nuclear or chemical reactor to the chemical process fluid or the to the power cycle. In the nuclear designs described above, heat is transferred from the primary to the secondary loop via an intermediate heat exchanger (IHX) and then from the intermediate loop to either a working process or a power cycle via a secondary heat exchanger (SHX). The IHX is a component in the primary coolant loop which will be classified as "safety related." The intermediate loop will likely be classified as "not safety related but important to safety." These safety classifications have a direct bearing on heat exchanger design approaches for the IHX and SHX. The very high temperatures being considered for the VHTR will require the use of very high temperature alloys for the IHX and SHX. Material cost considerations alone will dictate that the IHX and SHX be highly effective; that is, provide high heat transfer area in a small volume. This feature must be accompanied by low pressure drop and mechanical reliability and robustness. Classic shell and tube designs will be large and costly, and may only be appropriate in steam generator service in the SHX where boiling inside the tubes occurs. For other energy conversion systems, all of these features can be met in a compact heat exchanger design. This report will examine some of the ASME Code issues that will need to be addressed to allow use of a Code-qualified compact heat exchanger in IHX or SHX nuclear service. Most effort will focus on the IHX, since the safety-related (Class A) design rules are more extensive than those for important-to-safety (Class B) or commercial rules that are relevant to the SHX.« less

Science based integrated approach to advanced nuclear fuel development - integrated multi-scale multi-physics hierarchical modeling and simulation framework Part III: cladding

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

Tome, Carlos N; Caro, J A; Lebensohn, R A

2010-01-01

Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Reactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems to develop predictive tools is critical. Not only are fabrication and performance models needed to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating themore » phase and microstructural behavior of the nuclear fuel system materials and matrices. In this paper we review the current status of the advanced modeling and simulation of nuclear reactor cladding, with emphasis on what is available and what is to be developed in each scale of the project, how we propose to pass information from one scale to the next, and what experimental information is required for benchmarking and advancing the modeling at each scale level.« less

Modeling and simulation of CANDU reactor and its regulating system

NASA Astrophysics Data System (ADS)

Javidnia, Hooman

Analytical computer codes are indispensable tools in design, optimization, and control of nuclear power plants. Numerous codes have been developed to perform different types of analyses related to the nuclear power plants. A large number of these codes are designed to perform safety analyses. In the context of safety analyses, the control system is often neglected. Although there are good reasons for such a decision, that does not mean that the study of control systems in the nuclear power plants should be neglected altogether. In this thesis, a proof of concept code is developed as a tool that can be used in the design. optimization. and operation stages of the control system. The main objective in the design of this computer code is providing a tool that is easy to use by its target audience and is capable of producing high fidelity results that can be trusted to design the control system and optimize its performance. Since the overall plant control system covers a very wide range of processes, in this thesis the focus has been on one particular module of the the overall plant control system, namely, the reactor regulating system. The center of the reactor regulating system is the CANDU reactor. A nodal model for the reactor is used to represent the spatial neutronic kinetics of the core. The nodal model produces better results compared to the point kinetics model which is often used in the design and analysis of control system for nuclear reactors. The model can capture the spatial effects to some extent. although it is not as detailed as the finite difference methods. The criteria for choosing a nodal model of the core are: (1) the model should provide more detail than point kinetics and capture spatial effects, (2) it should not be too complex or overly detailed to slow down the simulation and provide details that are extraneous or unnecessary for a control engineer. Other than the reactor itself, there are auxiliary models that describe dynamics of different phenomena related to the transfer of the energy from the core. The main function of the reactor regulating system is to control the power of the reactor. This is achieved by using a set of detectors. reactivity devices. and digital control algorithms. Three main reactivity devices that are activated during short-term or intermediate-term transients are modeled in this thesis. The main elements of the digital control system are implemented in accordance to the program specifications for the actual control system in CANDU reactors. The simulation results are validated against requirements of the reactor regulating system. actual plant data. and pre-validated data from other computer codes. The validation process shows that the simulation results can be trusted in making engineering decisions regarding the reactor regulating system and prediction of the system performance in response to upset conditions or disturbances. KEYWORDS: CANDU reactors. reactor regulating system. nodal model. spatial kinetics. reactivity devices. simulation.

Minimizing human error in radiopharmaceutical preparation and administration via a bar code-enhanced nuclear pharmacy management system.

PubMed

Hakala, John L; Hung, Joseph C; Mosman, Elton A

2012-09-01

The objective of this project was to ensure correct radiopharmaceutical administration through the use of a bar code system that links patient and drug profiles with on-site information management systems. This new combined system would minimize the amount of manual human manipulation, which has proven to be a primary source of error. The most common reason for dosing errors is improper patient identification when a dose is obtained from the nuclear pharmacy or when a dose is administered. A standardized electronic transfer of information from radiopharmaceutical preparation to injection will further reduce the risk of misadministration. Value stream maps showing the flow of the patient dose information, as well as potential points of human error, were developed. Next, a future-state map was created that included proposed corrections for the most common critical sites of error. Transitioning the current process to the future state will require solutions that address these sites. To optimize the future-state process, a bar code system that links the on-site radiology management system with the nuclear pharmacy management system was proposed. A bar-coded wristband connects the patient directly to the electronic information systems. The bar code-enhanced process linking the patient dose with the electronic information reduces the number of crucial points for human error and provides a framework to ensure that the prepared dose reaches the correct patient. Although the proposed flowchart is designed for a site with an in-house central nuclear pharmacy, much of the framework could be applied by nuclear medicine facilities using unit doses. An electronic connection between information management systems to allow the tracking of a radiopharmaceutical from preparation to administration can be a useful tool in preventing the mistakes that are an unfortunate reality for any facility.

The FLUKA Code: An Overview

NASA Technical Reports Server (NTRS)

Ballarini, F.; Battistoni, G.; Campanella, M.; Carboni, M.; Cerutti, F.; Empl, A.; Fasso, A.; Ferrari, A.; Gadioli, E.; Garzelli, M. V.;

RELAP-7 Code Assessment Plan and Requirement Traceability Matrix

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

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

2016-10-01

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

Design and Analysis of Boiler Pressure Vessels based on IBR codes

NASA Astrophysics Data System (ADS)

Balakrishnan, B.; Kanimozhi, B.

2017-05-01

Pressure vessels components are widely used in the thermal and nuclear power plants for generating steam using the philosophy of heat transfer. In Thermal power plant, Coal is burnt inside the boiler furnace for generating the heat. The amount of heat produced through the combustion of pulverized coal is used in changing the phase transfer (i.e. Water into Super-Heated Steam) in the Pressure Parts Component. Pressure vessels are designed as per the Standards and Codes of the country, where the boiler is to be installed. One of the Standards followed in designing Pressure Parts is ASME (American Society of Mechanical Engineers). The mandatory requirements of ASME code must be satisfied by the manufacturer. In our project case, A Shell/pipe which has been manufactured using ASME code has an issue during the drilling of hole. The Actual Size of the drilled holes must be, as per the drawing, but due to error, the size has been differentiate from approved design calculation (i.e. the diameter size has been exceeded). In order to rectify this error, we have included an additional reinforcement pad to the drilled and modified the design of header in accordance with the code requirements.

Mapping Nuclear Fallout Using the Weather Research & Forecasting (WRF) Model

DTIC Science & Technology

2012-09-01

relevant modules, originally designed to predict the settling of volcanic ash, such that a stabilized cloud of nuclear particulate is initialized...within the model. This modified code is then executed for various atmospheric test explosions and the results are qualitatively and quantitatively...HYSPLIT Simulation ....................................... 44  Figure 7. WRF Fallout Prediction for Test Shot George, 0.8 R/h at H+1

Reformation of Regulatory Technical Standards for Nuclear Power Generation Equipments in Japan

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

Mikio Kurihara; Masahiro Aoki; Yu Maruyama

2006-07-01

Comprehensive reformation of the regulatory system has been introduced in Japan in order to apply recent technical progress in a timely manner. 'The Technical Standards for Nuclear Power Generation Equipments', known as the Ordinance No.622) of the Ministry of International Trade and Industry, which is used for detailed design, construction and operating stage of Nuclear Power Plants, was being modified to performance specifications with the consensus codes and standards being used as prescriptive specifications, in order to facilitate prompt review of the Ordinance with response to technological innovation. The activities on modification were performed by the Nuclear and Industrial Safetymore » Agency (NISA), the regulatory body in Japan, with support of the Japan Nuclear Energy Safety Organization (JNES), a technical support organization. The revised Ordinance No.62 was issued on July 1, 2005 and is enforced from January 1 2006. During the period from the issuance to the enforcement, JNES carried out to prepare enforceable regulatory guide which complies with each provisions of the Ordinance No.62, and also made technical assessment to endorse the applicability of consensus codes and standards, in response to NISA's request. Some consensus codes and standards were re-assessed since they were already used in regulatory review of the construction plan submitted by licensee. Other consensus codes and standards were newly assessed for endorsement. In case that proper consensus code or standards were not prepared, details of regulatory requirements were described in the regulatory guide as immediate measures. At the same time, appropriate standards developing bodies were requested to prepare those consensus code or standards. Supplementary note which provides background information on the modification, applicable examples etc. was prepared for convenience to the users of the Ordinance No. 62. This paper shows the activities on modification and the results, following the NISA's presentation at ICONE-13 that introduced the framework of the performance specifications and the modification process of the Ordinance NO. 62. (authors)« less

A Dynamic/Anisotropic Low Earth Orbit (LEO) Ionizing Radiation Model

NASA Technical Reports Server (NTRS)

Badavi, Francis F.; West, Katie J.; Nealy, John E.; Wilson, John W.; Abrahms, Briana L.; Luetke, Nathan J.

2006-01-01

The International Space Station (ISS) provides the proving ground for future long duration human activities in space. Ionizing radiation measurements in ISS form the ideal tool for the experimental validation of ionizing radiation environmental models, nuclear transport code algorithms, and nuclear reaction cross sections. Indeed, prior measurements on the Space Transportation System (STS; Shuttle) have provided vital information impacting both the environmental models and the nuclear transport code development by requiring dynamic models of the Low Earth Orbit (LEO) environment. Previous studies using Computer Aided Design (CAD) models of the evolving ISS configurations with Thermo Luminescent Detector (TLD) area monitors, demonstrated that computational dosimetry requires environmental models with accurate non-isotropic as well as dynamic behavior, detailed information on rack loading, and an accurate 6 degree of freedom (DOF) description of ISS trajectory and orientation.

RELAP-7 Software Verification and Validation Plan

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

Smith, Curtis L.; Choi, Yong-Joon; Zou, Ling

This INL plan comprehensively describes the software for RELAP-7 and documents the software, interface, and software design requirements for the application. The plan also describes the testing-based software verification and validation (SV&V) process—a set of specially designed software models used to test RELAP-7. The RELAP-7 (Reactor Excursion and Leak Analysis Program) code is a nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on the INL’s modern scientific software development framework – MOOSE (Multi-Physics Object-Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty yearsmore » of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5’s capability and extends the analysis capability for all reactor system simulation scenarios.« less

Neutronic calculation of fast reactors by the EUCLID/V1 integrated code

NASA Astrophysics Data System (ADS)

Koltashev, D. A.; Stakhanova, A. A.

2017-01-01

This article considers neutronic calculation of a fast-neutron lead-cooled reactor BREST-OD-300 by the EUCLID/V1 integrated code. The main goal of development and application of integrated codes is a nuclear power plant safety justification. EUCLID/V1 is integrated code designed for coupled neutronics, thermomechanical and thermohydraulic fast reactor calculations under normal and abnormal operating conditions. EUCLID/V1 code is being developed in the Nuclear Safety Institute of the Russian Academy of Sciences. The integrated code has a modular structure and consists of three main modules: thermohydraulic module HYDRA-IBRAE/LM/V1, thermomechanical module BERKUT and neutronic module DN3D. In addition, the integrated code includes databases with fuel, coolant and structural materials properties. Neutronic module DN3D provides full-scale simulation of neutronic processes in fast reactors. Heat sources distribution, control rods movement, reactivity level changes and other processes can be simulated. Neutron transport equation in multigroup diffusion approximation is solved. This paper contains some calculations implemented as a part of EUCLID/V1 code validation. A fast-neutron lead-cooled reactor BREST-OD-300 transient simulation (fuel assembly floating, decompression of passive feedback system channel) and cross-validation with MCU-FR code results are presented in this paper. The calculations demonstrate EUCLID/V1 code application for BREST-OD-300 simulating and safety justification.

The Design of PSB-VVER Experiments Relevant to Accident Management

NASA Astrophysics Data System (ADS)

Nevo, Alessandro Del; D'Auria, Francesco; Mazzini, Marino; Bykov, Michael; Elkin, Ilya V.; Suslov, Alexander

Experimental programs carried-out in integral test facilities are relevant for validating the best estimate thermal-hydraulic codes(1), which are used for accident analyses, design of accident management procedures, licensing of nuclear power plants, etc. The validation process, in fact, is based on well designed experiments. It consists in the comparison of the measured and calculated parameters and the determination whether a computer code has an adequate capability in predicting the major phenomena expected to occur in the course of transient and/or accidents. University of Pisa was responsible of the numerical design of the 12 experiments executed in PSB-VVER facility (2), operated at Electrogorsk Research and Engineering Center (Russia), in the framework of the TACIS 2.03/97 Contract 3.03.03 Part A, EC financed (3). The paper describes the methodology adopted at University of Pisa, starting form the scenarios foreseen in the final test matrix until the execution of the experiments. This process considers three key topics: a) the scaling issue and the simulation, with unavoidable distortions, of the expected performance of the reference nuclear power plants; b) the code assessment process involving the identification of phenomena challenging the code models; c) the features of the concerned integral test facility (scaling limitations, control logics, data acquisition system, instrumentation, etc.). The activities performed in this respect are discussed, and emphasis is also given to the relevance of the thermal losses to the environment. This issue affects particularly the small scaled facilities and has relevance on the scaling approach related to the power and volume of the facility.

User input verification and test driven development in the NJOY21 nuclear data processing code

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

Trainer, Amelia Jo; Conlin, Jeremy Lloyd; McCartney, Austin Paul

Before physically-meaningful data can be used in nuclear simulation codes, the data must be interpreted and manipulated by a nuclear data processing code so as to extract the relevant quantities (e.g. cross sections and angular distributions). Perhaps the most popular and widely-trusted of these processing codes is NJOY, which has been developed and improved over the course of 10 major releases since its creation at Los Alamos National Laboratory in the mid-1970’s. The current phase of NJOY development is the creation of NJOY21, which will be a vast improvement from its predecessor, NJOY2016. Designed to be fast, intuitive, accessible, andmore » capable of handling both established and modern formats of nuclear data, NJOY21 will address many issues that many NJOY users face, while remaining functional for those who prefer the existing format. Although early in its development, NJOY21 is quickly providing input validation to check user input. By providing rapid and helpful responses to users while writing input files, NJOY21 will prove to be more intuitive and easy to use than any of its predecessors. Furthermore, during its development, NJOY21 is subject to regular testing, such that its test coverage must strictly increase with the addition of any production code. This thorough testing will allow developers and NJOY users to establish confidence in NJOY21 as it gains functionality. This document serves as a discussion regarding the current state input checking and testing practices of NJOY21.« less

Current and anticipated uses of thermal hydraulic codes in Korea

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

Kim, Kyung-Doo; Chang, Won-Pyo

1997-07-01

In Korea, the current uses of thermal hydraulic codes are categorized into 3 areas. The first application is in designing both nuclear fuel and NSSS. The codes have usually been introduced based on the technology transfer programs agreed between KAERI and the foreign vendors. Another area is in the supporting of the plant operations and licensing by the utility. The third category is research purposes. In this area assessments and some applications to the safety issue resolutions are major activities using the best estimate thermal hydraulic codes such as RELAP5/MOD3 and CATHARE2. Recently KEPCO plans to couple thermal hydraulic codesmore » with a neutronics code for the design of the evolutionary type reactor by 2004. KAERI also plans to develop its own best estimate thermal hydraulic code, however, application range is different from KEPCO developing code. Considering these activities, it is anticipated that use of the best estimate hydraulic analysis code developed in Korea may be possible in the area of safety evaluation within 10 years.« less

Thermal and flow analysis of the Fluor Daniel, Inc., Nuclear Material Storage Facility renovation design (initial 30% effort of Title 1)

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

Steinke, R.G.; Mueller, C.; Knight, T.D.

1998-03-01

The computational fluid dynamics code CFX4.2 was used to evaluate steady-state thermal-hydraulic conditions in the Fluor Daniel, Inc., Nuclear Material Storage Facility renovation design (initial 30% of Title 1). Thirteen facility cases were evaluated with varying temperature dependence, drywell-array heat-source magnitude and distribution, location of the inlet tower, and no-flow curtains in the drywell-array vault. Four cases of a detailed model of the inlet-tower top fixture were evaluated to show the effect of the canopy-cruciform fixture design on the air pressure and flow distributions.

Integrated nuclear data utilisation system for innovative reactors.

PubMed

Yamano, N; Hasegawa, A; Kato, K; Igashira, M

2005-01-01

A five-year research and development project on an integrated nuclear data utilisation system was initiated in 2002, for developing innovative nuclear energy systems such as accelerator-driven systems. The integrated nuclear data utilisation system will be constructed as a modular code system, which consists of two sub-systems: the nuclear data search and plotting sub-system, and the nuclear data processing and utilisation sub-system. The system will be operated with a graphical user interface in order to enable easy utilisation through the Internet by both nuclear design engineers and nuclear data evaluators. This paper presents an overview of the integrated nuclear data utilisation system, describes the development of a prototype system to examine the operability of the user interface and discusses specifications of the two sub-systems.

Progress Towards a Rad-Hydro Code for Modern Computing Architectures LA-UR-10-02825

NASA Astrophysics Data System (ADS)

Wohlbier, J. G.; Lowrie, R. B.; Bergen, B.; Calef, M.

2010-11-01

We are entering an era of high performance computing where data movement is the overwhelming bottleneck to scalable performance, as opposed to the speed of floating-point operations per processor. All multi-core hardware paradigms, whether heterogeneous or homogeneous, be it the Cell processor, GPGPU, or multi-core x86, share this common trait. In multi-physics applications such as inertial confinement fusion or astrophysics, one may be solving multi-material hydrodynamics with tabular equation of state data lookups, radiation transport, nuclear reactions, and charged particle transport in a single time cycle. The algorithms are intensely data dependent, e.g., EOS, opacity, nuclear data, and multi-core hardware memory restrictions are forcing code developers to rethink code and algorithm design. For the past two years LANL has been funding a small effort referred to as Multi-Physics on Multi-Core to explore ideas for code design as pertaining to inertial confinement fusion and astrophysics applications. The near term goals of this project are to have a multi-material radiation hydrodynamics capability, with tabular equation of state lookups, on cartesian and curvilinear block structured meshes. In the longer term we plan to add fully implicit multi-group radiation diffusion and material heat conduction, and block structured AMR. We will report on our progress to date.

Methods for nuclear air-cleaning-system accident-consequence assessment

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

Andrae, R.W.; Bolstad, J.W.; Gregory, W.S.

1982-01-01

This paper describes a multilaboratory research program that is directed toward addressing many questions that analysts face when performing air cleaning accident consequence assessments. The program involves developing analytical tools and supportive experimental data that will be useful in making more realistic assessments of accident source terms within and up to the atmospheric boundaries of nuclear fuel cycle facilities. The types of accidents considered in this study includes fires, explosions, spills, tornadoes, criticalities, and equipment failures. The main focus of the program is developing an accident analysis handbook (AAH). We will describe the contents of the AAH, which include descriptionsmore » of selected nuclear fuel cycle facilities, process unit operations, source-term development, and accident consequence analyses. Three computer codes designed to predict gas and material propagation through facility air cleaning systems are described. These computer codes address accidents involving fires (FIRAC), explosions (EXPAC), and tornadoes (TORAC). The handbook relies on many illustrative examples to show the analyst how to approach accident consequence assessments. We will use the FIRAC code and a hypothetical fire scenario to illustrate the accident analysis capability.« less

Convergence studies of deterministic methods for LWR explicit reflector methodology

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

Canepa, S.; Hursin, M.; Ferroukhi, H.

2013-07-01

The standard approach in modem 3-D core simulators, employed either for steady-state or transient simulations, is to use Albedo coefficients or explicit reflectors at the core axial and radial boundaries. In the latter approach, few-group homogenized nuclear data are a priori produced with lattice transport codes using 2-D reflector models. Recently, the explicit reflector methodology of the deterministic CASMO-4/SIMULATE-3 code system was identified to potentially constitute one of the main sources of errors for core analyses of the Swiss operating LWRs, which are all belonging to GII design. Considering that some of the new GIII designs will rely on verymore » different reflector concepts, a review and assessment of the reflector methodology for various LWR designs appeared as relevant. Therefore, the purpose of this paper is to first recall the concepts of the explicit reflector modelling approach as employed by CASMO/SIMULATE. Then, for selected reflector configurations representative of both GII and GUI designs, a benchmarking of the few-group nuclear data produced with the deterministic lattice code CASMO-4 and its successor CASMO-5, is conducted. On this basis, a convergence study with regards to geometrical requirements when using deterministic methods with 2-D homogenous models is conducted and the effect on the downstream 3-D core analysis accuracy is evaluated for a typical GII deflector design in order to assess the results against available plant measurements. (authors)« less

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

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.

Safety system augmentation at Russian nuclear power plants

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

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

1996-12-31

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

Summary of papers on current and anticipated uses of thermal-hydraulic codes

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

Caruso, R.

1997-07-01

The author reviews a range of recent papers which discuss possible uses and future development needs for thermal/hydraulic codes in the nuclear industry. From this review, eight common recommendations are extracted. They are: improve the user interface so that more people can use the code, so that models are easier and less expensive to prepare and maintain, and so that the results are scrutable; design the code so that it can easily be coupled to other codes, such as core physics, containment, fission product behaviour during severe accidents; improve the numerical methods to make the code more robust and especiallymore » faster running, particularly for low pressure transients; ensure that future code development includes assessment of code uncertainties as integral part of code verification and validation; provide extensive user guidelines or structure the code so that the `user effect` is minimized; include the capability to model multiple fluids (gas and liquid phase); design the code in a modular fashion so that new models can be added easily; provide the ability to include detailed or simplified component models; build on work previously done with other codes (RETRAN, RELAP, TRAC, CATHARE) and other code validation efforts (CSAU, CSNI SET and IET matrices).« less

Monte Carlo simulations in Nuclear Medicine

NASA Astrophysics Data System (ADS)

Loudos, George K.

2007-11-01

Molecular imaging technologies provide unique abilities to localise signs of disease before symptoms appear, assist in drug testing, optimize and personalize therapy, and assess the efficacy of treatment regimes for different types of cancer. Monte Carlo simulation packages are used as an important tool for the optimal design of detector systems. In addition they have demonstrated potential to improve image quality and acquisition protocols. Many general purpose (MCNP, Geant4, etc) or dedicated codes (SimSET etc) have been developed aiming to provide accurate and fast results. Special emphasis will be given to GATE toolkit. The GATE code currently under development by the OpenGATE collaboration is the most accurate and promising code for performing realistic simulations. The purpose of this article is to introduce the non expert reader to the current status of MC simulations in nuclear medicine and briefly provide examples of current simulated systems, and present future challenges that include simulation of clinical studies and dosimetry applications.

Extremely accurate sequential verification of RELAP5-3D

DOE PAGES

Mesina, George L.; Aumiller, David L.; Buschman, Francis X.

2015-11-19

Large computer programs like RELAP5-3D solve complex systems of governing, closure and special process equations to model the underlying physics of nuclear power plants. Further, these programs incorporate many other features for physics, input, output, data management, user-interaction, and post-processing. For software quality assurance, the code must be verified and validated before being released to users. For RELAP5-3D, verification and validation are restricted to nuclear power plant applications. Verification means ensuring that the program is built right by checking that it meets its design specifications, comparing coding to algorithms and equations and comparing calculations against analytical solutions and method ofmore » manufactured solutions. Sequential verification performs these comparisons initially, but thereafter only compares code calculations between consecutive code versions to demonstrate that no unintended changes have been introduced. Recently, an automated, highly accurate sequential verification method has been developed for RELAP5-3D. The method also provides to test that no unintended consequences result from code development in the following code capabilities: repeating a timestep advancement, continuing a run from a restart file, multiple cases in a single code execution, and modes of coupled/uncoupled operation. In conclusion, mathematical analyses of the adequacy of the checks used in the comparisons are provided.« less

Design-Load Basis for LANL Structures, Systems, and Components

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

I. Cuesta

2004-09-01

This document supports the recommendations in the Los Alamos National Laboratory (LANL) Engineering Standard Manual (ESM), Chapter 5--Structural providing the basis for the loads, analysis procedures, and codes to be used in the ESM. It also provides the justification for eliminating the loads to be considered in design, and evidence that the design basis loads are appropriate and consistent with the graded approach required by the Department of Energy (DOE) Code of Federal Regulation Nuclear Safety Management, 10, Part 830. This document focuses on (1) the primary and secondary natural phenomena hazards listed in DOE-G-420.1-2, Appendix C, (2) additional loadsmore » not related to natural phenomena hazards, and (3) the design loads on structures during construction.« less

Re-evaluation of Spent Nuclear Fuel Assay Data for the Three Mile Island Unit 1 Reactor and Application to Code Validation

DOE PAGES

Gauld, Ian C.; Giaquinto, J. M.; Delashmitt, J. S.; ...

2016-01-01

Destructive radiochemical assay measurements of spent nuclear fuel rod segments from an assembly irradiated in the Three Mile Island unit 1 (TMI-1) pressurized water reactor have been performed at Oak Ridge National Laboratory (ORNL). Assay data are reported for five samples from two fuel rods of the same assembly. The TMI-1 assembly was a 15 X 15 design with an initial enrichment of 4.013 wt% 235U, and the measured samples achieved burnups between 45.5 and 54.5 gigawatt days per metric ton of initial uranium (GWd/t). Measurements were performed mainly using inductively coupled plasma mass spectrometry after elemental separation via highmore » performance liquid chromatography. High precision measurements were achieved using isotope dilution techniques for many of the lanthanides, uranium, and plutonium isotopes. Measurements are reported for more than 50 different isotopes and 16 elements. One of the two TMI-1 fuel rods measured in this work had been measured previously by Argonne National Laboratory (ANL), and these data have been widely used to support code and nuclear data validation. Recently, ORNL provided an important opportunity to independently cross check results against previous measurements performed at ANL. The measured nuclide concentrations are used to validate burnup calculations using the SCALE nuclear systems modeling and simulation code suite. These results show that the new measurements provide reliable benchmark data for computer code validation.« less

Security Hardened Cyber Components for Nuclear Power Plants: Phase I SBIR Final Technical Report

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

Franusich, Michael D.

SpiralGen, Inc. built a proof-of-concept toolkit for enhancing the cyber security of nuclear power plants and other critical infrastructure with high-assurance instrumentation and control code. The toolkit is based on technology from the DARPA High-Assurance Cyber Military Systems (HACMS) program, which has focused on applying the science of formal methods to the formidable set of problems involved in securing cyber physical systems. The primary challenges beyond HACMS in developing this toolkit were to make the new technology usable by control system engineers and compatible with the regulatory and commercial constraints of the nuclear power industry. The toolkit, packaged as amore » Simulink add-on, allows a system designer to assemble a high-assurance component from formally specified and proven blocks and generate provably correct control and monitor code for that subsystem.« less

Two-dimensional over-all neutronics analysis of the ITER device

NASA Astrophysics Data System (ADS)

Zimin, S.; Takatsu, Hideyuki; Mori, Seiji; Seki, Yasushi; Satoh, Satoshi; Tada, Eisuke; Maki, Koichi

1993-07-01

The present work attempts to carry out a comprehensive neutronics analysis of the International Thermonuclear Experimental Reactor (ITER) developed during the Conceptual Design Activities (CDA). The two-dimensional cylindrical over-all calculational models of ITER CDA device including the first wall, blanket, shield, vacuum vessel, magnets, cryostat and support structures were developed for this purpose with a help of the DOGII code. Two dimensional DOT 3.5 code with the FUSION-40 nuclear data library was employed for transport calculations of neutron and gamma ray fluxes, tritium breeding ratio (TBR), and nuclear heating in reactor components. The induced activity calculational code CINAC was employed for the calculations of exposure dose rate after reactor shutdown around the ITER CDA device. The two-dimensional over-all calculational model includes the design specifics such as the pebble bed Li2O/Be layered blanket, the thin double wall vacuum vessel, the concrete cryostat integrated with the over-all ITER design, the top maintenance shield plug, the additional ring biological shield placed under the top cryostat lid around the above-mentioned top maintenance shield plug etc. All the above-mentioned design specifics were included in the employed calculational models. Some alternative design options, such as the water-rich shielding blanket instead of lithium-bearing one, the additional biological shield plug at the top zone between the poloidal field (PF) coil No. 5, and the maintenance shield plug, were calculated as well. Much efforts have been focused on analyses of obtained results. These analyses aimed to obtain necessary recommendations on improving the ITER CDA design.

Computer program optimizes design of nuclear radiation shields

NASA Technical Reports Server (NTRS)

Lahti, G. P.

1971-01-01

Computer program, OPEX 2, determines minimum weight, volume, or cost for shields. Program incorporates improved coding, simplified data input, spherical geometry, and an expanded output. Method is capable of altering dose-thickness relationship when a shield layer has been removed.

Evaluation of CFETR as a Fusion Nuclear Science Facility using multiple system codes

NASA Astrophysics Data System (ADS)

Chan, V. S.; Costley, A. E.; Wan, B. N.; Garofalo, A. M.; Leuer, J. A.

2015-02-01

This paper presents the results of a multi-system codes benchmarking study of the recently published China Fusion Engineering Test Reactor (CFETR) pre-conceptual design (Wan et al 2014 IEEE Trans. Plasma Sci. 42 495). Two system codes, General Atomics System Code (GASC) and Tokamak Energy System Code (TESC), using different methodologies to arrive at CFETR performance parameters under the same CFETR constraints show that the correlation between the physics performance and the fusion performance is consistent, and the computed parameters are in good agreement. Optimization of the first wall surface for tritium breeding and the minimization of the machine size are highly compatible. Variations of the plasma currents and profiles lead to changes in the required normalized physics performance, however, they do not significantly affect the optimized size of the machine. GASC and TESC have also been used to explore a lower aspect ratio, larger volume plasma taking advantage of the engineering flexibility in the CFETR design. Assuming the ITER steady-state scenario physics, the larger plasma together with a moderately higher BT and Ip can result in a high gain Qfus ˜ 12, Pfus ˜ 1 GW machine approaching DEMO-like performance. It is concluded that the CFETR baseline mode can meet the minimum goal of the Fusion Nuclear Science Facility (FNSF) mission and advanced physics will enable it to address comprehensively the outstanding critical technology gaps on the path to a demonstration reactor (DEMO). Before proceeding with CFETR construction steady-state operation has to be demonstrated, further development is needed to solve the divertor heat load issue, and blankets have to be designed with tritium breeding ratio (TBR) >1 as a target.

ITER's Tokamak Cooling Water System and the the Use of ASME Codes to Comply with French Regulations of Nuclear Pressure Equipment

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

Berry, Jan; Ferrada, Juan J; Curd, Warren

During inductive plasma operation of ITER, fusion power will reach 500 MW with an energy multiplication factor of 10. The heat will be transferred by the Tokamak Cooling Water System (TCWS) to the environment using the secondary cooling system. Plasma operations are inherently safe even under the most severe postulated accident condition a large, in-vessel break that results in a loss-of-coolant accident. A functioning cooling water system is not required to ensure safe shutdown. Even though ITER is inherently safe, TCWS equipment (e.g., heat exchangers, piping, pressurizers) are classified as safety important components. This is because the water is predictedmore » to contain low-levels of radionuclides (e.g., activated corrosion products, tritium) with activity levels high enough to require the design of components to be in accordance with French regulations for nuclear pressure equipment, i.e., the French Order dated 12 December 2005 (ESPN). ESPN has extended the practical application of the methodology established by the Pressure Equipment Directive (97/23/EC) to nuclear pressure equipment, under French Decree 99-1046 dated 13 December 1999, and Order dated 21 December 1999 (ESP). ASME codes and supplementary analyses (e.g., Failure Modes and Effects Analysis) will be used to demonstrate that the TCWS equipment meets these essential safety requirements. TCWS is being designed to provide not only cooling, with a capacity of approximately 1 GW energy removal, but also elevated temperature baking of first-wall/blanket, vacuum vessel, and divertor. Additional TCWS functions include chemical control of water, draining and drying for maintenance, and facilitation of leak detection/localization. The TCWS interfaces with the majority of ITER systems, including the secondary cooling system. U.S. ITER is responsible for design, engineering, and procurement of the TCWS with industry support from an Engineering Services Organization (ESO) (AREVA Federal Services, with support from Northrop Grumman, and OneCIS). ITER International Organization (ITER-IO) is responsible for design oversight and equipment installation in Cadarache, France. TCWS equipment will be fabricated using ASME design codes with quality assurance and oversight by an Agreed Notified Body (approved by the French regulator) that will ensure regulatory compliance. This paper describes the TCWS design and how U.S. ITER and fabricators will use ASME codes to comply with EU Directives and French Orders and Decrees.« less

Lecture Notes on Criticality Safety Validation Using MCNP & Whisper

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

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

Training classes for nuclear criticality safety, MCNP documentation. The need for, and problems surrounding, validation of computer codes and data area considered first. Then some background for MCNP & Whisper is given--best practices for Monte Carlo criticality calculations, neutron spectra, S(α,β) thermal neutron scattering data, nuclear data sensitivities, covariance data, and correlation coefficients. Whisper is computational software designed to assist the nuclear criticality safety analyst with validation studies with the Monte Carlo radiation transport package MCNP. Whisper's methodology (benchmark selection – C k's, weights; extreme value theory – bias, bias uncertainty; MOS for nuclear data uncertainty – GLLS) and usagemore » are discussed.« less

ESAS Deliverable PS 1.1.2.3: Customer Survey on Code Generations in Safety-Critical Applications

NASA Technical Reports Server (NTRS)

Schumann, Johann; Denney, Ewen

2006-01-01

Automated code generators (ACG) are tools that convert a (higher-level) model of a software (sub-)system into executable code without the necessity for a developer to actually implement the code. Although both commercially supported and in-house tools have been used in many industrial applications, little data exists on how these tools are used in safety-critical domains (e.g., spacecraft, aircraft, automotive, nuclear). The aims of the survey, therefore, were threefold: 1) to determine if code generation is primarily used as a tool for prototyping, including design exploration and simulation, or for fiight/production code; 2) to determine the verification issues with code generators relating, in particular, to qualification and certification in safety-critical domains; and 3) to determine perceived gaps in functionality of existing tools.

Encoded physics knowledge in checking codes for nuclear cross section libraries at Los Alamos

NASA Astrophysics Data System (ADS)

Parsons, D. Kent

2017-09-01

Checking procedures for processed nuclear data at Los Alamos are described. Both continuous energy and multi-group nuclear data are verified by locally developed checking codes which use basic physics knowledge and common-sense rules. A list of nuclear data problems which have been identified with help of these checking codes is also given.

Optimization and parallelization of the thermal–hydraulic subchannel code CTF for high-fidelity multi-physics applications

DOE PAGES

Salko, Robert K.; Schmidt, Rodney C.; Avramova, Maria N.

2014-11-23

This study describes major improvements to the computational infrastructure of the CTF subchannel code so that full-core, pincell-resolved (i.e., one computational subchannel per real bundle flow channel) simulations can now be performed in much shorter run-times, either in stand-alone mode or as part of coupled-code multi-physics calculations. These improvements support the goals of the Department Of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL) Energy Innovation Hub to develop high fidelity multi-physics simulation tools for nuclear energy design and analysis.

The help of simulation codes in designing waste assay systems using neutron measurement methods: Application to the alpha low level waste assay system PROMETHEE 6

NASA Astrophysics Data System (ADS)

Mariani, A.; Passard, C.; Jallu, F.; Toubon, H.

2003-11-01

The design of a specific nuclear assay system for a dedicated application begins with a phase of development, which relies on information from the literature or on knowledge resulting from experience, and on specific experimental verifications. The latter ones may require experimental devices which can be restricting in terms of deadline, cost and safety. One way generally chosen to bypass these difficulties is to use simulation codes to study particular aspects. This paper deals with the potentialities offered by the simulation in the case of a passive-active neutron (PAN) assay system for alpha low level waste characterization; this system has been carried out at the Nuclear Measurements Development Laboratory of the French Atomic Energy Commission. Due to the high number of parameters to be taken into account for its development, this is a particularly sophisticated example. Since the PAN assay system, called PROMETHEE (prompt epithermal and thermal interrogation experiment), must have a detection efficiency of more than 20% and preserve a high level of modularity for various applications, an improved version has been studied using the MCNP4 (Monte Carlo N-Particle) transport code. Parameters such as the dimensions of the assay system, of the cavity and of the detection blocks, and the thicknesses of the nuclear materials of neutronic interest have been optimised. Therefore, the number of necessary experiments was reduced.

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)

Interface design of VSOP'94 computer code for safety analysis

NASA Astrophysics Data System (ADS)

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

2014-09-01

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

NASA Radiation Protection Research for Exploration Missions

NASA Technical Reports Server (NTRS)

Wilson, John W.; Cucinotta, Francis A.; Tripathi, Ram K.; Heinbockel, John H.; Tweed, John; Mertens, Christopher J.; Walker, Steve A.; Blattnig, Steven R.; Zeitlin, Cary J.

2006-01-01

The HZETRN code was used in recent trade studies for renewed lunar exploration and currently used in engineering development of the next generation of space vehicles, habitats, and EVA equipment. A new version of the HZETRN code capable of simulating high charge and energy (HZE) ions, light-ions and neutrons with either laboratory or space boundary conditions with enhanced neutron and light-ion propagation is under development. Atomic and nuclear model requirements to support that development will be discussed. Such engineering design codes require establishing validation processes using laboratory ion beams and space flight measurements in realistic geometries. We discuss limitations of code validation due to the currently available data and recommend priorities for new data sets.

Multi-Constraint Multi-Variable Optimization of Source-Driven Nuclear Systems

NASA Astrophysics Data System (ADS)

Watkins, Edward Francis

1995-01-01

A novel approach to the search for optimal designs of source-driven nuclear systems is investigated. Such systems include radiation shields, fusion reactor blankets and various neutron spectrum-shaping assemblies. The novel approach involves the replacement of the steepest-descents optimization algorithm incorporated in the code SWAN by a significantly more general and efficient sequential quadratic programming optimization algorithm provided by the code NPSOL. The resulting SWAN/NPSOL code system can be applied to more general, multi-variable, multi-constraint shield optimization problems. The constraints it accounts for may include simple bounds on variables, linear constraints, and smooth nonlinear constraints. It may also be applied to unconstrained, bound-constrained and linearly constrained optimization. The shield optimization capabilities of the SWAN/NPSOL code system is tested and verified in a variety of optimization problems: dose minimization at constant cost, cost minimization at constant dose, and multiple-nonlinear constraint optimization. The replacement of the optimization part of SWAN with NPSOL is found feasible and leads to a very substantial improvement in the complexity of optimization problems which can be efficiently handled.

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

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

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

2007-07-01

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

The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS

DOE PAGES

Ward, Andrew; Downar, Thomas J.; Xu, Y.; ...

2015-04-22

The PATHS (PARCS Advanced Thermal Hydraulic Solver) code was developed at the University of Michigan in support of U.S. Nuclear Regulatory Commission research to solve the steady-state, two-phase, thermal-hydraulic equations for a boiling water reactor (BWR) and to provide thermal-hydraulic feedback for BWR depletion calculations with the neutronics code PARCS (Purdue Advanced Reactor Core Simulator). The simplified solution methodology, including a three-equation drift flux formulation and an optimized iteration scheme, yields very fast run times in comparison to conventional thermal-hydraulic systems codes used in the industry, while still retaining sufficient accuracy for applications such as BWR depletion calculations. Lastly, themore » capability to model advanced BWR fuel designs with part-length fuel rods and heterogeneous axial channel flow geometry has been implemented in PATHS, and the code has been validated against previously benchmarked advanced core simulators as well as BWR plant and experimental data. We describe the modifications to the codes and the results of the validation in this paper.« less

Identification and Analysis of Critical Gaps in Nuclear Fuel Cycle Codes Required by the SINEMA Program

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

Adrian Miron; Joshua Valentine; John Christenson

2009-10-01

The current state of the art in nuclear fuel cycle (NFC) modeling is an eclectic mixture of codes with various levels of applicability, flexibility, and availability. In support of the advanced fuel cycle systems analyses, especially those by the Advanced Fuel Cycle Initiative (AFCI), Unviery of Cincinnati in collaboration with Idaho State University carried out a detailed review of the existing codes describing various aspects of the nuclear fuel cycle and identified the research and development needs required for a comprehensive model of the global nuclear energy infrastructure and the associated nuclear fuel cycles. Relevant information obtained on the NFCmore » codes was compiled into a relational database that allows easy access to various codes' properties. Additionally, the research analyzed the gaps in the NFC computer codes with respect to their potential integration into programs that perform comprehensive NFC analysis.« less

The Italian experience on T/H best estimate codes: Achievements and perspectives

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

Alemberti, A.; D`Auria, F.; Fiorino, E.

1997-07-01

Themalhydraulic system codes are complex tools developed to simulate the power plants behavior during off-normal conditions. Among the objectives of the code calculations the evaluation of safety margins, the operator training, the optimization of the plant design and of the emergency operating procedures, are mostly considered in the field of the nuclear safety. The first generation of codes was developed in the United States at the end of `60s. Since that time, different research groups all over the world started the development of their own codes. At the beginning of the `80s, the second generation codes were proposed; these differmore » from the first generation codes owing to the number of balance equations solved (six instead of three), the sophistication of the constitutive models and of the adopted numerics. The capabilities of available computers have been fully exploited during the years. The authors then summarize some of the major steps in the process of developing, modifying, and advancing the capabilities of the codes. They touch on the fact that Italian, and for that matter non-American, researchers have not been intimately involved in much of this work. They then describe the application of these codes in Italy, even though there are no operating or under construction nuclear power plants at this time. Much of this effort is directed at the general question of plant safety in the face of transient type events.« less

Brayton Power Conversion System Parametric Design Modelling for Nuclear Electric Propulsion

NASA Technical Reports Server (NTRS)

Ashe, Thomas L.; Otting, William D.

1993-01-01

The parametrically based closed Brayton cycle (CBC) computer design model was developed for inclusion into the NASA LeRC overall Nuclear Electric Propulsion (NEP) end-to-end systems model. The code is intended to provide greater depth to the NEP system modeling which is required to more accurately predict the impact of specific technology on system performance. The CBC model is parametrically based to allow for conducting detailed optimization studies and to provide for easy integration into an overall optimizer driver routine. The power conversion model includes the modeling of the turbines, alternators, compressors, ducting, and heat exchangers (hot-side heat exchanger and recuperator). The code predicts performance to significant detail. The system characteristics determined include estimates of mass, efficiency, and the characteristic dimensions of the major power conversion system components. These characteristics are parametrically modeled as a function of input parameters such as the aerodynamic configuration (axial or radial), turbine inlet temperature, cycle temperature ratio, power level, lifetime, materials, and redundancy.

Overview of Particle and Heavy Ion Transport Code System PHITS

NASA Astrophysics Data System (ADS)

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

2014-06-01

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

Trade Studies for a Manned High-Power Nuclear Electric Propulsion Vehicle

NASA Technical Reports Server (NTRS)

SanSoucie, Michael; Hull, Patrick V.; Irwin, Ryan W.; TInker, Michael L.; Patton, Bruce W.

2005-01-01

Nuclear electric propulsion (NEP) vehicles will be needed for future manned missions to Mars and beyond. Candidate vehicles must be identified through trade studies for further detailed design from a large array of possibilities. Genetic algorithms have proven their utility in conceptual design studies by effectively searching a large design space to pinpoint unique optimal designs. This research combines analysis codes for NEP subsystems with genetic algorithm-based optimization. Trade studies for a NEP reference mission to the asteroids were conducted to identify important trends, and to determine the effects of various technologies and subsystems on vehicle performance. It was found that the electric thruster type and thruster performance have a major impact on the achievable system performance, and that significant effort in thruster research and development is merited.

EMPIRE: A code for nuclear astrophysics

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

Palumbo, A.

The nuclear reaction code EMPIRE is presented as a useful tool for nuclear astrophysics. EMPIRE combines a variety of the reaction models with a comprehensive library of input parameters providing a diversity of options for the user. With exclusion of the directsemidirect capture all reaction mechanisms relevant to the nuclear astrophysics energy range of interest are implemented in the code. Comparison to experimental data show consistent agreement for all relevant channels.

SFCOMPO 2.0 - A relational database of spent fuel isotopic measurements, reactor operational histories, and design data

NASA Astrophysics Data System (ADS)

Michel-Sendis, Franco; Martinez-González, Jesus; Gauld, Ian

2017-09-01

SFCOMPO-2.0 is a database of experimental isotopic concentrations measured in destructive radiochemical analysis of spent nuclear fuel (SNF) samples. The database includes corresponding design description of the fuel rods and assemblies, relevant operating conditions and characteristics of the host reactors necessary for modelling and simulation. Aimed at establishing a thorough, reliable, and publicly available resource for code and data validation of safety-related applications, SFCOMPO-2.0 is developed and maintained by the OECD Nuclear Energy Agency (NEA). The SFCOMPO-2.0 database is a Java application which is downloadable from the NEA website.

Proceedings of the 21st DOE/NRC Nuclear Air Cleaning Conference; Sessions 1--8

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

First, M.W.

1991-02-01

Separate abstracts have been prepared for the papers presented at the meeting on nuclear facility air cleaning technology in the following specific areas of interest: air cleaning technologies for the management and disposal of radioactive wastes; Canadian waste management program; radiological health effects models for nuclear power plant accident consequence analysis; filter testing; US standard codes on nuclear air and gas treatment; European community nuclear codes and standards; chemical processing off-gas cleaning; incineration and vitrification; adsorbents; nuclear codes and standards; mathematical modeling techniques; filter technology; safety; containment system venting; and nuclear air cleaning programs around the world. (MB)

The Advanced Software Development and Commercialization Project

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

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

1990-09-01

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

Current and anticipated uses of the thermal hydraulics codes at the NRC

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

Caruso, R.

1997-07-01

The focus of Thermal-Hydraulic computer code usage in nuclear regulatory organizations has undergone a considerable shift since the codes were originally conceived. Less work is being done in the area of {open_quotes}Design Basis Accidents,{close_quotes}, and much more emphasis is being placed on analysis of operational events, probabalistic risk/safety assessment, and maintenance practices. All of these areas need support from Thermal-Hydraulic computer codes to model the behavior of plant fluid systems, and they all need the ability to perform large numbers of analyses quickly. It is therefore important for the T/H codes of the future to be able to support thesemore » needs, by providing robust, easy-to-use, tools that produce easy-to understand results for a wider community of nuclear professionals. These tools need to take advantage of the great advances that have occurred recently in computer software, by providing users with graphical user interfaces for both input and output. In addition, reduced costs of computer memory and other hardware have removed the need for excessively complex data structures and numerical schemes, which make the codes more difficult and expensive to modify, maintain, and debug, and which increase problem run-times. Future versions of the T/H codes should also be structured in a modular fashion, to allow for the easy incorporation of new correlations, models, or features, and to simplify maintenance and testing. Finally, it is important that future T/H code developers work closely with the code user community, to ensure that the code meet the needs of those users.« less

Simulation Enabled Safeguards Assessment Methodology

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

Robert Bean; Trond Bjornard; Thomas Larson

2007-09-01

It is expected that nuclear energy will be a significant component of future supplies. New facilities, operating under a strengthened international nonproliferation regime will be needed. There is good reason to believe virtual engineering applied to the facility design, as well as to the safeguards system design will reduce total project cost and improve efficiency in the design cycle. Simulation Enabled Safeguards Assessment MEthodology (SESAME) has been developed as a software package to provide this capability for nuclear reprocessing facilities. The software architecture is specifically designed for distributed computing, collaborative design efforts, and modular construction to allow step improvements inmore » functionality. Drag and drop wireframe construction allows the user to select the desired components from a component warehouse, render the system for 3D visualization, and, linked to a set of physics libraries and/or computational codes, conduct process evaluations of the system they have designed.« less

Code qualification of structural materials for AFCI advanced recycling reactors.

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

Natesan, K.; Li, M.; Majumdar, S.

2012-05-31

This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Codemore » Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP) and the Power Reactor Innovative Small Module (PRISM), the NRC/Advisory Committee on Reactor Safeguards (ACRS) raised numerous safety-related issues regarding elevated-temperature structural integrity criteria. Most of these issues remained unresolved today. These critical licensing reviews provide a basis for the evaluation of underlying technical issues for future advanced sodium-cooled reactors. Major materials performance issues and high temperature design methodology issues pertinent to the ARR are addressed in the report. The report is organized as follows: the ARR reference design concepts proposed by the Argonne National Laboratory and four industrial consortia were reviewed first, followed by a summary of the major code qualification and licensing issues for the ARR structural materials. The available database is presented for the ASME Code-qualified structural alloys (e.g. 304, 316 stainless steels, 2.25Cr-1Mo, and mod.9Cr-1Mo), including physical properties, tensile properties, impact properties and fracture toughness, creep, fatigue, creep-fatigue interaction, microstructural stability during long-term thermal aging, material degradation in sodium environments and effects of neutron irradiation for both base metals and weld metals. An assessment of modified versions of Type 316 SS, i.e. Type 316LN and its Japanese version, 316FR, was conducted to provide a perspective for codification of 316LN or 316FR in Subsection NH. Current status and data availability of four new advanced alloys, i.e. NF616, NF616+TMT, NF709, and HT-UPS, are also addressed to identify the R&D needs for their code qualification for ARR applications. For both conventional and new alloys, issues related to high temperature design methodology are described to address the needs for improvements for the ARR design and licensing. Assessments have shown that there are significant data gaps for the full qualification and licensing of the ARR structural materials. Development and evaluation of structural materials require a variety of experimental facilities that have been seriously degraded in the past. The availability and additional needs for the key experimental facilities are summarized at the end of the report. Detailed information covered in each Chapter is given.« less

Status of VICTORIA: NRC peer review and recent code applications

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

Bixler, N.E.; Schaperow, J.H.

1997-12-01

VICTORIA is a mechanistic computer code designed to analyze fission product behavior within a nuclear reactor coolant system (RCS) during a severe accident. It provides detailed predictions of the release of radioactive and nonradioactive materials from the reactor core and transport and deposition of these materials within the RCS. A summary of the results and recommendations of an independent peer review of VICTORIA by the US Nuclear Regulatory Commission (NRC) is presented, along with recent applications of the code. The latter include analyses of a temperature-induced steam generator tube rupture sequence and post-test analyses of the Phebus FPT-1 test. Themore » next planned Phebus test, FTP-4, will focus on fission product releases from a rubble bed, especially those of the less-volatile elements, and on the speciation of the released elements. Pretest analyses using VICTORIA to estimate the magnitude and timing of releases are presented. The predicted release of uranium is a matter of particular importance because of concern about filter plugging during the test.« less

Creep and Creep-Fatigue Crack Growth at Structural Discontinuities and Welds

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

Dr. F. W. Brust; Dr. G. M. Wilkowski; Dr. P. Krishnaswamy

2010-01-27

The subsection ASME NH high temperature design procedure does not admit crack-like defects into the structural components. The US NRC identified the lack of treatment of crack growth within NH as a limitation of the code and thus this effort was undertaken. This effort is broken into two parts. Part 1, summarized here, involved examining all high temperature creep-fatigue crack growth codes being used today and from these, the task objective was to choose a methodology that is appropriate for possible implementation within NH. The second part of this task, which has just started, is to develop design rules formore » possible implementation within NH. This second part is a challenge since all codes require step-by-step analysis procedures to be undertaken in order to assess the crack growth and life of the component. Simple rules for design do not exist in any code at present. The codes examined in this effort included R5, RCC-MR (A16), BS 7910, API 579, and ATK (and some lesser known codes). There are several reasons that the capability for assessing cracks in high temperature nuclear components is desirable. These include: (1) Some components that are part of GEN IV reactors may have geometries that have sharp corners - which are essentially cracks. Design of these components within the traditional ASME NH procedure is quite challenging. It is natural to ensure adequate life design by modeling these features as cracks within a creep-fatigue crack growth procedure. (2) Workmanship flaws in welds sometimes occur and are accepted in some ASME code sections. It can be convenient to consider these as flaws when making a design life assessment. (3) Non-destructive Evaluation (NDE) and inspection methods after fabrication are limited in the size of the crack or flaw that can be detected. It is often convenient to perform a life assessment using a flaw of a size that represents the maximum size that can elude detection. (4) Flaws that are observed using in-service detection methods often need to be addressed as plants age. Shutdown inspection intervals can only be designed using creep and creep-fatigue crack growth techniques. (5) The use of crack growth procedures can aid in examining the seriousness of creep damage in structural components. How cracks grow can be used to assess margins on components and lead to further safe operation. After examining the pros and cons of all these methods, the R5 code was chosen as the most up-to-date and validated high temperature creep and creep fatigue code currently used in the world at present. R5 is considered the leader because the code: (1) has well established and validated rules, (2) has a team of experts continually improving and updating it, (3) has software that can be used by designers, (4) extensive validation in many parts with available data from BE resources as well as input from Imperial college's database, and (5) was specifically developed for use in nuclear plants. R5 was specifically developed for use in gas cooled nuclear reactors which operate in the UK and much of the experience is based on materials and temperatures which are experienced in these reactors. If the next generation advanced reactors to be built in the US used these same materials within the same temperature ranges as these reactors, then R5 may be appropriate for consideration of direct implementation within ASME code NH or Section XI. However, until more verification and validation of these creep/fatigue crack growth rules for the specific materials and temperatures to be used in the GEN IV reactors is complete, ASME should consider delaying this implementation. With this in mind, it is this authors opinion that R5 methods are the best available for code use today. The focus of this work was to examine the literature for creep and creep-fatigue crack growth procedures that are well established in codes in other countries and choose a procedure to consider implementation into ASME NH. It is very important to recognize that all creep and creep fatigue crack growth procedures that are part of high temperature design codes are related and very similar. This effort made no attempt to develop a new creep-fatigue crack growth predictive methodology. Rather examination of current procedures was the only goal. The uncertainties in the R5 crack growth methods and recommendations for more work are summarized here also.« less

Requirements for migration of NSSD code systems from LTSS to NLTSS

NASA Technical Reports Server (NTRS)

Pratt, M.

1984-01-01

The purpose of this document is to address the requirements necessary for a successful conversion of the Nuclear Design (ND) application code systems to the NLTSS environment. The ND application code system community can be characterized as large-scale scientific computation carried out on supercomputers. NLTSS is a distributed operating system being developed at LLNL to replace the LTSS system currently in use. The implications of change are examined including a description of the computational environment and users in ND. The discussion then turns to requirements, first in a general way, followed by specific requirements, including a proposal for managing the transition.

New developments and prospects on COSI, the simulation software for fuel cycle analysis

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

Eschbach, R.; Meyer, M.; Coquelet-Pascal, C.

2013-07-01

COSI, software developed by the Nuclear Energy Direction of the CEA, is a code simulating a pool of nuclear power plants with its associated fuel cycle facilities. This code has been designed to study various short, medium and long term options for the introduction of various types of nuclear reactors and for the use of associated nuclear materials. In the frame of the French Act for waste management, scenario studies are carried out with COSI, to compare different options of evolution of the French reactor fleet and options of partitioning and transmutation of plutonium and minor actinides. Those studies aimmore » in particular at evaluating the sustainability of Sodium cooled Fast Reactors (SFR) deployment and the possibility to transmute minor actinides. The COSI6 version is a completely renewed software released in 2006. COSI6 is now coupled with the last version of CESAR (CESAR5.3 based on JEFF3.1.1 nuclear data) allowing the calculations on irradiated fuel with 200 fission products and 100 heavy nuclides. A new release is planned in 2013, including in particular the coupling with a recommended database of reactors. An exercise of validation of COSI6, carried out on the French PWR historic nuclear fleet, has been performed. During this exercise quantities like cumulative natural uranium consumption, or cumulative depleted uranium, or UOX/MOX spent fuel storage, or stocks of reprocessed uranium, or plutonium content in fresh MOX fuel, or the annual production of high level waste, have been computed by COSI6 and compared to industrial data. The results have allowed us to validate the essential phases of the fuel cycle computation, and reinforces the credibility of the results provided by the code.« less

Nuclear Power Plant Cyber Security Discrete Dynamic Event Tree Analysis (LDRD 17-0958) FY17 Report

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

Wheeler, Timothy A.; Denman, Matthew R.; Williams, R. A.

Instrumentation and control of nuclear power is transforming from analog to modern digital assets. These control systems perform key safety and security functions. This transformation is occurring in new plant designs as well as in the existing fleet of plants as the operation of those plants is extended to 60 years. This transformation introduces new and unknown issues involving both digital asset induced safety issues and security issues. Traditional nuclear power risk assessment tools and cyber security assessment methods have not been modified or developed to address the unique nature of cyber failure modes and of cyber security threat vulnerabilities.more » iii This Lab-Directed Research and Development project has developed a dynamic cyber-risk in- formed tool to facilitate the analysis of unique cyber failure modes and the time sequencing of cyber faults, both malicious and non-malicious, and impose those cyber exploits and cyber faults onto a nuclear power plant accident sequence simulator code to assess how cyber exploits and cyber faults could interact with a plants digital instrumentation and control (DI&C) system and defeat or circumvent a plants cyber security controls. This was achieved by coupling an existing Sandia National Laboratories nuclear accident dynamic simulator code with a cyber emulytics code to demonstrate real-time simulation of cyber exploits and their impact on automatic DI&C responses. Studying such potential time-sequenced cyber-attacks and their risks (i.e., the associated impact and the associated degree of difficulty to achieve the attack vector) on accident management establishes a technical risk informed framework for developing effective cyber security controls for nuclear power.« less

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

Merzari, E.; Yuan, Haomin; Kraus, A.

The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to-plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable componentsmore » in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. In addition to being more cost- and space-efficient, such steam generators need to be more reliable, since failure of the primary vessel represents a potential loss of coolant and a safety concern. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict their occurrence based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental programs, especially at the design stage. Over the past five years, the Reactor Product Line has developed the integrated multi-physics code suite SHARP. The goal of developing such a tool is to perform multi-physics neutronics, thermal/fluid, and structural mechanics modeling of the components inside the full reactor core or portions of it with a user-specified fidelity. In particular SHARP contains high-fidelity single-physics codes Diablo for structural mechanics and Nek5000 for fluid mechanics calculations. Both codes are state-of-the-art, highly scalable tools that have been extensively validated. These tools form a strong basis on which to build a flow-induced vibration modeling capability. In this report we discuss one-way coupled calculations performed with Nek5000 and Diablo aimed at simulating available FIV experiments in helical steam generators in the turbulent buffeting regime. In this regime one-way coupling is judged sufficient because the pressure loads do not cause substantial displacements. It is also the most common source of vibration in helical steam generators at the low flows expected in integral PWRs. The legacy data is obtained from two datasets developed at Argonne and B&W.« less

SPOC Benchmark Case: SNRE Model

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

Vishal Patel; Michael Eades; Claude Russel Joyner II

The Small Nuclear Rocket Engine (SNRE) was modeled in the Center for Space Nuclear Research’s (CSNR) Space Propulsion Optimization Code (SPOC). SPOC aims to create nuclear thermal propulsion (NTP) geometries quickly to perform parametric studies on design spaces of historic and new NTP designs. The SNRE geometry was modeled in SPOC and a critical core with a reasonable amount of criticality margin was found. The fuel, tie-tubes, reflector, and control drum masses were predicted rather well. These are all very important for neutronics calculations so the active reactor geometries created with SPOC can continue to be trusted. Thermal calculations ofmore » the average and hot fuel channels agreed very well. The specific impulse calculations used historically and in SPOC disagree so mass flow rates and impulses differed. Modeling peripheral and power balance components that do not affect nuclear characteristics of the core is not a feature of SPOC and as such, these components should continue to be designed using other tools. A full paper detailing the available SNRE data and comparisons with SPOC outputs will be submitted as a follow-up to this abstract.« less

Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

NASA Technical Reports Server (NTRS)

Emrich, William

2013-01-01

A key technology element in Nuclear Thermal Propulsion is the development of fuel materials and components which can withstand extremely high temperatures while being exposed to flowing hydrogen. NTREES provides a cost effective method for rapidly screening of candidate fuel components with regard to their viability for use in NTR systems. The NTREES is designed to mimic the conditions (minus the radiation) to which nuclear rocket fuel elements and other components would be subjected to during reactor operation. The NTREES consists of a water cooled ASME code stamped pressure vessel and its associated control hardware and instrumentation coupled with inductive heaters to simulate the heat provided by the fission process. The NTREES has been designed to safely allow hydrogen gas to be injected into internal flow passages of an inductively heated test article mounted in the chamber.

JASMIN: Japanese-American study of muon interactions and neutron detection

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

Nakashima, Hiroshi; /JAEA, Ibaraki; Mokhov, N.V.

Experimental studies of shielding and radiation effects at Fermi National Accelerator Laboratory (FNAL) have been carried out under collaboration between FNAL and Japan, aiming at benchmarking of simulation codes and study of irradiation effects for upgrade and design of new high-energy accelerator facilities. The purposes of this collaboration are (1) acquisition of shielding data in a proton beam energy domain above 100GeV; (2) further evaluation of predictive accuracy of the PHITS and MARS codes; (3) modification of physics models and data in these codes if needed; (4) establishment of irradiation field for radiation effect tests; and (5) development of amore » code module for improved description of radiation effects. A series of experiments has been performed at the Pbar target station and NuMI facility, using irradiation of targets with 120 GeV protons for antiproton and neutrino production, as well as the M-test beam line (M-test) for measuring nuclear data and detector responses. Various nuclear and shielding data have been measured by activation methods with chemical separation techniques as well as by other detectors such as a Bonner ball counter. Analyses with the experimental data are in progress for benchmarking the PHITS and MARS15 codes. In this presentation recent activities and results are reviewed.« less

Modeling the UO2 ex-AUC pellet process and predicting the fuel rod temperature distribution under steady-state operating condition

NASA Astrophysics Data System (ADS)

Hung, Nguyen Trong; Thuan, Le Ba; Thanh, Tran Chi; Nhuan, Hoang; Khoai, Do Van; Tung, Nguyen Van; Lee, Jin-Young; Jyothi, Rajesh Kumar

2018-06-01

Modeling uranium dioxide pellet process from ammonium uranyl carbonate - derived uranium dioxide powder (UO2 ex-AUC powder) and predicting fuel rod temperature distribution were reported in the paper. Response surface methodology (RSM) and FRAPCON-4.0 code were used to model the process and to predict the fuel rod temperature under steady-state operating condition. Fuel rod design of AP-1000 designed by Westinghouse Electric Corporation, in these the pellet fabrication parameters are from the study, were input data for the code. The predictive data were suggested the relationship between the fabrication parameters of UO2 pellets and their temperature image in nuclear reactor.

RESRAD model presentation

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

Yu, C.; Faillace, E.; Chen, S.Y.

RESRAD was one of the multimedia models selected by the US Nuclear Regulatory Commission (NRC) to include in its workshop on radiation dose modeling and demonstration of compliance with the radiological criteria for license termination. This paper is a summary of the presentation made at the workshop and focuses on the 10 questions the NRC distributed to all participants prior to the workshop. The code selection criteria, which were solicited by the NRC, for demonstrating compliance with the license termination rule are also included. Among the RESRAD family of codes, RESRAD and RESRAD-BUILD are designed for evaluating radiological contamination inmore » soils and in buildings. Many documents have been published to support the use of these codes. This paper focuses on these two codes. The pathways considered, the databases and parameters used, quality control and quality assurance, benchmarking, verification and validation of these codes, and capabilities as well as limitations of these codes are discussed in detail.« less

FY17 Status Report on NEAMS Neutronics Activities

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

Lee, C. H.; Jung, Y. S.; Smith, M. A.

2017-09-30

Under the U.S. DOE NEAMS program, the high-fidelity neutronics code system has been developed to support the multiphysics modeling and simulation capability named SHARP. The neutronics code system includes the high-fidelity neutronics code PROTEUS, the cross section library and preprocessing tools, the multigroup cross section generation code MC2-3, the in-house meshing generation tool, the perturbation and sensitivity analysis code PERSENT, and post-processing tools. The main objectives of the NEAMS neutronics activities in FY17 are to continue development of an advanced nodal solver in PROTEUS for use in nuclear reactor design and analysis projects, implement a simplified sub-channel based thermal-hydraulic (T/H)more » capability into PROTEUS to efficiently compute the thermal feedback, improve the performance of PROTEUS-MOCEX using numerical acceleration and code optimization, improve the cross section generation tools including MC2-3, and continue to perform verification and validation tests for PROTEUS.« less

Creep of A508/533 Pressure Vessel Steel

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

Richard Wright

2014-08-01

ABSTRACT Evaluation of potential Reactor Pressure Vessel (RPV) steels has been carried out as part of the pre-conceptual Very High Temperature Reactor (VHTR) design studies. These design studies have generally focused on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Initially, three candidate materials were identified by this process: conventional light water reactor (LWR) RPV steels A508 and A533, 2¼Cr-1Mo in the annealed condition, and Grade 91 steel. The low strength of 2¼Cr-1Mo at elevated temperature has eliminated this steel from serious consideration as the VHTR RPV candidate material. Discussions with themore » very few vendors that can potentially produce large forgings for nuclear pressure vessels indicate a strong preference for conventional LWR steels. This preference is based in part on extensive experience with forging these steels for nuclear components. It is also based on the inability to cast large ingots of the Grade 91 steel due to segregation during ingot solidification, thus restricting the possible mass of forging components and increasing the amount of welding required for completion of the RPV. Grade 91 steel is also prone to weld cracking and must be post-weld heat treated to ensure adequate high-temperature strength. There are also questions about the ability to produce, and very importantly, verify the through thickness properties of thick sections of Grade 91 material. The availability of large components, ease of fabrication, and nuclear service experience with the A508 and A533 steels strongly favor their use in the RPV for the VHTR. Lowering the gas outlet temperature for the VHTR to 750°C from 950 to 1000°C, proposed in early concept studies, further strengthens the justification for this material selection. This steel is allowed in the ASME Boiler and Pressure Vessel Code for nuclear service up to 371°C (700°F); certain excursions above that temperature are allowed by Code Case N-499-2 (now incorporated as an appendix to Section III Division 5 of the Code). This Code Case was developed with a rather sparse data set and focused primarily on rolled plate material (A533 specification). Confirmatory tests of creep behavior of both A508 and A533 are described here that are designed to extend the database in order to build higher confidence in ensuring the structural integrity of the VHTR RPV during off-normal conditions. A number of creep-rupture tests were carried out at temperatures above the 371°C (700°F) Code limit; longer term tests designed to evaluate minimum creep behavior are ongoing. A limited amount of rupture testing was also carried out on welded material. All of the rupture data from the current experiments is compared to historical values from the testing carried out to develop Code Case N-499-2. It is shown that the A508/533 basemetal tested here fits well with the rupture behavior reported from the historical testing. The presence of weldments significantly reduces the time to rupture. The primary purpose of this report is to summarize and record the experimental results in a single document.« less

Development of Yield and Tensile Strength Design Curves for Alloy 617

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

Nancy Lybeck; T. -L. Sham

2013-10-01

The U.S. Department of Energy Very High Temperature Reactor Program is acquiring data in preparation for developing an Alloy 617 Code Case for inclusion in the nuclear section of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code. A draft code case was previously developed, but effort was suspended before acceptance by ASME. As part of the draft code case effort, a database was compiled of yield and tensile strength data from tests performed in air. Yield strength and tensile strength at temperature are used to set time independent allowable stress for construction materials in B&PVmore » Code, Section III, Subsection NH. The yield and tensile strength data used for the draft code case has been augmented with additional data generated by Idaho National Laboratory and Oak Ridge National Laboratory in the U.S. and CEA in France. The standard ASME Section II procedure for generating yield and tensile strength at temperature is presented, along with alternate methods that accommodate the change in temperature trends seen at high temperatures, resulting in a more consistent design margin over the temperature range of interest.« less

RELAP-7 Closure Correlations

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

Zou, Ling; Berry, R. A.; Martineau, R. C.

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The code is based on the INL’s modern scientific software development framework, MOOSE (Multi-Physics Object Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5’s and TRACE’s capabilities and extends their analysis capabilities for all reactor system simulation scenarios. The RELAP-7 codemore » utilizes the well-posed 7-equation two-phase flow model for compressible two-phase flow. Closure models used in the TRACE code has been reviewed and selected to reflect the progress made during the past decades and provide a basis for the colure correlations implemented in the RELAP-7 code. This document provides a summary on the closure correlations that are currently implemented in the RELAP-7 code. The closure correlations include sub-grid models that describe interactions between the fluids and the flow channel, and interactions between the two phases.« less

Numerical simulation of tornado wind loading on structures

NASA Technical Reports Server (NTRS)

Maiden, D. E.

1976-01-01

A numerical simulation of a tornado interacting with a building was undertaken in order to compare the pressures due to a rotational unsteady wind with that due to steady straight winds used in design of nuclear facilities. The numerical simulations were performed on a two-dimensional compressible hydrodynamics code. Calculated pressure profiles for a typical building were then subjected to a tornado wind field and the results were compared with current quasisteady design calculations. The analysis indicates that current design practices are conservative.

Computation of a Canadian SCWR unit cell with deterministic and Monte Carlo codes

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

Harrisson, G.; Marleau, G.

2012-07-01

The Canadian SCWR has the potential to achieve the goals that the generation IV nuclear reactors must meet. As part of the optimization process for this design concept, lattice cell calculations are routinely performed using deterministic codes. In this study, the first step (self-shielding treatment) of the computation scheme developed with the deterministic code DRAGON for the Canadian SCWR has been validated. Some options available in the module responsible for the resonance self-shielding calculation in DRAGON 3.06 and different microscopic cross section libraries based on the ENDF/B-VII.0 evaluated nuclear data file have been tested and compared to a reference calculationmore » performed with the Monte Carlo code SERPENT under the same conditions. Compared to SERPENT, DRAGON underestimates the infinite multiplication factor in all cases. In general, the original Stammler model with the Livolant-Jeanpierre approximations are the most appropriate self-shielding options to use in this case of study. In addition, the 89 groups WIMS-AECL library for slight enriched uranium and the 172 groups WLUP library for a mixture of plutonium and thorium give the most consistent results with those of SERPENT. (authors)« less

Fundamental approaches for analysis thermal hydraulic parameter for Puspati Research Reactor

NASA Astrophysics Data System (ADS)

Hashim, Zaredah; Lanyau, Tonny Anak; Farid, Mohamad Fairus Abdul; Kassim, Mohammad Suhaimi; Azhar, Noraishah Syahirah

2016-01-01

The 1-MW PUSPATI Research Reactor (RTP) is the one and only nuclear pool type research reactor developed by General Atomic (GA) in Malaysia. It was installed at Malaysian Nuclear Agency and has reached the first criticality on 8 June 1982. Based on the initial core which comprised of 80 standard TRIGA fuel elements, the very fundamental thermal hydraulic model was investigated during steady state operation using the PARET-code. The main objective of this paper is to determine the variation of temperature profiles and Departure of Nucleate Boiling Ratio (DNBR) of RTP at full power operation. The second objective is to confirm that the values obtained from PARET-code are in agreement with Safety Analysis Report (SAR) for RTP. The code was employed for the hot and average channels in the core in order to calculate of fuel's center and surface, cladding, coolant temperatures as well as DNBR's values. In this study, it was found that the results obtained from the PARET-code showed that the thermal hydraulic parameters related to safety for initial core which was cooled by natural convection was in agreement with the designed values and safety limit in SAR.

Leadership Class Configuration Interaction Code - Status and Opportunities

NASA Astrophysics Data System (ADS)

Vary, James

2011-10-01

With support from SciDAC-UNEDF (www.unedf.org) nuclear theorists have developed and are continuously improving a Leadership Class Configuration Interaction Code (LCCI) for forefront nuclear structure calculations. The aim of this project is to make state-of-the-art nuclear structure tools available to the entire community of researchers including graduate students. The project includes codes such as NuShellX, MFDn and BIGSTICK that run a range of computers from laptops to leadership class supercomputers. Codes, scripts, test cases and documentation have been assembled, are under continuous development and are scheduled for release to the entire research community in November 2011. A covering script that accesses the appropriate code and supporting files is under development. In addition, a Data Base Management System (DBMS) that records key information from large production runs and archived results of those runs has been developed (http://nuclear.physics.iastate.edu/info/) and will be released. Following an outline of the project, the code structure, capabilities, the DBMS and current efforts, I will suggest a path forward that would benefit greatly from a significant partnership between researchers who use the codes, code developers and the National Nuclear Data efforts. This research is supported in part by DOE under grant DE-FG02-87ER40371 and grant DE-FC02-09ER41582 (SciDAC-UNEDF).

Generalized Nuclear Data: A New Structure (with Supporting Infrastructure) for Handling Nuclear Data

NASA Astrophysics Data System (ADS)

Mattoon, C. M.; Beck, B. R.; Patel, N. R.; Summers, N. C.; Hedstrom, G. W.; Brown, D. A.

2012-12-01

The Evaluated Nuclear Data File (ENDF) format was designed in the 1960s to accommodate neutron reaction data to support nuclear engineering applications in power, national security and criticality safety. Over the years, the scope of the format has been extended to handle many other kinds of data including charged particle, decay, atomic, photo-nuclear and thermal neutron scattering. Although ENDF has wide acceptance and support for many data types, its limited support for correlated particle emission, limited numeric precision, and general lack of extensibility mean that the nuclear data community cannot take advantage of many emerging opportunities. More generally, the ENDF format provides an unfriendly environment that makes it difficult for new data evaluators and users to create and access nuclear data. The Cross Section Evaluation Working Group (CSEWG) has begun the design of a new Generalized Nuclear Data (or 'GND') structure, meant to replace older formats with a hierarchy that mirrors the underlying physics, and is aligned with modern coding and database practices. In support of this new structure, Lawrence Livermore National Laboratory (LLNL) has updated its nuclear data/reactions management package Fudge to handle GND structured nuclear data. Fudge provides tools for converting both the latest ENDF format (ENDF-6) and the LLNL Evaluated Nuclear Data Library (ENDL) format to and from GND, as well as for visualizing, modifying and processing (i.e., converting evaluated nuclear data into a form more suitable to transport codes) GND structured nuclear data. GND defines the structure needed for storing nuclear data evaluations and the type of data that needs to be stored. But unlike ENDF and ENDL, GND does not define how the data are to be stored in a file. Currently, Fudge writes the structured GND data to a file using the eXtensible Markup Language (XML), as it is ASCII based and can be viewed with any text editor. XML is a meta-language, meaning that it has a primitive set of definitions for representing hierarchical data/text in a file. Other meta-languages, like HDF5 which stores the data in binary form, can also be used to store GND in a file. In this paper, we will present an overview of the new GND data structures along with associated tools in Fudge.

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

Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

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

McDeavitt, Sean; Shao, Lin; Tsvetkov, Pavel

2014-04-07

Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development suchmore » that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.« less

The RNA Exosome Adaptor ZFC3H1 Functionally Competes with Nuclear Export Activity to Retain Target Transcripts.

PubMed

Silla, Toomas; Karadoulama, Evdoxia; Mąkosa, Dawid; Lubas, Michal; Jensen, Torben Heick

2018-05-15

Mammalian genomes are promiscuously transcribed, yielding protein-coding and non-coding products. Many transcripts are short lived due to their nuclear degradation by the ribonucleolytic RNA exosome. Here, we show that abolished nuclear exosome function causes the formation of distinct nuclear foci, containing polyadenylated (pA + ) RNA secluded from nucleocytoplasmic export. We asked whether exosome co-factors could serve such nuclear retention. Co-localization studies revealed the enrichment of pA + RNA foci with "pA-tail exosome targeting (PAXT) connection" components MTR4, ZFC3H1, and PABPN1 but no overlap with known nuclear structures such as Cajal bodies, speckles, paraspeckles, or nucleoli. Interestingly, ZFC3H1 is required for foci formation, and in its absence, selected pA + RNAs, including coding and non-coding transcripts, are exported to the cytoplasm in a process dependent on the mRNA export factor AlyREF. Our results establish ZFC3H1 as a central nuclear pA + RNA retention factor, counteracting nuclear export activity. Copyright © 2018 The Author(s). Published by Elsevier Inc. All rights reserved.

Reduced Order Modeling Methods for Turbomachinery Design

DTIC Science & Technology

2009-03-01

and Ma- terials Conference, May 2006. [45] A. Gelman , J. B. Carlin, H. S. Stern, and D. B. Rubin, Bayesian Data Analysis. New York, NY: Chapman I& Hall...Macian- Juan , and R. Chawla, “A statistical methodology for quantif ca- tion of uncertainty in best estimate code physical models,” Annals of Nuclear En

78 FR 48727 - Proposed Revisions to Design of Structures, Components, Equipment and Systems

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-09

... Analysis Reports for Nuclear Power Plants: LWR Edition,'' Section 3.9.3 ``ASME Code Class 1, 2, and 3...'s Agencywide Documents Access and Management System (ADAMS): You may access publicly available... operational readiness of snubbers (ADAMS Accession No. ML070720041), and review interfaces have been updated...

Methodology for the nuclear design validation of an Alternate Emergency Management Centre (CAGE)

NASA Astrophysics Data System (ADS)

Hueso, César; Fabbri, Marco; de la Fuente, Cristina; Janés, Albert; Massuet, Joan; Zamora, Imanol; Gasca, Cristina; Hernández, Héctor; Vega, J. Ángel

2017-09-01

The methodology is devised by coupling different codes. The study of weather conditions as part of the data of the site will determine the relative concentrations of radionuclides in the air using ARCON96. The activity in the air is characterized depending on the source and release sequence specified in NUREG-1465 by RADTRAD code, which provides results of the inner cloud source term contribution. Known activities, energy spectra are inferred using ORIGEN-S, which are used as input for the models of the outer cloud, filters and containment generated with MCNP5. The sum of the different contributions must meet the conditions of habitability specified by the CSN (Spanish Nuclear Regulatory Body) (TEDE <50 mSv and equivalent dose to the thyroid <500 mSv within 30 days following the accident doses) so that the dose is optimized by varying parameters such as CAGE location, flow filtering need for recirculation, thicknesses and compositions of the walls, etc. The results for the most penalizing area meet the established criteria, and therefore the CAGE building design based on the methodology presented is radiologically validated.

Establishment and assessment of code scaling capability

NASA Astrophysics Data System (ADS)

Lim, Jaehyok

In this thesis, a method for using RELAP5/MOD3.3 (Patch03) code models is described to establish and assess the code scaling capability and to corroborate the scaling methodology that has been used in the design of the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR applications (PUMA-E) facility. It was sponsored by the United States Nuclear Regulatory Commission (USNRC) under the program "PUMA ESBWR Tests". PUMA-E facility was built for the USNRC to obtain data on the performance of the passive safety systems of the General Electric (GE) Nuclear Energy Economic Simplified Boiling Water Reactor (ESBWR). Similarities between the prototype plant and the scaled-down test facility were investigated for a Gravity-Driven Cooling System (GDCS) Drain Line Break (GDLB). This thesis presents the results of the GDLB test, i.e., the GDLB test with one Isolation Condenser System (ICS) unit disabled. The test is a hypothetical multi-failure small break loss of coolant (SB LOCA) accident scenario in the ESBWR. The test results indicated that the blow-down phase, Automatic Depressurization System (ADS) actuation, and GDCS injection processes occurred as expected. The GDCS as an emergency core cooling system provided adequate supply of water to keep the Reactor Pressure Vessel (RPV) coolant level well above the Top of Active Fuel (TAF) during the entire GDLB transient. The long-term cooling phase, which is governed by the Passive Containment Cooling System (PCCS) condensation, kept the reactor containment system that is composed of Drywell (DW) and Wetwell (WW) below the design pressure of 414 kPa (60 psia). In addition, the ICS continued participating in heat removal during the long-term cooling phase. A general Code Scaling, Applicability, and Uncertainty (CSAU) evaluation approach was discussed in detail relative to safety analyses of Light Water Reactor (LWR). The major components of the CSAU methodology that were highlighted particularly focused on the scaling issues of experiments and models and their applicability to the nuclear power plant transient and accidents. The major thermal-hydraulic phenomena to be analyzed were identified and the predictive models adopted in RELAP5/MOD3.3 (Patch03) code were briefly reviewed.

Use of the Fracture Continuum Model for Numerical Modeling of Flow and Transport of Deep Geologic Disposal of Nuclear Waste in Crystalline Rock

NASA Astrophysics Data System (ADS)

Hadgu, T.; Kalinina, E.; Klise, K. A.; Wang, Y.

2015-12-01

Numerical modeling of disposal of nuclear waste in a deep geologic repository in fractured crystalline rock requires robust characterization of fractures. Various methods for fracture representation in granitic rocks exist. In this study we used the fracture continuum model (FCM) to characterize fractured rock for use in the simulation of flow and transport in the far field of a generic nuclear waste repository located at 500 m depth. The FCM approach is a stochastic method that maps the permeability of discrete fractures onto a regular grid. The method generates permeability fields using field observations of fracture sets. The original method described in McKenna and Reeves (2005) was designed for vertical fractures. The method has since then been extended to incorporate fully three-dimensional representations of anisotropic permeability, multiple independent fracture sets, and arbitrary fracture dips and orientations, and spatial correlation (Kalinina et al. 20012, 2014). For this study the numerical code PFLOTRAN (Lichtner et al., 2015) has been used to model flow and transport. PFLOTRAN solves a system of generally nonlinear partial differential equations describing multiphase, multicomponent and multiscale reactive flow and transport in porous materials. The code is designed to run on massively parallel computing architectures as well as workstations and laptops (e.g. Hammond et al., 2011). Benchmark tests were conducted to simulate flow and transport in a specified model domain. Distributions of fracture parameters were used to generate a selected number of realizations. For each realization, the FCM method was used to generate a permeability field of the fractured rock. The PFLOTRAN code was then used to simulate flow and transport in the domain. Simulation results and analysis are presented. The results indicate that the FCM approach is a viable method to model fractured crystalline rocks. The FCM is a computationally efficient way to generate realistic representation of complex fracture systems. This approach is of interest for nuclear waste disposal models applied over large domains.

Review of current nuclear fallout codes.

PubMed

Auxier, Jerrad P; Auxier, John D; Hall, Howard L

2017-05-01

The importance of developing a robust nuclear forensics program to combat the illicit use of nuclear material that may be used as an improvised nuclear device is widely accepted. In order to decrease the threat to public safety and improve governmental response, government agencies have developed fallout-analysis codes to predict the fallout particle size, dose, and dispersion and dispersion following a detonation. This paper will review the different codes that have been developed for predicting fallout from both chemical and nuclear weapons. This will decrease the response time required for the government to respond to the event. Copyright © 2017 The Authors. Published by Elsevier Ltd.. All rights reserved.

77 FR 3073 - American Society of Mechanical Engineers (ASME) Codes and New and Revised ASME Code Cases...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-23

... NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 [NRC-2008-0554] RIN 3150-AI35 American Society of Mechanical Engineers (ASME) Codes and New and Revised ASME Code Cases; Corrections AGENCY: Nuclear Regulatory... the American Society of Mechanical Engineers, Three Park Avenue, New York, NY 10016, phone (800) 843...

Comparison of deterministic and stochastic approaches for isotopic concentration and decay heat uncertainty quantification on elementary fission pulse

NASA Astrophysics Data System (ADS)

Lahaye, S.; Huynh, T. D.; Tsilanizara, A.

2016-03-01

Uncertainty quantification of interest outputs in nuclear fuel cycle is an important issue for nuclear safety, from nuclear facilities to long term deposits. Most of those outputs are functions of the isotopic vector density which is estimated by fuel cycle codes, such as DARWIN/PEPIN2, MENDEL, ORIGEN or FISPACT. CEA code systems DARWIN/PEPIN2 and MENDEL propagate by two different methods the uncertainty from nuclear data inputs to isotopic concentrations and decay heat. This paper shows comparisons between those two codes on a Uranium-235 thermal fission pulse. Effects of nuclear data evaluation's choice (ENDF/B-VII.1, JEFF-3.1.1 and JENDL-2011) is inspected in this paper. All results show good agreement between both codes and methods, ensuring the reliability of both approaches for a given evaluation.

CESAR: A Code for Nuclear Fuel and Waste Characterisation

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

Vidal, J.M.; Grouiller, J.P.; Launay, A.

2006-07-01

CESAR (Simplified Evolution Code Applied to Reprocessing) is a depletion code developed through a joint program between CEA and COGEMA. In the late 1980's, the first use of this code dealt with nuclear measurement at the Laboratories of the La Hague reprocessing plant. The use of CESAR was then extended to characterizations of all entrance materials and for characterisation, via tracer, of all produced waste. The code can distinguish more than 100 heavy nuclides, 200 fission products and 100 activation products, and it can characterise both the fuel and the structural material of the fuel. CESAR can also make depletionmore » calculations from 3 months to 1 million years of cooling time. Between 2003-2005, the 5. version of the code was developed. The modifications were related to the harmonisation of the code's nuclear data with the JEF2.2 nuclear data file. This paper describes the code and explains the extensive use of this code at the La Hague reprocessing plant and also for prospective studies. The second part focuses on the modifications of the latest version, and describes the application field and the qualification of the code. Many companies and the IAEA use CESAR today. CESAR offers a Graphical User Interface, which is very user-friendly. (authors)« less

Nuclear modules for space electric propulsion

NASA Technical Reports Server (NTRS)

Difilippo, F. C.

1998-01-01

Analysis of interplanetary cargo and piloted missions requires calculations of the performances and masses of subsystems to be integrated in a final design. In a preliminary and scoping stage the designer needs to evaluate options iteratively by using fast computer simulations. The Oak Ridge National Laboratory (ORNL) has been involved in the development of models and calculational procedures for the analysis (neutronic and thermal hydraulic) of power sources for nuclear electric propulsion. The nuclear modules will be integrated into the whole simulation of the nuclear electric propulsion system. The vehicles use either a Brayton direct-conversion cycle, using the heated helium from a NERVA-type reactor, or a potassium Rankine cycle, with the working fluid heated on the secondary side of a heat exchanger and lithium on the primary side coming from a fast reactor. Given a set of input conditions, the codes calculate composition. dimensions, volumes, and masses of the core, reflector, control system, pressure vessel, neutron and gamma shields, as well as the thermal hydraulic conditions of the coolant, clad and fuel. Input conditions are power, core life, pressure and temperature of the coolant at the inlet of the core, either the temperature of the coolant at the outlet of the core or the coolant mass flow and the fluences and integrated doses at the cargo area. Using state-of-the-art neutron cross sections and transport codes, a database was created for the neutronic performance of both reactor designs. The free parameters of the models are the moderator/fuel mass ratio for the NERVA reactor and the enrichment and the pitch of the lattice for the fast reactor. Reactivity and energy balance equations are simultaneously solved to find the reactor design. Thermalhydraulic conditions are calculated by solving the one-dimensional versions of the equations of conservation of mass, energy, and momentum with compressible flow.

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

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

LaSalle, F.R.; Golbeg, P.R.; Chenault, D.M.

For reactor and nuclear facilities, both Title 10, Code of Federal Regulations, Part 50, and US Department of Energy Order 6430.1A require assessments of the interaction of non-Safety Class 1 piping and equipment with Safety Class 1 piping and equipment during a seismic event to maintain the safety function. The safety class systems of nuclear reactors or nuclear facilities are designed to the applicable American Society of Mechanical Engineers standards and Seismic Category 1 criteria that require rigorous analysis, construction, and quality assurance. Because non-safety class systems are generally designed to lesser standards and seismic criteria, they may become missilesmore » during a safe shutdown earthquake. The resistance of piping, tubing, and equipment to seismically generated missiles is addressed in the paper. Gross plastic and local penetration failures are considered with applicable test verification. Missile types and seismic zones of influence are discussed. Field qualification data are also developed for missile evaluation.« less

Warthog: Progress on Coupling BISON and PROTEUS

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

Hart, Shane W.D.

The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program from the Office of Nuclear Energy at the Department of Energy (DOE) provides a robust toolkit for modeling and simulation of current and future advanced nuclear reactor designs. This toolkit provides these technologies organized across product lines, with two divisions targeted at fuels and end-to-end reactor modeling, and a third for integration, coupling, and high-level workflow management. The Fuels Product Line (FPL) and the Reactor Product Line (RPL) provide advanced computational technologies that serve each respective field effectively. There is currently a lack of integration between the product lines, impeding futuremore » improvements of simulation solution fidelity. In order to mix and match tools across the product lines, a new application called Warthog was produced. Warthog is built on the Multi-physics Object-Oriented Simulation Environment (MOOSE) framework developed at Idaho National Laboratory (INL). This report details the continuing efforts to provide the Integration Product Line (IPL) with interoperability using the Warthog code. Currently, this application strives to couple the BISON fuel performance application from the FPL using the PROTEUS Core Neutronics application from the RPL. Warthog leverages as much prior work from the NEAMS program as possible, enabling interoperability between the independently developed MOOSE and SHARP frameworks, and the libMesh and MOAB mesh data formats. Previous work performed on Warthog allowed it to couple a pin cell between the two codes. However, as the temperature changed due to the BISON calculation, the cross sections were not recalculated, leading to errors as the temperature got further away from the initial conditions. XSProc from the SCALE code suite was used to calculate the cross sections as needed. The remainder of this report discusses the changes to Warthog to allow for the implementation of XSProc as an external code. It also discusses the changes made to Warthog to allow it to fit more cleanly into the MultiApp syntax of the MOOSE framework. The capabilities, design, and limitations of Warthog will be described, in addition to some of the test cases that were used to demonstrate the code. Future plans for Warthog will be discussed, including continuation of the modifications to the input and coupling to other SHARP codes such as Nek5000.« less

CATS, continuous automated testing of seismological, hydroacoustic, and infrasound (SHI) processing software.

NASA Astrophysics Data System (ADS)

Brouwer, Albert; Brown, David; Tomuta, Elena

2017-04-01

To detect nuclear explosions, waveform data from over 240 SHI stations world-wide flows into the International Data Centre (IDC) of the Comprehensive Nuclear-Test-Ban Treaty Organization (CTBTO), located in Vienna, Austria. A complex pipeline of software applications processes this data in numerous ways to form event hypotheses. The software codebase comprises over 2 million lines of code, reflects decades of development, and is subject to frequent enhancement and revision. Since processing must run continuously and reliably, software changes are subjected to thorough testing before being put into production. To overcome the limitations and cost of manual testing, the Continuous Automated Testing System (CATS) has been created. CATS provides an isolated replica of the IDC processing environment, and is able to build and test different versions of the pipeline software directly from code repositories that are placed under strict configuration control. Test jobs are scheduled automatically when code repository commits are made. Regressions are reported. We present the CATS design choices and test methods. Particular attention is paid to how the system accommodates the individual testing of strongly interacting software components that lack test instrumentation.

Neutronic Calculation Analysis for CN HCCB TBM-Set

NASA Astrophysics Data System (ADS)

Cao, Qixiang; Zhao, Fengchao; Zhao, Zhou; Wu, Xinghua; Li, Zaixin; Wang, Xiaoyu; Feng, Kaiming

2015-07-01

Using the Monte Carlo transport code MCNP, neutronic calculation analysis for China helium cooled ceramic breeder test blanket module (CN HCCB TBM) and the associated shield block (together called TBM-set) has been carried out based on the latest design of HCCB TBM-set and C-lite model. Key nuclear responses of HCCB TBM-set, such as the neutron flux, tritium production rate, nuclear heating and radiation damage, have been obtained and discussed. These nuclear performance data can be used as the basic input data for other analyses of HCCB TBM-set, such as thermal-hydraulics, thermal-mechanics and safety analysis. supported by the Major State Basic Research Development Program of China (973 Program) (No. 2013GB108000)

Evaluation of Spacecraft Shielding Effectiveness for Radiation Protection

NASA Technical Reports Server (NTRS)

Cucinotta, Francis A.; Wilson, John W.

1999-01-01

The potential for serious health risks from solar particle events (SPE) and galactic cosmic rays (GCR) is a critical issue in the NASA strategic plan for the Human Exploration and Development of Space (HEDS). The excess cost to protect against the GCR and SPE due to current uncertainties in radiation transmission properties and cancer biology could be exceedingly large based on the excess launch costs to shield against uncertainties. The development of advanced shielding concepts is an important risk mitigation area with the potential to significantly reduce risk below conventional mission designs. A key issue in spacecraft material selection is the understanding of nuclear reactions on the transmission properties of materials. High-energy nuclear particles undergo nuclear reactions in passing through materials and tissue altering their composition and producing new radiation types. Spacecraft and planetary habitat designers can utilize radiation transport codes to identify optimal materials for lowering exposures and to optimize spacecraft design to reduce astronaut exposures. To reach these objectives will require providing design engineers with accurate data bases and computationally efficient software for describing the transmission properties of space radiation in materials. Our program will reduce the uncertainty in the transmission properties of space radiation by improving the theoretical description of nuclear reactions and radiation transport, and provide accurate physical descriptions of the track structure of microscopic energy deposition.

Etude des performances de solveurs deterministes sur un coeur rapide a caloporteur sodium

NASA Astrophysics Data System (ADS)

Bay, Charlotte

The reactors of next generation, in particular SFR model, represent a true challenge for current codes and solvers, used mainly for thermic cores. There is no guarantee that their competences could be straight adapted to fast neutron spectrum, or to major design differences. Thus it is necessary to assess the validity of solvers and their potential shortfall in the case of fast neutron reactors. As part of an internship with CEA (France), and at the instigation of EPM Nuclear Institute, this study concerns the following codes : DRAGON/DONJON, ERANOS, PARIS and APOLLO3. The precision assessment has been performed using Monte Carlo code TRIPOLI4. Only core calculation was of interest, namely numerical methods competences in precision and rapidity. Lattice code was not part of the study, that is to say nuclear data, self-shielding, or isotopic compositions. Nor was tackled burnup or time evolution effects. The study consists in two main steps : first evaluating the sensitivity of each solver to calculation parameters, and obtain its optimal calculation set ; then compare their competences in terms of precision and rapidity, by collecting usual quantities (effective multiplication factor, reaction rates map), but also more specific quantities which are crucial to the SFR design, namely control rod worth and sodium void effect. The calculation time is also a key factor. Whatever conclusion or recommendation that could be drawn from this study, they must first of all be applied within similar frameworks, that is to say small fast neutron cores with hexagonal geometry. Eventual adjustments for big cores will have to be demonstrated in developments of this study.

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

NASA Astrophysics Data System (ADS)

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

2004-02-01

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

Cyclotron production of 48V via natTi(d,x)48V nuclear reaction; a promising radionuclide

NASA Astrophysics Data System (ADS)

Usman, A. R.; Khandaker, M. U.; Haba, H.

2017-06-01

In this experimental work, we studied the excitation function of natTi(d,x)48V nuclear reactions from 24 MeV down to threshold energy. Natural titanium foils were arranged in the popular stacked-foil method and activated with deuteron beam generated from an AVF cyclotron at RIKEN, Wako, Japan. The emitted γ activities from the activated foils were measured using an offline γ-ray spectrometry. The present results were analyzed, compared with earlier published experimental data and also with the evaluated data of Talys code. Our new measured data agree with some of the earlier reported experimental data while a partial agreement is found with the evaluated theoretical data. In addition to the use of 48V as a beam intensity monitor, recent studies indicate its potentials as calibrating source in PET cameras and also as a (radioactive) label for medical applications. The results are also expected to further enrich the experimental database and also to play an important role in nuclear reactions model codes design.

Tornado and extreme wind design criteria for nuclear power plants

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

None

1973-12-01

Nuclear power plant design criteria for tornadoes and extreme winds are presented. Data, formulas, and procedures for determining maximum wind loading on structures and parts of structures are included. Extreme wind loading is applied to structures using methods and procedures consistent with ANSI Building Code A58.1- 1972. The design wind velocities specified generally exceed 100-year recurrent interval winds. Tornado wind loading is applied to structures using procedures paralleling those for extrene winds with additional criteria resulting from the atmospheric pressure change accompanying tornadoes and tornado missile inipact effects. Tornado loading for the 48 contiguous United States is specified for twomore » major zones separated by the Continental Divide. A cross reference listing items related to Atomic Energy Commission Safety Analysis Report format is provided. Development supporting tornado criteria is included. (auth)« less

Gap Analysis of Material Properties Data for Ferritic/Martensitic HT-9 Steel

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

Brown, Neil R.; Serrano De Caro, Magdalena; Rodriguez, Edward A.

2012-08-28

The US Department of Energy (DOE), Office of Nuclear Energy (NE), is supporting the development of an ASME Code Case for adoption of 12Cr-1Mo-VW ferritic/martensitic (F/M) steel, commonly known as HT-9, primarily for use in elevated temperature design of liquid-metal fast reactors (LMFR) and components. In 2011, Los Alamos National Laboratory (LANL) nuclear engineering staff began assisting in the development of a small modular reactor (SMR) design concept, previously known as the Hyperion Module, now called the Gen4 Module. LANL staff immediately proposed HT-9 for the reactor vessel and components, as well as fuel clad and ducting, due to itsmore » superior thermal qualities. Although the ASME material Code Case, for adoption of HT-9 as an approved elevated temperature material for LMFR service, is the ultimate goal of this project, there are several key deliverables that must first be successfully accomplished. The most important key deliverable is the research, accumulation, and documentation of specific material parameters; physical, mechanical, and environmental, which becomes the basis for an ASME Code Case. Time-independent tensile and ductility data and time-dependent creep and creep-rupture behavior are some of the material properties required for a successful ASME Code case. Although this report provides a cursory review of the available data, a much more comprehensive study of open-source data would be necessary. This report serves three purposes: (a) provides a list of already existing material data information that could ultimately be made available to the ASME Code, (b) determines the HT-9 material properties data missing from available sources that would be required and (c) estimates the necessary material testing required to close the gap. Ultimately, the gap analysis demonstrates that certain material properties testing will be required to fulfill the necessary information package for an ASME Code Case.« less

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.

Development of 3D pseudo pin-by-pin calculation methodology in ANC

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

Zhang, B.; Mayhue, L.; Huria, H.

2012-07-01

Advanced cores and fuel assembly designs have been developed to improve operational flexibility, economic performance and further enhance safety features of nuclear power plants. The simulation of these new designs, along with strong heterogeneous fuel loading, have brought new challenges to the reactor physics methodologies currently employed in the industrial codes for core analyses. Control rod insertion during normal operation is one operational feature in the AP1000{sup R} plant of Westinghouse next generation Pressurized Water Reactor (PWR) design. This design improves its operational flexibility and efficiency but significantly challenges the conventional reactor physics methods, especially in pin power calculations. Themore » mixture loading of fuel assemblies with significant neutron spectrums causes a strong interaction between different fuel assembly types that is not fully captured with the current core design codes. To overcome the weaknesses of the conventional methods, Westinghouse has developed a state-of-the-art 3D Pin-by-Pin Calculation Methodology (P3C) and successfully implemented in the Westinghouse core design code ANC. The new methodology has been qualified and licensed for pin power prediction. The 3D P3C methodology along with its application and validation will be discussed in the paper. (authors)« less

Overview and Current Status of Analyses of Potential LEU Design Concepts for TREAT

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

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

2015-10-01

Neutronic and thermal-hydraulic analyses have been performed to evaluate the performance of different low-enriched uranium (LEU) fuel design concepts for the conversion of the Transient Reactor Test Facility (TREAT) from its current high-enriched uranium (HEU) fuel. TREAT is an experimental reactor developed to generate high neutron flux transients for the testing of nuclear fuels. The goal of this work was to identify an LEU design which can maintain the performance of the existing HEU core while continuing to operate safely. A wide variety of design options were considered, with a focus on minimizing peak fuel temperatures and optimizing the powermore » coupling between the TREAT core and test samples. Designs were also evaluated to ensure that they provide sufficient reactivity and shutdown margin for each control rod bank. Analyses were performed using the core loading and experiment configuration of historic M8 Power Calibration experiments (M8CAL). The Monte Carlo code MCNP was utilized for steady-state analyses, and transient calculations were performed with the point kinetics code TREKIN. Thermal analyses were performed with the COMSOL multi-physics code. Using the results of this study, a new LEU Baseline design concept is being established, which will be evaluated in detail in a future report.« less

Study on Utilization of Super Grade Plutonium in Molten Salt Reactor FUJI-U3 using CITATION Code

NASA Astrophysics Data System (ADS)

Wulandari, Cici; Waris, Abdul; Pramuditya, Syeilendra; Asril, Pramutadi AM; Novitrian

2017-07-01

FUJI-U3 type of Molten Salt Reactor (MSR) has a unique design since it consists of three core regions in order to avoid the replacement of graphite as moderator. MSR uses floride as a nuclear fuel salt with the most popular chemical composition is LiF-BeF2-ThF4-233UF4. ThF4 and 233UF4 are the fertile and fissile materials, respectively. On the other hand, LiF and BeF2 working as both fuel and heat transfer medium. In this study, the super grade plutonium will be utilized as substitution of 233U since plutonium is easier to be obtained compared to 233U as main fuel. Neutronics calculation was performed by using PIJ and CITATION modules of SRAC 2002 code with JENDL 3.2 as nuclear data library.

Decay heat uncertainty quantification of MYRRHA

NASA Astrophysics Data System (ADS)

Fiorito, Luca; Buss, Oliver; Hoefer, Axel; Stankovskiy, Alexey; Eynde, Gert Van den

2017-09-01

MYRRHA is a lead-bismuth cooled MOX-fueled accelerator driven system (ADS) currently in the design phase at SCK·CEN in Belgium. The correct evaluation of the decay heat and of its uncertainty level is very important for the safety demonstration of the reactor. In the first part of this work we assessed the decay heat released by the MYRRHA core using the ALEPH-2 burnup code. The second part of the study focused on the nuclear data uncertainty and covariance propagation to the MYRRHA decay heat. Radioactive decay data, independent fission yield and cross section uncertainties/covariances were propagated using two nuclear data sampling codes, namely NUDUNA and SANDY. According to the results, 238U cross sections and fission yield data are the largest contributors to the MYRRHA decay heat uncertainty. The calculated uncertainty values are deemed acceptable from the safety point of view as they are well within the available regulatory limits.

Neutronics Analysis of SMART Small Modular Reactor using SRAC 2006 Code

NASA Astrophysics Data System (ADS)

Ramdhani, Rahmi N.; Prastyo, Puguh A.; Waris, Abdul; Widayani; Kurniadi, Rizal

2017-07-01

Small modular reactors (SMRs) are part of a new generation of nuclear reactor being developed worldwide. One of the advantages of SMR is the flexibility to adopt the advanced design concepts and technology. SMART (System integrated Modular Advanced ReacTor) is a small sized integral type PWR with a thermal power of 330 MW that has been developed by KAERI (Korea Atomic Energy Research Institute). SMART core consists of 57 fuel assemblies which are based on the well proven 17×17 array that has been used in Korean commercial PWRs. SMART is soluble boron free, and the high initial reactivity is mainly controlled by burnable absorbers. The goal of this study is to perform neutronics evaluation of SMART core with UO2 as main fuel. Neutronics calculation was performed by using PIJ and CITATION modules of SRAC 2006 code with JENDL 3.3 as nuclear data library.

Decoding the function of nuclear long non-coding RNAs.

PubMed

Chen, Ling-Ling; Carmichael, Gordon G

2010-06-01

Long non-coding RNAs (lncRNAs) are mRNA-like, non-protein-coding RNAs that are pervasively transcribed throughout eukaryotic genomes. Rather than silently accumulating in the nucleus, many of these are now known or suspected to play important roles in nuclear architecture or in the regulation of gene expression. In this review, we highlight some recent progress in how lncRNAs regulate these important nuclear processes at the molecular level. Copyright 2010 Elsevier Ltd. All rights reserved.

Development of a Coded Aperture X-Ray Backscatter Imager for Explosive Device Detection

NASA Astrophysics Data System (ADS)

Faust, Anthony A.; Rothschild, Richard E.; Leblanc, Philippe; McFee, John Elton

2009-02-01

Defence R&D Canada has an active research and development program on detection of explosive devices using nuclear methods. One system under development is a coded aperture-based X-ray backscatter imaging detector designed to provide sufficient speed, contrast and spatial resolution to detect antipersonnel landmines and improvised explosive devices. The successful development of a hand-held imaging detector requires, among other things, a light-weight, ruggedized detector with low power requirements, supplying high spatial resolution. The University of California, San Diego-designed HEXIS detector provides a modern, large area, high-temperature CZT imaging surface, robustly packaged in a light-weight housing with sound mechanical properties. Based on the potential for the HEXIS detector to be incorporated as the detection element of a hand-held imaging detector, the authors initiated a collaborative effort to demonstrate the capability of a coded aperture-based X-ray backscatter imaging detector. This paper will discuss the landmine and IED detection problem and review the coded aperture technique. Results from initial proof-of-principle experiments will then be reported.

Overview of codes and tools for nuclear engineering education

NASA Astrophysics Data System (ADS)

Yakovlev, D.; Pryakhin, A.; Medvedeva, L.

2017-01-01

The recent world trends in nuclear education have been developed in the direction of social education, networking, virtual tools and codes. MEPhI as a global leader on the world education market implements new advanced technologies for the distance and online learning and for student research work. MEPhI produced special codes, tools and web resources based on the internet platform to support education in the field of nuclear technology. At the same time, MEPhI actively uses codes and tools from the third parties. Several types of the tools are considered: calculation codes, nuclear data visualization tools, virtual labs, PC-based educational simulators for nuclear power plants (NPP), CLP4NET, education web-platforms, distance courses (MOOCs and controlled and managed content systems). The university pays special attention to integrated products such as CLP4NET, which is not a learning course, but serves to automate the process of learning through distance technologies. CLP4NET organizes all tools in the same information space. Up to now, MEPhI has achieved significant results in the field of distance education and online system implementation.

Integrated System Modeling for Nuclear Thermal Propulsion (NTP)

NASA Technical Reports Server (NTRS)

Ryan, Stephen W.; Borowski, Stanley K.

2014-01-01

Nuclear thermal propulsion (NTP) has long been identified as a key enabling technology for space exploration beyond LEO. From Wernher Von Braun's early concepts for crewed missions to the Moon and Mars to the current Mars Design Reference Architecture (DRA) 5.0 and recent lunar and asteroid mission studies, the high thrust and specific impulse of NTP opens up possibilities such as reusability that are just not feasible with competing approaches. Although NTP technology was proven in the Rover / NERVA projects in the early days of the space program, an integrated spacecraft using NTP has never been developed. Such a spacecraft presents a challenging multidisciplinary systems integration problem. The disciplines that must come together include not only nuclear propulsion and power, but also thermal management, power, structures, orbital dynamics, etc. Some of this integration logic was incorporated into a vehicle sizing code developed at NASA's Glenn Research Center (GRC) in the early 1990s called MOMMA, and later into an Excel-based tool called SIZER. Recently, a team at GRC has developed an open source framework for solving Multidisciplinary Design, Analysis and Optimization (MDAO) problems called OpenMDAO. A modeling approach is presented that builds on previous work in NTP vehicle sizing and mission analysis by making use of the OpenMDAO framework to enable modular and reconfigurable representations of various NTP vehicle configurations and mission scenarios. This approach is currently applied to vehicle sizing, but is extensible to optimization of vehicle and mission designs. The key features of the code will be discussed and examples of NTP transfer vehicles and candidate missions will be presented.

Utilization of recently developed codes for high power Brayton and Rankine cycle power systems

NASA Technical Reports Server (NTRS)

Doherty, Michael P.

1993-01-01

Two recently developed FORTRAN computer codes for high power Brayton and Rankine thermodynamic cycle analysis for space power applications are presented. The codes were written in support of an effort to develop a series of subsystem models for multimegawatt Nuclear Electric Propulsion, but their use is not limited just to nuclear heat sources or to electric propulsion. Code development background, a description of the codes, some sample input/output from one of the codes, and state future plans/implications for the use of these codes by NASA's Lewis Research Center are provided.

Experimental investigations, modeling, and analyses of high-temperature devices for space applications: Part 1. Final report, June 1996--December 1998

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

Tournier, J.; El-Genk, M.S.; Huang, L.

1999-01-01

The Institute of Space and Nuclear Power Studies at the University of New Mexico has developed a computer simulation of cylindrical geometry alkali metal thermal-to-electric converter cells using a standard Fortran 77 computer code. The objective and use of this code was to compare the experimental measurements with computer simulations, upgrade the model as appropriate, and conduct investigations of various methods to improve the design and performance of the devices for improved efficiency, durability, and longer operational lifetime. The Institute of Space and Nuclear Power Studies participated in vacuum testing of PX series alkali metal thermal-to-electric converter cells and developedmore » the alkali metal thermal-to-electric converter Performance Evaluation and Analysis Model. This computer model consisted of a sodium pressure loss model, a cell electrochemical and electric model, and a radiation/conduction heat transfer model. The code closely predicted the operation and performance of a wide variety of PX series cells which led to suggestions for improvements to both lifetime and performance. The code provides valuable insight into the operation of the cell, predicts parameters of components within the cell, and is a useful tool for predicting both the transient and steady state performance of systems of cells.« less

Experimental investigations, modeling, and analyses of high-temperature devices for space applications: Part 2. Final report, June 1996--December 1998

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

Tournier, J.; El-Genk, M.S.; Huang, L.

1999-01-01

The Institute of Space and Nuclear Power Studies at the University of New Mexico has developed a computer simulation of cylindrical geometry alkali metal thermal-to-electric converter cells using a standard Fortran 77 computer code. The objective and use of this code was to compare the experimental measurements with computer simulations, upgrade the model as appropriate, and conduct investigations of various methods to improve the design and performance of the devices for improved efficiency, durability, and longer operational lifetime. The Institute of Space and Nuclear Power Studies participated in vacuum testing of PX series alkali metal thermal-to-electric converter cells and developedmore » the alkali metal thermal-to-electric converter Performance Evaluation and Analysis Model. This computer model consisted of a sodium pressure loss model, a cell electrochemical and electric model, and a radiation/conduction heat transfer model. The code closely predicted the operation and performance of a wide variety of PX series cells which led to suggestions for improvements to both lifetime and performance. The code provides valuable insight into the operation of the cell, predicts parameters of components within the cell, and is a useful tool for predicting both the transient and steady state performance of systems of cells.« less

Technology Innovation for the CTBT, the National Laboratory Contribution

NASA Astrophysics Data System (ADS)

Goldstein, W. H.

2016-12-01

The Comprehensive Nuclear-Test-Ban Treaty (CTBT) and its Protocol are the result of a long history of scientific engagement and international technical collaboration. The U.S. Department of Energy National Laboratories have been conducting nuclear explosive test-ban research for over 50 years and have made significant contributions to this legacy. Recent examples include the RSTT (regional seismic travel time) computer code and the Smart Sampler—both of these products are the result of collaborations among Livermore, Sandia, Los Alamos, and Pacific Northwest National Laboratories. The RSTT code enables fast and accurate seismic event locations using regional data. This code solves the long-standing problem of using teleseismic and regional seismic data together to locate events. The Smart Sampler is designed for use in On-site Inspections to sample soil gases to look for noble gas fission products from a potential underground nuclear explosive test. The Smart Sampler solves the long-standing problem of collecting soil gases without contaminating the sample with gases from the atmosphere by operating only during atmospheric low-pressure events. Both these products are being evaluated by the Preparatory Commission for the CTBT Organization and the international community. In addition to R&D, the National Laboratories provide experts to support U.S. policy makers in ongoing discussions such as CTBT Working Group B, which sets policy for the development of the CTBT monitoring and verification regime.

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

Gauld, Ian C.; Giaquinto, J. M.; Delashmitt, J. S.

Destructive radiochemical assay measurements of spent nuclear fuel rod segments from an assembly irradiated in the Three Mile Island unit 1 (TMI-1) pressurized water reactor have been performed at Oak Ridge National Laboratory (ORNL). Assay data are reported for five samples from two fuel rods of the same assembly. The TMI-1 assembly was a 15 X 15 design with an initial enrichment of 4.013 wt% 235U, and the measured samples achieved burnups between 45.5 and 54.5 gigawatt days per metric ton of initial uranium (GWd/t). Measurements were performed mainly using inductively coupled plasma mass spectrometry after elemental separation via highmore » performance liquid chromatography. High precision measurements were achieved using isotope dilution techniques for many of the lanthanides, uranium, and plutonium isotopes. Measurements are reported for more than 50 different isotopes and 16 elements. One of the two TMI-1 fuel rods measured in this work had been measured previously by Argonne National Laboratory (ANL), and these data have been widely used to support code and nuclear data validation. Recently, ORNL provided an important opportunity to independently cross check results against previous measurements performed at ANL. The measured nuclide concentrations are used to validate burnup calculations using the SCALE nuclear systems modeling and simulation code suite. These results show that the new measurements provide reliable benchmark data for computer code validation.« less

Cross-sections of residual nuclei from deuteron irradiation of thin thorium target at energy 7 GeV

NASA Astrophysics Data System (ADS)

Vespalec, Radek; Adam, Jindrich; Baldin, Anton Alexandrovich; Khushvaktov, Jurabek; Solnyshkin, Alexander Alexandrovich; Tsoupko-Sitnikov, Vsevolod Mikhailovich; Tyutyunikov, Sergey Ivanovich; Vrzalova, Jitka; Zavorka, Lukas; Zeman, Miroslav

2017-09-01

The residual nuclei yields are of great importance for the estimation of basic radiation-technology characteristics (like a total target activity, production of long-lived nuclides etc.) of accelerator driven systems planned for transmutation of spent nuclear fuel and for a design of radioisotopes production facilities. Experimental data are also essential for validation of nuclear codes describing various stages of a spallation reaction. Therefore, the main aim of this work is to add new experimental data in energy region of relativistic deuterons, as similar data are missing in nuclear databases. The sample made of thin natural thorium foil was irradiated at JINR Nuclotron accelerator with a deuteron beam of the total kinetic energy 7 GeV. Integral number of deuterons was determined with the use of aluminum activation detectors. Products of deuteron induced spallation reaction were qualified and quantified by means of gamma-ray spectroscopy method. Several important spectroscopic corrections were applied to obtain results of high accuracy. Experimental cumulative and independent cross-sections were determined for more than 80 isotopes including meta-stable isomers. The total uncertainty of results rarely exceeded 9%. Experimental results were compared with MCNP6.1 Monte-Carlo code predictions. Generally, experimental and calculated cross-sections are in a reasonably good agreement, with the exception of a few light isotopes in a fragmentation region, where the calculations are highly under-estimated. Measured data will be useful for future development of high-energy nuclear codes. After completion, final data will be added into the EXFOR database.

Systems for the Intermodal Routing of Spent Nuclear Fuel

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

Peterson, Steven K; Liu, Cheng

The safe and secure movement of spent nuclear fuel from shutdown and active reactor facilities to intermediate or long term storage sites may, in some instances, require the use of several modes of transportation to accomplish the move. To that end, a fully operable multi-modal routing system is being developed within Oak Ridge National Laboratory s (ORNL) WebTRAGIS (Transportation Routing Analysis Geographic Information System). This study aims to provide an overview of multi-modal routing, the existing state of the TRAGIS networks, the source data needs, and the requirements for developing structural relationships between various modes to create a suitable systemmore » for modeling the transport of spent nuclear fuel via a multimodal network. Modern transportation systems are comprised of interconnected, yet separate, modal networks. Efficient transportation networks rely upon the smooth transfer of cargoes at junction points that serve as connectors between modes. A key logistical impediment to the shipment of spent nuclear fuel is the absence of identified or designated transfer locations between transport modes. Understanding the potential network impacts on intermodal transportation of spent nuclear fuel is vital for planning transportation routes from origin to destination. By identifying key locations where modes intersect, routing decisions can be made to prioritize cost savings, optimize transport times and minimize potential risks to the population and environment. In order to facilitate such a process, ORNL began the development of a base intermodal network and associated routing code. The network was developed using previous intermodal networks and information from publicly available data sources to construct a database of potential intermodal transfer locations with likely capability to handle spent nuclear fuel casks. The coding development focused on modifying the existing WebTRAGIS routing code to accommodate intermodal transfers and the selection of prioritization constraints and modifiers to determine route selection. The limitations of the current model and future directions for development are discussed, including the current state of information on possible intermodal transfer locations for spent fuel.« less

Nuclear shell model code CRUNCHER

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

Resler, D.A.; Grimes, S.M.

1988-05-01

A new nuclear shell model code CRUNCHER, patterned after the code VLADIMIR, has been developed. While CRUNCHER and VLADIMIR employ the techniques of an uncoupled basis and the Lanczos process, improvements in the new code allow it to handle much larger problems than the previous code and to perform them more efficiently. Tests involving a moderately sized calculation indicate that CRUNCHER running on a SUN 3/260 workstation requires approximately one-half the central processing unit (CPU) time required by VLADIMIR running on a CRAY-1 supercomputer.

EMPIRE: Nuclear Reaction Model Code System for Data Evaluation

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

Herman, M.; Capote, R.; Carlson, B.V.

EMPIRE is a modular system of nuclear reaction codes, comprising various nuclear models, and designed for calculations over a broad range of energies and incident particles. A projectile can be a neutron, proton, any ion (including heavy-ions) or a photon. The energy range extends from the beginning of the unresolved resonance region for neutron-induced reactions ({approx} keV) and goes up to several hundred MeV for heavy-ion induced reactions. The code accounts for the major nuclear reaction mechanisms, including direct, pre-equilibrium and compound nucleus ones. Direct reactions are described by a generalized optical model (ECIS03) or by the simplified coupled-channels approachmore » (CCFUS). The pre-equilibrium mechanism can be treated by a deformation dependent multi-step direct (ORION + TRISTAN) model, by a NVWY multi-step compound one or by either a pre-equilibrium exciton model with cluster emission (PCROSS) or by another with full angular momentum coupling (DEGAS). Finally, the compound nucleus decay is described by the full featured Hauser-Feshbach model with {gamma}-cascade and width-fluctuations. Advanced treatment of the fission channel takes into account transmission through a multiple-humped fission barrier with absorption in the wells. The fission probability is derived in the WKB approximation within the optical model of fission. Several options for nuclear level densities include the EMPIRE-specific approach, which accounts for the effects of the dynamic deformation of a fast rotating nucleus, the classical Gilbert-Cameron approach and pre-calculated tables obtained with a microscopic model based on HFB single-particle level schemes with collective enhancement. A comprehensive library of input parameters covers nuclear masses, optical model parameters, ground state deformations, discrete levels and decay schemes, level densities, fission barriers, moments of inertia and {gamma}-ray strength functions. The results can be converted into ENDF-6 formatted files using the accompanying code EMPEND and completed with neutron resonances extracted from the existing evaluations. The package contains the full EXFOR (CSISRS) library of experimental reaction data that are automatically retrieved during the calculations. Publication quality graphs can be obtained using the powerful and flexible plotting package ZVView. The graphic user interface, written in Tcl/Tk, provides for easy operation of the system. This paper describes the capabilities of the code, outlines physical models and indicates parameter libraries used by EMPIRE to predict reaction cross sections and spectra, mainly for nucleon-induced reactions. Selected applications of EMPIRE are discussed, the most important being an extensive use of the code in evaluations of neutron reactions for the new US library ENDF/B-VII.0. Future extensions of the system are outlined, including neutron resonance module as well as capabilities of generating covariances, using both KALMAN and Monte-Carlo methods, that are still being advanced and refined.« less

Three-dimensional Monte Carlo calculation of some nuclear parameters

NASA Astrophysics Data System (ADS)

Günay, Mehtap; Şeker, Gökmen

2017-09-01

In this study, a fusion-fission hybrid reactor system was designed by using 9Cr2WVTa Ferritic steel structural material and the molten salt-heavy metal mixtures 99-95% Li20Sn80 + 1-5% RG-Pu, 99-95% Li20Sn80 + 1-5% RG-PuF4, and 99-95% Li20Sn80 + 1-5% RG-PuO2, as fluids. The fluids were used in the liquid first wall, blanket and shield zones of a fusion-fission hybrid reactor system. Beryllium (Be) zone with the width of 3 cm was used for the neutron multiplication between the liquid first wall and blanket. This study analyzes the nuclear parameters such as tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate, fission reaction rate in liquid first wall, blanket and shield zones and investigates effects of reactor grade Pu content in the designed system on these nuclear parameters. Three-dimensional analyses were performed by using the Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0.

Modelisation and distribution of neutron flux in radium-beryllium source (226Ra-Be)

NASA Astrophysics Data System (ADS)

Didi, Abdessamad; Dadouch, Ahmed; Jai, Otman

2017-09-01

Using the Monte Carlo N-Particle code (MCNP-6), to analyze the thermal, epithermal and fast neutron fluxes, of 3 millicuries of radium-beryllium, for determine the qualitative and quantitative of many materials, using method of neutron activation analysis. Radium-beryllium source of neutron is established to practical work and research in nuclear field. The main objective of this work was to enable us harness the profile flux of radium-beryllium irradiation, this theoretical study permits to discuss the design of the optimal irradiation and performance for increased the facility research and education of nuclear physics.

Alignment-based and alignment-free methods converge with experimental data on amino acids coded by stop codons at split between nuclear and mitochondrial genetic codes.

PubMed

Seligmann, Hervé

2018-05-01

Genetic codes mainly evolve by reassigning punctuation codons, starts and stops. Previous analyses assuming that undefined amino acids translate stops showed greater divergence between nuclear and mitochondrial genetic codes. Here, three independent methods converge on which amino acids translated stops at split between nuclear and mitochondrial genetic codes: (a) alignment-free genetic code comparisons inserting different amino acids at stops; (b) alignment-based blast analyses of hypothetical peptides translated from non-coding mitochondrial sequences, inserting different amino acids at stops; (c) biases in amino acid insertions at stops in proteomic data. Hence short-term protein evolution models reconstruct long-term genetic code evolution. Mitochondria reassign stops to amino acids otherwise inserted at stops by codon-anticodon mismatches (near-cognate tRNAs). Hence dual function (translation termination and translation by codon-anticodon mismatch) precedes mitochondrial reassignments of stops to amino acids. Stop ambiguity increases coded information, compensates endocellular mitogenome reduction. Mitochondrial codon reassignments might prevent viral infections. Copyright © 2018 Elsevier B.V. All rights reserved.

Integration of Dakota into the NEAMS Workbench

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

Swiler, Laura Painton; Lefebvre, Robert A.; Langley, Brandon R.

2017-07-01

This report summarizes a NEAMS (Nuclear Energy Advanced Modeling and Simulation) project focused on integrating Dakota into the NEAMS Workbench. The NEAMS Workbench, developed at Oak Ridge National Laboratory, is a new software framework that provides a graphical user interface, input file creation, parsing, validation, job execution, workflow management, and output processing for a variety of nuclear codes. Dakota is a tool developed at Sandia National Laboratories that provides a suite of uncertainty quantification and optimization algorithms. Providing Dakota within the NEAMS Workbench allows users of nuclear simulation codes to perform uncertainty and optimization studies on their nuclear codes frommore » within a common, integrated environment. Details of the integration and parsing are provided, along with an example of Dakota running a sampling study on the fuels performance code, BISON, from within the NEAMS Workbench.« less

A dose assessment method for arbitrary geometries with virtual reality in the nuclear facilities decommissioning

NASA Astrophysics Data System (ADS)

Chao, Nan; Liu, Yong-kuo; Xia, Hong; Ayodeji, Abiodun; Bai, Lu

2018-03-01

During the decommissioning of nuclear facilities, a large number of cutting and demolition activities are performed, which results in a frequent change in the structure and produce many irregular objects. In order to assess dose rates during the cutting and demolition process, a flexible dose assessment method for arbitrary geometries and radiation sources was proposed based on virtual reality technology and Point-Kernel method. The initial geometry is designed with the three-dimensional computer-aided design tools. An approximate model is built automatically in the process of geometric modeling via three procedures namely: space division, rough modeling of the body and fine modeling of the surface, all in combination with collision detection of virtual reality technology. Then point kernels are generated by sampling within the approximate model, and when the material and radiometric attributes are inputted, dose rates can be calculated with the Point-Kernel method. To account for radiation scattering effects, buildup factors are calculated with the Geometric-Progression formula in the fitting function. The effectiveness and accuracy of the proposed method was verified by means of simulations using different geometries and the dose rate results were compared with that derived from CIDEC code, MCNP code and experimental measurements.

HZETRN: Description of a free-space ion and nucleon transport and shielding computer program

NASA Technical Reports Server (NTRS)

Wilson, John W.; Badavi, Francis F.; Cucinotta, Francis A.; Shinn, Judy L.; Badhwar, Gautam D.; Silberberg, R.; Tsao, C. H.; Townsend, Lawrence W.; Tripathi, Ram K.

1995-01-01

The high-charge-and energy (HZE) transport computer program HZETRN is developed to address the problems of free-space radiation transport and shielding. The HZETRN program is intended specifically for the design engineer who is interested in obtaining fast and accurate dosimetric information for the design and construction of space modules and devices. The program is based on a one-dimensional space-marching formulation of the Boltzmann transport equation with a straight-ahead approximation. The effect of the long-range Coulomb force and electron interaction is treated as a continuous slowing-down process. Atomic (electronic) stopping power coefficients with energies above a few A MeV are calculated by using Bethe's theory including Bragg's rule, Ziegler's shell corrections, and effective charge. Nuclear absorption cross sections are obtained from fits to quantum calculations and total cross sections are obtained with a Ramsauer formalism. Nuclear fragmentation cross sections are calculated with a semiempirical abrasion-ablation fragmentation model. The relation of the final computer code to the Boltzmann equation is discussed in the context of simplifying assumptions. A detailed description of the flow of the computer code, input requirements, sample output, and compatibility requirements for non-VAX platforms are provided.

Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS)

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

Banerjee, Kaushik; Clarity, Justin B; Cumberland, Riley M

This will be licensed via RSICC. A new, integrated data and analysis system has been designed to simplify and automate the performance of accurate and efficient evaluations for characterizing the input to the overall nuclear waste management system -UNF-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). A relational database within UNF-ST&DARDS provides a standard means by which UNF-ST&DARDS can succinctly store and retrieve modeling and simulation (M&S) parameters for specific spent nuclear fuel analysis. A library of various analysis model templates provides the ability to communicate the various set of M&S parameters to the most appropriate M&S application.more » Interactive visualization capabilities facilitate data analysis and results interpretation. UNF-ST&DARDS current analysis capabilities include (1) assembly-specific depletion and decay, (2) and spent nuclear fuel cask-specific criticality and shielding. Currently, UNF-ST&DARDS uses SCALE nuclear analysis code system for performing nuclear analysis.« less

Fuel burnup analysis for IRIS reactor using MCNPX and WIMS-D5 codes

NASA Astrophysics Data System (ADS)

Amin, E. A.; Bashter, I. I.; Hassan, Nabil M.; Mustafa, S. S.

2017-02-01

International Reactor Innovative and Secure (IRIS) reactor is a compact power reactor designed with especial features. It contains Integral Fuel Burnable Absorber (IFBA). The core is heterogeneous both axially and radially. This work provides the full core burn up analysis for IRIS reactor using MCNPX and WIMDS-D5 codes. Criticality calculations, radial and axial power distributions and nuclear peaking factor at the different stages of burnup were studied. Effective multiplication factor values for the core were estimated by coupling MCNPX code with WIMS-D5 code and compared with SAS2H/KENO-V code values at different stages of burnup. The two calculation codes show good agreement and correlation. The values of radial and axial powers for the full core were also compared with published results given by SAS2H/KENO-V code (at the beginning and end of reactor operation). The behavior of both radial and axial power distribution is quiet similar to the other data published by SAS2H/KENO-V code. The peaking factor values estimated in the present work are close to its values calculated by SAS2H/KENO-V code.

International Maritime Organizational Code For Safe Carriage Of Irradiated Nuclear Fuel, Plutonium and High-Level Radioactive Wastes in Flasks On Board Ships -- IMO Resolution A.748(18)

DOT National Transportation Integrated Search

1994-05-26

This Circular calls the attention of Coast Guard field units, marine surveyors, shippers and carriers of nuclear materials to the International Maritime Organization (IMO) Code for the Safe Carriage of Irradiated Nuclear Fuel, Plutonium and High-Leve...

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

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

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

1997-07-01

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

TRAC-PF1 code verification with data from the OTIS test facility. [Once-Through Intergral System

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

Childerson, M.T.; Fujita, R.K.

1985-01-01

A computer code (TRAC-PF1/MOD1) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, designated the One-Through Integral System (OTIS), were obtained for the Babcock and Wilcox NSSS and compared to TRAC predictions. The OTIS tests provided a challenging small break LOCA data set for TRAC verification. The major phases of a small break LOCA observed in the OTIS tests included pressurizer draining and loop saturation, intermittent reactor coolant system circulation, boiler-condenser mode, and the initial stages of refill. The TRAC code wasmore » successful in predicting OTIS loop conditions (system pressures and temperatures) after modification of the steam generator model. In particular, the code predicted both pool and auxiliary-feedwater initiated boiler-condenser mode heat transfer.« less

TOOKUIL: A case study in user interface development for safety code application

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

Gray, D.L.; Harkins, C.K.; Hoole, J.G.

1997-07-01

Traditionally, there has been a very high learning curve associated with using nuclear power plant (NPP) analysis codes. Even for seasoned plant analysts and engineers, the process of building or modifying an input model for present day NPP analysis codes is tedious, error prone, and time consuming. Current cost constraints and performance demands place an additional burden on today`s safety analysis community. Advances in graphical user interface (GUI) technology have been applied to obtain significant productivity and quality assurance improvements for the Transient Reactor Analysis Code (TRAC) input model development. KAPL Inc. has developed an X Windows-based graphical user interfacemore » named TOOKUIL which supports the design and analysis process, acting as a preprocessor, runtime editor, help system, and post processor for TRAC. This paper summarizes the objectives of the project, the GUI development process and experiences, and the resulting end product, TOOKUIL.« less

Simulation of Nuclear Reactor Kinetics by the Monte Carlo Method

NASA Astrophysics Data System (ADS)

Gomin, E. A.; Davidenko, V. D.; Zinchenko, A. S.; Kharchenko, I. K.

2017-12-01

The KIR computer code intended for calculations of nuclear reactor kinetics using the Monte Carlo method is described. The algorithm implemented in the code is described in detail. Some results of test calculations are given.

The gene coding for small ribosomal subunit RNA in the basidiomycete Ustilago maydis contains a group I intron.

PubMed Central

De Wachter, R; Neefs, J M; Goris, A; Van de Peer, Y

1992-01-01

The nucleotide sequence of the gene coding for small ribosomal subunit RNA in the basidiomycete Ustilago maydis was determined. It revealed the presence of a group I intron with a length of 411 nucleotides. This is the third occurrence of such an intron discovered in a small subunit rRNA gene encoded by a eukaryotic nuclear genome. The other two occurrences are in Pneumocystis carinii, a fungus of uncertain taxonomic status, and Ankistrodesmus stipitatus, a green alga. The nucleotides of the conserved core structure of 101 group I intron sequences present in different genes and genome types were aligned and their evolutionary relatedness was examined. This revealed a cluster including all group I introns hitherto found in eukaryotic nuclear genes coding for small and large subunit rRNAs. A secondary structure model was designed for the area of the Ustilago maydis small ribosomal subunit RNA precursor where the intron is situated. It shows that the internal guide sequence pairing with the intron boundaries fits between two helices of the small subunit rRNA, and that minimal rearrangement of base pairs suffices to achieve the definitive secondary structure of the 18S rRNA upon splicing. PMID:1561081

Radiation transport calculations for cosmic radiation.

PubMed

Endo, A; Sato, T

2012-01-01

The radiation environment inside and near spacecraft consists of various components of primary radiation in space and secondary radiation produced by the interaction of the primary radiation with the walls and equipment of the spacecraft. Radiation fields inside astronauts are different from those outside them, because of the body's self-shielding as well as the nuclear fragmentation reactions occurring in the human body. Several computer codes have been developed to simulate the physical processes of the coupled transport of protons, high-charge and high-energy nuclei, and the secondary radiation produced in atomic and nuclear collision processes in matter. These computer codes have been used in various space radiation protection applications: shielding design for spacecraft and planetary habitats, simulation of instrument and detector responses, analysis of absorbed doses and quality factors in organs and tissues, and study of biological effects. This paper focuses on the methods and computer codes used for radiation transport calculations on cosmic radiation, and their application to the analysis of radiation fields inside spacecraft, evaluation of organ doses in the human body, and calculation of dose conversion coefficients using the reference phantoms defined in ICRP Publication 110. Copyright © 2012. Published by Elsevier Ltd.

Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

NASA Astrophysics Data System (ADS)

Ilham, Muhammad; Su'ud, Zaki

2017-01-01

Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

Academic Radiologist Subspecialty Identification Using a Novel Claims-Based Classification System.

PubMed

Rosenkrantz, Andrew B; Wang, Wenyi; Hughes, Danny R; Ginocchio, Luke A; Rosman, David A; Duszak, Richard

2017-06-01

The objective of the present study is to assess the feasibility of a novel claims-based classification system for payer identification of academic radiologist subspecialties. Using a categorization scheme based on the Neiman Imaging Types of Service (NITOS) system, we mapped the Medicare Part B services billed by all radiologists from 2012 to 2014, assigning them to the following subspecialty categories: abdominal imaging, breast imaging, cardiothoracic imaging, musculoskeletal imaging, nuclear medicine, interventional radiology, and neuroradiology. The percentage of subspecialty work relative value units (RVUs) to total billed work RVUs was calculated for each radiologist nationwide. For radiologists at the top 20 academic departments funded by the National Institutes of Health, those percentages were compared with subspecialties designated on faculty websites. NITOS-based subspecialty assignments were also compared with the only radiologist subspecialty classifications currently recognized by Medicare (i.e., nuclear medicine and interventional radiology). Of 1012 academic radiologists studied, the median percentage of Medicare-billed NITOS-based subspecialty work RVUs matching the subspecialty designated on radiologists' own websites ranged from 71.3% (for nuclear medicine) to 98.9% (for neuroradiology). A NITOS-based work RVU threshold of 50% correctly classified 89.8% of radiologists (5.9% were not mapped to any subspecialty; subspecialty error rate, 4.2%). In contrast, existing Medicare provider codes identified only 46.7% of nuclear medicine physicians and 39.4% of interventional radiologists. Using a framework based on a recently established imaging health services research tool that maps service codes based on imaging modality and body region, Medicare claims data can be used to consistently identify academic radiologists by subspecialty in a manner not possible with the use of existing Medicare physician specialty identifiers. This method may facilitate more appropriate performance metrics for subspecialty academic physicians under emerging value-based payment models.

Impact of thorium based molten salt reactor on the closure of the nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Jaradat, Safwan Qasim Mohammad

Molten salt reactor (MSR) is one of six reactors selected by the Generation IV International Forum (GIF). The liquid fluoride thorium reactor (LFTR) is a MSR concept based on thorium fuel cycle. LFTR uses liquid fluoride salts as a nuclear fuel. It uses 232Th and 233U as the fertile and fissile materials, respectively. Fluoride salt of these nuclides is dissolved in a mixed carrier salt of lithium and beryllium (FLiBe). The objective of this research was to complete feasibility studies of a small commercial thermal LFTR. The focus was on neutronic calculations in order to prescribe core design parameter such as core size, fuel block pitch (p), fuel channel radius, fuel path, reflector thickness, fuel salt composition, and power. In order to achieve this objective, the applicability of Monte Carlo N-Particle Transport Code (MCNP) to MSR modeling was verified. Then, a prescription for conceptual small thermal reactor LFTR and relevant calculations were performed using MCNP to determine the main neutronic parameters of the core reactor. The MCNP code was used to study the reactor physics characteristics for the FUJI-U3 reactor. The results were then compared with the results obtained from the original FUJI-U3 using the reactor physics code SRAC95 and the burnup analysis code ORIPHY2. The results were comparable with each other. Based on the results, MCNP was found to be a reliable code to model a small thermal LFTR and study all the related reactor physics characteristics. The results of this study were promising and successful in demonstrating a prefatory small commercial LFTR design. The outcome of using a small core reactor with a diameter/height of 280/260 cm that would operate for more than five years at a power level of 150 MWth was studied. The fuel system 7LiF - BeF2 - ThF4 - UF4 with a (233U/ 232Th) = 2.01 % was the candidate fuel for this reactor core.

PACER -- A fast running computer code for the calculation of short-term containment/confinement loads following coolant boundary failure. Volume 2: User information

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

Sienicki, J.J.

A fast running and simple computer code has been developed to calculate pressure loadings inside light water reactor containments/confinements under loss-of-coolant accident conditions. PACER was originally developed to calculate containment/confinement pressure and temperature time histories for loss-of-coolant accidents in Soviet-designed VVER reactors and is relevant to the activities of the US International Nuclear Safety Center. The code employs a multicompartment representation of the containment volume and is focused upon application to early time containment phenomena during and immediately following blowdown. PACER has been developed for FORTRAN 77 and earlier versions of FORTRAN. The code has been successfully compiled and executedmore » on SUN SPARC and Hewlett-Packard HP-735 workstations provided that appropriate compiler options are specified. The code incorporates both capabilities built around a hardwired default generic VVER-440 Model V230 design as well as fairly general user-defined input. However, array dimensions are hardwired and must be changed by modifying the source code if the number of compartments/cells differs from the default number of nine. Detailed input instructions are provided as well as a description of outputs. Input files and selected output are presented for two sample problems run on both HP-735 and SUN SPARC workstations.« less

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

Samohyl, P.

The application of the LBB requires also fatigue flaw growth assessment. This analysis was performed for PWR nuclear power plants types VVER 440/230, VVER 440/213c, VVER 1000/320. Respecting that these NPP`s were designed according to Russian codes that differ from US codes it was needed to compare these approaches. Comparison with our experimental data was accomplished, too. Margins of applicability of the US methods and their modifications for the materials used for construction of Czech and Slovak NPP`s are shown. Computer code accomplishing the analysis according to described method is presented. Some measurement and calculations show that thermal stratifications inmore » horizontal pipelines can lead to additive loads that are not negligible and can be dangerous. An attempt to include these loads induced by steady-state stratification was made.« less

An integrated radiation physics computer code system.

NASA Technical Reports Server (NTRS)

Steyn, J. J.; Harris, D. W.

1972-01-01

An integrated computer code system for the semi-automatic and rapid analysis of experimental and analytic problems in gamma photon and fast neutron radiation physics is presented. Such problems as the design of optimum radiation shields and radioisotope power source configurations may be studied. The system codes allow for the unfolding of complex neutron and gamma photon experimental spectra. Monte Carlo and analytic techniques are used for the theoretical prediction of radiation transport. The system includes a multichannel pulse-height analyzer scintillation and semiconductor spectrometer coupled to an on-line digital computer with appropriate peripheral equipment. The system is geometry generalized as well as self-contained with respect to material nuclear cross sections and the determination of the spectrometer response functions. Input data may be either analytic or experimental.

An overview of the ENEA activities in the field of coupled codes NPP simulation

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

Parisi, C.; Negrenti, E.; Sepielli, M.

2012-07-01

In the framework of the nuclear research activities in the fields of safety, training and education, ENEA (the Italian National Agency for New Technologies, Energy and the Sustainable Development) is in charge of defining and pursuing all the necessary steps for the development of a NPP engineering simulator at the 'Casaccia' Research Center near Rome. A summary of the activities in the field of the nuclear power plants simulation by coupled codes is here presented with the long term strategy for the engineering simulator development. Specifically, results from the participation in international benchmarking activities like the OECD/NEA 'Kalinin-3' benchmark andmore » the 'AER-DYN-002' benchmark, together with simulations of relevant events like the Fukushima accident, are here reported. The ultimate goal of such activities performed using state-of-the-art technology is the re-establishment of top level competencies in the NPP simulation field in order to facilitate the development of Enhanced Engineering Simulators and to upgrade competencies for supporting national energy strategy decisions, the nuclear national safety authority, and the R and D activities on NPP designs. (authors)« less

Subgroup A : nuclear model codes report to the Sixteenth Meeting of the WPEC

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

Talou, P.; Chadwick, M. B.; Dietrich, F. S.

2004-01-01

The Subgroup A activities focus on the development of nuclear reaction models and codes, used in evaluation work for nuclear reactions from the unresolved energy region up to the pion threshold production limit, and for target nuclides from the low teens and heavier. Much of the efforts are devoted by each participant to the continuing development of their own Institution codes. Progresses in this arena are reported in detail for each code in the present document. EMPIRE-II is of public access. The release of the TALYS code has been announced for the ND2004 Conference in Santa Fe, NM, October 2004.more » McGNASH is still under development and is not expected to be released in the very near future. In addition, Subgroup A members have demonstrated a growing interest in working on common modeling and codes capabilities, which would significantly reduce the amount of duplicate work, help manage efficiently the growing lines of existing codes, and render codes inter-comparison much easier. A recent and important activity of the Subgroup A has therefore been to develop the framework and the first bricks of the ModLib library, which is constituted of mostly independent pieces of codes written in Fortran 90 (and above) to be used in existing and future nuclear reaction codes. Significant progresses in the development of ModLib have been made during the past year. Several physics modules have been added to the library, and a few more have been planned in detail for the coming year.« less

76 FR 17160 - Office of New Reactors; Final Interim Staff Guidance on the Review of Nuclear Power Plant Designs...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-28

...The NRC staff is issuing its Final Interim Staff Guidance (ISG) DC/COL-ISG-021 titled ``Interim Staff Guidance on the Review of Nuclear Power Plant Designs Using a Gas Turbine Driven Standby Emergency Alternating Current Power System,'' Agencywide Documents Access and Management System (ADAMS) Accession No. ML102510119 for DC/ COL-ISG-021 and ADAMS Accession No. ML102510164 for Attachment 1 to DC/ COL-ISG-021. This ISG provides new guidance for applicants submitting a combined license (COL) or design certification (DC) application for new nuclear power reactors under Title 10 of the Code of Federal Regulations, part 52. In addition, it supplements the guidance provided to the NRC staff in NUREG-0800, ``Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,'' March 2007, Standard Review Plan (SRP), Section 8.3.1 and Sections 9.5.4 through 9.5.8. The NRC staff issues DC/COL-ISGs to facilitate activities associated with NRC review of applications for DCs and COLs. The NRC staff intends to incorporate DC/COL-ISG-021 into the next revision of SRP Section 8.3.1 and Sections 9.5.4 through 9.5.8 and Regulatory Guide 1.206, ``Combined License Applications for Nuclear Power Plants (LWR Edition),'' June 2007. Disposition: On February 3, 2010, the NRC staff issued proposed DC/ COL-ISG-021 on ``Review of Nuclear Power Plant Designs Using a Gas Turbine Driven Standby Emergency Alternating Current Power System,'' ADAMS Accession No. ML092640035. The NRC staff received comments on the proposed guidance. This final issuance resolves the majority of the comments. The NRC staff responses to these comments can be found in ADAMS Accession No. ML102510176.

TRANSURANUS: a fuel rod analysis code ready for use

NASA Astrophysics Data System (ADS)

Lassmann, K.

1992-06-01

TRANSURANUS is a computer program for the thermal and mechanical analysis of fuel rods in nuclear reactors and was developed at the European Institute for Transuranium Elements (TUI). The TRANSURANUS code consists of a clearly defined mechanical-mathematical framework into which physical models can easily be incorporated. Besides its flexibility for different fuel rod designs the TRANSURANUS code can deal with very different situations, as given for instance in an experiment, under normal, off-normal and accident conditions. The time scale of the problems to be treated may range from milliseconds to years. The code has a comprehensive material data bank for oxide, mixed oxide, carbide and nitride fuels, Zircaloy and steel claddings and different coolants. During its development great effort was spent on obtaining an extremely flexible tool which is easy to handle, exhibiting very fast running times. The total development effort is approximately 40 man-years. In recent years the interest to use this code grew and the code is in use in several organisations, both research and private industry. The code is now available to all interested parties. The paper outlines the main features and capabilities of the TRANSURANUS code, its validation and treats also some practical aspects.

The Nuclear Energy Knowledge and Validation Center Summary of Activities Conducted in FY16

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

Gougar, Hans David

The Nuclear Energy Knowledge and Validation Center (NEKVaC) is a new initiative by the Department of Energy (DOE) and Idaho National Laboratory (INL) to coordinate and focus the resources and expertise that exist with the DOE toward solving issues in modern nuclear code validation and knowledge management. In time, code owners, users, and developers will view the NEKVaC as a partner and essential resource for acquiring the best practices and latest techniques for validating codes, providing guidance in planning and executing experiments, facilitating access to and maximizing the usefulness of existing data, and preserving knowledge for continual use by nuclearmore » professionals and organizations for their own validation needs. The scope of the NEKVaC covers many interrelated activities that will need to be cultivated carefully in the near term and managed properly once the NEKVaC is fully functional. Three areas comprise the principal mission: (1) identify and prioritize projects that extend the field of validation science and its application to modern codes, (2) develop and disseminate best practices and guidelines for high-fidelity multiphysics/multiscale analysis code development and associated experiment design, and (3) define protocols for data acquisition and knowledge preservation and provide a portal for access to databases currently scattered among numerous organizations. These mission areas, while each having a unique focus, are interdependent and complementary. Likewise, all activities supported by the NEKVaC, both near term and long term, must possess elements supporting all three areas. This cross cutting nature is essential to ensuring that activities and supporting personnel do not become “stove piped” (i.e., focused a specific function that the activity itself becomes the objective rather than achieving the larger vision). This report begins with a description of the mission areas; specifically, the role played by each major committee and the types of activities for which they are responsible. It then lists and describes the proposed near term tasks upon which future efforts can build.« less

Calculations of skyshine from an intense portable electron linac

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

Estes, G.P.; Hughes, H.G.; Fry, D.A.

1994-12-31

The MCNP Monte carlo code has been used at Los Alamos to calculate skyshine and terrain albedo efects from an intense portable electron linear accelerator that is to be used by the Russian Federation to radiograph nuclear weapons that may have been damaged by accidents. Relative dose rate profiles have been calculated. The design of the accelerator, along with a diagram, is presented.

Development of Innovative Design Processor

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

Park, Y.S.; Park, C.O.

2004-07-01

The nuclear design analysis requires time-consuming and erroneous model-input preparation, code run, output analysis and quality assurance process. To reduce human effort and improve design quality and productivity, Innovative Design Processor (IDP) is being developed. Two basic principles of IDP are the document-oriented design and the web-based design. The document-oriented design is that, if the designer writes a design document called active document and feeds it to a special program, the final document with complete analysis, table and plots is made automatically. The active documents can be written with ordinary HTML editors or created automatically on the web, which ismore » another framework of IDP. Using the proper mix-up of server side and client side programming under the LAMP (Linux/Apache/MySQL/PHP) environment, the design process on the web is modeled as a design wizard style so that even a novice designer makes the design document easily. This automation using the IDP is now being implemented for all the reload design of Korea Standard Nuclear Power Plant (KSNP) type PWRs. The introduction of this process will allow large reduction in all reload design efforts of KSNP and provide a platform for design and R and D tasks of KNFC. (authors)« less

Potential advantages associated with implementing a risk-based inspection program by a nuclear facility

NASA Astrophysics Data System (ADS)

McNeill, Alexander, III; Balkey, Kenneth R.

1995-05-01

The current inservice inspection activities at a U.S. nuclear facility are based upon the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI. The Code selects examination locations based upon a sampling criteria which includes component geometry, stress, and usage among other criteria. This can result in a significant number of required examinations. As a result of regulatory action each nuclear facility has conducted probabilistic risk assessments (PRA) or individual plant examinations (IPE), producing plant specific risk-based information. Several initiatives have been introduced to apply this new plant risk information. Among these initiatives is risk-based inservice inspection. A code case has been introduced for piping inspections based upon this new risk- based technology. This effort brought forward to the ASME Section XI Code committee, has been initiated and championed by the ASME Research Task Force on Risk-Based Inspection Guidelines -- LWR Nuclear Power Plant Application. Preliminary assessments associated with the code case have revealed that potential advantages exist in a risk-based inservice inspection program with regard to a number of exams, risk, personnel exposure, and cost.

System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

NASA Astrophysics Data System (ADS)

Porter, Ian Edward

A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several additional fuels will also be analyzed, including uranium nitride (UN), uranium carbide (UC) and uranium silicide (U3Si2). Focusing on the system response in an accident scenario, an emphasis is placed on the fracture mechanics of the ceramic cladding by design the fuel rods to eliminate pellet cladding mechanical interaction (PCMI). The time to failure and how much of the fuel in the reactor fails with an advanced fuel design will be analyzed and compared to the current UO2/Zircaloy design using a full scale reactor model.

Acceleration of MCNP calculations for small pipes configurations by using Weigth Windows Importance cards created by the SN-3D ATTILA

NASA Astrophysics Data System (ADS)

Castanier, Eric; Paterne, Loic; Louis, Céline

2017-09-01

In the nuclear engineering, you have to manage time and precision. Especially in shielding design, you have to be more accurate and efficient to reduce cost (shielding thickness optimization), and for this, you use 3D codes. In this paper, we want to see if we can easily applicate the CADIS methods for design shielding of small pipes which go through large concrete walls. We assess the impact of the WW generated by the 3D-deterministic code ATTILA versus WW directly generated by MCNP (iterative and manual process). The comparison is based on the quality of the convergence (estimated relative error (σ), Variance of Variance (VOV) and Figure of Merit (FOM)), on time (computer time + modelling) and on the implement for the engineer.

A proposed benchmark problem for cargo nuclear threat monitoring

NASA Astrophysics Data System (ADS)

Wesley Holmes, Thomas; Calderon, Adan; Peeples, Cody R.; Gardner, Robin P.

2011-10-01

There is currently a great deal of technical and political effort focused on reducing the risk of potential attacks on the United States involving radiological dispersal devices or nuclear weapons. This paper proposes a benchmark problem for gamma-ray and X-ray cargo monitoring with results calculated using MCNP5, v1.51. The primary goal is to provide a benchmark problem that will allow researchers in this area to evaluate Monte Carlo models for both speed and accuracy in both forward and inverse calculational codes and approaches for nuclear security applications. A previous benchmark problem was developed by one of the authors (RPG) for two similar oil well logging problems (Gardner and Verghese, 1991, [1]). One of those benchmarks has recently been used by at least two researchers in the nuclear threat area to evaluate the speed and accuracy of Monte Carlo codes combined with variance reduction techniques. This apparent need has prompted us to design this benchmark problem specifically for the nuclear threat researcher. This benchmark consists of conceptual design and preliminary calculational results using gamma-ray interactions on a system containing three thicknesses of three different shielding materials. A point source is placed inside the three materials lead, aluminum, and plywood. The first two materials are in right circular cylindrical form while the third is a cube. The entire system rests on a sufficiently thick lead base so as to reduce undesired scattering events. The configuration was arranged in such a manner that as gamma-ray moves from the source outward it first passes through the lead circular cylinder, then the aluminum circular cylinder, and finally the wooden cube before reaching the detector. A 2 in.×4 in.×16 in. box style NaI (Tl) detector was placed 1 m from the point source located in the center with the 4 in.×16 in. side facing the system. The two sources used in the benchmark are 137Cs and 235U.

Implicit time-integration method for simultaneous solution of a coupled non-linear system

NASA Astrophysics Data System (ADS)

Watson, Justin Kyle

Historically large physical problems have been divided into smaller problems based on the physics involved. This is no different in reactor safety analysis. The problem of analyzing a nuclear reactor for design basis accidents is performed by a handful of computer codes each solving a portion of the problem. The reactor thermal hydraulic response to an event is determined using a system code like TRAC RELAP Advanced Computational Engine (TRACE). The core power response to the same accident scenario is determined using a core physics code like Purdue Advanced Core Simulator (PARCS). Containment response to the reactor depressurization in a Loss Of Coolant Accident (LOCA) type event is calculated by a separate code. Sub-channel analysis is performed with yet another computer code. This is just a sample of the computer codes used to solve the overall problems of nuclear reactor design basis accidents. Traditionally each of these codes operates independently from each other using only the global results from one calculation as boundary conditions to another. Industry's drive to uprate power for reactors has motivated analysts to move from a conservative approach to design basis accident towards a best estimate method. To achieve a best estimate calculation efforts have been aimed at coupling the individual physics models to improve the accuracy of the analysis and reduce margins. The current coupling techniques are sequential in nature. During a calculation time-step data is passed between the two codes. The individual codes solve their portion of the calculation and converge to a solution before the calculation is allowed to proceed to the next time-step. This thesis presents a fully implicit method of simultaneous solving the neutron balance equations, heat conduction equations and the constitutive fluid dynamics equations. It discusses the problems involved in coupling different physics phenomena within multi-physics codes and presents a solution to these problems. The thesis also outlines the basic concepts behind the nodal balance equations, heat transfer equations and the thermal hydraulic equations, which will be coupled to form a fully implicit nonlinear system of equations. The coupling of separate physics models to solve a larger problem and improve accuracy and efficiency of a calculation is not a new idea, however implementing them in an implicit manner and solving the system simultaneously is. Also the application to reactor safety codes is new and has not be done with thermal hydraulics and neutronics codes on realistic applications in the past. The coupling technique described in this thesis is applicable to other similar coupled thermal hydraulic and core physics reactor safety codes. This technique is demonstrated using coupled input decks to show that the system is solved correctly and then verified by using two derivative test problems based on international benchmark problems the OECD/NRC Three mile Island (TMI) Main Steam Line Break (MSLB) problem (representative of pressurized water reactor analysis) and the OECD/NRC Peach Bottom (PB) Turbine Trip (TT) benchmark (representative of boiling water reactor analysis).

Scaling Studies for Advanced High Temperature Reactor Concepts, Final Technical Report: October 2014—December 2017

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

Woods, Brian; Gutowska, Izabela; Chiger, Howard

Computer simulations of nuclear reactor thermal-hydraulic phenomena are often used in the design and licensing of nuclear reactor systems. In order to assess the accuracy of these computer simulations, computer codes and methods are often validated against experimental data. This experimental data must be of sufficiently high quality in order to conduct a robust validation exercise. In addition, this experimental data is generally collected at experimental facilities that are of a smaller scale than the reactor systems that are being simulated due to cost considerations. Therefore, smaller scale test facilities must be designed and constructed in such a fashion tomore » ensure that the prototypical behavior of a particular nuclear reactor system is preserved. The work completed through this project has resulted in scaling analyses and conceptual design development for a test facility capable of collecting code validation data for the following high temperature gas reactor systems and events— 1. Passive natural circulation core cooling system, 2. pebble bed gas reactor concept, 3. General Atomics Energy Multiplier Module reactor, and 4. prismatic block design steam-water ingress event. In the event that code validation data for these systems or events is needed in the future, significant progress in the design of an appropriate integral-type test facility has already been completed as a result of this project. Where applicable, the next step would be to begin the detailed design development and material procurement. As part of this project applicable scaling analyses were completed and test facility design requirements developed. Conceptual designs were developed for the implementation of these design requirements at the Oregon State University (OSU) High Temperature Test Facility (HTTF). The original HTTF is based on a ¼-scale model of a high temperature gas reactor concept with the capability for both forced and natural circulation flow through a prismatic core with an electrical heat source. The peak core region temperature capability is 1400°C. As part of this project, an inventory of test facilities that could be used for these experimental programs was completed. Several of these facilities showed some promise, however, upon further investigation it became clear that only the OSU HTTF had the power and/or peak temperature limits that would allow for the experimental programs envisioned herein. Thus the conceptual design and feasibility study development focused on examining the feasibility of configuring the current HTTF to collect validation data for these experimental programs. In addition to the scaling analyses and conceptual design development, a test plan was developed for the envisioned modified test facility. This test plan included a discussion on an appropriate shakedown test program as well as the specific matrix tests. Finally, a feasibility study was completed to determine the cost and schedule considerations that would be important to any test program developed to investigate these designs and events.« less

Requirements for construction of nuclear system components at elevated temperatures (supplement to ASME Code Cases 1592, 1593, 1594, 1595, and 1596)

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

Not Available

This standard provides rules for the construction of Class 1 nuclear components, parts, and appurtenances for use at elevated temperatures. This standard is a complete set of requirements only when used in conjunction with Section III of the ASME Boiler and Pressure Vessel Code (ASME Code) and addenda, ASME Code Cases 1592, 1593, 1594, 1595, and 1596, and RDT E 15-2NB. Unmodified paragraphs of the referenced Code Cases are not repeated in this standard but are a part of the requirements of this standard.

EMPIRE: A Reaction Model Code for Nuclear Astrophysics

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

Palumbo, A., E-mail: apalumbo@bnl.gov; Herman, M.; Capote, R.

The correct modeling of abundances requires knowledge of nuclear cross sections for a variety of neutron, charged particle and γ induced reactions. These involve targets far from stability and are therefore difficult (or currently impossible) to measure. Nuclear reaction theory provides the only way to estimate values of such cross sections. In this paper we present application of the EMPIRE reaction code to nuclear astrophysics. Recent measurements are compared to the calculated cross sections showing consistent agreement for n-, p- and α-induced reactions of strophysical relevance.

Study of components and statistical reaction mechanism in simulation of nuclear process for optimized production of {sup 64}Cu and {sup 67}Ga medical radioisotopes using TALYS, EMPIRE and LISE++ nuclear reaction and evaporation codes

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

Nasrabadi, M. N., E-mail: mnnasrabadi@ast.ui.ac.ir; Sepiani, M.

2015-03-30

Production of medical radioisotopes is one of the most important tasks in the field of nuclear technology. These radioactive isotopes are mainly produced through variety nuclear process. In this research, excitation functions and nuclear reaction mechanisms are studied for simulation of production of these radioisotopes in the TALYS, EMPIRE and LISE++ reaction codes, then parameters and different models of nuclear level density as one of the most important components in statistical reaction models are adjusted for optimum production of desired radioactive yields.

Study of components and statistical reaction mechanism in simulation of nuclear process for optimized production of 64Cu and 67Ga medical radioisotopes using TALYS, EMPIRE and LISE++ nuclear reaction and evaporation codes

NASA Astrophysics Data System (ADS)

Nasrabadi, M. N.; Sepiani, M.

2015-03-01

Production of medical radioisotopes is one of the most important tasks in the field of nuclear technology. These radioactive isotopes are mainly produced through variety nuclear process. In this research, excitation functions and nuclear reaction mechanisms are studied for simulation of production of these radioisotopes in the TALYS, EMPIRE & LISE++ reaction codes, then parameters and different models of nuclear level density as one of the most important components in statistical reaction models are adjusted for optimum production of desired radioactive yields.

Seismic assessment of Technical Area V (TA-V).

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

Medrano, Carlos S.

The Technical Area V (TA-V) Seismic Assessment Report was commissioned as part of Sandia National Laboratories (SNL) Self Assessment Requirement per DOE O 414.1, Quality Assurance, for seismic impact on existing facilities at Technical Area-V (TA-V). SNL TA-V facilities are located on an existing Uniform Building Code (UBC) Seismic Zone IIB Site within the physical boundary of the Kirtland Air Force Base (KAFB). The document delineates a summary of the existing facilities with their safety-significant structure, system and components, identifies DOE Guidance, conceptual framework, past assessments and the present Geological and Seismic conditions. Building upon the past information and themore » evolution of the new seismic design criteria, the document discusses the potential impact of the new standards and provides recommendations based upon the current International Building Code (IBC) per DOE O 420.1B, Facility Safety and DOE G 420.1-2, Guide for the Mitigation of Natural Phenomena Hazards for DOE Nuclear Facilities and Non-Nuclear Facilities.« less

Capabilities, methodologies, and use of the cambio file-translation application.

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

Lasche, George P.

2007-03-01

This report describes the capabilities, methodologies, and uses of the Cambio computer application, designed to automatically read and display nuclear spectral data files of any known format in the world and to convert spectral data to one of several commonly used analysis formats. To further assist responders, Cambio incorporates an analysis method based on non-linear fitting techniques found in open literature and implemented in openly published source code in the late 1980s. A brief description is provided of how Cambio works, of what basic formats it can currently read, and how it can be used. Cambio was developed at Sandiamore » National Laboratories and is provided as a free service to assist nuclear emergency response analysts anywhere in the world in the fight against nuclear terrorism.« less

Recent Progress in the Development of a Multi-Layer Green's Function Code for Ion Beam Transport

NASA Technical Reports Server (NTRS)

Tweed, John; Walker, Steven A.; Wilson, John W.; Tripathi, Ram K.

2008-01-01

To meet the challenge of future deep space programs, an accurate and efficient engineering code for analyzing the shielding requirements against high-energy galactic heavy radiation is needed. To address this need, a new Green's function code capable of simulating high charge and energy ions with either laboratory or space boundary conditions is currently under development. The computational model consists of combinations of physical perturbation expansions based on the scales of atomic interaction, multiple scattering, and nuclear reactive processes with use of the Neumann-asymptotic expansions with non-perturbative corrections. The code contains energy loss due to straggling, nuclear attenuation, nuclear fragmentation with energy dispersion and downshifts. Previous reports show that the new code accurately models the transport of ion beams through a single slab of material. Current research efforts are focused on enabling the code to handle multiple layers of material and the present paper reports on progress made towards that end.

The NJOY Nuclear Data Processing System, Version 2016

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

Macfarlane, Robert; Muir, Douglas W.; Boicourt, R. M.

The NJOY Nuclear Data Processing System, version 2016, is a comprehensive computer code package for producing pointwise and multigroup cross sections and related quantities from evaluated nuclear data in the ENDF-4 through ENDF-6 legacy card-image formats. NJOY works with evaluated files for incident neutrons, photons, and charged particles, producing libraries for a wide variety of particle transport and reactor analysis codes.

OWL: A code for the two-center shell model with spherical Woods-Saxon potentials

NASA Astrophysics Data System (ADS)

Diaz-Torres, Alexis

2018-03-01

A Fortran-90 code for solving the two-center nuclear shell model problem is presented. The model is based on two spherical Woods-Saxon potentials and the potential separable expansion method. It describes the single-particle motion in low-energy nuclear collisions, and is useful for characterizing a broad range of phenomena from fusion to nuclear molecular structures.

Progress on China nuclear data processing code system

NASA Astrophysics Data System (ADS)

Liu, Ping; Wu, Xiaofei; Ge, Zhigang; Li, Songyang; Wu, Haicheng; Wen, Lili; Wang, Wenming; Zhang, Huanyu

2017-09-01

China is developing the nuclear data processing code Ruler, which can be used for producing multi-group cross sections and related quantities from evaluated nuclear data in the ENDF format [1]. The Ruler includes modules for reconstructing cross sections in all energy range, generating Doppler-broadened cross sections for given temperature, producing effective self-shielded cross sections in unresolved energy range, calculating scattering cross sections in thermal energy range, generating group cross sections and matrices, preparing WIMS-D format data files for the reactor physics code WIMS-D [2]. Programming language of the Ruler is Fortran-90. The Ruler is tested for 32-bit computers with Windows-XP and Linux operating systems. The verification of Ruler has been performed by comparison with calculation results obtained by the NJOY99 [3] processing code. The validation of Ruler has been performed by using WIMSD5B code.

Magnet Design Considerations for Fusion Nuclear Science Facility

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

Zhai, Y.; Kessel, C.; El-Guebaly, L.

2016-06-01

The Fusion Nuclear Science Facility (FNSF) is a nuclear confinement facility that provides a fusion environment with components of the reactor integrated together to bridge the technical gaps of burning plasma and nuclear science between the International Thermonuclear Experimental Reactor (ITER) and the demonstration power plant (DEMO). Compared with ITER, the FNSF is smaller in size but generates much higher magnetic field, i.e., 30 times higher neutron fluence with three orders of magnitude longer plasma operation at higher operating temperatures for structures surrounding the plasma. Input parameters to the magnet design from system code analysis include magnetic field of 7.5more » T at the plasma center with a plasma major radius of 4.8 m and a minor radius of 1.2 m and a peak field of 15.5 T on the toroidal field (TF) coils for the FNSF. Both low-temperature superconductors (LTS) and high-temperature superconductors (HTS) are considered for the FNSF magnet design based on the state-of-the-art fusion magnet technology. The higher magnetic field can be achieved by using the high-performance ternary restacked-rod process Nb3Sn strands for TF magnets. The circular cable-in-conduit conductor (CICC) design similar to ITER magnets and a high-aspect-ratio rectangular CICC design are evaluated for FNSF magnets, but low-activation-jacket materials may need to be selected. The conductor design concept and TF coil winding pack composition and dimension based on the horizontal maintenance schemes are discussed. Neutron radiation limits for the LTS and HTS superconductors and electrical insulation materials are also reviewed based on the available materials previously tested. The material radiation limits for FNSF magnets are defined as part of the conceptual design studies for FNSF magnets.« less

Magnet design considerations for Fusion Nuclear Science Facility

DOE PAGES

Zhai, Yuhu; Kessel, Chuck; El-guebaly, Laila; ...

2016-02-25

The Fusion Nuclear Science Facility (FNSF) is a nuclear confinement facility to provide a fusion environment with components of the reactor integrated together to bridge the technical gaps of burning plasma and nuclear science between ITER and the demonstration power plant (DEMO). Compared to ITER, the FNSF is smaller in size but generates much higher magnetic field, 30 times higher neutron fluence with 3 orders of magnitude longer plasma operation at higher operating temperatures for structures surrounding the plasma. Input parameters to the magnet design from system code analysis include magnetic field of 7.5 T at the plasma center withmore » plasma major radius of 4.8 m and minor radius of 1.2 m, and a peak field of 15.5 T on the TF coils for FNSF. Both low temperature superconductor (LTS) and high temperature superconductor (HTS) are considered for the FNSF magnet design based on the state-of-the-art fusion magnet technology. The higher magnetic field can be achieved by using the high performance ternary Restack Rod Process (RRP) Nb3Sn strands for toroidal field (TF) magnets. The circular cable-in-conduit conductor (CICC) design similar to ITER magnets and a high aspect ratio rectangular CICC design are evaluated for FNSF magnets but low activation jacket materials may need to be selected. The conductor design concept and TF coil winding pack composition and dimension based on the horizontal maintenance schemes are discussed. Neutron radiation limits for the LTS and HTS superconductors and electrical insulation materials are also reviewed based on the available materials previously tested. As a result, the material radiation limits for FNSF magnets are defined as part of the conceptual design studies for FNSF magnets.« 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.

National Nuclear Data Center

Science.gov Websites

reaction data Sigma Retrieval & Plotting Nuclear structure & decay Data Nuclear Science References Experimental Unevaluated Nuclear Data List Evaluated Nuclear Structure Data File NNDC databases Ground and isomeric states properties Nuclear structure & decay data journal Nuclear reaction model code Tools and

Comprehensive Report For Proposed Elevated Temperature Elastic Perfectly Plastic (EPP) Code Cases Representative Example Problems

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

Hollinger, Greg L.

Background: The current rules in the nuclear section of the ASME Boiler and Pressure Vessel (B&PV) Code , Section III, Subsection NH for the evaluation of strain limits and creep-fatigue damage using simplified methods based on elastic analysis have been deemed inappropriate for Alloy 617 at temperatures above 1200F (650C)1. To address this issue, proposed code rules have been developed which are based on the use of elastic-perfectly plastic (E-PP) analysis methods and which are expected to be applicable to very high temperatures. The proposed rules for strain limits and creep-fatigue evaluation were initially documented in the technical literature 2,more » 3, and have been recently revised to incorporate comments and simplify their application. The revised code cases have been developed. Task Objectives: The goal of the Sample Problem task is to exercise these code cases through example problems to demonstrate their feasibility and, also, to identify potential corrections and improvements should problems be encountered. This will provide input to the development of technical background documents for consideration by the applicable B&PV committees considering these code cases for approval. This task has been performed by Hollinger and Pease of Becht Engineering Co., Inc., Nuclear Services Division and a report detailing the results of the E-PP analyses conducted on example problems per the procedures of the E-PP strain limits and creep-fatigue draft code cases is enclosed as Enclosure 1. Conclusions: The feasibility of the application of the E-PP code cases has been demonstrated through example problems that consist of realistic geometry (a nozzle attached to a semi-hemispheric shell with a circumferential weld) and load (pressure; pipe reaction load applied at the end of the nozzle, including axial and shear forces, bending and torsional moments; through-wall transient temperature gradient) and design and operating conditions (Levels A, B and C).« less

CESAR5.3: Isotopic depletion for Research and Testing Reactor decommissioning

NASA Astrophysics Data System (ADS)

Ritter, Guillaume; Eschbach, Romain; Girieud, Richard; Soulard, Maxime

2018-05-01

CESAR stands in French for "simplified depletion applied to reprocessing". The current version is now number 5.3 as it started 30 years ago from a long lasting cooperation with ORANO, co-owner of the code with CEA. This computer code can characterize several types of nuclear fuel assemblies, from the most regular PWR power plants to the most unexpected gas cooled and graphite moderated old timer research facility. Each type of fuel can also include numerous ranges of compositions like UOX, MOX, LEU or HEU. Such versatility comes from a broad catalog of cross section libraries, each corresponding to a specific reactor and fuel matrix design. CESAR goes beyond fuel characterization and can also provide an evaluation of structural materials activation. The cross-sections libraries are generated using the most refined assembly or core level transport code calculation schemes (CEA APOLLO2 or ERANOS), based on the European JEFF3.1.1 nuclear data base. Each new CESAR self shielded cross section library benefits all most recent CEA recommendations as for deterministic physics options. Resulting cross sections are organized as a function of burn up and initial fuel enrichment which allows to condensate this costly process into a series of Legendre polynomials. The final outcome is a fast, accurate and compact CESAR cross section library. Each library is fully validated, against a stochastic transport code (CEA TRIPOLI 4) if needed and against a reference depletion code (CEA DARWIN). Using CESAR does not require any of the neutron physics expertise implemented into cross section libraries generation. It is based on top quality nuclear data (JEFF3.1.1 for ˜400 isotopes) and includes up to date Bateman equation solving algorithms. However, defining a CESAR computation case can be very straightforward. Most results are only 3 steps away from any beginner's ambition: Initial composition, in core depletion and pool decay scenario. On top of a simple utilization architecture, CESAR includes a portable Graphical User Interface which can be broadly deployed in R&D or industrial facilities. Aging facilities currently face decommissioning and dismantling issues. This way to the end of the nuclear fuel cycle requires a careful assessment of source terms in the fuel, core structures and all parts of a facility that must be disposed of with "industrial nuclear" constraints. In that perspective, several CESAR cross section libraries were constructed for early CEA Research and Testing Reactors (RTR's). The aim of this paper is to describe how CESAR operates and how it can be used to help these facilities care for waste disposal, nuclear materials transport or basic safety cases. The test case will be based on the PHEBUS Facility located at CEA - Cadarache.

Assuring Structural Integrity in Army Systems

DTIC Science & Technology

1985-02-28

power plants are* I. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code , Section III - Rules for Construction of Nuclear...Power Plant Components; 2. ASNE Boiler and Pressure Vessel Code , Section XI, Rules for In-Service Inspection of Nuclear Power Plant Components; and 3

Scoping Calculations of Power Sources for Nuclear Electric Propulsion

NASA Technical Reports Server (NTRS)

Difilippo, F. C.

1994-01-01

This technical memorandum describes models and calculational procedures to fully characterize the nuclear island of power sources for nuclear electric propulsion. Two computer codes were written: one for the gas-cooled NERVA derivative reactor and the other for liquid metal-cooled fuel pin reactors. These codes are going to be interfaced by NASA with the balance of plant in order to make scoping calculations for mission analysis.

Guide to Using Sierra

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

Shaw, Ryan Phillip; Agelastos, Anthony Michael; Miller, Joel D.

2015-03-01

Sierra is an engineering mechanics simulation code suite supporting the Nation's Nuclear Weapons mission as well as other customers. It has explicit ties to Sandia National Labs' workfow, including geometry and meshing, design and optimization, and visualization. Dis- tinguishing strengths include "application aware" development, scalability, SQA and V&V, multiple scales, and multi-physics coupling. This document is intended to help new and existing users of Sierra as a user manual and troubleshooting guide.

Determination of neutron flux distribution in an Am-Be irradiator using the MCNP.

PubMed

Shtejer-Diaz, K; Zamboni, C B; Zahn, G S; Zevallos-Chávez, J Y

2003-10-01

A neutron irradiator has been assembled at IPEN facilities to perform qualitative-quantitative analysis of many materials using thermal and fast neutrons outside the nuclear reactor premises. To establish the prototype specifications, the neutron flux distribution and the absorbed dose rates were calculated using the MCNP computer code. These theoretical predictions then allow one to discuss the optimum irradiator design and its performance.

Guide to Using Sierra

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

Shaw, Ryan Phillip; Agelastos, Anthony Michael; Miller, Joel D.

2017-04-01

Sierra is an engineering mechanics simulation code suite supporting the Nation's Nuclear Weapons mission as well as other customers. It has explicit ties to Sandia National Labs' workfow, including geometry and meshing, design and optimization, and visualization. Dis- tinguishing strengths include "application aware" development, scalability, SQA and V&V, multiple scales, and multi-physics coupling. This document is intended to help new and existing users of Sierra as a user manual and troubleshooting guide.

Verification and validation of RADMODL Version 1.0

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

Kimball, K.D.

1993-03-01

RADMODL is a system of linked computer codes designed to calculate the radiation environment following an accident in which nuclear materials are released. The RADMODL code and the corresponding Verification and Validation (V&V) calculations (Appendix A), were developed for Westinghouse Savannah River Company (WSRC) by EGS Corporation (EGS). Each module of RADMODL is an independent code and was verified separately. The full system was validated by comparing the output of the various modules with the corresponding output of a previously verified version of the modules. The results of the verification and validation tests show that RADMODL correctly calculates the transportmore » of radionuclides and radiation doses. As a result of this verification and validation effort, RADMODL Version 1.0 is certified for use in calculating the radiation environment following an accident.« 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

Square lattice honeycomb reactor for space power and propulsion

NASA Astrophysics Data System (ADS)

Gouw, Reza; Anghaie, Samim

2000-01-01

The most recent nuclear design study at the Innovative Nuclear Space Power and Propulsion Institute (INSPI) is the Moderated Square-Lattice Honeycomb (M-SLHC) reactor design utilizing the solid solution of ternary carbide fuels. The reactor is fueled with solid solution of 93% enriched (U,Zr,Nb)C. The square-lattice honeycomb design provides high strength and is amenable to the processing complexities of these ultrahigh temperature fuels. The optimum core configuration requires a balance between high specific impulse and thrust level performance, and maintaining the temperature and strength limits of the fuel. The M-SLHC design is based on a cylindrical core that has critical radius and length of 37 cm and 50 cm, respectively. This design utilized zirconium hydrate to act as moderator. The fuel sub-assemblies are designed as cylindrical tubes with 12 cm in diameter and 10 cm in length. Five fuel subassemblies are stacked up axially to form one complete fuel assembly. These fuel assemblies are then arranged in the circular arrangement to form two fuel regions. The first fuel region consists of six fuel assemblies, and 18 fuel assemblies for the second fuel region. A 10-cm radial beryllium reflector in addition to 10-cm top axial beryllium reflector is used to reduce neutron leakage from the system. To perform nuclear design analysis of the M-SLHC design, a series of neutron transport and diffusion codes are used. To optimize the system design, five axial regions are specified. In each axial region, temperature and fuel density are varied. The axial and radial power distributions for the system are calculated, as well as the axial and radial flux distributions. Temperature coefficients of the system are also calculated. A water submersion accident scenario is also analyzed for these systems. Results of the nuclear design analysis indicate that a compact core can be designed based on ternary uranium carbide square-lattice honeycomb fuel, which provides a relatively high thrust to weight ratio. .

Transmutation Fuel Performance Code Thermal Model Verification

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

Gregory K. Miller; Pavel G. Medvedev

2007-09-01

FRAPCON fuel performance code is being modified to be able to model performance of the nuclear fuels of interest to the Global Nuclear Energy Partnership (GNEP). The present report documents the effort for verification of the FRAPCON thermal model. It was found that, with minor modifications, FRAPCON thermal model temperature calculation agrees with that of the commercial software ABAQUS (Version 6.4-4). This report outlines the methodology of the verification, code input, and calculation results.

Uncertainty analysis on reactivity and discharged inventory for a pressurized water reactor fuel assembly due to {sup 235,238}U nuclear data uncertainties

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

Da Cruz, D. F.; Rochman, D.; Koning, A. J.

2012-07-01

This paper discusses the uncertainty analysis on reactivity and inventory for a typical PWR fuel element as a result of uncertainties in {sup 235,238}U nuclear data. A typical Westinghouse 3-loop fuel assembly fuelled with UO{sub 2} fuel with 4.8% enrichment has been selected. The Total Monte-Carlo method has been applied using the deterministic transport code DRAGON. This code allows the generation of the few-groups nuclear data libraries by directly using data contained in the nuclear data evaluation files. The nuclear data used in this study is from the JEFF3.1 evaluation, and the nuclear data files for {sup 238}U and {supmore » 235}U (randomized for the generation of the various DRAGON libraries) are taken from the nuclear data library TENDL. The total uncertainty (obtained by randomizing all {sup 238}U and {sup 235}U nuclear data in the ENDF files) on the reactor parameters has been split into different components (different nuclear reaction channels). Results show that the TMC method in combination with a deterministic transport code constitutes a powerful tool for performing uncertainty and sensitivity analysis of reactor physics parameters. (authors)« less

Improvement of Modeling HTGR Neutron Physics by Uncertainty Analysis with the Use of Cross-Section Covariance Information

NASA Astrophysics Data System (ADS)

Boyarinov, V. F.; Grol, A. V.; Fomichenko, P. A.; Ternovykh, M. Yu

2017-01-01

This work is aimed at improvement of HTGR neutron physics design calculations by application of uncertainty analysis with the use of cross-section covariance information. Methodology and codes for preparation of multigroup libraries of covariance information for individual isotopes from the basic 44-group library of SCALE-6 code system were developed. A 69-group library of covariance information in a special format for main isotopes and elements typical for high temperature gas cooled reactors (HTGR) was generated. This library can be used for estimation of uncertainties, associated with nuclear data, in analysis of HTGR neutron physics with design codes. As an example, calculations of one-group cross-section uncertainties for fission and capture reactions for main isotopes of the MHTGR-350 benchmark, as well as uncertainties of the multiplication factor (k∞) for the MHTGR-350 fuel compact cell model and fuel block model were performed. These uncertainties were estimated by the developed technology with the use of WIMS-D code and modules of SCALE-6 code system, namely, by TSUNAMI, KENO-VI and SAMS. Eight most important reactions on isotopes for MHTGR-350 benchmark were identified, namely: 10B(capt), 238U(n,γ), ν5, 235U(n,γ), 238U(el), natC(el), 235U(fiss)-235U(n,γ), 235U(fiss).

INL Results for Phases I and III of the OECD/NEA MHTGR-350 Benchmark

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

Gerhard Strydom; Javier Ortensi; Sonat Sen

2013-09-01

The Idaho National Laboratory (INL) Very High Temperature Reactor (VHTR) Technology Development Office (TDO) Methods Core Simulation group led the construction of the Organization for Economic Cooperation and Development (OECD) Modular High Temperature Reactor (MHTGR) 350 MW benchmark for comparing and evaluating prismatic VHTR analysis codes. The benchmark is sponsored by the OECD's Nuclear Energy Agency (NEA), and the project will yield a set of reference steady-state, transient, and lattice depletion problems that can be used by the Department of Energy (DOE), the Nuclear Regulatory Commission (NRC), and vendors to assess their code suits. The Methods group is responsible formore » defining the benchmark specifications, leading the data collection and comparison activities, and chairing the annual technical workshops. This report summarizes the latest INL results for Phase I (steady state) and Phase III (lattice depletion) of the benchmark. The INSTANT, Pronghorn and RattleSnake codes were used for the standalone core neutronics modeling of Exercise 1, and the results obtained from these codes are compared in Section 4. Exercise 2 of Phase I requires the standalone steady-state thermal fluids modeling of the MHTGR-350 design, and the results for the systems code RELAP5-3D are discussed in Section 5. The coupled neutronics and thermal fluids steady-state solution for Exercise 3 are reported in Section 6, utilizing the newly developed Parallel and Highly Innovative Simulation for INL Code System (PHISICS)/RELAP5-3D code suit. Finally, the lattice depletion models and results obtained for Phase III are compared in Section 7. The MHTGR-350 benchmark proved to be a challenging simulation set of problems to model accurately, and even with the simplifications introduced in the benchmark specification this activity is an important step in the code-to-code verification of modern prismatic VHTR codes. A final OECD/NEA comparison report will compare the Phase I and III results of all other international participants in 2014, while the remaining Phase II transient case results will be reported in 2015.« less

RELAP5-3D Resolution of Known Restart/Backup Issues

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

Mesina, George L.; Anderson, Nolan A.

2014-12-01

The state-of-the-art nuclear reactor system safety analysis computer program developed at the Idaho National Laboratory (INL), RELAP5-3D, continues to adapt to changes in computer hardware and software and to develop to meet the ever-expanding needs of the nuclear industry. To continue at the forefront, code testing must evolve with both code and industry developments, and it must work correctly. To best ensure this, the processes of Software Verification and Validation (V&V) are applied. Verification compares coding against its documented algorithms and equations and compares its calculations against analytical solutions and the method of manufactured solutions. A form of this, sequentialmore » verification, checks code specifications against coding only when originally written then applies regression testing which compares code calculations between consecutive updates or versions on a set of test cases to check that the performance does not change. A sequential verification testing system was specially constructed for RELAP5-3D to both detect errors with extreme accuracy and cover all nuclear-plant-relevant code features. Detection is provided through a “verification file” that records double precision sums of key variables. Coverage is provided by a test suite of input decks that exercise code features and capabilities necessary to model a nuclear power plant. A matrix of test features and short-running cases that exercise them is presented. This testing system is used to test base cases (called null testing) as well as restart and backup cases. It can test RELAP5-3D performance in both standalone and coupled (through PVM to other codes) runs. Application of verification testing revealed numerous restart and backup issues in both standalone and couple modes. This document reports the resolution of these issues.« less

Extreme Scale Computing to Secure the Nation

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

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

2009-11-10

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

Recent Improvements of Particle and Heavy Ion Transport code System: PHITS

NASA Astrophysics Data System (ADS)

Sato, Tatsuhiko; Niita, Koji; Iwamoto, Yosuke; Hashimoto, Shintaro; Ogawa, Tatsuhiko; Furuta, Takuya; Abe, Shin-ichiro; Kai, Takeshi; Matsuda, Norihiro; Okumura, Keisuke; Kai, Tetsuya; Iwase, Hiroshi; Sihver, Lembit

2017-09-01

The Particle and Heavy Ion Transport code System, PHITS, has been developed under the collaboration of several research institutes in Japan and Europe. This system can simulate the transport of most particles with energy levels up to 1 TeV (per nucleon for ion) using different nuclear reaction models and data libraries. More than 2,500 registered researchers and technicians have used this system for various applications such as accelerator design, radiation shielding and protection, medical physics, and space- and geo-sciences. This paper summarizes the physics models and functions recently implemented in PHITS, between versions 2.52 and 2.88, especially those related to source generation useful for simulating brachytherapy and internal exposures of radioisotopes.

15 CFR Appendix B to Part 30 - AES Filing Codes

Code of Federal Regulations, 2014 CFR

2014-01-01

... charity FS—Foreign Military Sales ZD—North American Free Trade Agreements (NAFTA) duty deferral shipments...—Validated End User Authorization C58CCD—Consumer Communication Devices C59STA—Strategic Trade Authorization Department of Energy/National Nuclear Security Administration (DOE/NNSA) Codes E01—DOE/NNSA Nuclear...

Reactivity effects in VVER-1000 of the third unit of the kalinin nuclear power plant at physical start-up. Computations in ShIPR intellectual code system with library of two-group cross sections generated by UNK code

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

Zizin, M. N.; Zimin, V. G.; Zizina, S. N., E-mail: zizin@adis.vver.kiae.ru

2010-12-15

The ShIPR intellectual code system for mathematical simulation of nuclear reactors includes a set of computing modules implementing the preparation of macro cross sections on the basis of the two-group library of neutron-physics cross sections obtained for the SKETCH-N nodal code. This library is created by using the UNK code for 3D diffusion computation of first VVER-1000 fuel loadings. Computation of neutron fields in the ShIPR system is performed using the DP3 code in the two-group diffusion approximation in 3D triangular geometry. The efficiency of all groups of control rods for the first fuel loading of the third unit ofmore » the Kalinin Nuclear Power Plant is computed. The temperature, barometric, and density effects of reactivity as well as the reactivity coefficient due to the concentration of boric acid in the reactor were computed additionally. Results of computations are compared with the experiment.« less

Reactivity effects in VVER-1000 of the third unit of the kalinin nuclear power plant at physical start-up. Computations in ShIPR intellectual code system with library of two-group cross sections generated by UNK code

NASA Astrophysics Data System (ADS)

Zizin, M. N.; Zimin, V. G.; Zizina, S. N.; Kryakvin, L. V.; Pitilimov, V. A.; Tereshonok, V. A.

2010-12-01

The ShIPR intellectual code system for mathematical simulation of nuclear reactors includes a set of computing modules implementing the preparation of macro cross sections on the basis of the two-group library of neutron-physics cross sections obtained for the SKETCH-N nodal code. This library is created by using the UNK code for 3D diffusion computation of first VVER-1000 fuel loadings. Computation of neutron fields in the ShIPR system is performed using the DP3 code in the two-group diffusion approximation in 3D triangular geometry. The efficiency of all groups of control rods for the first fuel loading of the third unit of the Kalinin Nuclear Power Plant is computed. The temperature, barometric, and density effects of reactivity as well as the reactivity coefficient due to the concentration of boric acid in the reactor were computed additionally. Results of computations are compared with the experiment.

Effect of Light Water Reactor Water Environments on the Fatigue Life of Reactor Materials

DOE PAGES

Chopra, O. K.; Stevens, G. L.; Tregoning, R.; ...

2017-10-06

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) provides rules for the design of Class 1 components of nuclear power plants. Figures I-9.1 through I-9.6 of Appendix I to Section III of the Code specify fatigue design curves for applicable structural materials. However, the Code design curves do not explicitly address the effects of light water reactor (LWR) water environments. Existing fatigue strain-vs.-life (ε-N) laboratory data illustrate potentially significant effects of LWR water environments on the fatigue resistance of pressure vessel and piping steels. Extensive studies have been conducted at Argonne National Laboratory and elsewheremore » since 1990 to investigate the effects of LWR environments on the fatigue life of piping and pressure vessel steels. This article summarizes the results of these studies. Existing fatigue ε-N data were evaluated to identify the various material, environmental, and loading conditions that influence fatigue crack initiation; a methodology for estimating fatigue lives as a function of these parameters was developed. The effects were incorporated into the ASME Code Section III fatigue evaluations in terms of an environmental correction factor, F en, which is defined as the ratio of fatigue life in air at room temperature to the fatigue life in the LWR water environment at reactor operating temperatures. Available fatigue data were used to develop fatigue design curves for carbon and low-alloy steels, austenitic stainless steels, and nickel-chromium-iron (NiCr-Fe) alloys and their weld metals in air at room temperature. A review of the Code Section III fatigue adjustment factors of 2 on strain and 20 on life is also presented and the possible conservatism inherent in the choice of these adjustment factors is evaluated. A brief description of potential effects of neutron irradiation on fatigue crack initiation for these structural materials is also presented.« less

Effect of Light Water Reactor Water Environments on the Fatigue Life of Reactor Materials

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

Chopra, O. K.; Stevens, G. L.; Tregoning, R.

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) provides rules for the design of Class 1 components of nuclear power plants. Figures I-9.1 through I-9.6 of Appendix I to Section III of the Code specify fatigue design curves for applicable structural materials. However, the Code design curves do not explicitly address the effects of light water reactor (LWR) water environments. Existing fatigue strain-vs.-life (ε-N) laboratory data illustrate potentially significant effects of LWR water environments on the fatigue resistance of pressure vessel and piping steels. Extensive studies have been conducted at Argonne National Laboratory and elsewheremore » since 1990 to investigate the effects of LWR environments on the fatigue life of piping and pressure vessel steels. This article summarizes the results of these studies. Existing fatigue ε-N data were evaluated to identify the various material, environmental, and loading conditions that influence fatigue crack initiation; a methodology for estimating fatigue lives as a function of these parameters was developed. The effects were incorporated into the ASME Code Section III fatigue evaluations in terms of an environmental correction factor, F en, which is defined as the ratio of fatigue life in air at room temperature to the fatigue life in the LWR water environment at reactor operating temperatures. Available fatigue data were used to develop fatigue design curves for carbon and low-alloy steels, austenitic stainless steels, and nickel-chromium-iron (NiCr-Fe) alloys and their weld metals in air at room temperature. A review of the Code Section III fatigue adjustment factors of 2 on strain and 20 on life is also presented and the possible conservatism inherent in the choice of these adjustment factors is evaluated. A brief description of potential effects of neutron irradiation on fatigue crack initiation for these structural materials is also presented.« less

Verification and Validation Strategy for LWRS Tools

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

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

2012-09-01

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

Mitochondrial genome evolution in the Saccharomyces sensu stricto complex.

PubMed

Ruan, Jiangxing; Cheng, Jian; Zhang, Tongcun; Jiang, Huifeng

2017-01-01

Exploring the evolutionary patterns of mitochondrial genomes is important for our understanding of the Saccharomyces sensu stricto (SSS) group, which is a model system for genomic evolution and ecological analysis. In this study, we first obtained the complete mitochondrial sequences of two important species, Saccharomyces mikatae and Saccharomyces kudriavzevii. We then compared the mitochondrial genomes in the SSS group with those of close relatives, and found that the non-coding regions evolved rapidly, including dramatic expansion of intergenic regions, fast evolution of introns and almost 20-fold higher rearrangement rates than those of the nuclear genomes. However, the coding regions, and especially the protein-coding genes, are more conserved than those in the nuclear genomes of the SSS group. The different evolutionary patterns of coding and non-coding regions in the mitochondrial and nuclear genomes may be related to the origin of the aerobic fermentation lifestyle in this group. Our analysis thus provides novel insights into the evolution of mitochondrial genomes.

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

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

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

1999-09-01

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

Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

NASA Astrophysics Data System (ADS)

Grodzki, Marcin; Darnowski, Piotr; Niewiński, Grzegorz

2017-12-01

The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an `early design' variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit). A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

Gamma-Ray Imaging Probes.

NASA Astrophysics Data System (ADS)

Wild, Walter James

1988-12-01

External nuclear medicine diagnostic imaging of early primary and metastatic lung cancer tumors is difficult due to the poor sensitivity and resolution of existing gamma cameras. Nonimaging counting detectors used for internal tumor detection give ambiguous results because distant background variations are difficult to discriminate from neighboring tumor sites. This suggests that an internal imaging nuclear medicine probe, particularly an esophageal probe, may be advantageously used to detect small tumors because of the ability to discriminate against background variations and the capability to get close to sites neighboring the esophagus. The design, theory of operation, preliminary bench tests, characterization of noise behavior and optimization of such an imaging probe is the central theme of this work. The central concept lies in the representation of the aperture shell by a sequence of binary digits. This, coupled with the mode of operation which is data encoding within an axial slice of space, leads to the fundamental imaging equation in which the coding operation is conveniently described by a circulant matrix operator. The coding/decoding process is a classic coded-aperture problem, and various estimators to achieve decoding are discussed. Some estimators require a priori information about the object (or object class) being imaged; the only unbiased estimator that does not impose this requirement is the simple inverse-matrix operator. The effects of noise on the estimate (or reconstruction) is discussed for general noise models and various codes/decoding operators. The choice of an optimal aperture for detector count times of clinical relevance is examined using a statistical class-separability formalism.

System Simulation of Nuclear Power Plant by Coupling RELAP5 and Matlab/Simulink

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

Meng Lin; Dong Hou; Zhihong Xu

2006-07-01

Since RELAP5 code has general and advanced features in thermal-hydraulic computation, it has been widely used in transient and accident safety analysis, experiment planning analysis, and system simulation, etc. So we wish to design, analyze, verify a new Instrumentation And Control (I and C) system of Nuclear Power Plant (NPP) based on the best-estimated code, and even develop our engineering simulator. But because of limited function of simulating control and protection system in RELAP5, it is necessary to expand the function for high efficient, accurate, flexible design and simulation of I and C system. Matlab/Simulink, a scientific computation software, justmore » can compensate the limitation, which is a powerful tool in research and simulation of plant process control. The software is selected as I and C part to be coupled with RELAP5 code to realize system simulation of NPPs. There are two key techniques to be solved. One is the dynamic data exchange, by which Matlab/Simulink receives plant parameters and returns control results. Database is used to communicate the two codes. Accordingly, Dynamic Link Library (DLL) is applied to link database in RELAP5, while DLL and S-Function is applied in Matlab/Simulink. The other problem is synchronization between the two codes for ensuring consistency in global simulation time. Because Matlab/Simulink always computes faster than RELAP5, the simulation time is sent by RELAP5 and received by Matlab/Simulink. A time control subroutine is added into the simulation procedure of Matlab/Simulink to control its simulation advancement. Through these ways, Matlab/Simulink is dynamically coupled with RELAP5. Thus, in Matlab/Simulink, we can freely design control and protection logic of NPPs and test it with best-estimated plant model feedback. A test will be shown to illuminate that results of coupling calculation are nearly the same with one of single RELAP5 with control logic. In practice, a real Pressurized Water Reactor (PWR) is modeled by RELAP5 code, and its main control and protection system is duplicated by Matlab/Simulink. Some steady states and transients are calculated under control of these I and C systems, and the results are compared with the plant test curves. The application showed that it can do exact system simulation of NPPs by coupling RELAP5 and Matlab/Simulink. This paper will mainly focus on the coupling method, plant thermal-hydraulic model, main control logics, test and application results. (authors)« less

Gas Core Reactor Numerical Simulation Using a Coupled MHD-MCNP Model

NASA Technical Reports Server (NTRS)

Kazeminezhad, F.; Anghaie, S.

2008-01-01

Analysis is provided in this report of using two head-on magnetohydrodynamic (MHD) shocks to achieve supercritical nuclear fission in an axially elongated cylinder filled with UF4 gas as an energy source for deep space missions. The motivation for each aspect of the design is explained and supported by theory and numerical simulations. A subsequent report will provide detail on relevant experimental work to validate the concept. Here the focus is on the theory of and simulations for the proposed gas core reactor conceptual design from the onset of shock generations to the supercritical state achieved when the shocks collide. The MHD model is coupled to a standard nuclear code (MCNP) to observe the neutron flux and fission power attributed to the supercritical state brought about by the shock collisions. Throughout the modeling, realistic parameters are used for the initial ambient gaseous state and currents to ensure a resulting supercritical state upon shock collisions.

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 Efficiency Nuclear Power Plants Using Liquid Fluoride Thorium Reactor Technology

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

2009-01-01

An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITs of 950 and 1200 K are presented. Power plant performance data were obtained for TITs ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo-generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

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

Dakota Uncertainty Quantification Methods Applied to the CFD code Nek5000

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

Delchini, Marc-Olivier; Popov, Emilian L.; Pointer, William David

This report presents the state of advancement of a Nuclear Energy Advanced Modeling and Simulation (NEAMS) project to characterize the uncertainty of the computational fluid dynamics (CFD) code Nek5000 using the Dakota package for flows encountered in the nuclear engineering industry. Nek5000 is a high-order spectral element CFD code developed at Argonne National Laboratory for high-resolution spectral-filtered large eddy simulations (LESs) and unsteady Reynolds-averaged Navier-Stokes (URANS) simulations.

Survey of Codes Employing Nuclear Damage Assessment

DTIC Science & Technology

1977-10-01

surveyed codes were com- DO 73Mu 1473 ETN OF 1NOVSSSOLETE UNCLASSIFIED 1 SECURITY CLASSIFICATION OF THIS f AGE (Wh*11 Date Efntered)S<>-~C. I UNCLASSIFIED...level and above) TALLEY/TOTEM not nuclear TARTARUS too highly aggregated (battalion level and above) UNICORN highly aggregated force allocation code...vulnerability data can bq input by the user as he receives them, and there is the abil ’ity to replay any situation using hindsight. The age of target

Knowledge management: Role of the the Radiation Safety Information Computational Center (RSICC)

NASA Astrophysics Data System (ADS)

Valentine, Timothy

2017-09-01

The Radiation Safety Information Computational Center (RSICC) at Oak Ridge National Laboratory (ORNL) is an information analysis center that collects, archives, evaluates, synthesizes and distributes information, data and codes that are used in various nuclear technology applications. RSICC retains more than 2,000 software packages that have been provided by code developers from various federal and international agencies. RSICC's customers (scientists, engineers, and students from around the world) obtain access to such computing codes (source and/or executable versions) and processed nuclear data files to promote on-going research, to ensure nuclear and radiological safety, and to advance nuclear technology. The role of such information analysis centers is critical for supporting and sustaining nuclear education and training programs both domestically and internationally, as the majority of RSICC's customers are students attending U.S. universities. Additionally, RSICC operates a secure CLOUD computing system to provide access to sensitive export-controlled modeling and simulation (M&S) tools that support both domestic and international activities. This presentation will provide a general review of RSICC's activities, services, and systems that support knowledge management and education and training in the nuclear field.

Navigation and vessel inspection circular No. 3-94. International maritime organization code for the safe carriage of irradiated nuclear fuel, plutonium and high-level radioactive wastes in flasks on board ships (IMO resolution a.748(18)). Final report

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

NONE

1994-05-26

The Circular calls the attention of Coast Guard field units, marine surveyors, shippers and carriers of nuclear materials to the International Maritime Organization (IMO) Code for the Safe Carriage of Irradiated Nuclear Fuel, Plutonium and High-Level Radioactive Wastes in Flasks on Board Ships (IMO Resolution A.748(18)).

Deterministic and probabilistic analysis of damping device resistance under impact loads from nuclear fuel container drop

NASA Astrophysics Data System (ADS)

Kala, J.; Bajer, M.; Barnat, J.; Smutný, J.

2010-12-01

Pedestrian-induced vibrations are a criterion for serviceability. This loading is significant for light-weight footbridge structures, but was established as a basic loading for the ceilings of various ordinary buildings. Wide variations of this action exist. To verify the different conclusions of various authors, vertical pressure measurements invoked during walking were performed. In the article the approaches of different design codes are also shown.

Analysis of PANDA Passive Containment Cooling Steady-State Tests with the Spectra Code

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

Stempniewicz, Marek M

2000-07-15

Results of post test simulation of the PANDA passive containment cooling (PCC) steady-state tests (S-series tests), performed at the PANDA facility at the Paul Scherrer Institute, Switzerland, are presented. The simulation has been performed using the computer code SPECTRA, a thermal-hydraulic code, designed specifically for analyzing containment behavior of nuclear power plants.Results of the present calculations are compared to the measurement data as well as the results obtained earlier with the codes MELCOR, TRAC-BF1, and TRACG. The calculated PCC efficiencies are somewhat lower than the measured values. Similar underestimation of PCC efficiencies had been obtained in the past, with themore » other computer codes. To explain this difference, it is postulated that condensate coming into the tubes forms a stream of liquid in one or two tubes, leaving most of the tubes unaffected. The condensate entering the water box is assumed to fall down in the form of droplets. With these assumptions, the results calculated with SPECTRA are close to the experimental data.It is concluded that the SPECTRA code is a suitable tool for analyzing containments of advanced reactors, equipped with passive containment cooling systems.« less

Modified Laser and Thermos cell calculations on microcomputers

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

Shapiro, A.; Huria, H.C.

1987-01-01

In the course of designing and operating nuclear reactors, many fuel pin cell calculations are required to obtain homogenized cell cross sections as a function of burnup. In the interest of convenience and cost, it would be very desirable to be able to make such calculations on microcomputers. In addition, such a microcomputer code would be very helpful for educational course work in reactor computations. To establish the feasibility of making detailed cell calculations on a microcomputer, a mainframe cell code was compiled and run on a microcomputer. The computer code Laser, originally written in Fortran IV for the IBM-7090more » class of mainframe computers, is a cylindrical, one-dimensional, multigroup lattice cell program that includes burnup. It is based on the MUFT code for epithermal and fast group calculations, and Thermos for the thermal calculations. There are 50 fast and epithermal groups and 35 thermal groups. Resonances are calculated assuming a homogeneous system and then corrected for self-shielding, Dancoff, and Doppler by self-shielding factors. The Laser code was converted to run on a microcomputer. In addition, the Thermos portion of Laser was extracted and compiled separately to have available a stand alone thermal code.« less

Magnetic shield for turbomolecular pump of the Magnetized Plasma Linear Experimental device at Saha Institute of Nuclear Physics.

PubMed

Biswas, Subir; Chattopadhyay, Monobir; Pal, Rabindranath

2011-01-01

The turbo molecular pump of the Magnetized Plasma Linear Experimental device is protected from damage by a magnetic shield. As the pump runs continuously in a magnetic field environment during a plasma physics experiment, it may get damaged owing to eddy current effect. For design and testing of the shield, first we simulate in details various aspects of magnetic shield layouts using a readily available field design code. The performance of the shield made from two half cylinders of soft iron material, is experimentally observed to agree very well with the simulation results.

Reference commercial high-level waste glass and canister definition

NASA Astrophysics Data System (ADS)

Slate, S. C.; Ross, W. A.; Partain, W. L.

1981-09-01

Technical data and performance characteristics of a high level waste glass and canister intended for use in the design of a complete waste encapsulation package suitable for disposal in a geologic repository are presented. The borosilicate glass contained in the stainless steel canister represents the probable type of high level waste product that is produced in a commercial nuclear-fuel reprocessing plant. Development history is summarized for high level liquid waste compositions, waste glass composition and characteristics, and canister design. The decay histories of the fission products and actinides (plus daughters) calculated by the ORIGEN-II code are presented.

HUFF, a One-Dimensional Hydrodynamics Code for Strong Shocks

DTIC Science & Technology

1978-12-01

results for two sample problems. The first problem discussed is a one-kiloton nuclear burst in infinite sea level air. The second problem is the one...of HUFF as an effective first order hydro- dynamic computer code. 1 KT Explosion The one-kiloton nuclear explosion in infinite sea level air was

CSEWG

Science.gov Websites

Nuclear Data Evaluation Co-operation (WPEC) Nuclear Reaction Data Centers, NRDC (IAEA Vienna) EMPIRE , Nuclear Reaction Model Code Atlas of Neutron Resonances The Cross Section Evaluation Working Group (CSEWG

Thermal-hydraulic modeling needs for passive reactors

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

Kelly, J.M.

1997-07-01

The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered,more » but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken.« less

FRENDY: A new nuclear data processing system being developed at JAEA

NASA Astrophysics Data System (ADS)

Tada, Kenichi; Nagaya, Yasunobu; Kunieda, Satoshi; Suyama, Kenya; Fukahori, Tokio

2017-09-01

JAEA has provided an evaluated nuclear data library JENDL and nuclear application codes such as MARBLE, SRAC, MVP and PHITS. These domestic codes have been widely used in many universities and industrial companies in Japan. However, we sometimes find problems in imported processing systems and need to revise them when the new JENDL is released. To overcome such problems and immediately process the nuclear data when it is released, JAEA started developing a new nuclear data processing system, FRENDY in 2013. This paper describes the outline of the development of FRENDY and both its capabilities and performances by the analyses of criticality experiments. The verification results indicate that FRENDY properly generates ACE files.

Development of Fuel Shuffling Module for PHISICS

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

Allan Mabe; Andrea Alfonsi; Cristian Rabiti

2013-06-01

PHISICS (Parallel and Highly Innovative Simulation for the INL Code System) [4] code toolkit has been in development at the Idaho National Laboratory. This package is intended to provide a modern analysis tool for reactor physics investigation. It is designed with the mindset to maximize accuracy for a given availability of computational resources and to give state of the art tools to the modern nuclear engineer. This is obtained by implementing several different algorithms and meshing approaches among which the user will be able to choose, in order to optimize his computational resources and accuracy needs. The software is completelymore » modular in order to simplify the independent development of modules by different teams and future maintenance. The package is coupled with the thermo-hydraulic code RELAP5-3D [3]. In the following the structure of the different PHISICS modules is briefly recalled, focusing on the new shuffling module (SHUFFLE), object of this paper.« less

WE-AB-204-11: Development of a Nuclear Medicine Dosimetry Module for the GPU-Based Monte Carlo Code ARCHER

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

Liu, T; Lin, H; Xu, X

Purpose: To develop a nuclear medicine dosimetry module for the GPU-based Monte Carlo code ARCHER. Methods: We have developed a nuclear medicine dosimetry module for the fast Monte Carlo code ARCHER. The coupled electron-photon Monte Carlo transport kernel included in ARCHER is built upon the Dose Planning Method code (DPM). The developed module manages the radioactive decay simulation by consecutively tracking several types of radiation on a per disintegration basis using the statistical sampling method. Optimization techniques such as persistent threads and prefetching are studied and implemented. The developed module is verified against the VIDA code, which is based onmore » Geant4 toolkit and has previously been verified against OLINDA/EXM. A voxelized geometry is used in the preliminary test: a sphere made of ICRP soft tissue is surrounded by a box filled with water. Uniform activity distribution of I-131 is assumed in the sphere. Results: The self-absorption dose factors (mGy/MBqs) of the sphere with varying diameters are calculated by ARCHER and VIDA respectively. ARCHER’s result is in agreement with VIDA’s that are obtained from a previous publication. VIDA takes hours of CPU time to finish the computation, while it takes ARCHER 4.31 seconds for the 12.4-cm uniform activity sphere case. For a fairer CPU-GPU comparison, more effort will be made to eliminate the algorithmic differences. Conclusion: The coupled electron-photon Monte Carlo code ARCHER has been extended to radioactive decay simulation for nuclear medicine dosimetry. The developed code exhibits good performance in our preliminary test. The GPU-based Monte Carlo code is developed with grant support from the National Institute of Biomedical Imaging and Bioengineering through an R01 grant (R01EB015478)« less

Successful Recovery of Nuclear Protein-Coding Genes from Small Insects in Museums Using Illumina Sequencing.

PubMed

Kanda, Kojun; Pflug, James M; Sproul, John S; Dasenko, Mark A; Maddison, David R

2015-01-01

In this paper we explore high-throughput Illumina sequencing of nuclear protein-coding, ribosomal, and mitochondrial genes in small, dried insects stored in natural history collections. We sequenced one tenebrionid beetle and 12 carabid beetles ranging in size from 3.7 to 9.7 mm in length that have been stored in various museums for 4 to 84 years. Although we chose a number of old, small specimens for which we expected low sequence recovery, we successfully recovered at least some low-copy nuclear protein-coding genes from all specimens. For example, in one 56-year-old beetle, 4.4 mm in length, our de novo assembly recovered about 63% of approximately 41,900 nucleotides in a target suite of 67 nuclear protein-coding gene fragments, and 70% using a reference-based assembly. Even in the least successfully sequenced carabid specimen, reference-based assembly yielded fragments that were at least 50% of the target length for 34 of 67 nuclear protein-coding gene fragments. Exploration of alternative references for reference-based assembly revealed few signs of bias created by the reference. For all specimens we recovered almost complete copies of ribosomal and mitochondrial genes. We verified the general accuracy of the sequences through comparisons with sequences obtained from PCR and Sanger sequencing, including of conspecific, fresh specimens, and through phylogenetic analysis that tested the placement of sequences in predicted regions. A few possible inaccuracies in the sequences were detected, but these rarely affected the phylogenetic placement of the samples. Although our sample sizes are low, an exploratory regression study suggests that the dominant factor in predicting success at recovering nuclear protein-coding genes is a high number of Illumina reads, with success at PCR of COI and killing by immersion in ethanol being secondary factors; in analyses of only high-read samples, the primary significant explanatory variable was body length, with small beetles being more successfully sequenced.

Successful Recovery of Nuclear Protein-Coding Genes from Small Insects in Museums Using Illumina Sequencing

PubMed Central

Dasenko, Mark A.

2015-01-01

In this paper we explore high-throughput Illumina sequencing of nuclear protein-coding, ribosomal, and mitochondrial genes in small, dried insects stored in natural history collections. We sequenced one tenebrionid beetle and 12 carabid beetles ranging in size from 3.7 to 9.7 mm in length that have been stored in various museums for 4 to 84 years. Although we chose a number of old, small specimens for which we expected low sequence recovery, we successfully recovered at least some low-copy nuclear protein-coding genes from all specimens. For example, in one 56-year-old beetle, 4.4 mm in length, our de novo assembly recovered about 63% of approximately 41,900 nucleotides in a target suite of 67 nuclear protein-coding gene fragments, and 70% using a reference-based assembly. Even in the least successfully sequenced carabid specimen, reference-based assembly yielded fragments that were at least 50% of the target length for 34 of 67 nuclear protein-coding gene fragments. Exploration of alternative references for reference-based assembly revealed few signs of bias created by the reference. For all specimens we recovered almost complete copies of ribosomal and mitochondrial genes. We verified the general accuracy of the sequences through comparisons with sequences obtained from PCR and Sanger sequencing, including of conspecific, fresh specimens, and through phylogenetic analysis that tested the placement of sequences in predicted regions. A few possible inaccuracies in the sequences were detected, but these rarely affected the phylogenetic placement of the samples. Although our sample sizes are low, an exploratory regression study suggests that the dominant factor in predicting success at recovering nuclear protein-coding genes is a high number of Illumina reads, with success at PCR of COI and killing by immersion in ethanol being secondary factors; in analyses of only high-read samples, the primary significant explanatory variable was body length, with small beetles being more successfully sequenced. PMID:26716693

Nuclear Test Depth Determination with Synthetic Modelling: Global Analysis from PNEs to DPRK-2016

NASA Astrophysics Data System (ADS)

Rozhkov, Mikhail; Stachnik, Joshua; Baker, Ben; Epiphansky, Alexey; Bobrov, Dmitry

2016-04-01

Seismic event depth determination is critical for the event screening process at the International Data Center, CTBTO. A thorough determination of the event depth can be conducted mostly through additional special analysis because the IDC's Event Definition Criteria is based, in particular, on depth estimation uncertainties. This causes a large number of events in the Reviewed Event Bulletin to have depth constrained to the surface making the depth screening criterion not applicable. Further it may result in a heavier workload to manually distinguish between subsurface and deeper crustal events. Since the shape of the first few seconds of signal of very shallow events is very sensitive to the depth phases, cross correlation between observed and theoretic seismograms can provide a basis for the event depth estimation, and so an expansion to the screening process. We applied this approach mostly to events at teleseismic and partially regional distances. The approach was found efficient for the seismic event screening process, with certain caveats related mostly to poorly defined source and receiver crustal models which can shift the depth estimate. An adjustable teleseismic attenuation model (t*) for synthetics was used since this characteristic is not known for most of the rays we studied. We studied a wide set of historical records of nuclear explosions, including so called Peaceful Nuclear Explosions (PNE) with presumably known depths, and recent DPRK nuclear tests. The teleseismic synthetic approach is based on the stationary phase approximation with hudson96 program, and the regional modelling was done with the generalized ray technique by Vlastislav Cerveny modified to account for the complex source topography. The software prototype is designed to be used for the Expert Technical Analysis at the IDC. With this, the design effectively reuses the NDC-in-a-Box code and can be comfortably utilized by the NDC users. The package uses Geotool as a front-end for data retrieval and pre-processing. After the event database is compiled, the control is passed to the driver software, running the external processing and plotting toolboxes, which controls the final stage and produces the final result. The modules are mostly Python coded, C-coded (Raysynth3D complex topography regional synthetics) and FORTRAN coded synthetics from the CPS330 software package by Robert Herrmann of Saint Louis University. The extension of this single station depth determination method is under development and uses joint information from all stations participating in processing. It is based on simultaneous depth and moment tensor determination for both short and long period seismic phases. A novel approach recently developed for microseismic event location utilizing only phase waveform information was migrated to a global scale. It should provide faster computation as it does not require intensive synthetic modelling, and might benefit processing noisy signals. A consistent depth estimate for all recent nuclear tests was produced for the vast number of IMS stations (primary and auxiliary) used in processing.

Neutron flux measurements on a mock-up of a storage cask for high-level nuclear waste using 2.5 MeV neutrons.

PubMed

Suárez, H Saurí; Becker, F; Klix, A; Pang, B; Döring, T

2018-06-07

To store and dispose spent nuclear fuel, shielding casks are employed to reduce the emitted radiation. To evaluate the exposure of employees handling such casks, Monte Carlo radiation transport codes can be employed. Nevertheless, to assess the reliability of these codes and nuclear data, experimental checks are required. In this study, a neutron generator (NG) producing neutrons of 2.5 MeV was employed to simulate neutrons produced in spent nuclear fuel. Different configurations of shielding layers of steel and polyethylene were positioned between the target of the NG and a NE-213 detector. The results of the measurements of neutron and γ radiation and the corresponding simulations with the code MCNP6 are presented. Details of the experimental set-up as well as neutron and photon flux spectra are provided as reference points for such NG investigations with shielding structures.

Evaluation of Production Cross Sections of Li, Be, B in CR

NASA Technical Reports Server (NTRS)

Moskalenko, I. V.; Mashnik, S. G.

2003-01-01

Accurate evaluation of the production cross section of light elements is important for models of cosmic ray (CR) propagation, galactic chemical evolution, and cosmological studies. However, the experimental spallation cross section data are scarce and often unavailable to CR community while semi-empirical systematics are frequently wrong by a significant factor. Running sophisticated nuclear codes is not an option of choice for everyone either. We use the Los Alamos versions of the Quark-Gluon String Model code LAQGSM and the improved Cascade-Exciton Model code CEM2k together with all available data from Los Alamos Nuclear Laboratory (LANL) nuclear database to produce evaluated production cross sections of isotopes of Li, Be, and B suitable for astrophysical applications. The LAQGSM and CEM2k models have been shown to reproduce well nuclear reactions and hadronic data in the range 0.01-800 GeV/nucleon.

Assessment of the Effects of Entrainment and Wind Shear on Nuclear Cloud Rise Modeling

NASA Astrophysics Data System (ADS)

Zalewski, Daniel; Jodoin, Vincent

2001-04-01

Accurate modeling of nuclear cloud rise is critical in hazard prediction following a nuclear detonation. This thesis recommends improvements to the model currently used by DOD. It considers a single-term versus a three-term entrainment equation, the value of the entrainment and eddy viscous drag parameters, as well as the effect of wind shear in the cloud rise following a nuclear detonation. It examines departures from the 1979 version of the Department of Defense Land Fallout Interpretive Code (DELFIC) with the current code used in the Hazard Prediction and Assessment Capability (HPAC) code version 3.2. The recommendation for a single-term entrainment equation, with constant value parameters, without wind shear corrections, and without cloud oscillations is based on both a statistical analysis using 67 U.S. nuclear atmospheric test shots and the physical representation of the modeling. The statistical analysis optimized the parameter values of interest for four cases: the three-term entrainment equation with wind shear and without wind shear as well as the single-term entrainment equation with and without wind shear. The thesis then examines the effect of cloud oscillations as a significant departure in the code. Modifications to user input atmospheric tables are identified as a potential problem in the calculation of stabilized cloud dimensions in HPAC.

Fission Activities of the Nuclear Reactions Group in Uppsala

NASA Astrophysics Data System (ADS)

Al-Adili, A.; Alhassan, E.; Gustavsson, C.; Helgesson, P.; Jansson, K.; Koning, A.; Lantz, M.; Mattera, A.; Prokofiev, A. V.; Rakopoulos, V.; Sjöstrand, H.; Solders, A.; Tarrío, D.; Österlund, M.; Pomp, S.

This paper highlights some of the main activities related to fission of the nuclear reactions group at Uppsala University. The group is involved for instance in fission yield experiments at the IGISOL facility, cross-section measurements at the NFS facility, as well as fission dynamics studies at the IRMM JRC-EC. Moreover, work is ongoing on the Total Monte Carlo (TMC) methodology and on including the GEF fission code into the TALYS nuclear reaction code. Selected results from these projects are discussed.

Structural mechanics simulations

NASA Technical Reports Server (NTRS)

Biffle, Johnny H.

1992-01-01

Sandia National Laboratory has a very broad structural capability. Work has been performed in support of reentry vehicles, nuclear reactor safety, weapons systems and components, nuclear waste transport, strategic petroleum reserve, nuclear waste storage, wind and solar energy, drilling technology, and submarine programs. The analysis environment contains both commercial and internally developed software. Included are mesh generation capabilities, structural simulation codes, and visual codes for examining simulation results. To effectively simulate a wide variety of physical phenomena, a large number of constitutive models have been developed.

Preliminary Study on LiF4-ThF4-PuF4 Utilization as Fuel Salt of miniFUJI Molten Salt Reactor

NASA Astrophysics Data System (ADS)

Waris, Abdul; Aji, Indarta K.; Pramuditya, Syeilendra; Widayani; Irwanto, Dwi

2016-08-01

miniFUJI reactor is molten salt reactor (MSR) which is one type of the Generation IV nuclear energy systems. The original miniFUJI reactor design uses LiF-BeF2-ThF4-233UF4 as a fuel salt. In the present study, the use of LiF4-ThF4-PuF4 as fuel salt instead of LiF-BeF2-ThF4-UF4 will be discussed. The neutronics cell calculation has been performed by using PIJ (collision probability method code) routine of SRAC 2006 code, with the nuclear data library is JENDL-4.0. The results reveal that the reactor can attain the criticality condition with the plutonium concentration in the fuel salt is equal to 9.16% or more. The conversion ratio diminishes with the enlarging of plutonium concentration in the fuel. The neutron spectrum of miniFUJI MSR with plutonium fuel becomes harder compared to that of the 233U fuel.

Development of fission-products transport model in severe-accident scenarios for Scdap/Relap5

NASA Astrophysics Data System (ADS)

Honaiser, Eduardo Henrique Rangel

The understanding and estimation of the release of fission products during a severe accident became one of the priorities of the nuclear community after 1980, with the events of the Three-mile Island unit 2 (TMI-2), in 1979, and Chernobyl accidents, in 1986. Since this time, theoretical developments and experiments have shown that the primary circuit systems of light water reactors (LWR) have the potential to attenuate the release of fission products, a fact that had been neglected before. An advanced tool, compatible with nuclear thermal-hydraulics integral codes, is developed to predict the retention and physical evolution of the fission products in the primary circuit of LWRs, without considering the chemistry effects. The tool embodies the state-of-the-art models for the involved phenomena as well as develops new models. The capabilities acquired after the implementation of this tool in the Scdap/Relap5 code can be used to increase the accuracy of probability safety assessment (PSA) level 2, enhance the reactor accident management procedures and design new emergency safety features.

Transient analysis of ”2 inch Direct Vessel Injection line break” in SPES-2 facility by using TRACE code

NASA Astrophysics Data System (ADS)

D'Amico, S.; Lombardo, C.; Moscato, I.; Polidori, M.; Vella, G.

2015-11-01

In the past few decades a lot of theoretical and experimental researches have been done to understand the physical phenomena characterizing nuclear accidents. In particular, after the Three Miles Island accident, several reactors have been designed to handle successfully LOCA events. This paper presents a comparison between experimental and numerical results obtained for the “2 inch Direct Vessel Injection line break” in SPES-2. This facility is an integral test facility built in Piacenza at the SIET laboratories and simulating the primary circuit, the relevant parts of the secondary circuits and the passive safety systems typical of the AP600 nuclear power plant. The numerical analysis here presented was performed by using TRACE and CATHARE thermal-hydraulic codes with the purpose of evaluating their prediction capability. The main results show that the TRACE model well predicts the overall behaviour of the plant during the transient, in particular it is able to simulate the principal thermal-hydraulic phenomena related to all passive safety systems. The performance of the presented CATHARE noding has suggested some possible improvements of the model.

U.S. Commercial Spent Nuclear Fuel Assembly Characteristics - 1968-2013

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

Hu, Jianwei; Peterson, Joshua L.; Gauld, Ian C.

2016-09-01

Activities related to management of spent nuclear fuel (SNF) are increasing in the US and many other countries. Over 240,000 SNF assemblies have been discharged from US commercial reactors since the late 1960s. The enrichment and burnup of SNF have changed significantly over the past 40 years, and fuel assembly designs have also evolved. Understanding the general characteristics of SNF helps regulators and other stakeholders form overall strategies towards the final disposal of US SNF. This report documents a survey of all US commercial SNF assemblies in the GC-859 database and provides reference SNF source terms (e.g., nuclide inventories, decaymore » heat, and neutron/photon emission) at various cooling times up to 200 years after fuel discharge. This study reviews the distribution and evolution of fuel parameters of all SNF assemblies discharged over the past 40 years. Assemblies were categorized into three groups based on discharge year, and the median burnups and enrichments of each group were used to establish representative cases. An extended burnup case was created for boiling water reactor (BWR) fuels, and another was created for the pressurized water reactor (PWR) fuels. Two additional cases were developed to represent the eight mixed oxide (MOX) fuel assemblies in the database. Burnup calculations were performed for each representative case. Realistic parameters for fuel design and operations were used to model the SNF and to provide reference fuel characteristics representative of the current inventory. Burnup calculations were performed using the ORIGEN code, which is part of the SCALE nuclear modeling and simulation code system. Results include total activity, decay heat, photon emission, neutron flux, gamma heat, and plutonium content, as well as concentrations for 115 significant nuclides. These quantities are important in the design, regulation, and operations of SNF storage, transportation, and disposal systems.« less

Addressing Fission Product Validation in MCNP Burnup Credit Criticality Calculations

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

Mueller, Don; Bowen, Douglas G; Marshall, William BJ J

2015-01-01

The US Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation issued Interim Staff Guidance (ISG) 8, Revision 3 in September 2012. This ISG provides guidance for NRC staff members’ review of burnup credit (BUC) analyses supporting transport and dry storage of pressurized water reactor spent nuclear fuel (SNF) in casks. The ISG includes guidance for addressing validation of criticality (k eff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MAs). Based on previous work documented in NRC Regulatory Guide (NUREG) Contractor Report (CR)-7109, the ISG recommends that NRC staff members acceptmore » the use of either 1.5 or 3% of the FP&MA worth—in addition to bias and bias uncertainty resulting from validation of k eff calculations for the major actinides in SNF—to conservatively account for the bias and bias uncertainty associated with the specified unvalidated FP&MAs. The ISG recommends (1) use of 1.5% of the FP&MA worth if a modern version of SCALE and its nuclear data are used and (2) 3% of the FP&MA worth for well qualified, industry standard code systems other than SCALE with the Evaluated Nuclear Data Files, Part B (ENDF/B),-V, ENDF/B-VI, or ENDF/B-VII cross sections libraries. The work presented in this paper provides a basis for extending the use of the 1.5% of the FP&MA worth bias to BUC criticality calculations performed using the Monte Carlo N-Particle (MCNP) code. The extended use of the 1.5% FP&MA worth bias is shown to be acceptable by comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII–based nuclear data. The comparison supports use of the 1.5% FP&MA worth bias when the MCNP code is used for criticality calculations, provided that the cask design is similar to the hypothetical generic BUC-32 cask model and that the credited FP&MA worth is no more than 0.1 Δk eff (ISG-8, Rev. 3, Recommendation 4).« less

The Monte Carlo code MCPTV--Monte Carlo dose calculation in radiation therapy with carbon ions.

PubMed

Karg, Juergen; Speer, Stefan; Schmidt, Manfred; Mueller, Reinhold

2010-07-07

The Monte Carlo code MCPTV is presented. MCPTV is designed for dose calculation in treatment planning in radiation therapy with particles and especially carbon ions. MCPTV has a voxel-based concept and can perform a fast calculation of the dose distribution on patient CT data. Material and density information from CT are taken into account. Electromagnetic and nuclear interactions are implemented. Furthermore the algorithm gives information about the particle spectra and the energy deposition in each voxel. This can be used to calculate the relative biological effectiveness (RBE) for each voxel. Depth dose distributions are compared to experimental data giving good agreement. A clinical example is shown to demonstrate the capabilities of the MCPTV dose calculation.

An analytical study on excitation of nuclear-coupled thermal-hydraulic instability due to seismically induced resonance in BWR

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

Hirano, Masashi

1997-07-01

This paper describes the results of a scoping study on seismically induced resonance of nuclear-coupled thermal-hydraulic instability in BWRs, which was conducted by using TRAC-BF1 within a framework of a point kinetics model. As a result of the analysis, it is shown that a reactivity insertion could occur accompanied by in-surge of coolant into the core resulted from the excitation of the nuclear-coupled instability by the external acceleration. In order to analyze this phenomenon more in detail, it is necessary to couple a thermal-hydraulic code with a three-dimensional nuclear kinetics code.

Karyopherin-mediated nuclear import of the homing endonuclease VMA1-derived endonuclease is required for self-propagation of the coding region.

PubMed

Nagai, Yuri; Nogami, Satoru; Kumagai-Sano, Fumi; Ohya, Yoshikazu

2003-03-01

VMA1-derived endonuclease (VDE), a site-specific endonuclease in Saccharomyces cerevisiae, enters the nucleus to generate a double-strand break in the VDE-negative allelic locus, mediating the self-propagating gene conversion called homing. Although VDE is excluded from the nucleus in mitotic cells, it relocalizes at premeiosis, becoming localized in both the nucleus and the cytoplasm in meiosis. The nuclear localization of VDE is induced by inactivation of TOR kinases, which constitute central regulators of cell differentiation in S. cerevisiae, and by nutrient depletion. A functional genomic approach revealed that at least two karyopherins, Srp1p and Kap142p, are required for the nuclear localization pattern. Genetic and physical interactions between Srp1p and VDE imply direct involvement of karyopherin-mediated nuclear transport in this process. Inactivation of TOR signaling or acquisition of an extra nuclear localization signal in the VDE coding region leads to artificial nuclear localization of VDE and thereby induces homing even during mitosis. These results serve as evidence that VDE utilizes the host systems of nutrient signal transduction and nucleocytoplasmic transport to ensure the propagation of its coding region.

Karyopherin-Mediated Nuclear Import of the Homing Endonuclease VMA1-Derived Endonuclease Is Required for Self-Propagation of the Coding Region

PubMed Central

Nagai, Yuri; Nogami, Satoru; Kumagai-Sano, Fumi; Ohya, Yoshikazu

2003-01-01

VMA1-derived endonuclease (VDE), a site-specific endonuclease in Saccharomyces cerevisiae, enters the nucleus to generate a double-strand break in the VDE-negative allelic locus, mediating the self-propagating gene conversion called homing. Although VDE is excluded from the nucleus in mitotic cells, it relocalizes at premeiosis, becoming localized in both the nucleus and the cytoplasm in meiosis. The nuclear localization of VDE is induced by inactivation of TOR kinases, which constitute central regulators of cell differentiation in S. cerevisiae, and by nutrient depletion. A functional genomic approach revealed that at least two karyopherins, Srp1p and Kap142p, are required for the nuclear localization pattern. Genetic and physical interactions between Srp1p and VDE imply direct involvement of karyopherin-mediated nuclear transport in this process. Inactivation of TOR signaling or acquisition of an extra nuclear localization signal in the VDE coding region leads to artificial nuclear localization of VDE and thereby induces homing even during mitosis. These results serve as evidence that VDE utilizes the host systems of nutrient signal transduction and nucleocytoplasmic transport to ensure the propagation of its coding region. PMID:12588991

Maestro and Castro: Simulation Codes for Astrophysical Flows

NASA Astrophysics Data System (ADS)

Zingale, Michael; Almgren, Ann; Beckner, Vince; Bell, John; Friesen, Brian; Jacobs, Adam; Katz, Maximilian P.; Malone, Christopher; Nonaka, Andrew; Zhang, Weiqun

2017-01-01

Stellar explosions are multiphysics problems—modeling them requires the coordinated input of gravity solvers, reaction networks, radiation transport, and hydrodynamics together with microphysics recipes to describe the physics of matter under extreme conditions. Furthermore, these models involve following a wide range of spatial and temporal scales, which puts tough demands on simulation codes. We developed the codes Maestro and Castro to meet the computational challenges of these problems. Maestro uses a low Mach number formulation of the hydrodynamics to efficiently model convection. Castro solves the fully compressible radiation hydrodynamics equations to capture the explosive phases of stellar phenomena. Both codes are built upon the BoxLib adaptive mesh refinement library, which prepares them for next-generation exascale computers. Common microphysics shared between the codes allows us to transfer a problem from the low Mach number regime in Maestro to the explosive regime in Castro. Importantly, both codes are freely available (https://github.com/BoxLib-Codes). We will describe the design of the codes and some of their science applications, as well as future development directions.Support for development was provided by NSF award AST-1211563 and DOE/Office of Nuclear Physics grant DE-FG02-87ER40317 to Stony Brook and by the Applied Mathematics Program of the DOE Office of Advance Scientific Computing Research under US DOE contract DE-AC02-05CH11231 to LBNL.

Nuclear design analysis of square-lattice honeycomb space nuclear rocket engine

NASA Astrophysics Data System (ADS)

Widargo, Reza; Anghaie, Samim

1999-01-01

The square-lattice honeycomb reactor is designed based on a cylindrical core that is determined to have critical diameter and length of 0.50 m and 0.50 c, respectively. A 0.10-cm thick radial graphite reflector, in addition to a 0.20-m thick axial graphite reflector are used to reduce neutron leakage from the reactor. The core is fueled with solid solution of 93% enriched (U, Zr, Nb)C, which is one of several ternary uranium carbides that are considered for this concept. The fuel is to be fabricated as 2 mm grooved (U, Zr, Nb)C wafers. The fuel wafers are used to form square-lattice honeycomb fuel assemblies, 0.10 m in length with 30% cross-sectional flow area. Five fuel assemblies are stacked up axially to form the reactor core. Based on the 30% void fraction, the width of the square flow channel is about 1.3 mm. The hydrogen propellant is passed through these flow channels and removes the heat from the reactor core. To perform nuclear design analysis, a series of neutron transport and diffusion codes are used. The preliminary results are obtained using a simple four-group cross-section model. To optimize the nuclear design, the fuel densities are varied for each assembly. Tantalum, hafnium and tungsten are considered and used as a replacement for niobium in fuel material to provide water submersion sub-criticality for the reactor. Axial and radial neutron flux and power density distributions are calculated for the core. Results of the neutronic analysis indicate that the core has a relatively fast spectrum. From the results of the thermal hydraulic analyses, eight axial temperature zones are chosen for the calculation of group average cross-sections. An iterative process is conducted to couple the neutronic calculations with the thermal hydraulics calculations. Results of the nuclear design analysis indicate that a compact core can be designed based on ternary uranium carbide square-lattice honeycomb fuel. This design provides a relatively high thrust to weight ratio.

World Energy Data System (WENDS). Volume X. Nuclear facility profiles, PO--ZA. [Brief tabulated information

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

Not Available

1979-06-01

In this compendium each profile of a nuclear facility is a capsule summary of pertinent facts regarding that particular installation. The facilities described include the entire fuel cycle in the broadest sense, encompassing resource recovery through waste management. Power plants and all US facilities have been excluded. To facilitate comparison the profiles have been recorded in a standard format. Because of the breadth of the undertaking some data fields do not apply to the establishment under discussion and accordingly are blank. The set of nuclear facility profiles occupies four volumes; the profiles are ordered by country name, and then bymore » facility code. Each nuclear facility profile volume contains two complete indexes to the information. The first index aggregates the facilities alphabetically by country. It is further organized by category of facility, and then by the four-character facility code. It provides a quick summary of the nuclear energy capability or interest in each country and also an identifier, the facility code, which can be used to access the information contained in the profile.« less

Summary of comparison and analysis of results from exercises 1 and 2 of the OECD PBMR coupled neutronics/thermal hydraulics transient benchmark

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

Mkhabela, P.; Han, J.; Tyobeka, B.

2006-07-01

The Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) has accepted, through the Nuclear Science Committee (NSC), the inclusion of the Pebble-Bed Modular Reactor 400 MW design (PBMR-400) coupled neutronics/thermal hydraulics transient benchmark problem as part of their official activities. The scope of the benchmark is to establish a well-defined problem, based on a common given library of cross sections, to compare methods and tools in core simulation and thermal hydraulics analysis with a specific focus on transient events through a set of multi-dimensional computational test problems. The benchmark includes three steady state exercises andmore » six transient exercises. This paper describes the first two steady state exercises, their objectives and the international participation in terms of organization, country and computer code utilized. This description is followed by a comparison and analysis of the participants' results submitted for these two exercises. The comparison of results from different codes allows for an assessment of the sensitivity of a result to the method employed and can thus help to focus the development efforts on the most critical areas. The two first exercises also allow for removing of user-related modeling errors and prepare core neutronics and thermal-hydraulics models of the different codes for the rest of the exercises in the benchmark. (authors)« less

Nuclear Energy Advanced Modeling and Simulation (NEAMS) Waste Integrated Performance and Safety Codes (IPSC) : FY10 development and integration.

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

Criscenti, Louise Jacqueline; Sassani, David Carl; Arguello, Jose Guadalupe, Jr.

2011-02-01

This report describes the progress in fiscal year 2010 in developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs,more » and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. Waste IPSC activities in fiscal year 2010 focused on specifying a challenge problem to demonstrate proof of concept, developing a verification and validation plan, and performing an initial gap analyses to identify candidate codes and tools to support the development and integration of the Waste IPSC. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. This year-end progress report documents the FY10 status of acquisition, development, and integration of thermal-hydrologic-chemical-mechanical (THCM) code capabilities, frameworks, and enabling tools and infrastructure.« less

The ITER ICRF Antenna Design with TOPICA

NASA Astrophysics Data System (ADS)

Milanesio, Daniele; Maggiora, Riccardo; Meneghini, Orso; Vecchi, Giuseppe

2007-11-01

TOPICA (Torino Polytechnic Ion Cyclotron Antenna) code is an innovative tool for the 3D/1D simulation of Ion Cyclotron Radio Frequency (ICRF), i.e. accounting for antennas in a realistic 3D geometry and with an accurate 1D plasma model [1]. The TOPICA code has been deeply parallelized and has been already proved to be a reliable tool for antennas design and performance prediction. A detailed analysis of the 24 straps ITER ICRF antenna geometry has been carried out, underlining the strong dependence and asymmetries of the antenna input parameters due to the ITER plasma response. We optimized the antenna array geometry dimensions to maximize loading, lower mutual couplings and mitigate sheath effects. The calculated antenna input impedance matrices are TOPICA results of a paramount importance for the tuning and matching system design. Electric field distributions have been also calculated and they are used as the main input for the power flux estimation tool. The designed optimized antenna is capable of coupling 20 MW of power to plasma in the 40 -- 55 MHz frequency range with a maximum voltage of 45 kV in the feeding coaxial cables. [1] V. Lancellotti et al., Nuclear Fusion, 46 (2006) S476-S499

GASFLOW: A Computational Fluid Dynamics Code for Gases, Aerosols, and Combustion, Volume 3: Assessment Manual

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

Müller, C.; Hughes, E. D.; Niederauer, G. F.

1998-10-01

Los Alamos National Laboratory (LANL) and Forschungszentrum Karlsruhe (FzK) are developing GASFLOW, a three-dimensional (3D) fluid dynamics field code as a best- estimate tool to characterize local phenomena within a flow field. Examples of 3D phenomena include circulation patterns; flow stratification; hydrogen distribution mixing and stratification; combustion and flame propagation; effects of noncondensable gas distribution on local condensation and evaporation; and aerosol entrainment, transport, and deposition. An analysis with GASFLOW will result in a prediction of the gas composition and discrete particle distribution in space and time throughout the facility and the resulting pressure and temperature loadings on the wallsmore » and internal structures with or without combustion. A major application of GASFLOW is for predicting the transport, mixing, and combustion of hydrogen and other gases in nuclear reactor containment and other facilities. It has been applied to situations involving transporting and distributing combustible gas mixtures. It has been used to study gas dynamic behavior in low-speed, buoyancy-driven flows, as well as sonic flows or diffusion dominated flows; and during chemically reacting flows, including deflagrations. The effects of controlling such mixtures by safety systems can be analyzed. The code version described in this manual is designated GASFLOW 2.1, which combines previous versions of the United States Nuclear Regulatory Commission code HMS (for Hydrogen Mixing Studies) and the Department of Energy and FzK versions of GASFLOW. The code was written in standard Fortran 90. This manual comprises three volumes. Volume I describes the governing physical equations and computational model. Volume II describes how to use the code to set up a model geometry, specify gas species and material properties, define initial and boundary conditions, and specify different outputs, especially graphical displays. Sample problems are included. Volume III contains some of the assessments performed by LANL and FzK« less

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

Not Available

This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation,more » boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.« less

PRIC320, a transcription coactivator, isolated from peroxisome proliferator-binding protein complex.

PubMed

Surapureddi, Sailesh; Viswakarma, Navin; Yu, Songtao; Guo, Dongsheng; Rao, M Sambasiva; Reddy, Janardan K

2006-05-05

Ciprofibrate, a potent peroxisome proliferator, induces pleiotropic responses in liver by activating peroxisome proliferator-activated receptor alpha (PPARalpha), a nuclear receptor. Transcriptional regulation by liganded nuclear receptors involves the participation of coregulators that form multiprotein complexes possibly to achieve cell and gene specific transcription. SDS-PAGE and matrix-assisted laser desorption/ionization reflection time-of-flight mass spectrometric analyses of ciprofibrate-binding proteins from liver nuclear extracts obtained using ciprofibrate-Sepharose affinity matrix resulted in the identification of a new high molecular weight nuclear receptor coactivator, which we designated PRIC320. The full-length human cDNA encoding this protein has an open-reading frame that codes for a 320kDa protein containing 2882 amino acids. PRIC320 contains five LXXLL signature motifs that mediate interaction with nuclear receptors. PRIC320 binds avidly to nuclear receptors PPARalpha, CAR, ERalpha, and RXR, but only minimally with PPARgamma. PRIC320 also interacts with transcription cofactors CBP, PRIP, and PBP. Immunoprecipitation-immunoblotting as well as cellular localization studies confirmed the interaction between PPARalpha and PRIC320. PRIC320 acts as a transcription coactivator by stimulating PPARalpha-mediated transcription. We conclude that ciprofibrate, a PPARalpha ligand, binds a multiprotein complex and PRIC320 cloned from this complex functions as a nuclear receptor coactivator.

Design and testing of the reactor-internal hydraulic control rod drive for the nuclear heating plant

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

Batheja, P.; Meier, W.J.; Rau, P.J.

A hydraulically driven control rod is being developed at Kraftwerk Union for integration in the primary system of a small nuclear district heating reactor. An elaborate test program, under way for --3 yr, was initiated with a plexiglass rig to understand the basic principles. A design specification list was prepared, taking reactor boundary conditions and relevant German rules and regulations into account. Subsequently, an atmospheric loop for testing of components at 20 to 90/sup 0/C was erected. The objectives involved optimization of individual components such as a piston/cylinder drive unit, electromagnetic valves, and an ultrasonic position indication system as wellmore » as verification of computer codes. Based on the results obtained, full-scale components were designed and fabricated for a prototype test rig, which is currently in operation. Thus far, all atmospheric tests in this rig have been completed. Investigations under reactor temperature and pressure, followed by endurance tests, are under way. All tests to date have shown a reliable functioning of the hydraulic drive, including a novel ultrasonic position indication system.« less

Using NJOY to Create MCNP ACE Files and Visualize Nuclear Data

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

Kahler, Albert Comstock

We provide lecture materials that describe the input requirements to create various MCNP ACE files (Fast, Thermal, Dosimetry, Photo-nuclear and Photo-atomic) with the NJOY Nuclear Data Processing code system. Input instructions to visualize nuclear data with NJOY are also provided.

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...

Study of steam condensation at sub-atmospheric pressure: setting a basic research using MELCOR code

NASA Astrophysics Data System (ADS)

Manfredini, A.; Mazzini, M.

2017-11-01

One of the most serious accidents that can occur in the experimental nuclear fusion reactor ITER is the break of one of the headers of the refrigeration system of the first wall of the Tokamak. This results in water-steam mixture discharge in vacuum vessel (VV), with consequent pressurization of this container. To prevent the pressure in the VV exceeds 150 KPa absolute, a system discharges the steam inside a suppression pool, at an absolute pressure of 4.2 kPa. The computer codes used to analyze such incident (eg. RELAP 5 or MELCOR) are not validated experimentally for such conditions. Therefore, we planned a basic research, in order to have experimental data useful to validate the heat transfer correlations used in these codes. After a thorough literature search on this topic, ACTA, in collaboration with the staff of ITER, defined the experimental matrix and performed the design of the experimental apparatus. For the thermal-hydraulic design of the experiments, we executed a series of calculations by MELCOR. This code, however, was used in an unconventional mode, with the development of models suited respectively to low and high steam flow-rate tests. The article concludes with a discussion of the placement of experimental data within the map featuring the phenomenon characteristics, showing the importance of the new knowledge acquired, particularly in the case of chugging.

KEWPIE2: A cascade code for the study of dynamical decay of excited nuclei

NASA Astrophysics Data System (ADS)

Lü, Hongliang; Marchix, Anthony; Abe, Yasuhisa; Boilley, David

2016-03-01

KEWPIE-a cascade code devoted to investigating the dynamical decay of excited nuclei, specially designed for treating very low probability events related to the synthesis of super-heavy nuclei formed in fusion-evaporation reactions-has been improved and rewritten in C++ programming language to become KEWPIE2. The current version of the code comprises various nuclear models concerning the light-particle emission, fission process and statistical properties of excited nuclei. General features of the code, such as the numerical scheme and the main physical ingredients, are described in detail. Some typical calculations having been performed in the present paper clearly show that theoretical predictions are generally in accordance with experimental data. Furthermore, since the values of some input parameters cannot be determined neither theoretically nor experimentally, a sensibility analysis is presented. To this end, we systematically investigate the effects of using different parameter values and reaction models on the final results. As expected, in the case of heavy nuclei, the fission process has the most crucial role to play in theoretical predictions. This work would be essential for numerical modeling of fusion-evaporation reactions.

Impacts-BRC (below regulatory concern): The microcomputer version

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

Campbell, J.E.; O'Neal, B.L.

1989-01-01

The IMPACTS-BRC computer code was designed for use by the Nuclear Regulatory Commission and industry to evaluate petitions to classify specific waste streams as below regulatory concern (BRC). The code provides a capability for calculating radiation doses to a maximal individual, critical group, and the general population as a result of transportation, treatment, disposal, and post-disposal activities involving low level radioactive waste. Since IMPACTS-BRC is expected to be widely used, the code has been adapted for use on a microcomputer. The microcomputer version of the code provides several features that simplify its use and broaden its applicability. These features includemore » (1) a menu-driven environment, (2) an input editor to simplify creation and editing of input files, (3) default input values and help screens to guide the user in analyzing a particular problem, (4) the ability to perform both parametric studies and Monte Carlo analysis to examine uncertainties, and (5) interactive graphics and statistics output. This paper describes the microcomputer version of IMPACTS-BRC and illustrates its use through an example application. 5 refs., 5 figs., 3 tabs.« less

Assessment of the TRACE Reactor Analysis Code Against Selected PANDA Transient Data

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

Zavisca, M.; Ghaderi, M.; Khatib-Rahbar, M.

2006-07-01

The TRACE (TRAC/RELAP Advanced Computational Engine) code is an advanced, best-estimate thermal-hydraulic program intended to simulate the transient behavior of light-water reactor systems, using a two-fluid (steam and water, with non-condensable gas), seven-equation representation of the conservation equations and flow-regime dependent constitutive relations in a component-based model with one-, two-, or three-dimensional elements, as well as solid heat structures and logical elements for the control system. The U.S. Nuclear Regulatory Commission is currently supporting the development of the TRACE code and its assessment against a variety of experimental data pertinent to existing and evolutionary reactor designs. This paper presents themore » results of TRACE post-test prediction of P-series of experiments (i.e., tests comprising the ISP-42 blind and open phases) conducted at the PANDA large-scale test facility in 1990's. These results show reasonable agreement with the reported test results, indicating good performance of the code and relevant underlying thermal-hydraulic and heat transfer models. (authors)« less

Determination of the NPP Kr\\vsko spent fuel decay heat

NASA Astrophysics Data System (ADS)

Kromar, Marjan; Kurinčič, Bojan

2017-07-01

Nuclear fuel is designed to support fission process in a reactor core. Some of the isotopes, formed during the fission, decay and produce decay heat and radiation. Accurate knowledge of the nuclide inventory producing decay heat is important after reactor shut down, during the fuel storage and subsequent reprocessing or disposal. In this paper possibility to calculate the fuel isotopic composition and determination of the fuel decay heat with the Serpent code is investigated. Serpent is a well-known Monte Carlo code used primarily for the calculation of the neutron transport in a reactor. It has been validated for the burn-up calculations. In the calculation of the fuel decay heat different set of isotopes is important than in the neutron transport case. Comparison with the Origen code is performed to verify that the Serpent is taking into account all isotopes important to assess the fuel decay heat. After the code validation, a sensitivity study is carried out. Influence of several factors such as enrichment, fuel temperature, moderator temperature (density), soluble boron concentration, average power, burnable absorbers, and burnup is analyzed.

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

Jordan, Amy B.; Zyvoloski, George Anthony; Weaver, Douglas James

The simulation work presented in this report supports DOE-NE Used Fuel Disposition Campaign (UFDC) goals related to the development of drift scale in-situ field testing of heat-generating nuclear waste (HGNW) in salt formations. Numerical code verification and validation is an important part of the lead-up to field testing, allowing exploration of potential heater emplacement designs, monitoring locations, and perhaps most importantly the ability to predict heat and mass transfer around an evolving test. Such predictions are crucial for the design and location of sampling and monitoring that can be used to validate our understanding of a drift scale test thatmore » is likely to span several years.« less

The Initial Atmospheric Transport (IAT) Code: Description and Validation

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

Morrow, Charles W.; Bartel, Timothy James

The Initial Atmospheric Transport (IAT) computer code was developed at Sandia National Laboratories as part of their nuclear launch accident consequences analysis suite of computer codes. The purpose of IAT is to predict the initial puff/plume rise resulting from either a solid rocket propellant or liquid rocket fuel fire. The code generates initial conditions for subsequent atmospheric transport calculations. The Initial Atmospheric Transfer (IAT) code has been compared to two data sets which are appropriate to the design space of space launch accident analyses. The primary model uncertainties are the entrainment coefficients for the extended Taylor model. The Titan 34Dmore » accident (1986) was used to calibrate these entrainment settings for a prototypic liquid propellant accident while the recent Johns Hopkins University Applied Physics Laboratory (JHU/APL, or simply APL) large propellant block tests (2012) were used to calibrate the entrainment settings for prototypic solid propellant accidents. North American Meteorology (NAM )formatted weather data profiles are used by IAT to determine the local buoyancy force balance. The IAT comparisons for the APL solid propellant tests illustrate the sensitivity of the plume elevation to the weather profiles; that is, the weather profile is a dominant factor in determining the plume elevation. The IAT code performed remarkably well and is considered validated for neutral weather conditions.« less

Calculation of Local Stress and Fatigue Resistance due to Thermal Stratification on Pressurized Surge Line Pipe

NASA Astrophysics Data System (ADS)

Bandriyana, B.; Utaja

2010-06-01

Thermal stratification introduces thermal shock effect which results in local stress and fatique problems that must be considered in the design of nuclear power plant components. Local stress and fatique calculation were performed on the Pressurize Surge Line piping system of the Pressurize Water Reactor of the Nuclear Power Plant. Analysis was done on the operating temperature between 177 to 343° C and the operating pressure of 16 MPa (160 Bar). The stagnant and transient condition with two kinds of stratification model has been evaluated by the two dimensional finite elements method using the ANSYS program. Evaluation of fatigue resistance is developed based on the maximum local stress using the ASME standard Code formula. Maximum stress of 427 MPa occurred at the upper side of the top half of hot fluid pipe stratification model in the transient case condition. The evaluation of the fatigue resistance is performed on 500 operating cycles in the life time of 40 years and giving the usage value of 0,64 which met to the design requirement for class 1 of nuclear component. The out surge transient were the most significant case in the localized effects due to thermal stratification.

Recent optimization of the beam-optical characteristics of the 6 MV van de Graaff accelerator for high brightness beams at the iThemba LABS NMP facility

NASA Astrophysics Data System (ADS)

Conradie, J. L.; Eisa, M. E. M.; Celliers, P. J.; Delsink, J. L. G.; Fourie, D. T.; de Villiers, J. G.; Maine, P. M.; Springhorn, K. A.; Pineda-Vargas, C. A.

2005-04-01

With the aim of improving the reliability and stability of the beams delivered to the nuclear microprobe at iThemba LABS, as well as optimization of the beam characteristics along the van de Graaff accelerator beamlines in general, relevant modifications were implemented since the beginning of 2003. The design and layout of the beamlines were revised. The beam-optical characteristics through the accelerator, from the ion source up to the analysing magnet directly after the accelerator, were calculated and the design optimised, using the computer codes TRANSPORT, IGUN and TOSCA. The ion source characteristics and optimal operating conditions were determined on an ion source test bench. The measured optimal emittance for 90% of the beam intensity was about 50π mm mrad for an extraction voltage of 6 kV. These changes allow operation of the Nuclear Microprobe at proton energies in the range 1 MeV-4 MeV with beam intensities of tenths of a pA at the target surface. The capabilities of the nuclear microprobe facility were evaluated in the improved beamline, with particular emphasis to bio-medical samples.

Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS

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

Connaway, H. M.; Lee, C. H.

The neutron transport code PROTEUS has been used to perform preliminary simulations of the Transient Reactor Test Facility (TREAT). TREAT is an experimental reactor designed for the testing of nuclear fuels and other materials under transient conditions. It operated from 1959 to 1994, when it was placed on non-operational standby. The restart of TREAT to support the U.S. Department of Energy’s resumption of transient testing is currently underway. Both single assembly and assembly-homogenized full core models have been evaluated. Simulations were performed using a historic set of WIMS-ANL-generated cross-sections as well as a new set of Serpent-generated cross-sections. To supportmore » this work, further analyses were also performed using additional codes in order to investigate particular aspects of TREAT modeling. DIF3D and the Monte-Carlo codes MCNP and Serpent were utilized in these studies. MCNP and Serpent were used to evaluate the effect of geometry homogenization on the simulation results and to support code-to-code comparisons. New meshes for the PROTEUS simulations were created using the CUBIT toolkit, with additional meshes generated via conversion of selected DIF3D models to support code-to-code verifications. All current analyses have focused on code-to-code verifications, with additional verification and validation studies planned. The analysis of TREAT with PROTEUS-SN is an ongoing project. This report documents the studies that have been performed thus far, and highlights key challenges to address in future work.« less

Two-phase reduced gravity experiments for a space reactor design

NASA Technical Reports Server (NTRS)

Antoniak, Zenen I.

1987-01-01

Future space missions researchers envision using large nuclear reactors with either a single or a two-phase alkali-metal working fluid. The design and analysis of such reactors require state-of-the-art computer codes that can properly treat alkali-metal flow and heat transfer in a reduced-gravity environment. New flow regime maps, models, and correlations are required if the codes are to be successfully applied to reduced-gravity flow and heat transfer. General plans are put forth for the reduced-gravity experiments which will have to be performed, at NASA facilities, with benign fluids. Data from the reduced-gravity experiments with innocuous fluids are to be combined with normal gravity data from two-phase alkali-metal experiments. Because these reduced-gravity experiments will be very basic, and will employ small test loops of simple geometry, a large measure of commonality exists between them and experiments planned by other organizations. It is recommended that a committee be formed to coordinate all ongoing and planned reduced gravity flow experiments.

Does CTCF mediate between nuclear organization and gene expression?

PubMed

Ohlsson, Rolf; Lobanenkov, Victor; Klenova, Elena

2010-01-01

The multifunctional zinc-finger protein CCCTC-binding factor (CTCF) is a very strong candidate for the role of coordinating the expression level of coding sequences with their three-dimensional position in the nucleus, apparently responding to a "code" in the DNA itself. Dynamic interactions between chromatin fibers in the context of nuclear architecture have been implicated in various aspects of genome functions. However, the molecular basis of these interactions still remains elusive and is a subject of intense debate. Here we discuss the nature of CTCF-DNA interactions, the CTCF-binding specificity to its binding sites and the relationship between CTCF and chromatin, and we examine data linking CTCF with gene regulation in the three-dimensional nuclear space. We discuss why these features render CTCF a very strong candidate for the role and propose a unifying model, the "CTCF code," explaining the mechanistic basis of how the information encrypted in DNA may be interpreted by CTCF into diverse nuclear functions.

Nuclear Resonance Fluorescence for Materials Assay

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

Quiter, Brian; Ludewigt, Bernhard; Mozin, Vladimir

This paper discusses the use of nuclear resonance fluorescence (NRF) techniques for the isotopic and quantitative assaying of radioactive material. Potential applications include age-dating of an unknown radioactive source, pre- and post-detonation nuclear forensics, and safeguards for nuclear fuel cycles Examples of age-dating a strong radioactive source and assaying a spent fuel pin are discussed. The modeling work has ben performed with the Monte Carlo radiation transport computer code MCNPX, and the capability to simulate NRF has bee added to the code. Discussed are the limitations in MCNPX's photon transport physics for accurately describing photon scattering processes that are importantmore » contributions to the background and impact the applicability of the NRF assay technique.« less

75 FR 69137 - Southern Nuclear Operating Company Inc. Edwin I. Hatch Nuclear Plant, Unit No. 2 Environmental...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-10

... NUCLEAR REGULATORY COMMISSION [Docket No. 50-366; NRC-2010-0345] Southern Nuclear Operating Company Inc. Edwin I. Hatch Nuclear Plant, Unit No. 2 Environmental Assessment and Finding of No Significant Impact The U.S. Nuclear Regulatory Commission (NRC) is considering the issuance of an exemption from Title 10 of the Code of Federal Regulations, ...

75 FR 73135 - Southern Nuclear Operating Company, Inc. Joseph M. Farley Nuclear Plant, Environmental Assessment...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-29

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-348 and 50-364; NRC-2009-0375] Southern Nuclear Operating Company, Inc. Joseph M. Farley Nuclear Plant, Environmental Assessment and Finding of No Significant Impact The U.S. Nuclear Regulatory Commission (NRC) is considering issuance of an Exemption, pursuant to Title 10 of the Code of Federal...

Improvements to the nuclear model code GNASH for cross section calculations at higher energies

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

Young, P.G.; Chadwick, M.B.

1994-05-01

The nuclear model code GNASH, which in the past has been used predominantly for incident particle energies below 20 MeV, has been modified extensively for calculations at higher energies. The model extensions and improvements are described in this paper, and their significance is illustrated by comparing calculations with experimental data for incident energies up to 160 MeV.

Calculation of the temperature in the container unit with a modified design for the production of {sup 99}Mo at the VVR-Ts research reactor facility (IVV.10M)

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

Kazantsev, A. A., E-mail: kazantsevanatoly@gmail.com; Sergeev, V. V.; Kochnov, O. Yu.

The temperature regime is calculated for two different designs of containers with uranium-bearing material for the upgraded VVR-Ts research reactor facility (IVV.10M). The containers are to be used in the production of {sup 99}Mo. It is demonstrated that the modification of the container design leads to a considerable temperature reduction and an increase in the near-wall boiling margin and allows one to raise the amount of material loaded into the container. The calculations were conducted using the international thermohydraulic contour code TRAC intended to analyze the technical safety of water-cooled nuclear power units.

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

Sublet, J.-Ch., E-mail: jean-christophe.sublet@ukaea.uk; Eastwood, J.W.; Morgan, J.G.

Fispact-II is a code system and library database for modelling activation-transmutation processes, depletion-burn-up, time dependent inventory and radiation damage source terms caused by nuclear reactions and decays. The Fispact-II code, written in object-style Fortran, follows the evolution of material irradiated by neutrons, alphas, gammas, protons, or deuterons, and provides a wide range of derived radiological output quantities to satisfy most needs for nuclear applications. It can be used with any ENDF-compliant group library data for nuclear reactions, particle-induced and spontaneous fission yields, and radioactive decay (including but not limited to TENDL-2015, ENDF/B-VII.1, JEFF-3.2, JENDL-4.0u, CENDL-3.1 processed into fine-group-structure files, GEFY-5.2more » and UKDD-16), as well as resolved and unresolved resonance range probability tables for self-shielding corrections and updated radiological hazard indices. The code has many novel features including: extension of the energy range up to 1 GeV; additional neutron physics including self-shielding effects, temperature dependence, thin and thick target yields; pathway analysis; and sensitivity and uncertainty quantification and propagation using full covariance data. The latest ENDF libraries such as TENDL encompass thousands of target isotopes. Nuclear data libraries for Fispact-II are prepared from these using processing codes PREPRO, NJOY and CALENDF. These data include resonance parameters, cross sections with covariances, probability tables in the resonance ranges, PKA spectra, kerma, dpa, gas and radionuclide production and energy-dependent fission yields, supplemented with all 27 decay types. All such data for the five most important incident particles are provided in evaluated data tables. The Fispact-II simulation software is described in detail in this paper, together with the nuclear data libraries. The Fispact-II system also includes several utility programs for code-use optimisation, visualisation and production of secondary radiological quantities. Included in the paper are summaries of results from the suite of verification and validation reports available with the code.« less

FISPACT-II: An Advanced Simulation System for Activation, Transmutation and Material Modelling

NASA Astrophysics Data System (ADS)

Sublet, J.-Ch.; Eastwood, J. W.; Morgan, J. G.; Gilbert, M. R.; Fleming, M.; Arter, W.

2017-01-01

Fispact-II is a code system and library database for modelling activation-transmutation processes, depletion-burn-up, time dependent inventory and radiation damage source terms caused by nuclear reactions and decays. The Fispact-II code, written in object-style Fortran, follows the evolution of material irradiated by neutrons, alphas, gammas, protons, or deuterons, and provides a wide range of derived radiological output quantities to satisfy most needs for nuclear applications. It can be used with any ENDF-compliant group library data for nuclear reactions, particle-induced and spontaneous fission yields, and radioactive decay (including but not limited to TENDL-2015, ENDF/B-VII.1, JEFF-3.2, JENDL-4.0u, CENDL-3.1 processed into fine-group-structure files, GEFY-5.2 and UKDD-16), as well as resolved and unresolved resonance range probability tables for self-shielding corrections and updated radiological hazard indices. The code has many novel features including: extension of the energy range up to 1 GeV; additional neutron physics including self-shielding effects, temperature dependence, thin and thick target yields; pathway analysis; and sensitivity and uncertainty quantification and propagation using full covariance data. The latest ENDF libraries such as TENDL encompass thousands of target isotopes. Nuclear data libraries for Fispact-II are prepared from these using processing codes PREPRO, NJOY and CALENDF. These data include resonance parameters, cross sections with covariances, probability tables in the resonance ranges, PKA spectra, kerma, dpa, gas and radionuclide production and energy-dependent fission yields, supplemented with all 27 decay types. All such data for the five most important incident particles are provided in evaluated data tables. The Fispact-II simulation software is described in detail in this paper, together with the nuclear data libraries. The Fispact-II system also includes several utility programs for code-use optimisation, visualisation and production of secondary radiological quantities. Included in the paper are summaries of results from the suite of verification and validation reports available with the code.

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

NASA Astrophysics Data System (ADS)

Qvist, Staffan Alexander

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

Integral nuclear data validation using experimental spent nuclear fuel compositions

DOE PAGES

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

2017-07-19

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

Integral nuclear data validation using experimental spent nuclear fuel compositions

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

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

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

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

High Efficiency Nuclear Power Plants using Liquid Fluoride Thorium Reactor Technology

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

2009-01-01

An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITS of 950 K and 1200 K are presented. Power plant performance data were obtained for TITS ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo -generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

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...

QRAP: A numerical code for projected (Q)uasiparticle (RA)ndom (P)hase approximation

NASA Astrophysics Data System (ADS)

Samana, A. R.; Krmpotić, F.; Bertulani, C. A.

2010-06-01

A computer code for quasiparticle random phase approximation - QRPA and projected quasiparticle random phase approximation - PQRPA models of nuclear structure is explained in details. The residual interaction is approximated by a simple δ-force. An important application of the code consists in evaluating nuclear matrix elements involved in neutrino-nucleus reactions. As an example, cross sections for 56Fe and 12C are calculated and the code output is explained. The application to other nuclei and the description of other nuclear and weak decay processes are also discussed. Program summaryTitle of program: QRAP ( Quasiparticle RAndom Phase approximation) Computers: The code has been created on a PC, but also runs on UNIX or LINUX machines Operating systems: WINDOWS or UNIX Program language used: Fortran-77 Memory required to execute with typical data: 16 Mbytes of RAM memory and 2 MB of hard disk space No. of lines in distributed program, including test data, etc.: ˜ 8000 No. of bytes in distributed program, including test data, etc.: ˜ 256 kB Distribution format: tar.gz Nature of physical problem: The program calculates neutrino- and antineutrino-nucleus cross sections as a function of the incident neutrino energy, and muon capture rates, using the QRPA or PQRPA as nuclear structure models. Method of solution: The QRPA, or PQRPA, equations are solved in a self-consistent way for even-even nuclei. The nuclear matrix elements for the neutrino-nucleus interaction are treated as the beta inverse reaction of odd-odd nuclei as function of the transfer momentum. Typical running time: ≈ 5 min on a 3 GHz processor for Data set 1.

Verification of BWR Turbine Skyshine Dose with the MCNP5 Code Based on an Experiment Made at SHIMANE Nuclear Power Station

NASA Astrophysics Data System (ADS)

Tayama, Ryuichi; Wakasugi, Kenichi; Kawanaka, Ikunori; Kadota, Yoshinobu; Murakami, Yasuhiro

We measured the skyshine dose from turbine buildings at Shimane Nuclear Power Station Unit 1 (NS-1) and Unit 2 (NS-2), and then compared it with the dose calculated with the Monte Carlo transport code MCNP5. The skyshine dose values calculated with the MCNP5 code agreed with the experimental data within a factor of 2.8, when the roof of the turbine building was precisely modeled. We concluded that our MCNP5 calculation was valid for BWR turbine skyshine dose evaluation.

Three-dimensional time-dependent STAR reactor kinetics analyses coupled with RETRAN and MCPWR system response

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

Feltus, M.A.

1989-11-01

The operation of a nuclear power plant must be regularly supported by various reactor dynamics and thermal-hydraulic analyses, which may include final safety analysis report (FSAR) design-basis calculations, and conservative and best-estimate analyses. The development and improvement of computer codes and analysis methodologies provide many advantages, including the ability to evaluate the effect of modeling simplifications and assumptions made in previous reactor kinetics and thermal-hydraulic calculations. This paper describes the results of using the RETRAN, MCPWR, and STAR codes in a tandem, predictive-corrective manner for three pressurized water reactor (PWR) transients: (a) loss of feedwater (LOF) anticipated transient without scrammore » (ATWS), (b) station blackout ATWS, and (c) loss of total reactor coolant system (RCS) flow with a scram.« less

Posttest analysis of international standard problem 10 using RELAP4/MOD7. [PWR

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

Hsu, M.; Davis, C.B.; Peterson, A.C. Jr.

RELAP4/MOD7, a best estimate computer code for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This paper evaluates the capability of RELAP4/MOD7 to calculate refill/reflood phenomena. This evaluation uses the data of International Standard Problem 10, which is based on West Germany's KWU PKL refill/reflood experiment K9A. The PKL test facility represents a typical West German four-loop, 1300 MW pressurized water reactor (PWR) in reduced scale while maintaining prototypical volume-to-power ratio. The PKL facility was designed to specifically simulate the refill/reflood phase of amore » hypothetical loss-of-coolant accident (LOCA).« less

Integral experiments on thorium assemblies with D-T neutron source

NASA Astrophysics Data System (ADS)

Liu, Rong; Yang, Yiwei; Feng, Song; Zheng, Lei; Lai, Caifeng; Lu, Xinxin; Wang, Mei; Jiang, Li

2017-09-01

To validate nuclear data and code in the neutronics design of a hybrid reactor with thorium, integral experiments in two kinds of benchmark thorium assemblies with a D-T fusion neutron source have been performed. The one kind of 1D assemblies consists of polyethylene and depleted uranium shells. The other kind of 2D assemblies consists of three thorium oxide cylinders. The capture reaction rates, fission reaction rates, and (n, 2n) reaction rates in 232Th in the assemblies are measured by ThO2 foils. The leakage neutron spectra from the ThO2 cylinders are measured by a liquid scintillation detector. The experimental uncertainties in all the results are analyzed. The measured results are compared to the calculated ones with MCNP code and ENDF/B-VII.0 library data.

Neutron flux and power in RTP core-15

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

Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Zin, Muhammad Rawi Md; Usang, Mark Dennis

PUSPATI TRIGA Reactor achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutron flux distribution of TRIGA core. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core withmore » literally no physical approximation. The consistency and accuracy of the developed RTP MCNP model was established by comparing calculations to the available experimental results and TRIGLAV code calculation.« less

75 FR 3761 - Southern Nuclear Operating Company, Inc., Edwin I. Hatch Nuclear Plant, Units 1 and 2...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-22

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-321 and 50-366; NRC-2010-0024] Southern Nuclear Operating Company, Inc., Edwin I. Hatch Nuclear Plant, Units 1 and 2; Environmental Assessment and Finding of No Significant Impact The U.S. Nuclear Regulatory Commission (NRC) is considering issuance of an Exemption, pursuant to Title 10 of the Code of...

High-performance superconductors for Fusion Nuclear Science Facility

DOE PAGES

Zhai, Yuhu; Kessel, Chuck; Barth, Christian; ...

2016-11-09

High-performance superconducting magnets play an important role in the design of the next step large-scale, high-field fusion reactors such as the fusion nuclear science facility (FNSF) and the spherical tokamak (ST) pilot plant beyond ITER. Here, Princeton Plasma Physics Laboratory is currently leading the design studies of the FNSF and the ST pilot plant study. ITER, which is under construction in the south of France, utilizes the state-of-the-art low temperature superconducting magnet technology based on the cable-in-conduit conductor design, where over a thousand multifilament Nb 3Sn superconducting strands are twisted together to form a high-current-carrying cable inserted into a steelmore » jacket for coil windings. We present design options of the high-performance superconductors in the winding pack for the FNSF toroidal field magnet system based on the toroidal field radial build from the system code. For the low temperature superconductor options, the advanced J cNb 3Sn RRP strands (J c > 1000 A/mm 2 at 16 T, 4 K) from Oxford Superconducting Technology are under consideration. For the high-temperature superconductor options, the rectangular-shaped high-current HTS cable made of stacked YBCO tapes will be considered to validate feasibility of TF coil winding pack design for the ST-FNSF magnets.« less

High-performance superconductors for Fusion Nuclear Science Facility

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

Zhai, Yuhu; Kessel, Chuck; Barth, Christian

High-performance superconducting magnets play an important role in the design of the next step large-scale, high-field fusion reactors such as the fusion nuclear science facility (FNSF) and the spherical tokamak (ST) pilot plant beyond ITER. Here, Princeton Plasma Physics Laboratory is currently leading the design studies of the FNSF and the ST pilot plant study. ITER, which is under construction in the south of France, utilizes the state-of-the-art low temperature superconducting magnet technology based on the cable-in-conduit conductor design, where over a thousand multifilament Nb 3Sn superconducting strands are twisted together to form a high-current-carrying cable inserted into a steelmore » jacket for coil windings. We present design options of the high-performance superconductors in the winding pack for the FNSF toroidal field magnet system based on the toroidal field radial build from the system code. For the low temperature superconductor options, the advanced J cNb 3Sn RRP strands (J c > 1000 A/mm 2 at 16 T, 4 K) from Oxford Superconducting Technology are under consideration. For the high-temperature superconductor options, the rectangular-shaped high-current HTS cable made of stacked YBCO tapes will be considered to validate feasibility of TF coil winding pack design for the ST-FNSF magnets.« less

Computational Nuclear Physics and Post Hartree-Fock Methods

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

Lietz, Justin; Sam, Novario; Hjorth-Jensen, M.

We present a computational approach to infinite nuclear matter employing Hartree-Fock theory, many-body perturbation theory and coupled cluster theory. These lectures are closely linked with those of chapters 9, 10 and 11 and serve as input for the correlation functions employed in Monte Carlo calculations in chapter 9, the in-medium similarity renormalization group theory of dense fermionic systems of chapter 10 and the Green's function approach in chapter 11. We provide extensive code examples and benchmark calculations, allowing thereby an eventual reader to start writing her/his own codes. We start with an object-oriented serial code and end with discussions onmore » strategies for porting the code to present and planned high-performance computing facilities.« 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

A novel nuclear genetic code alteration in yeasts and the evolution of codon reassignment in eukaryotes

PubMed Central

Mühlhausen, Stefanie; Findeisen, Peggy; Plessmann, Uwe; Urlaub, Henning; Kollmar, Martin

2016-01-01

The genetic code is the cellular translation table for the conversion of nucleotide sequences into amino acid sequences. Changes to the meaning of sense codons would introduce errors into almost every translated message and are expected to be highly detrimental. However, reassignment of single or multiple codons in mitochondria and nuclear genomes, although extremely rare, demonstrates that the code can evolve. Several models for the mechanism of alteration of nuclear genetic codes have been proposed (including “codon capture,” “genome streamlining,” and “ambiguous intermediate” theories), but with little resolution. Here, we report a novel sense codon reassignment in Pachysolen tannophilus, a yeast related to the Pichiaceae. By generating proteomics data and using tRNA sequence comparisons, we show that Pachysolen translates CUG codons as alanine and not as the more usual leucine. The Pachysolen tRNACAG is an anticodon-mutated tRNAAla containing all major alanine tRNA recognition sites. The polyphyly of the CUG-decoding tRNAs in yeasts is best explained by a tRNA loss driven codon reassignment mechanism. Loss of the CUG-tRNA in the ancient yeast is followed by gradual decrease of respective codons and subsequent codon capture by tRNAs whose anticodon is not part of the aminoacyl-tRNA synthetase recognition region. Our hypothesis applies to all nuclear genetic code alterations and provides several testable predictions. We anticipate more codon reassignments to be uncovered in existing and upcoming genome projects. PMID:27197221

MEASUREMENTS OF NEUTRON SPECTRA IN 0.8-GEV AND 1.6-GEV PROTON-IRRADIATED<2 OF 2>NA THICK TARGETS

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

Titarenko, Y. E.; Batyaev, V. F.; Zhivun, V. M.

2001-01-01

Measurements of neutron spectra in W, and Na targets irradiated by 0.8 GeV and 1.6 GeV protons are presented. Measurements were made by the TOF techniques using the proton beam from ITEP U-10 synchrotron. Neutrons were detected with BICRON-511 liquid scintillator-based detectors. The neutron detection efficiency was calculated via the SCINFUL and CECIL codes. The W results are compared with the similar data obtained elsewhere. The measured neutron spectra are compared with the LAHET and CEM2k code simulations results. Attempt is made to explain some observed disagreements between experiments and simulations. The presented results are of interest both in termsmore » of nuclear data buildup and as a benchmark of the up-to-date predictive power of the simulation codes used in designing the hybrid accelerator-driven system (ADS) facilities with sodium-cooled tungsten targets.« less

A comparative study of history-based versus vectorized Monte Carlo methods in the GPU/CUDA environment for a simple neutron eigenvalue problem

NASA Astrophysics Data System (ADS)

Liu, Tianyu; Du, Xining; Ji, Wei; Xu, X. George; Brown, Forrest B.

2014-06-01

For nuclear reactor analysis such as the neutron eigenvalue calculations, the time consuming Monte Carlo (MC) simulations can be accelerated by using graphics processing units (GPUs). However, traditional MC methods are often history-based, and their performance on GPUs is affected significantly by the thread divergence problem. In this paper we describe the development of a newly designed event-based vectorized MC algorithm for solving the neutron eigenvalue problem. The code was implemented using NVIDIA's Compute Unified Device Architecture (CUDA), and tested on a NVIDIA Tesla M2090 GPU card. We found that although the vectorized MC algorithm greatly reduces the occurrence of thread divergence thus enhancing the warp execution efficiency, the overall simulation speed is roughly ten times slower than the history-based MC code on GPUs. Profiling results suggest that the slow speed is probably due to the memory access latency caused by the large amount of global memory transactions. Possible solutions to improve the code efficiency are discussed.

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.

Existing Fortran interfaces to Trilinos in preparation for exascale ForTrilinos development

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

Evans, Katherine J.; Young, Mitchell T.; Collins, Benjamin S.

This report summarizes the current state of Fortran interfaces to the Trilinos library within several key applications of the Exascale Computing Program (ECP), with the aim of informing developers about strategies to develop ForTrilinos, an exascale-ready, Fortran interface software package within Trilinos. The two software projects assessed within are the DOE Office of Science's Accelerated Climate Model for Energy (ACME) atmosphere component, CAM, and the DOE Office of Nuclear Energy's core-simulator portion of VERA, a nuclear reactor simulation code. Trilinos is an object-oriented, C++ based software project, and spans a collection of algorithms and other enabling technologies such as uncertaintymore » quantification and mesh generation. To date, Trilinos has enabled these codes to achieve large-scale simulation results, however the simulation needs of CAM and VERA-CS will approach exascale over the next five years. A Fortran interface to Trilinos that enables efficient use of programming models and more advanced algorithms is necessary. Where appropriate, the needs of the CAM and VERA-CS software to achieve their simulation goals are called out specifically. With this report, a design document and execution plan for ForTrilinos development can proceed.« less

Nuclear Resonance Fluorescence for Materials Assay

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

Quiter, Brian J.; Ludewigt, Bernhard; Mozin, Vladimir

This paper discusses the use of nuclear resonance fluorescence (NRF) techniques for the isotopic and quantitative assaying of radioactive material. Potential applications include age-dating of an unknown radioactive source, pre- and post-detonation nuclear forensics, and safeguards for nuclear fuel cycles Examples of age-dating a strong radioactive source and assaying a spent fuel pin are discussed. The modeling work has ben performed with the Monte Carlo radiation transport computer code MCNPX, and the capability to simulate NRF has bee added to the code. Discussed are the limitations in MCNPX?s photon transport physics for accurately describing photon scattering processes that are importantmore » contributions to the background and impact the applicability of the NRF assay technique.« less

Mechanical Analysis of W78/88-1 Life Extension Program Warhead Design Options

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

Spencer, Nathan

2014-09-01

Life Extension Program (LEP) is a program to repair/replace components of nuclear weapons to ensure the ability to meet military requirements. The W78/88-1 LEP encompasses the modernization of two major nuclear weapon reentry systems into an interoperable warhead. Several design concepts exist to provide different options for robust safety and security themes, maximum non-nuclear commonality, and cost. Simulation is one capability used to evaluate the mechanical performance of the designs in various operational environments, plan for system and component qualification efforts, and provide insight into the survivability of the warhead in environments that are not currently testable. The simulation effortsmore » use several Sandia-developed tools through the Advanced Simulation and Computing program, including Cubit for mesh generation, the DART Model Manager, SIERRA codes running on the HPC TLCC2 platforms, DAKOTA, and ParaView. Several programmatic objectives were met using the simulation capability including: (1) providing early environmental specification estimates that may be used by component designers to understand the severity of the loads their components will need to survive, (2) providing guidance for load levels and configurations for subassembly tests intended to represent operational environments, and (3) recommending design options including modified geometry and material properties. These objectives were accomplished through regular interactions with component, system, and test engineers while using the laboratory's computational infrastructure to effectively perform ensembles of simulations. Because NNSA has decided to defer the LEP program, simulation results are being documented and models are being archived for future reference. However, some advanced and exploratory efforts will continue to mature key technologies, using the results from these and ongoing simulations for design insights, test planning, and model validation.« less

75 FR 11575 - James A. Fitzpatrick Nuclear Power Plant Environmental Assessment and Finding of No Significant...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-11

... Power Plant Environmental Assessment and Finding of No Significant Impact The U.S. Nuclear Regulatory... Code of Federal Regulations (10 CFR), Appendix R, ``Fire Protection Program for Nuclear Power...), for the operation of the James A. FitzPatrick Nuclear Power Plant (JAFNPP) located in Oswego County...

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 ...

Path Toward a Unified Geometry for Radiation Transport

NASA Astrophysics Data System (ADS)

Lee, Kerry

The Direct Accelerated Geometry for Radiation Analysis and Design (DAGRAD) element of the RadWorks Project under Advanced Exploration Systems (AES) within the Space Technology Mission Directorate (STMD) of NASA will enable new designs and concepts of operation for radiation risk assessment, mitigation and protection. This element is designed to produce a solution that will allow NASA to calculate the transport of space radiation through complex CAD models using the state-of-the-art analytic and Monte Carlo radiation transport codes. Due to the inherent hazard of astronaut and spacecraft exposure to ionizing radiation in low-Earth orbit (LEO) or in deep space, risk analyses must be performed for all crew vehicles and habitats. Incorporating these analyses into the design process can minimize the mass needed solely for radiation protection. Transport of the radiation fields as they pass through shielding and body materials can be simulated using Monte Carlo techniques or described by the Boltzmann equation, which is obtained by balancing changes in particle fluxes as they traverse a small volume of material with the gains and losses caused by atomic and nuclear collisions. Deterministic codes that solve the Boltzmann transport equation, such as HZETRN (high charge and energy transport code developed by NASA LaRC), are generally computationally faster than Monte Carlo codes such as FLUKA, GEANT4, MCNP(X) or PHITS; however, they are currently limited to transport in one dimension, which poorly represents the secondary light ion and neutron radiation fields. NASA currently uses HZETRN space radiation transport software, both because it is computationally efficient and because proven methods have been developed for using this software to analyze complex geometries. Although Monte Carlo codes describe the relevant physics in a fully three-dimensional manner, their computational costs have thus far prevented their widespread use for analysis of complex CAD models, leading to the creation and maintenance of toolkit specific simplistic geometry models. The work presented here builds on the Direct Accelerated Geometry Monte Carlo (DAGMC) toolkit developed for use with the Monte Carlo N-Particle (MCNP) transport code. The work-flow for doing radiation transport on CAD models using MCNP and FLUKA has been demonstrated and the results of analyses on realistic spacecraft/habitats will be presented. Future work is planned that will further automate this process and enable the use of multiple radiation transport codes on identical geometry models imported from CAD. This effort will enhance the modeling tools used by NASA to accurately evaluate the astronaut space radiation risk and accurately determine the protection provided by as-designed exploration mission vehicles and habitats.

Quick reproduction of blast-wave flow-field properties of nuclear, TNT, and ANFO explosions

NASA Astrophysics Data System (ADS)

Groth, C. P. T.

1986-04-01

In many instances, extensive blast-wave flow-field properties are required in gasdynamics research studies of blast-wave loading and structure response, and in evaluating the effects of explosions on their environment. This report provides a very useful computer code, which can be used in conjunction with the DNA Nuclear Blast Standard subroutines and code, to quickly reconstruct complete and fairly accurate blast-wave data for almost any free-air (spherical) and surface-burst (hemispherical) nuclear, trinitrotoluene (TNT), or ammonium nitrate-fuel oil (ANFO) explosion. This code is capable of computing all of the main flow properties as functions of radius and time, as well as providing additional information regarding air viscosity, reflected shock-wave properties, and the initial decay of the flow properties just behind the shock front. Both spatial and temporal distributions of the major blast-wave flow properties are also made readily available. Finally, provisions are also included in the code to provide additional information regarding the peak or shock-front flow properties over a range of radii, for a specific explosion of interest.

Optimization of beam shaping assembly based on D-T neutron generator and dose evaluation for BNCT

NASA Astrophysics Data System (ADS)

Naeem, Hamza; Chen, Chaobin; Zheng, Huaqing; Song, Jing

2017-04-01

The feasibility of developing an epithermal neutron beam for a boron neutron capture therapy (BNCT) facility based on a high intensity D-T fusion neutron generator (HINEG) and using the Monte Carlo code SuperMC (Super Monte Carlo simulation program for nuclear and radiation process) is proposed in this study. The Monte Carlo code SuperMC is used to determine and optimize the final configuration of the beam shaping assembly (BSA). The optimal BSA design in a cylindrical geometry which consists of a natural uranium sphere (14 cm) as a neutron multiplier, AlF3 and TiF3 as moderators (20 cm each), Cd (1 mm) as a thermal neutron filter, Bi (5 cm) as a gamma shield, and Pb as a reflector and collimator to guide neutrons towards the exit window. The epithermal neutron beam flux of the proposed model is 5.73 × 109 n/cm2s, and other dosimetric parameters for the BNCT reported by IAEA-TECDOC-1223 have been verified. The phantom dose analysis shows that the designed BSA is accurate, efficient and suitable for BNCT applications. Thus, the Monte Carlo code SuperMC is concluded to be capable of simulating the BSA and the dose calculation for BNCT, and high epithermal flux can be achieved using proposed BSA.

Long non-coding RNA nuclear paraspeckle assembly transcript 1 inhibits the apoptosis of retina Müller cells after diabetic retinopathy through regulating miR-497/brain-derived neurotrophic factor axis.

PubMed

Li, Xiu-Juan

2018-05-01

The role of long non-coding RNA in diabetic retinopathy, a serious complication of diabetes mellitus, has attracted increasing attention in recent years. The purpose of this study was to explore whether long non-coding RNA nuclear paraspeckle assembly transcript 1 was involved in the context of diabetic retinopathy and its underlying mechanisms. Our results revealed that nuclear paraspeckle assembly transcript 1 was significantly downregulated in the retina of diabetes mellitus rats. Meanwhile, miR-497 was significantly increased in diabetes mellitus rats' retina and high glucose-treated Müller cells, but brain-derived neurotrophic factor was increased. We also found that high glucose-induced apoptosis of Müller cells was accompanied by the significant downregulation of nuclear paraspeckle assembly transcript 1 in vitro. Further study demonstrated that high glucose-promoted Müller cells apoptosis through downregulating nuclear paraspeckle assembly transcript 1 and downregulated nuclear paraspeckle assembly transcript 1 mediated this effect via negative regulating miR-497. Moreover, brain-derived neurotrophic factor was negatively regulated by miR-497 and associated with the apoptosis of Müller cells under high glucose. Our results suggested that under diabetic conditions, downregulated nuclear paraspeckle assembly transcript 1 decreased the expression of brain-derived neurotrophic factor through elevating miR-497, thereby promoting Müller cells apoptosis and aggravating diabetic retinopathy.

The ENSDF Java Package

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

Sonzogni, A.A.

2005-05-24

A package of computer codes has been developed to process and display nuclear structure and decay data stored in the ENSDF (Evaluated Nuclear Structure Data File) library. The codes were written in an object-oriented fashion using the java language. This allows for an easy implementation across multiple platforms as well as deployment on web pages. The structure of the different java classes that make up the package is discussed as well as several different implementations.

Hexagonal Uniformly Redundant Arrays (HURAs) for scintillator based coded aperture neutron imaging

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

Gamage, K.A.A.; Zhou, Q.

2015-07-01

A series of Monte Carlo simulations have been conducted, making use of the EJ-426 neutron scintillator detector, to investigate the potential of using hexagonal uniformly redundant arrays (HURAs) for scintillator based coded aperture neutron imaging. This type of scintillator material has a low sensitivity to gamma rays, therefore, is of particular use in a system with a source that emits both neutrons and gamma rays. The simulations used an AmBe source, neutron images have been produced using different coded-aperture materials (boron- 10, cadmium-113 and gadolinium-157) and location error has also been estimated. In each case the neutron image clearly showsmore » the location of the source with a relatively small location error. Neutron images with high resolution can be easily used to identify and locate nuclear materials precisely in nuclear security and nuclear decommissioning applications. (authors)« less

Axial deformed solution of the Skyrme-Hartree-Fock-Bogolyubov equations using the transformed harmonic oscillator Basis

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

Perez, R. Navarro; Schunck, N.; Lasseri, R.

2017-03-09

HFBTHO is a physics computer code that is used to model the structure of the nucleus. It is an implementation of the nuclear energy Density Functional Theory (DFT), where the energy of the nucleus is obtained by integration over space of some phenomenological energy density, which is itself a functional of the neutron and proton densities. In HFBTHO, the energy density derives either from the zero-range Dkyrme or the finite-range Gogny effective two-body interaction between nucleons. Nuclear superfluidity is treated at the Hartree-Fock-Bogoliubov (HFB) approximation, and axial-symmetry of the nuclear shape is assumed. This version is the 3rd release ofmore » the program; the two previous versions were published in Computer Physics Communications [1,2]. The previous version was released at LLNL under GPL 3 Open Source License and was given release code LLNL-CODE-573953.« less

New French Regulation for NPPs and Code Consequences

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

Faidy, Claude

2006-07-01

On December 2005, the French regulator issued a new regulation for French nuclear power plants, in particular for pressure equipment (PE). This regulation need first to agree with non-nuclear PE regulation and add to that some specific requirements, in particular radiation protection requirements. Different advantages are in these proposal, it's more qualitative risk oriented and it's an important link with non-nuclear industry. Only few components are nuclear specific. But, the general philosophy of the existing Codes (RCC-M [15], KTA [16] or ASME [17]) have to be improved. For foreign Codes, it's plan to define the differences in the user specifications.more » In parallel to that, a new safety classification has been developed by French utility. The consequences is the need to cross all these specifications to define a minimum quality level for each components or systems. In the same time a new concept has been developed to replace the well known 'Leak Before Break methodology': the 'Break Exclusion' methodology. This paper will summarize the key aspects of these different topics. (authors)« less

In vitro cell irradiation systems based on 210Po alpha source: construction and characterisation

NASA Technical Reports Server (NTRS)

Szabo, J.; Feher, I.; Palfalvi, J.; Balashazy, I.; Dam, A. M.; Polonyi, I.; Bogdandi, E. N.

2002-01-01

One way of studying the risk to human health of low-level radiation exposure is to make biological experiments on living cell cultures. Two 210Po alpha-particle emitting devices, with 0.5 and 100 MBq activity, were designed and constructed to perform such experiments irradiating monolayers of cells. Estimates of dose rate at the cell surface were obtained from measurements by a PIPS alpha-particle spectrometer and from calculations by the SRIM 2000, Monte Carlo charged particle transport code. Particle fluence area distributions were measured by solid state nuclear track detectors. The design and dosimetric characterisation of the devices are discussed. c2002 Elsevier Science Ltd. All rights reserved.

Nuclear data uncertainty propagation by the XSUSA method in the HELIOS2 lattice code

NASA Astrophysics Data System (ADS)

Wemple, Charles; Zwermann, Winfried

2017-09-01

Uncertainty quantification has been extensively applied to nuclear criticality analyses for many years and has recently begun to be applied to depletion calculations. However, regulatory bodies worldwide are trending toward requiring such analyses for reactor fuel cycle calculations, which also requires uncertainty propagation for isotopics and nuclear reaction rates. XSUSA is a proven methodology for cross section uncertainty propagation based on random sampling of the nuclear data according to covariance data in multi-group representation; HELIOS2 is a lattice code widely used for commercial and research reactor fuel cycle calculations. This work describes a technique to automatically propagate the nuclear data uncertainties via the XSUSA approach through fuel lattice calculations in HELIOS2. Application of the XSUSA methodology in HELIOS2 presented some unusual challenges because of the highly-processed multi-group cross section data used in commercial lattice codes. Currently, uncertainties based on the SCALE 6.1 covariance data file are being used, but the implementation can be adapted to other covariance data in multi-group structure. Pin-cell and assembly depletion calculations, based on models described in the UAM-LWR Phase I and II benchmarks, are performed and uncertainties in multiplication factor, reaction rates, isotope concentrations, and delayed-neutron data are calculated. With this extension, it will be possible for HELIOS2 users to propagate nuclear data uncertainties directly from the microscopic cross sections to subsequent core simulations.

A Novel Multi-Scale Domain Overlapping CFD/STH Coupling Methodology for Multi-Dimensional Flows Relevant to Nuclear Applications

NASA Astrophysics Data System (ADS)

Grunloh, Timothy P.

The objective of this dissertation is to develop a 3-D domain-overlapping coupling method that leverages the superior flow field resolution of the Computational Fluid Dynamics (CFD) code STAR-CCM+ and the fast execution of the System Thermal Hydraulic (STH) code TRACE to efficiently and accurately model thermal hydraulic transport properties in nuclear power plants under complex conditions of regulatory and economic importance. The primary contribution is the novel Stabilized Inertial Domain Overlapping (SIDO) coupling method, which allows for on-the-fly correction of TRACE solutions for local pressures and velocity profiles inside multi-dimensional regions based on the results of the CFD simulation. The method is found to outperform the more frequently-used domain decomposition coupling methods. An STH code such as TRACE is designed to simulate large, diverse component networks, requiring simplifications to the fluid flow equations for reasonable execution times. Empirical correlations are therefore required for many sub-grid processes. The coarse grids used by TRACE diminish sensitivity to small scale geometric details such as Reactor Pressure Vessel (RPV) internals. A CFD code such as STAR-CCM+ uses much finer computational meshes that are sensitive to the geometric details of reactor internals. In turbulent flows, it is infeasible to fully resolve the flow solution, but the correlations used to model turbulence are at a low level. The CFD code can therefore resolve smaller scale flow processes. The development of a 3-D coupling method was carried out with the intention of improving predictive capabilities of transport properties in the downcomer and lower plenum regions of an RPV in reactor safety calculations. These regions are responsible for the multi-dimensional mixing effects that determine the distribution at the core inlet of quantities with reactivity implications, such as fluid temperature and dissolved neutron absorber concentration.

48 CFR 2001.104-1 - Publication and code arrangement.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 48 Federal Acquisition Regulations System 6 2011-10-01 2011-10-01 false Publication and code arrangement. 2001.104-1 Section 2001.104-1 Federal Acquisition Regulations System NUCLEAR REGULATORY... 2001.104-1 Publication and code arrangement. (a) The NRCAR and its subsequent changes are: (1...

Development of Northeast Asia Nuclear Power Plant Accident Simulator.

PubMed

Kim, Juyub; Kim, Juyoul; Po, Li-Chi Cliff

2017-06-15

A conclusion from the lessons learned after the March 2011 Fukushima Daiichi accident was that Korea needs a tool to estimate consequences from a major accident that could occur at a nuclear power plant located in a neighboring country. This paper describes a suite of computer-based codes to be used by Korea's nuclear emergency response staff for training and potentially operational support in Korea's national emergency preparedness and response program. The systems of codes, Northeast Asia Nuclear Accident Simulator (NANAS), consist of three modules: source-term estimation, atmospheric dispersion prediction and dose assessment. To quickly assess potential doses to the public in Korea, NANAS includes specific reactor data from the nuclear power plants in China, Japan and Taiwan. The completed simulator is demonstrated using data for a hypothetical release. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Test case for VVER-1000 complex modeling using MCU and ATHLET

NASA Astrophysics Data System (ADS)

Bahdanovich, R. B.; Bogdanova, E. V.; Gamtsemlidze, I. D.; Nikonov, S. P.; Tikhomirov, G. V.

2017-01-01

The correct modeling of processes occurring in the fuel core of the reactor is very important. In the design and operation of nuclear reactors it is necessary to cover the entire range of reactor physics. Very often the calculations are carried out within the framework of only one domain, for example, in the framework of structural analysis, neutronics (NT) or thermal hydraulics (TH). However, this is not always correct, as the impact of related physical processes occurring simultaneously, could be significant. Therefore it is recommended to spend the coupled calculations. The paper provides test case for the coupled neutronics-thermal hydraulics calculation of VVER-1000 using the precise neutron code MCU and system engineering code ATHLET. The model is based on the fuel assembly (type 2M). Test case for calculation of power distribution, fuel and coolant temperature, coolant density, etc. has been developed. It is assumed that the test case will be used for simulation of VVER-1000 reactor and in the calculation using other programs, for example, for codes cross-verification. The detailed description of the codes (MCU, ATHLET), geometry and material composition of the model and an iterative calculation scheme is given in the paper. Script in PERL language was written to couple the codes.

Preliminary results of Malaysian nuclear agency plasma focus (MNA-PF) as a slow focus mode device for argon and deuterium filling gas in correlation with Lee model code

NASA Astrophysics Data System (ADS)

Zin, M. F. M.; Baijan, A. H.; Damideh, V.; Hashim, S. A.; Sabri, R. M.

2017-03-01

In this work, preliminary results of MNA-PF device as a Slow Focus Mode device are presented. Four different kinds of Rogowski Coils which have been designed and constructed for dI/dt signals measurements show that response frequency of Rogowski Coil can affect signal time resolution and delay which can change the discharge circuit inductance. Experimental results for 10 to 20 mbar Deuterium and 0.5 mbar to 6 mbar Argon which are captured by 630 MHz Rogowski coil in correlation with Lee Model Code are presented. Proper current fitting using Lee Model Code shows that the speed factor for MNA-PF device working with 13 mbar Deuterium is 30 kA/cm.torr1/2 at 14 kV which indicates that the device is operating at slow focus mode. Model parameters fm and fmr predicted by Lee Model Code during current fitting for 13 mbar Deuterium at 14kV were 0.025 and 0.31 respectively. Microspec-4 Neutron Detector was used to obtain the dose rate which was found to be maximum at 4.78 uSv/hr and also the maximum neutron yield calculated from Lee Model Code is 7.5E+03 neutron per shot.

Performance assessment methodology and preliminary results for low-level radioactive waste disposal in Taiwan.

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

Arnold, Bill Walter; Chang, Fu-lin; Mattie, Patrick D.

2006-02-01

Sandia National Laboratories (SNL) and Taiwan's Institute for Nuclear Energy Research (INER) have teamed together to evaluate several candidate sites for Low-Level Radioactive Waste (LLW) disposal in Taiwan. Taiwan currently has three nuclear power plants, with another under construction. Taiwan also has a research reactor, as well as medical and industrial wastes to contend with. Eventually the reactors will be decomissioned. Operational and decommissioning wastes will need to be disposed in a licensed disposal facility starting in 2014. Taiwan has adopted regulations similar to the US Nuclear Regulatory Commission's (NRC's) low-level radioactive waste rules (10 CFR 61) to govern themore » disposal of LLW. Taiwan has proposed several potential sites for the final disposal of LLW that is now in temporary storage on Lanyu Island and on-site at operating nuclear power plants, and for waste generated in the future through 2045. The planned final disposal facility will have a capacity of approximately 966,000 55-gallon drums. Taiwan is in the process of evaluating the best candidate site to pursue for licensing. Among these proposed sites there are basically two disposal concepts: shallow land burial and cavern disposal. A representative potential site for shallow land burial is located on a small island in the Taiwan Strait with basalt bedrock and interbedded sedimentary rocks. An engineered cover system would be constructed to limit infiltration for shallow land burial. A representative potential site for cavern disposal is located along the southeastern coast of Taiwan in a tunnel system that would be about 500 to 800 m below the surface. Bedrock at this site consists of argillite and meta-sedimentary rocks. Performance assessment analyses will be performed to evaluate future performance of the facility and the potential dose/risk to exposed populations. Preliminary performance assessment analyses will be used in the site-selection process and to aid in design of the disposal system. Final performance assessment analyses will be used in the regulatory process of licensing a site. The SNL/INER team has developed a performance assessment methodology that is used to simulate processes associated with the potential release of radionuclides to evaluate these sites. The following software codes are utilized in the performance assessment methodology: GoldSim (to implement a probabilistic analysis that will explicitly address uncertainties); the NRC's Breach, Leach, and Transport - Multiple Species (BLT-MS) code (to simulate waste-container degradation, waste-form leaching, and transport through the host rock); the Finite Element Heat and Mass Transfer code (FEHM) (to simulate groundwater flow and estimate flow velocities); the Hydrologic Evaluation of Landfill performance Model (HELP) code (to evaluate infiltration through the disposal cover); the AMBER code (to evaluate human health exposures); and the NRC's Disposal Unit Source Term -- Multiple Species (DUST-MS) code (to screen applicable radionuclides). Preliminary results of the evaluations of the two disposal concept sites are presented.« less

Space Radiation Transport Codes: A Comparative Study for Galactic Cosmic Rays Environment

NASA Astrophysics Data System (ADS)

Tripathi, Ram; Wilson, John W.; Townsend, Lawrence W.; Gabriel, Tony; Pinsky, Lawrence S.; Slaba, Tony

For long duration and/or deep space human missions, protection from severe space radiation exposure is a challenging design constraint and may be a potential limiting factor. The space radiation environment consists of galactic cosmic rays (GCR), solar particle events (SPE), trapped radiation, and includes ions of all the known elements over a very broad energy range. These ions penetrate spacecraft materials producing nuclear fragments and secondary particles that damage biological tissues, microelectronic devices, and materials. In deep space missions, where the Earth's magnetic field does not provide protection from space radiation, the GCR environment is significantly enhanced due to the absence of geomagnetic cut-off and is a major component of radiation exposure. Accurate risk assessments critically depend on the accuracy of the input information as well as radiation transport codes used, and so systematic verification of codes is necessary. In this study, comparisons are made between the deterministic code HZETRN2006 and the Monte Carlo codes HETC-HEDS and FLUKA for an aluminum shield followed by a water target exposed to the 1977 solar minimum GCR spectrum. Interaction and transport of high charge ions present in GCR radiation environment provide a more stringent constraint in the comparison of the codes. Dose, dose equivalent and flux spectra are compared; details of the comparisons will be discussed, and conclusions will be drawn for future directions.

Report of the Secretary of Defense Task Force on DoD Nuclear Weapons Management. Phase 1. The Air Force’s Nuclear Mission

DTIC Science & Technology

2008-09-01

under- resourced. • Missile transfer vans /warhead transfer vans require upgrades. • ICBM weapon system test sets under-funded; the coding system...Air Force’s Nuclear Mission D-1 Appendix D. Current B-52 Basing Status Barksdale AFB, LA 64 B-52Hs Minot AFB, ND 27 B-52Hs Edwards AFB, CA 3...Barksdale – 64 B-52s 2 BW (ACC) 15 TF; 24 CC; 7 BAI 53 WG (ACC) 2 Test Coded 917 WG (AFRC) 8 CC; 1 BAI 7 Unfunded AR Edwards - 3 B-52s 412 TW 2 Test

Solving Problems With SINDA/FLUINT

NASA Technical Reports Server (NTRS)

2002-01-01

SINDA/FLUINT, the NASA standard software system for thermohydraulic analysis, provides computational simulation of interacting thermal and fluid effects in designs modeled as heat transfer and fluid flow networks. The product saves time and money by making the user's design process faster and easier, and allowing the user to gain a better understanding of complex systems. The code is completely extensible, allowing the user to choose the features, accuracy and approximation levels, and outputs. Users can also add their own customizations as needed to handle unique design tasks or to automate repetitive tasks. Applications for SINDA/FLUINT include the pharmaceutical, petrochemical, biomedical, electronics, and energy industries. The system has been used to simulate nuclear reactors, windshield wipers, and human windpipes. In the automotive industry, it simulates the transient liquid/vapor flows within air conditioning systems.

Generalized Nuclear Data

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

Conlin, Jeremy

2017-03-15

This software is code related to reading/writing/manipulating nuclear data in the Generalized Nuclear Data (GND) format, a new format for sharing nuclear data among institutions. In addition to the software and its documentation, notes and documentation from the WPEC Subgroup 43 will be included. WPEC Subgroup 43 is an international committee charged with creating the API for the GND format.

76 FR 26320 - Entergy Operations, Inc.; Biweekly Notice; Notice of Issuance of Amendment to Facility Operating...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-05-06

... Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone (301..., Office of Nuclear Reactor Regulation. [FR Doc. 2011-11107 Filed 5-5-11; 8:45 am] BILLING CODE 7590-01-P ... NUCLEAR REGULATORY COMMISSION [NRC-2011-0071; Docket No. 50-382] Entergy Operations, Inc...

76 FR 13676 - Exelon Generation Company, LLC; Notice of Withdrawal of Application for Amendment to Facility...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-14

.... Brown, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555..., Office of Nuclear Reactor Regulation. [FR Doc. 2011-5756 Filed 3-11-11; 8:45 am] BILLING CODE 7590-01-P ... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-373 and 50-374; NRC-2011-0051] Exelon Generation...

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 ...

49 CFR 176.2 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-10-01

..., spillage, or other accident. INF cargo means packaged irradiated nuclear fuel, plutonium or high-level... Irradiated Nuclear Fuel, Plutonium and High-Level Radioactive Wastes on Board Ships” (INF Code) contained in...

High-Energy Activation Simulation Coupling TENDL and SPACS with FISPACT-II

NASA Astrophysics Data System (ADS)

Fleming, Michael; Sublet, Jean-Christophe; Gilbert, Mark

2018-06-01

To address the needs of activation-transmutation simulation in incident-particle fields with energies above a few hundred MeV, the FISPACT-II code has been extended to splice TENDL standard ENDF-6 nuclear data with extended nuclear data forms. The JENDL-2007/HE and HEAD-2009 libraries were processed for FISPACT-II and used to demonstrate the capabilities of the new code version. Tests of the libraries and comparisons against both experimental yield data and the most recent intra-nuclear cascade model results demonstrate that there is need for improved nuclear data libraries up to and above 1 GeV. Simulations on lead targets show that important radionuclides, such as 148Gd, can vary by more than an order of magnitude where more advanced models find agreement within the experimental uncertainties.

Kiwi: An Evaluated Library of Uncertainties in Nuclear Data and Package for Nuclear Sensitivity Studies

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

Pruet, J

2007-06-23

This report describes Kiwi, a program developed at Livermore to enable mature studies of the relation between imperfectly known nuclear physics and uncertainties in simulations of complicated systems. Kiwi includes a library of evaluated nuclear data uncertainties, tools for modifying data according to these uncertainties, and a simple interface for generating processed data used by transport codes. As well, Kiwi provides access to calculations of k eigenvalues for critical assemblies. This allows the user to check implications of data modifications against integral experiments for multiplying systems. Kiwi is written in python. The uncertainty library has the same format and directorymore » structure as the native ENDL used at Livermore. Calculations for critical assemblies rely on deterministic and Monte Carlo codes developed by B division.« less

A novel nuclear genetic code alteration in yeasts and the evolution of codon reassignment in eukaryotes.

PubMed

Mühlhausen, Stefanie; Findeisen, Peggy; Plessmann, Uwe; Urlaub, Henning; Kollmar, Martin

2016-07-01

The genetic code is the cellular translation table for the conversion of nucleotide sequences into amino acid sequences. Changes to the meaning of sense codons would introduce errors into almost every translated message and are expected to be highly detrimental. However, reassignment of single or multiple codons in mitochondria and nuclear genomes, although extremely rare, demonstrates that the code can evolve. Several models for the mechanism of alteration of nuclear genetic codes have been proposed (including "codon capture," "genome streamlining," and "ambiguous intermediate" theories), but with little resolution. Here, we report a novel sense codon reassignment in Pachysolen tannophilus, a yeast related to the Pichiaceae. By generating proteomics data and using tRNA sequence comparisons, we show that Pachysolen translates CUG codons as alanine and not as the more usual leucine. The Pachysolen tRNACAG is an anticodon-mutated tRNA(Ala) containing all major alanine tRNA recognition sites. The polyphyly of the CUG-decoding tRNAs in yeasts is best explained by a tRNA loss driven codon reassignment mechanism. Loss of the CUG-tRNA in the ancient yeast is followed by gradual decrease of respective codons and subsequent codon capture by tRNAs whose anticodon is not part of the aminoacyl-tRNA synthetase recognition region. Our hypothesis applies to all nuclear genetic code alterations and provides several testable predictions. We anticipate more codon reassignments to be uncovered in existing and upcoming genome projects. © 2016 Mühlhausen et al.; Published by Cold Spring Harbor Laboratory Press.

MitoNuc: a database of nuclear genes coding for mitochondrial proteins. Update 2002.

PubMed

Attimonelli, Marcella; Catalano, Domenico; Gissi, Carmela; Grillo, Giorgio; Licciulli, Flavio; Liuni, Sabino; Santamaria, Monica; Pesole, Graziano; Saccone, Cecilia

2002-01-01

Mitochondria, besides their central role in energy metabolism, have recently been found to be involved in a number of basic processes of cell life and to contribute to the pathogenesis of many degenerative diseases. All functions of mitochondria depend on the interaction of nuclear and organelle genomes. Mitochondrial genomes have been extensively sequenced and analysed and data have been collected in several specialised databases. In order to collect information on nuclear coded mitochondrial proteins we developed MitoNuc, a database containing detailed information on sequenced nuclear genes coding for mitochondrial proteins in Metazoa. The MitoNuc database can be retrieved through SRS and is available via the web site http://bighost.area.ba.cnr.it/mitochondriome where other mitochondrial databases developed by our group, the complete list of the sequenced mitochondrial genomes, links to other mitochondrial sites and related information, are available. The MitoAln database, related to MitoNuc in the previous release, reporting the multiple alignments of the relevant homologous protein coding regions, is no longer supported in the present release. In order to keep the links among entries in MitoNuc from homologous proteins, a new field in the database has been defined: the cluster identifier, an alpha numeric code used to identify each cluster of homologous proteins. A comment field derived from the corresponding SWISS-PROT entry has been introduced; this reports clinical data related to dysfunction of the protein. The logic scheme of MitoNuc database has been implemented in the ORACLE DBMS. This will allow the end-users to retrieve data through a friendly interface that will be soon implemented.

MELCOR/CONTAIN LMR Implementation Report. FY14 Progress

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

Humphries, Larry L; Louie, David L.Y.

2014-10-01

This report describes the preliminary implementation of the sodium thermophysical properties and the design documentation for the sodium models of CONTAIN-LMR to be implemented into MELCOR 2.1. In the past year, the implementation included two separate sodium properties from two different sources. The first source is based on the previous work done by Idaho National Laboratory by modifying MELCOR to include liquid lithium equation of state as a working fluid to model the nuclear fusion safety research. To minimize the impact to MELCOR, the implementation of the fusion safety database (FSD) was done by utilizing the detection of the datamore » input file as a way to invoking the FSD. The FSD methodology has been adapted currently for this work, but it may subject modification as the project continues. The second source uses properties generated for the SIMMER code. Preliminary testing and results from this implementation of sodium properties are given. In this year, the design document for the CONTAIN-LMR sodium models, such as the two condensable option, sodium spray fire, and sodium pool fire is being developed. This design document is intended to serve as a guide for the MELCOR implementation. In addition, CONTAIN-LMR code used was based on the earlier version of CONTAIN code. Many physical models that were developed since this early version of CONTAIN may not be captured by the code. Although CONTAIN 2, which represents the latest development of CONTAIN, contains some sodium specific models, which are not complete, the utilizing CONTAIN 2 with all sodium models implemented from CONTAIN-LMR as a comparison code for MELCOR should be done. This implementation should be completed in early next year, while sodium models from CONTAIN-LMR are being integrated into MELCOR. For testing, CONTAIN decks have been developed for verification and validation use.« less

I-NERI Quarterly Technical Report (April 1 to June 30, 2005)

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

Chang Oh; Prof. Hee Cheon NO; Prof. John Lee

2005-06-01

The objective of this Korean/United States/laboratory/university collaboration is to develop new advanced computational methods for safety analysis codes for very-high-temperature gas-cooled reactors (VHTGRs) and numerical and experimental validation of these computer codes. This study consists of five tasks for FY-03: (1) development of computational methods for the VHTGR, (2) theoretical modification of aforementioned computer codes for molecular diffusion (RELAP5/ATHENA) and modeling CO and CO2 equilibrium (MELCOR), (3) development of a state-of-the-art methodology for VHTGR neutronic analysis and calculation of accurate power distributions and decay heat deposition rates, (4) reactor cavity cooling system experiment, and (5) graphite oxidation experiment. Second quartermore » of Year 3: (A) Prof. NO and Kim continued Task 1. As a further plant application of GAMMA code, we conducted two analyses: IAEA GT-MHR benchmark calculation for LPCC and air ingress analysis for PMR 600MWt. The GAMMA code shows comparable peak fuel temperature trend to those of other country codes. The analysis results for air ingress show much different trend from that of previous PBR analysis: later onset of natural circulation and less significant rise in graphite temperature. (B) Prof. Park continued Task 2. We have designed new separate effect test device having same heat transfer area and different diameter and total number of U-bands of air cooling pipe. New design has smaller pressure drop in the air cooling pipe than the previous one as designed with larger diameter and less number of U-bands. With the device, additional experiments have been performed to obtain temperature distributions of the water tank, the surface and the center of cooling pipe on axis. The results will be used to optimize the design of SNU-RCCS. (C) Prof. NO continued Task 3. The experimental work of air ingress is going on without any concern: With nuclear graphite IG-110, various kinetic parameters and reaction rates for the C/CO2 reaction were measured. Then, the rates of C/CO2 reaction were compared to the ones of C/O2 reaction. The rate equation for C/CO2 has been developed. (D) INL added models to RELAP5/ATHENA to cacilate the chemical reactions in a VHTR during an air ingress accident. Limited testing of the models indicate that they are calculating a correct special distribution in gas compositions. (E) INL benchmarked NACOK natural circulation data. (F) Professor Lee et al at the University of Michigan (UM) Task 5. The funding was received from the DOE Richland Office at the end of May and the subcontract paperwork was delivered to the UM on the sixth of June. The objective of this task is to develop a state of the art neutronics model for determining power distributions and decay heat deposition rates in a VHTGR core. Our effort during the reporting period covered reactor physics analysis of coated particles and coupled nuclear-thermal-hydraulic (TH) calculations, together with initial calculations for decay heat deposition rates in the core.« less

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

NONE

This report supplements the final safety evaluation report (FSER) for the System 80+ standard design. The FSER was issued by the US Nuclear Regulatory Commission (NRC) staff as NUREG-1462 in August 1994 to document the NRC staff`s review of the System 80+ design. The System 80+ design was submitted by Asea Brown Boveri-Combustion Engineering (ABB-CE), in accordance with the procedures of Subpart B to Part 52 of Title 10 of the Code of Federal Regulations. This supplement documents the NRC staff`s review of the changes to the System 80+ design documentation since the issuance of the FSER. ABB-CE made thesemore » changes as a result of its review of the System 80+ design details. The NRC staff concludes that the changes to the System 80+ design documentation are acceptable, and that ABB-CE`s application for design certification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the System 80+ design.« less

Transient Approximation of SAFE-100 Heat Pipe Operation

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Reid, Robert S.

2005-01-01

Engineers at Los Alamos National Laboratory (LANL) have designed several heat pipe cooled reactor concepts, ranging in power from 15 kWt to 800 kWt, for both surface power systems and nuclear electric propulsion systems. The Safe, Affordable Fission Engine (SAFE) is now being developed in a collaborative effort between LANL and NASA Marshall Space Flight Center (NASA/MSFC). NASA is responsible for fabrication and testing of non-nuclear, electrically heated modules in the Early Flight Fission Test Facility (EFF-TF) at MSFC. In-core heat pipes must be properly thawed as the reactor power starts. Computational models have been developed to assess the expected operation of a specific heat pipe design during start-up, steady state operation, and shutdown. While computationally intensive codes provide complete, detailed analyses of heat pipe thaw, a relatively simple. concise routine can also be applied to approximate the response of a heat pipe to changes in the evaporator heat transfer rate during start-up and power transients (e.g., modification of reactor power level) with reasonably accurate results. This paper describes a simplified model of heat pipe start-up that extends previous work and compares the results to experimental measurements for a SAFE-100 type heat pipe design.

78 FR 37721 - Approval of American Society of Mechanical Engineers' Code Cases

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-24

...-0359] RIN 3150-AI72 Approval of American Society of Mechanical Engineers' Code Cases AGENCY: Nuclear... mandatory American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code and... Guide'' series. In a notice of proposed rulemaking, ``Approval of American Society of Mechanical...

IEEE 1982. Proceedings of the international conference on cybernetics and society

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

Not Available

1982-01-01

The following topics were dealt with: knowledge-based systems; risk analysis; man-machine interactions; human information processing; metaphor, analogy and problem-solving; manual control modelling; transportation systems; simulation; adaptive and learning systems; biocybernetics; cybernetics; mathematical programming; robotics; decision support systems; analysis, design and validation of models; computer vision; systems science; energy systems; environmental modelling and policy; pattern recognition; nuclear warfare; technological forecasting; artificial intelligence; the Turin shroud; optimisation; workloads. Abstracts of individual papers can be found under the relevant classification codes in this or future issues.

Calculation evaluation of multiplying properties of LWR with thorium fuel

NASA Astrophysics Data System (ADS)

Shamanin, I. V.; Grachev, V. M.; Knyshev, V. V.; Bedenko, S. V.; Novikova, N. G.

2017-01-01

The results of multiplying properties design research of the unit cell and LWR fuel assembly with the high temperature gas-cooled thorium reactor fuel pellet are presented in the work. The calculation evaluation showed the possibility of using thorium in LWR effectively. In this case the amount of fissile isotope is 2.45 times smaller in comparison with the standard loading of LWR. The research and numerical experiments were carried out using the verified accounting code of the program MCU5, modern libraries of evaluated nuclear data and multigroup approximations.

Test Analysis Tools to Ensure Higher Quality of On-Board Real Time Software for Space Applications

NASA Astrophysics Data System (ADS)

Boudillet, O.; Mescam, J.-C.; Dalemagne, D.

2008-08-01

EADS Astrium Space Transportation, in its Les Mureaux premises, is responsible for the French M51 nuclear deterrent missile onboard SW. There was also developed over 1 million of line of code, mostly in ADA, for the Automated Transfer Vehicle (ATV) onboard SW and the flight control SW of the ARIANE5 launcher which has put it into orbit. As part of the ATV SW, ASTRIUM ST has developed the first Category A SW ever qualified for a European space application. To ensure that all these embedded SW have been developed with the highest quality and reliability level, specific development tools have been designed to cover the steps of source code verification, automated validation test or complete target instruction coverage verification. Three of such dedicated tools are presented here.

Analysis of dosimetry from the H.B. Robinson unit 2 pressure vessel benchmark using RAPTOR-M3G and ALPAN

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

Fischer, G.A.

2011-07-01

Document available in abstract form only, full text of document follows: The dosimetry from the H. B. Robinson Unit 2 Pressure Vessel Benchmark is analyzed with a suite of Westinghouse-developed codes and data libraries. The radiation transport from the reactor core to the surveillance capsule and ex-vessel locations is performed by RAPTOR-M3G, a parallel deterministic radiation transport code that calculates high-resolution neutron flux information in three dimensions. The cross-section library used in this analysis is the ALPAN library, an Evaluated Nuclear Data File (ENDF)/B-VII.0-based library designed for reactor dosimetry and fluence analysis applications. Dosimetry is evaluated with the industry-standard SNLRMLmore » reactor dosimetry cross-section data library. (authors)« less

Coupling procedure for TRANSURANUS and KTF codes

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

Jimenez, J.; Alglave, S.; Avramova, M.

2012-07-01

The nuclear industry aims to ensure safe and economic operation of each single fuel rod introduced in the reactor core. This goal is even more challenging nowadays due to the current strategy of going for higher burn-up (fuel cycles of 18 or 24 months) and longer residence time. In order to achieve that goal, fuel modeling is the key to predict the fuel rod behavior and lifetime under thermal and pressure loads, corrosion and irradiation. In this context, fuel performance codes, such as TRANSURANUS, are used to improve the fuel rod design. The modeling capabilities of the above mentioned toolsmore » can be significantly improved if they are coupled with a thermal-hydraulic code in order to have a better description of the flow conditions within the rod bundle. For LWR applications, a good representation of the two phase flow within the fuel assembly is necessary in order to have a best estimate calculation of the heat transfer inside the bundle. In this paper we present the coupling methodology of TRANSURANUS with KTF (Karlsruhe Two phase Flow subchannel code) as well as selected results of the coupling proof of principle. (authors)« less

Monte Carlo simulation of β γ coincidence system using plastic scintillators in 4π geometry

NASA Astrophysics Data System (ADS)

Dias, M. S.; Piuvezam-Filho, H.; Baccarelli, A. M.; Takeda, M. N.; Koskinas, M. F.

2007-09-01

A modified version of a Monte Carlo code called Esquema, developed at the Nuclear Metrology Laboratory in IPEN, São Paulo, Brazil, has been applied for simulating a 4 πβ(PS)-γ coincidence system designed for primary radionuclide standardisation. This system consists of a plastic scintillator in 4 π geometry, for alpha or electron detection, coupled to a NaI(Tl) counter for gamma-ray detection. The response curves for monoenergetic electrons and photons have been calculated previously by Penelope code and applied as input data to code Esquema. The latter code simulates all the disintegration processes, from the precursor nucleus to the ground state of the daughter radionuclide. As a result, the curve between the observed disintegration rate as a function of the beta efficiency parameter can be simulated. A least-squares fit between the experimental activity values and the Monte Carlo calculation provided the actual radioactive source activity, without need of conventional extrapolation procedures. Application of this methodology to 60Co and 133Ba radioactive sources is presented and showed results in good agreement with a conventional proportional counter 4 πβ(PC)-γ coincidence system.

Path Toward a Unifid Geometry for Radiation Transport

NASA Technical Reports Server (NTRS)

Lee, Kerry; Barzilla, Janet; Davis, Andrew; Zachmann

2014-01-01

The Direct Accelerated Geometry for Radiation Analysis and Design (DAGRAD) element of the RadWorks Project under Advanced Exploration Systems (AES) within the Space Technology Mission Directorate (STMD) of NASA will enable new designs and concepts of operation for radiation risk assessment, mitigation and protection. This element is designed to produce a solution that will allow NASA to calculate the transport of space radiation through complex computer-aided design (CAD) models using the state-of-the-art analytic and Monte Carlo radiation transport codes. Due to the inherent hazard of astronaut and spacecraft exposure to ionizing radiation in low-Earth orbit (LEO) or in deep space, risk analyses must be performed for all crew vehicles and habitats. Incorporating these analyses into the design process can minimize the mass needed solely for radiation protection. Transport of the radiation fields as they pass through shielding and body materials can be simulated using Monte Carlo techniques or described by the Boltzmann equation, which is obtained by balancing changes in particle fluxes as they traverse a small volume of material with the gains and losses caused by atomic and nuclear collisions. Deterministic codes that solve the Boltzmann transport equation, such as HZETRN [high charge and energy transport code developed by NASA Langley Research Center (LaRC)], are generally computationally faster than Monte Carlo codes such as FLUKA, GEANT4, MCNP(X) or PHITS; however, they are currently limited to transport in one dimension, which poorly represents the secondary light ion and neutron radiation fields. NASA currently uses HZETRN space radiation transport software, both because it is computationally efficient and because proven methods have been developed for using this software to analyze complex geometries. Although Monte Carlo codes describe the relevant physics in a fully three-dimensional manner, their computational costs have thus far prevented their widespread use for analysis of complex CAD models, leading to the creation and maintenance of toolkit-specific simplistic geometry models. The work presented here builds on the Direct Accelerated Geometry Monte Carlo (DAGMC) toolkit developed for use with the Monte Carlo N-Particle (MCNP) transport code. The workflow for achieving radiation transport on CAD models using MCNP and FLUKA has been demonstrated and the results of analyses on realistic spacecraft/habitats will be presented. Future work is planned that will further automate this process and enable the use of multiple radiation transport codes on identical geometry models imported from CAD. This effort will enhance the modeling tools used by NASA to accurately evaluate the astronaut space radiation risk and accurately determine the protection provided by as-designed exploration mission vehicles and habitats

A journey from nuclear criticality methods to high energy density radflow experiments

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

Urbatsch, Todd James

Los Alamos National Laboratory is a nuclear weapons laboratory supporting our nation's defense. In support of this mission is a high energy-density physics program in which we design and execute experiments to study radiationhydrodynamics phenomena and improve the predictive capability of our largescale multi-physics software codes on our big-iron computers. The Radflow project’s main experimental effort now is to understand why we haven't been able to predict opacities on Sandia National Laboratory's Z-machine. We are modeling an increasing fraction of the Z-machine's dynamic hohlraum to find multi-physics explanations for the experimental results. Further, we are building an entirely different opacitymore » platform on Lawrence Livermore National Laboratory's National Ignition Facility (NIF), which is set to get results early 2017. Will the results match our predictions, match the Z-machine, or give us something entirely different? The new platform brings new challenges such as designing hohlraums and spectrometers. The speaker will recount his history, starting with one-dimensional Monte Carlo nuclear criticality methods in graduate school, radiative transfer methods research and software development for his first 16 years at LANL, and, now, radflow technology and experiments. Who knew that the real world was more than just radiation transport? Experiments aren't easy, but they sure are fun.« less

High and low energy gamma beam dump designs for the gamma beam delivery system at ELI-NP

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

Yasin, Zafar, E-mail: zafar.yasin@eli-np.ro; Matei, Catalin; Ur, Calin A.

The Extreme Light Infrastructure - Nuclear Physics (ELI-NP) is under construction in Magurele, Bucharest, Romania. The facility will use two 10 PW lasers and a high intensity, narrow bandwidth gamma beam for stand-alone and combined laser-gamma experiments. The accurate estimation of particle doses and their restriction within the limits for both personel and general public is very important in the design phase of any nuclear facility. In the present work, Monte Carlo simulations are performed using FLUKA and MCNPX to design 19.4 and 4 MeV gamma beam dumps along with shielding of experimental areas. Dose rate contour plots from both FLUKAmore » and MCNPX along with numerical values of doses in experimental area E8 of the facility are performed. The calculated doses are within the permissible limits. Furthermore, a reasonable agreement between both codes enhances our confidence in using one or both of them for future calculations in beam dump designs, radiation shielding, radioactive inventory, and other calculations releated to radiation protection. Residual dose rates and residual activity calculations are also performed for high-energy beam dump and their effect is negligible in comparison to contributions from prompt radiation.« less

Final Report for Project 13-4791: New Mechanistic Models of Creep-Fatigue Crack Growth Interactions for Advanced High Temperature Reactor Components

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

Kruzic, Jamie J; Siegmund, Thomas; Tomar, Vikas

This project developed and validated a novel, multi-scale, mechanism-based model to quantitatively predict creep-fatigue crack growth and failure for Ni-based Alloy 617 at 800°C. Alloy 617 is a target material for intermediate heat exchangers in Generation IV very high temperature reactor designs, and it is envisioned that this model will aid in the design of safe, long lasting nuclear power plants. The technical effectiveness of the model was shown by demonstrating that experimentally observed crack growth rates can be predicted under both steady state and overload crack growth conditions. Feasibility was considered by incorporating our model into a commercially availablemore » finite element method code, ABAQUS, that is commonly used by design engineers. While the focus of the project was specifically on an alloy targeted for Generation IV nuclear reactors, the benefits to the public are expected to be wide reaching. Indeed, creep-fatigue failure is a design consideration for a wide range of high temperature mechanical systems that rely on Ni-based alloys, including industrial gas power turbines, advanced ultra-super critical steam turbines, and aerospace turbine engines. It is envisioned that this new model can be adapted to a wide range of engineering applications.« less

Analysis Code - Data Analysis in 'Leveraging Multiple Statistical Methods for Inverse Prediction in Nuclear Forensics Applications' (LMSMIPNFA) v. 1.0

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

Lewis, John R

R code that performs the analysis of a data set presented in the paper ‘Leveraging Multiple Statistical Methods for Inverse Prediction in Nuclear Forensics Applications’ by Lewis, J., Zhang, A., Anderson-Cook, C. It provides functions for doing inverse predictions in this setting using several different statistical methods. The data set is a publicly available data set from a historical Plutonium production experiment.

Nuclear Fuel Depletion Analysis Using Matlab Software

NASA Astrophysics Data System (ADS)

Faghihi, F.; Nematollahi, M. R.

Coupled first order IVPs are frequently used in many parts of engineering and sciences. In this article, we presented a code including three computer programs which are joint with the Matlab software to solve and plot the solutions of the first order coupled stiff or non-stiff IVPs. Some engineering and scientific problems related to IVPs are given and fuel depletion (production of the 239Pu isotope) in a Pressurized Water Nuclear Reactor (PWR) are computed by the present code.

10 CFR 820.20 - Purpose and scope.

Code of Federal Regulations, 2013 CFR

2013-01-01

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

10 CFR 820.20 - Purpose and scope.

Code of Federal Regulations, 2012 CFR

2012-01-01

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

10 CFR 820.20 - Purpose and scope.

Code of Federal Regulations, 2011 CFR

2011-01-01

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

10 CFR 820.20 - Purpose and scope.

Code of Federal Regulations, 2014 CFR

2014-01-01

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

FERRET data analysis code

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

Schmittroth, F.

1979-09-01

A documentation of the FERRET data analysis code is given. The code provides a way to combine related measurements and calculations in a consistent evaluation. Basically a very general least-squares code, it is oriented towards problems frequently encountered in nuclear data and reactor physics. A strong emphasis is on the proper treatment of uncertainties and correlations and in providing quantitative uncertainty estimates. Documentation includes a review of the method, structure of the code, input formats, and examples.

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

Harmony, S.C.; Steiner, J.L.; Stumpf, H.J.

The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. As part of the preapplication and eventual design certification process, advanced reactor applicants are required to submit neutronic and thermal-hydraulic safety analyses over a sufficient range of normal operation, transient conditions, and specified accident sequences. Los Alamos is supporting the US Nuclear Regulatory Commission`s preapplication review of the PIUS reactor. A fully one-dimensional modelmore » of the PIUS reactor has been developed for the Transient Reactor Analysis Code, TRACPF1/MOD2. Early in 1992, ABB submitted a Supplemental Information Package describing recent design modifications. An important feature of the PIUS Supplement design was the addition of an active scram system that will function for most transient and accident conditions. A one-dimensional Transient Reactor Analysis Code baseline calculation of the PIUS Supplement design were performed for a break in the main steam line at the outlet nozzle of the loop 3 steam generator. Sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions following a main steam line break. The sensitivity study results provide insights into the robustness of the design.« less

Computer optimization of reactor-thermoelectric space power systems

NASA Technical Reports Server (NTRS)

Maag, W. L.; Finnegan, P. M.; Fishbach, L. H.

1973-01-01

A computer simulation and optimization code that has been developed for nuclear space power systems is described. The results of using this code to analyze two reactor-thermoelectric systems are presented.

RADTRAD: A simplified model for RADionuclide Transport and Removal And Dose estimation

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

Humphreys, S.L.; Miller, L.A.; Monroe, D.K.

1998-04-01

This report documents the RADTRAD computer code developed for the U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Reactor Regulation (NRR) to estimate transport and removal of radionuclides and dose at selected receptors. The document includes a users` guide to the code, a description of the technical basis for the code, the quality assurance and code acceptance testing documentation, and a programmers` guide. The RADTRAD code can be used to estimate the containment release using either the NRC TID-14844 or NUREG-1465 source terms and assumptions, or a user-specified table. In addition, the code can account for a reduction in themore » quantity of radioactive material due to containment sprays, natural deposition, filters, and other natural and engineered safety features. The RADTRAD code uses a combination of tables and/or numerical models of source term reduction phenomena to determine the time-dependent dose at user-specified locations for a given accident scenario. The code system also provides the inventory, decay chain, and dose conversion factor tables needed for the dose calculation. The RADTRAD code can be used to assess occupational radiation exposures, typically in the control room; to estimate site boundary doses; and to estimate dose attenuation due to modification of a facility or accident sequence.« less

DYNSYL: a general-purpose dynamic simulator for chemical processes

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

Patterson, G.K.; Rozsa, R.B.

1978-09-05

Lawrence Livermore Laboratory is conducting a safeguards program for the Nuclear Regulatory Commission. The goal of the Material Control Project of this program is to evaluate material control and accounting (MCA) methods in plants that handle special nuclear material (SNM). To this end we designed and implemented the dynamic chemical plant simulation program DYNSYL. This program can be used to generate process data or to provide estimates of process performance; it simulates both steady-state and dynamic behavior. The MCA methods that may have to be evaluated range from sophisticated on-line material trackers such as Kalman filter estimators, to relatively simplemore » material balance procedures. This report describes the overall structure of DYNSYL and includes some example problems. The code is still in the experimental stage and revision is continuing.« less

Creep-Fatigue Damage Investigation and Modeling of Alloy 617 at High Temperatures

NASA Astrophysics Data System (ADS)

Tahir, Fraaz

The Very High Temperature Reactor (VHTR) is one of six conceptual designs proposed for Generation IV nuclear reactors. Alloy 617, a solid solution strengthened Ni-base superalloy, is currently the primary candidate material for the tubing of the Intermediate Heat Exchanger (IHX) in the VHTR design. Steady-state operation of the nuclear power plant at elevated temperatures leads to creep deformation, whereas loading transients including startup and shutdown generate fatigue. A detailed understanding of the creep-fatigue interaction in Alloy 617 is necessary before it can be considered as a material for nuclear construction in ASME Boiler and Pressure Vessel Code. Current design codes for components undergoing creep-fatigue interaction at elevated temperatures require creep-fatigue testing data covering the entire range from fatigue-dominant to creep-dominant loading. Classical strain-controlled tests, which produce stress relaxation during the hold period, show a saturation in cycle life with increasing hold periods due to the rapid stress-relaxation of Alloy 617 at high temperatures. Therefore, applying longer hold time in these tests cannot generate creep-dominated failure. In this study, uniaxial isothermal creep-fatigue tests with non-traditional loading waveforms were designed and performed at 850 and 950°C, with an objective of generating test data in the creep-dominant regime. The new loading waveforms are hybrid strain-controlled and force-controlled testing which avoid stress relaxation during the creep hold. The experimental data showed varying proportions of creep and fatigue damage, and provided evidence for the inadequacy of the widely-used time fraction rule for estimating creep damage under creep-fatigue conditions. Micro-scale damage features in failed test specimens, such as fatigue cracks and creep voids, were quantified using a Scanning Electron Microscope (SEM) to find a correlation between creep and fatigue damage. Quantitative statistical imaging analysis showed that the microstructural damage features (cracks and voids) are correlated with a new mechanical driving force parameter. The results from this image-based damage analysis were used to develop a phenomenological life-prediction methodology called the effective time fraction approach. Finally, the constitutive creep-fatigue response of the material at 950°C was modeled using a unified viscoplastic model coupled with a damage accumulation model. The simulation results were used to validate an energy-based constitutive life-prediction model, as a mechanistic model for potential component and structure level creep-fatigue analysis.

The use of the SRIM code for calculation of radiation damage induced by neutrons

NASA Astrophysics Data System (ADS)

Mohammadi, A.; Hamidi, S.; Asadabad, Mohsen Asadi

2017-12-01

Materials subjected to neutron irradiation will being evolve to structural changes by the displacement cascades initiated by nuclear reaction. This study discusses a methodology to compute primary knock-on atoms or PKAs information that lead to radiation damage. A program AMTRACK has been developed for assessing of the PKAs information. This software determines the specifications of recoil atoms (using PTRAC card of MCNPX code) and also the kinematics of interactions. The deterministic method was used for verification of the results of (MCNPX+AMTRACK). The SRIM (formely TRIM) code is capable to compute neutron radiation damage. The PKAs information was extracted by AMTRACK program, which can be used as an input of SRIM codes for systematic analysis of primary radiation damage. Then the Bushehr Nuclear Power Plant (BNPP) radiation damage on reactor pressure vessel is calculated.

Strategic need for a multi-purpose thermal hydraulic loop for support of advanced reactor technologies

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

O'Brien, James E.; Sabharwall, Piyush; Yoon, Su -Jong

2014-09-01

This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs)more » at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation needs. The experimental database will guide development of appropriate predictive methods and be available for code verification and validation (V&V) related to these systems.« less

Stochastic Modeling of Radioactive Material Releases

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

Andrus, Jason; Pope, Chad

2015-09-01

Nonreactor nuclear facilities operated under the approval authority of the U.S. Department of Energy use unmitigated hazard evaluations to determine if potential radiological doses associated with design basis events challenge or exceed dose evaluation guidelines. Unmitigated design basis events that sufficiently challenge dose evaluation guidelines or exceed the guidelines for members of the public or workers, merit selection of safety structures, systems, or components or other controls to prevent or mitigate the hazard. Idaho State University, in collaboration with Idaho National Laboratory, has developed a portable and simple to use software application called SODA (Stochastic Objective Decision-Aide) that stochastically calculatesmore » the radiation dose associated with hypothetical radiological material release scenarios. Rather than producing a point estimate of the dose, SODA produces a dose distribution result to allow a deeper understanding of the dose potential. SODA allows users to select the distribution type and parameter values for all of the input variables used to perform the dose calculation. SODA then randomly samples each distribution input variable and calculates the overall resulting dose distribution. In cases where an input variable distribution is unknown, a traditional single point value can be used. SODA was developed using the MATLAB coding framework. The software application has a graphical user input. SODA can be installed on both Windows and Mac computers and does not require MATLAB to function. SODA provides improved risk understanding leading to better informed decision making associated with establishing nuclear facility material-at-risk limits and safety structure, system, or component selection. It is important to note that SODA does not replace or compete with codes such as MACCS or RSAC, rather it is viewed as an easy to use supplemental tool to help improve risk understanding and support better informed decisions. The work was funded through a grant from the DOE Nuclear Safety Research and Development Program.« less

Design of an electron-accelerator-driven compact neutron source for non-destructive assay

NASA Astrophysics Data System (ADS)

Murata, A.; Ikeda, S.; Hayashizaki, N.

2017-09-01

The threat of nuclear and radiological terrorism remains one of the greatest challenges to international security, and the threat is constantly evolving. In order to prevent nuclear terrorism, it is important to avoid unlawful import of nuclear materials, such as uranium and plutonium. Development of technologies for non-destructive measurement, detection and recognition of nuclear materials is essential for control at national borders. At Tokyo Institute of Technology, a compact neutron source system driven by an electron-accelerator has been designed for non-destructive assay (NDA). This system is composed of a combination of an S-band (2.856 GHz) RF-gun, a tungsten target to produce photons by bremsstrahlung, a beryllium target, which is suitable for use in generating neutrons because of the low threshold energy of photonuclear reactions, and a moderator to thermalize the fast neutrons. The advantage of this system can accelerate a short pulse beam with a pulse width less than 1 μs which is difficult to produce by neutron generators. The amounts of photons and neutron produced by electron beams were simulated using the Monte Carlo simulation code PHITS 2.82. When the RF-gun is operated with an average electron beam current of 0.1 mA, it is expected that the neutron intensities are 1.19 × 109 n/s and 9.94 × 109 n/s for incident electron beam energies of 5 MeV and 10 MeV, respectively.

Power control of SAFE reactor using fuzzy logic

NASA Astrophysics Data System (ADS)

Irvine, Claude

2002-01-01

Controlling the 100 kW SAFE (Safe Affordable Fission Engine) reactor consists of design and implementation of a fuzzy logic process control system to regulate dynamic variables related to nuclear system power. The first phase of development concentrates primarily on system power startup and regulation, maintaining core temperature equilibrium, and power profile matching. This paper discusses the experimental work performed in those areas. Nuclear core power from the fuel elements is simulated using resistive heating elements while heat rejection is processed by a series of heat pipes. Both axial and radial nuclear power distributions are determined from neuronic modeling codes. The axial temperature profile of the simulated core is matched to the nuclear power profile by varying the resistance of the heating elements. The SAFE model establishes radial temperature profile equivalence by establishing 32 control zones as the nodal coordinates. Control features also allow for slow warm up, since complete shutoff can occur in the heat pipes if heat-source temperatures drop/rise below a certain minimum value, depending on the specific fluid and gas combination in the heat pipe. The entire system is expected to be self-adaptive, i.e., capable of responding to long-range changes in the space environment. Particular attention in the development of the fuzzy logic algorithm shall ensure that the system process remains at set point, virtually eliminating overshoot on start-up and during in-process disturbances. The controller design will withstand harsh environments and applications where it might come in contact with water, corrosive chemicals, radiation fields, etc. .

A MODEL BUILDING CODE ARTICLE ON FALLOUT SHELTERS WITH RECOMMENDATIONS FOR INCLUSION OF REQUIREMENTS FOR FALLOUT SHELTER CONSTRUCTION IN FOUR NATIONAL MODEL BUILDING CODES.

ERIC Educational Resources Information Center

American Inst. of Architects, Washington, DC.

A MODEL BUILDING CODE FOR FALLOUT SHELTERS WAS DRAWN UP FOR INCLUSION IN FOUR NATIONAL MODEL BUILDING CODES. DISCUSSION IS GIVEN OF FALLOUT SHELTERS WITH RESPECT TO--(1) NUCLEAR RADIATION, (2) NATIONAL POLICIES, AND (3) COMMUNITY PLANNING. FALLOUT SHELTER REQUIREMENTS FOR SHIELDING, SPACE, VENTILATION, CONSTRUCTION, AND SERVICES SUCH AS ELECTRICAL…

10 CFR 50.55a - Codes and standards.

Code of Federal Regulations, 2011 CFR

2011-01-01

... specified in § 50.55, except that each combined license for a boiling or pressurized water-cooled nuclear... boiling or pressurized water-cooled nuclear power facility is subject to the conditions in paragraphs (f... performed. (2) Systems and components of boiling and pressurized water-cooled nuclear power reactors must...

78 FR 9745 - Kewaunee Power Station; Application for Amendment to Facility Operating License

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-11

... FURTHER INFORMATION CONTACT: Karl Feintuch, Project Manager, Office of Nuclear Reactor Regulation, U.S... Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2013-03037 Filed 2-8-13; 8:45 am] BILLING CODE... NUCLEAR REGULATORY COMMISSION [Docket No. 50-305; NRC-2013-0028] Kewaunee Power Station...

The One-Dimensional Cryogenic Implosion Campaign on OMEGA: Modeling, Experiments, and a Statistical Approach to Predict and Understand Direct-Drive Implosions

NASA Astrophysics Data System (ADS)

Betti, R.

2017-10-01

The 1-D campaign on OMEGA is aimed at validating a novel approach to design cryogenic implosion experiments and provide valuable data to improve the accuracy of 1-D physics models. This new design methodology is being tested first on low-convergence, high-adiabat (α 6 to 7) implosions and will subsequently be applied to implosions with increasing convergence up to the level required for a hydro-equivalent demonstration of ignition. This design procedure assumes that the hydrodynamic codes used in implosion designs lack the necessary physics and that measurements of implosion properties are imperfect. It also assumes that while the measurements may have significant systematic errors, the shot-to-shot variations are small and that cryogenic implosion data are reproducible as observed on OMEGA. One of the goals of the 1-D campaign is to find a mapping of the data to the code results and use the mapping relations to design future implosions. In the 1-D campaign, this predictive methodology was used to design eight implosions using a simple two-shock pulse design, leading to pre-shot predictions of yields within 5% and ion temperatures within 4% of the experimental values. These implosions have also produced the highest neutron yield of 1014 in OMEGA cryogenic implosion experiments with an areal density of 100 mg/cm2. Furthermore, the results from this campaign have been used to test the validity of the 1-D physics models used in the radiation-hydrodynamics codes. This material is based upon work supported by the Department of Energy National Nuclear Security Administration under Award Number DENA0001944 and LLNL under Contract DE-AC52-07NA27344. * In collaboration with J.P. Knauer, V. Gopalaswamy, D. Patel, K.M. Woo, K.S. Anderson, A. Bose, A.R. Christopherson, V.Yu. Glebov, F.J. Marshall, S.P. Regan, P.B. Radha, C. Stoeckl, and E.M. Campbell.

Nuclear fuel management optimization using genetic algorithms

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

DeChaine, M.D.; Feltus, M.A.

1995-07-01

The code independent genetic algorithm reactor optimization (CIGARO) system has been developed to optimize nuclear reactor loading patterns. It uses genetic algorithms (GAs) and a code-independent interface, so any reactor physics code (e.g., CASMO-3/SIMULATE-3) can be used to evaluate the loading patterns. The system is compared to other GA-based loading pattern optimizers. Tests were carried out to maximize the beginning of cycle k{sub eff} for a pressurized water reactor core loading with a penalty function to limit power peaking. The CIGARO system performed well, increasing the k{sub eff} after lowering the peak power. Tests of a prototype parallel evaluation methodmore » showed the potential for a significant speedup.« less

Prognostics and Health Management in Nuclear Power Plants: A Review of Technologies and Applications

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

Coble, Jamie B.; Ramuhalli, Pradeep; Bond, Leonard J.

This report reviews the current state of the art of prognostics and health management (PHM) for nuclear power systems and related technology currently applied in field or under development in other technological application areas, as well as key research needs and technical gaps for increased use of PHM in nuclear power systems. The historical approach to monitoring and maintenance in nuclear power plants (NPPs), including the Maintenance Rule for active components and Aging Management Plans for passive components, are reviewed. An outline is given for the technical and economic challenges that make PHM attractive for both legacy plants through Lightmore » Water Reactor Sustainability (LWRS) and new plant designs. There is a general introduction to PHM systems for monitoring, fault detection and diagnostics, and prognostics in other, non-nuclear fields. The state of the art for health monitoring in nuclear power systems is reviewed. A discussion of related technologies that support the application of PHM systems in NPPs, including digital instrumentation and control systems, wired and wireless sensor technology, and PHM software architectures is provided. Appropriate codes and standards for PHM are discussed, along with a description of the ongoing work in developing additional necessary standards. Finally, an outline of key research needs and opportunities that must be addressed in order to support the application of PHM in legacy and new NPPs is presented.« less

SASS-1--SUBASSEMBLY STRESS SURVEY CODE

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

Friedrich, C.M.

1960-01-01

SASS-1, an IBM-704 FORTRAN code, calculates pressure, thermal, and combined stresses in a nuclear reactor core subassembly. In addition to cross- section stresses, the code calculates axial shear stresses needed to keep plane cross sections plane under axial variations of temperature. The input and output nomenclature, arrangement, and formats are described. (B.O.G.)

78 FR 37848 - ASME Code Cases Not Approved for Use

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-24

...The U.S. Nuclear Regulatory Commission (NRC) is issuing for public comment draft regulatory guide (DG), DG-1233, ``ASME Code Cases not Approved for Use.'' This regulatory guide lists the American Society of Mechanical Engineers (ASME) Code Cases that the NRC has determined not to be acceptable for use on a generic basis.

An automated approach to magnetic divertor configuration design

NASA Astrophysics Data System (ADS)

Blommaert, M.; Dekeyser, W.; Baelmans, M.; Gauger, N. R.; Reiter, D.

2015-01-01

Automated methods based on optimization can greatly assist computational engineering design in many areas. In this paper an optimization approach to the magnetic design of a nuclear fusion reactor divertor is proposed and applied to a tokamak edge magnetic configuration in a first feasibility study. The approach is based on reduced models for magnetic field and plasma edge, which are integrated with a grid generator into one sensitivity code. The design objective chosen here for demonstrative purposes is to spread the divertor target heat load as much as possible over the entire target area. Constraints on the separatrix position are introduced to eliminate physically irrelevant magnetic field configurations during the optimization cycle. A gradient projection method is used to ensure stable cost function evaluations during optimization. The concept is applied to a configuration with typical Joint European Torus (JET) parameters and it automatically provides plausible configurations with reduced heat load.

Globalization of ASME Nuclear Codes and Standards

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

Swayne, Rick; Erler, Bryan A.

2006-07-01

With the globalization of the nuclear industry, it is clear that the reactor suppliers are based in many countries around the world (such as United States, France, Japan, Canada, South Korea, South Africa) and they will be marketing their reactors to many countries around the world (such as US, China, South Korea, France, Canada, Finland, Taiwan). They will also be fabricating their components in many different countries around the world. With this situation, it is clear that the requirements of ASME Nuclear Codes and Standards need to be adjusted to accommodate the regulations, fabricating processes, and technology of various countriesmore » around the world. It is also very important for the American Society of Mechanical Engineers (ASME) to be able to assure that products meeting the applicable ASME Code requirements will provide the same level of safety and quality assurance as those products currently fabricated under the ASME accreditation process. To do this, many countries are in the process of establishing or changing their regulations, and it is important for ASME to interface with the appropriate organizations in those countries, in order to ensure there is effective use of ASME Codes and standards around the world. (authors)« less

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

BRISC is a developmental prototype for a nextgeneration “systems-level” integrated performance and safety code (IPSC) for nuclear reactors. Its development served to demonstrate how a lightweight multi-physics coupling approach can be used to tightly couple the physics models in several different physics codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled “burner” nuclear reactor. For example, the RIO Fluid Flow and Heat transfer code developed at Sandia (SNL: Chris Moen, Dept. 08005) is used in BRISC to model fluid flow and heat transfer, as well as conduction heat transfermore » in solids. Because BRISC is a prototype, its most practical application is as a foundation or starting point for developing a true production code. The sub-codes and the associated models and correlations currently employed within BRISC were chosen to cover the required application space and demonstrate feasibility, but were not optimized or validated against experimental data within the context of their use in BRISC.« less

A new code for modelling the near field diffusion releases from the final disposal of nuclear waste

NASA Astrophysics Data System (ADS)

Vopálka, D.; Vokál, A.

2003-01-01

The canisters with spent nuclear fuel produced during the operation of WWER reactors at the Czech power plants are planned, like in other countries, to be disposed of in an underground repository. Canisters will be surrounded by compacted bentonite that will retard the migration of safety-relevant radionuclides into the host rock. A new code that enables the modelling of the critical radionuclides transport from the canister through the bentonite layer in the cylindrical geometry was developed. The code enables to solve the diffusion equation for various types of initial and boundary conditions by means of the finite difference method and to take into account the non-linear shape of the sorption isotherm. A comparison of the code reported here with code PAGODA, which is based on analytical solution of the transport equation, was made for the actinide chain 4N+3 that includes 239Pu. A simple parametric study of the releases of 239Pu, 129I, and 14C into geosphere is discussed.

PHASE I MATERIALS PROPERTY DATABASE DEVELOPMENT FOR ASME CODES AND STANDARDS

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

Ren, Weiju; Lin, Lianshan

2013-01-01

To support the ASME Boiler and Pressure Vessel Codes and Standard (BPVC) in modern information era, development of a web-based materials property database is initiated under the supervision of ASME Committee on Materials. To achieve efficiency, the project heavily draws upon experience from development of the Gen IV Materials Handbook and the Nuclear System Materials Handbook. The effort is divided into two phases. Phase I is planned to deliver a materials data file warehouse that offers a depository for various files containing raw data and background information, and Phase II will provide a relational digital database that provides advanced featuresmore » facilitating digital data processing and management. Population of the database will start with materials property data for nuclear applications and expand to data covering the entire ASME Code and Standards including the piping codes as the database structure is continuously optimized. The ultimate goal of the effort is to establish a sound cyber infrastructure that support ASME Codes and Standards development and maintenance.« less

MCNP capabilities for nuclear well logging calculations

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

Forster, R.A.; Little, R.C.; Briesmeister, J.F.

The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. This paper discusses how the general-purpose continuous-energy Monte Carlo code MCNP ({und M}onte {und C}arlo {und n}eutron {und p}hoton), part of the LARTCS, provides a computational predictive capability for many applications of interest to the nuclear well logging community. The generalized three-dimensional geometry of MCNP is well suited for borehole-tool models. SABRINA, another component of the LARTCS, is a graphics code that can be used to interactively create a complex MCNP geometry. Users can define many source and tallymore » characteristics with standard MCNP features. The time-dependent capability of the code is essential when modeling pulsed sources. Problems with neutrons, photons, and electrons as either single particle or coupled particles can be calculated with MCNP. The physics of neutron and photon transport and interactions is modeled in detail using the latest available cross-section data.« less

The Paucity Problem: Where Have All the Space Reactor Experiments Gone?

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

Bess, John D.; Marshall, Margaret A.

2016-10-01

The Handbooks of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) together contain a plethora of documented and evaluated experiments essential in the validation of nuclear data, neutronics codes, and modeling of various nuclear systems. Unfortunately, only a minute selection of handbook data (twelve evaluations) are of actual experimental facilities and mockups designed specifically for space nuclear research. There is a paucity problem, such that the multitude of space nuclear experimental activities performed in the past several decades have yet to be recovered and made available in such detail that themore » international community could benefit from these valuable historical research efforts. Those experiments represent extensive investments in infrastructure, expertise, and cost, as well as constitute significantly valuable resources of data supporting past, present, and future research activities. The ICSBEP and IRPhEP were established to identify and verify comprehensive sets of benchmark data; evaluate the data, including quantification of biases and uncertainties; compile the data and calculations in a standardized format; and formally document the effort into a single source of verified benchmark data. See full abstract in attached document.« less

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

Thermal Neutron Radiography using a High-flux Compact Neutron Generator

NASA Astrophysics Data System (ADS)

Taylor, Michael; Sengbusch, Evan; Seyfert, Chris; Moll, Eli; Radel, Ross

A novel neutron imaging system has been designed and constructed by Phoenix Nuclear Labs to investigate specimens when conventional X-ray imaging will not suffice. A first-generation electronic neutron generator is actively being used by the United States Army and is coupled with activation films for neutron radiography to inspect munitions and other critical defence and aerospace components. A second-generation system has been designed to increase the total neutron output from an upgraded gaseous deuterium target to 5×1011 DD n/s, generating higher neutron flux at the imaging plane and dramatically reducing interrogation time, while maintaining high spatial resolution and low geometric unsharpness. A description of the neutron generator and imaging system, including the beamline, target and detector platform, is given in this paper. State of the art neutron moderators, collimators and imaging detector components are also discussed in the context of increasing specimen throughput and optimizing image quality. Neutron radiographs captured with the neutron radiography system will be further compared against simulated images using the MCNP nuclear simulation code.

Neutron Capture Gamma-Ray Libraries for Nuclear Applications

NASA Astrophysics Data System (ADS)

Sleaford, B. W.; Firestone, R. B.; Summers, N.; Escher, J.; Hurst, A.; Krticka, M.; Basunia, S.; Molnar, G.; Belgya, T.; Revay, Z.; Choi, H. D.

2011-06-01

The neutron capture reaction is useful in identifying and analyzing the gamma-ray spectrum from an unknown assembly as it gives unambiguous information on its composition. This can be done passively or actively where an external neutron source is used to probe an unknown assembly. There are known capture gamma-ray data gaps in the ENDF libraries used by transport codes for various nuclear applications. The Evaluated Gamma-ray Activation file (EGAF) is a new thermal neutron capture database of discrete line spectra and cross sections for over 260 isotopes that was developed as part of an IAEA Coordinated Research Project. EGAF is being used to improve the capture gamma production in ENDF libraries. For medium to heavy nuclei the quasi continuum contribution to the gamma cascades is not experimentally resolved. The continuum contains up to 90% of all the decay energy and is modeled here with the statistical nuclear structure code DICEBOX. This code also provides a consistency check of the level scheme nuclear structure evaluation. The calculated continuum is of sufficient accuracy to include in the ENDF libraries. This analysis also determines new total thermal capture cross sections and provides an improved RIPL database. For higher energy neutron capture there is less experimental data available making benchmarking of the modeling codes more difficult. We are investigating the capture spectra from higher energy neutrons experimentally using surrogate reactions and modeling this with Hauser-Feshbach codes. This can then be used to benchmark CASINO, a version of DICEBOX modified for neutron capture at higher energy. This can be used to simulate spectra from neutron capture at incident neutron energies up to 20 MeV to improve the gamma-ray spectrum in neutron data libraries used for transport modeling of unknown assemblies.

The Activities of the European Consortium on Nuclear Data Development and Analysis for Fusion

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

Fischer, U., E-mail: ulrich.fischer@kit.edu; Avrigeanu, M.; Avrigeanu, V.

This paper presents an overview of the activities of the European Consortium on Nuclear Data Development and Analysis for Fusion. The Consortium combines available European expertise to provide services for the generation, maintenance, and validation of nuclear data evaluations and data files relevant for ITER, IFMIF and DEMO, as well as codes and software tools required for related nuclear calculations.

75 FR 3946 - License Nos. DPR-42 and DPR-60; Northern States Power Company; Prairie Island Nuclear Generating...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-25

..., Office of Nuclear Reactor Regulation. [FR Doc. 2010-1309 Filed 1-22-10; 8:45 am] BILLING CODE 7590-01-P ... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-282 and 50-306; NRC-2010-0022] License Nos. DPR-42 and DPR-60; Northern States Power Company; Prairie Island Nuclear Generating Plant, Units 1 and 2...

Facility Targeting, Protection and Mission Decision Making Using the VISAC Code

NASA Technical Reports Server (NTRS)

Morris, Robert H.; Sulfredge, C. David

2011-01-01

The Visual Interactive Site Analysis Code (VISAC) has been used by DTRA and several other agencies to aid in targeting facilities and to predict the associated collateral effects for the go, no go mission decision making process. VISAC integrates the three concepts of target geometric modeling, damage assessment capabilities, and an event/fault tree methodology for evaluating accident/incident consequences. It can analyze a variety of accidents/incidents at nuclear or industrial facilities, ranging from simple component sabotage to an attack with military or terrorist weapons. For nuclear facilities, VISAC predicts the facility damage, estimated downtime, amount and timing of any radionuclides released. Used in conjunction with DTRA's HPAC code, VISAC also can analyze transport and dispersion of the radionuclides, levels of contamination of the surrounding area, and the population at risk. VISAC has also been used by the NRC to aid in the development of protective measures for nuclear facilities that may be subjected to attacks by car/truck bombs.

Nuclear code case development of printed-circuit heat exchangers with thermal and mechanical performance testing

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

Aakre, Shaun R.; Jentz, Ian W.; Anderson, Mark H.

The U.S. Department of Energy has agreed to fund a three-year integrated research project to close technical gaps involved with compact heat exchangers to be used in nuclear applications. This paper introduces the goals of the project, the research institutions, and industrial partners working in collaboration to develop a draft Boiler and Pressure Vessel Code Case for this technology. Heat exchanger testing, as well as non-destructive and destructive evaluation, will be performed by researchers across the country to understand the performance of compact heat exchangers. Testing will be performed using coolants and conditions proposed for Gen IV Reactor designs. Preliminarymore » observations of the mechanical failure mechanisms of the heat exchangers using destructive and non-destructive methods is presented. Unit-cell finite element models assembled to help predict the mechanical behavior of these high-temperature components are discussed as well. Performance testing methodology is laid out in this paper along with preliminary modeling results, an introduction to x-ray and neutron inspection techniques, and results from a recent pressurization test of a printed-circuit heat exchanger. The operational and quality assurance knowledge gained from these models and validation tests will be useful to developers of supercritical CO 2 systems, which commonly employ printed-circuit heat exchangers.« less

Conductor analysis of the ITER FEAT poloidal field coils during a plasma scenario

NASA Astrophysics Data System (ADS)

Nicollet, S.; Hertout, P.; Duchateau, J. L.; Bleyer, A.; Bessette, D.

2002-05-01

In the framework of the ITER (International Thermonuclear Experimental Reactor) FEAT (Fusion Energy Advanced Tokamak) project, a fully superconducting PF (Poloidal Field) system has been designed in detail. The Central Solenoid and the 6 equilibrium coils constituting the PF system provide the magnetic fields which develop, shape and control the 15 MA plasma during the 1800 s of a typical plasma scenario. The 6 PF coils will be wound two-in-hand from a 45 kA niobium-titanium CICC (Cable-In-Conduit-Conductor). These coils will experience severe heat loads specially during the 400 s of the plasma burn: nuclear heating due to the 400 MW of fusion power, thermal radiation and AC losses (30 to 300 kJ). The AC losses along the PF coil pancakes are deduced from accurate magnetic field computations performed with a 3D magnetostatic code, TRAPS. The nuclear heating and the thermal radiation are assumed to be uniform over a given face of the PF coils. These heat loads are used as input to perform the thermal and hydraulic analysis with a finite element code, GANDALF. The temperature increases (0.1 to 0.4 K) are computed, the margins and performances of the conductor are evaluated.

Mechanistic materials modeling for nuclear fuel performance

DOE PAGES

Tonks, Michael R.; Andersson, David; Phillpot, Simon R.; ...

2017-03-15

Fuel performance codes are critical tools for the design, certification, and safety analysis of nuclear reactors. However, their ability to predict fuel behavior under abnormal conditions is severely limited by their considerable reliance on empirical materials models correlated to burn-up (a measure of the number of fission events that have occurred, but not a unique measure of the history of the material). In this paper, we propose a different paradigm for fuel performance codes to employ mechanistic materials models that are based on the current state of the evolving microstructure rather than burn-up. In this approach, a series of statemore » variables are stored at material points and define the current state of the microstructure. The evolution of these state variables is defined by mechanistic models that are functions of fuel conditions and other state variables. The material properties of the fuel and cladding are determined from microstructure/property relationships that are functions of the state variables and the current fuel conditions. Multiscale modeling and simulation is being used in conjunction with experimental data to inform the development of these models. Finally, this mechanistic, microstructure-based approach has the potential to provide a more predictive fuel performance capability, but will require a team of researchers to complete the required development and to validate the approach.« less

Characterization of gamma field in the JSI TRIGA reactor

NASA Astrophysics Data System (ADS)

Ambrožič, Klemen; Radulović, Vladimir; Snoj, Luka; Gruel, Adrien; Guillou, Mael Le; Blaise, Patrick; Destouches, Christophe; Barbot, Loïc

2018-01-01

Research reactors such as the "Jožzef Stefan" Institute TRIGA reactor have primarily been designed for experimentation and sample irradiation with neutrons. However recent developments in incorporating additional instrumentation for nuclear power plant support and with novel high flux material testing reactor designs, γ field characterization has become of great interest for the characterization of the changes in operational parameters of electronic devices and for the evaluation of γ heating of MTR's structural materials in a representative reactor Γ spectrum. In this paper, we present ongoing work on γ field characterization both experimentally, by performing γ field measurements, and by simulations, using Monte Carlo particle transport codes in conjunction with R2S methodology for delayed γ field characterization.

Photoneutron Reaction Data for Nuclear Physics and Astrophysics

NASA Astrophysics Data System (ADS)

Utsunomiya, Hiroaki; Renstrøm, Therese; Tveten, Gry Merete; Gheorghe, Ioana; Filipescu, Dan Mihai; Belyshev, Sergey; Stopani, Konstantin; Wang, Hongwei; Fan, Gongtao; Lui, Yiu-Wing; Symochko, Dmytro; Goriely, Stephane; Larsen, Ann-Cecilie; Siem, Sunniva; Varlamov, Vladimir; Ishkhanov, Boris; Glodariu, Tudor; Krzysiek, Mateusz; Takenaka, Daiki; Ari-izumi, Takashi; Amano, Sho; Miyamoto, Shuji

2018-05-01

We discuss the role of photoneutron reaction data in nuclear physics and astrophysics in conjunction with the Coordinated Research Project of the International Atomic Energy Agency with the code F41032 (IAEA-CRP F41032).

Design of a portable dose rate detector based on a double Geiger-Mueller counter

NASA Astrophysics Data System (ADS)

Wang, Peng; Tang, Xiao-Bin; Gong, Pin; Huang, Xi; Wen, Liang-Sheng; Han, Zhen-Yang; He, Jian-Ping

2018-01-01

A portable dose rate detector was designed to monitor radioactive pollution and radioactive environments. The portable dose detector can measure background radiation levels (0.1 μSv/h) to nuclear accident radiation levels (>10 Sv/h). Both automatic switch technology of a double Geiger-Mueller counter and time-to-count technology were adopted to broaden the measurement range of the instrument. Global positioning systems and the 3G telecommunication protocol were installed to prevent radiation damage to the human body. In addition, the Monte Carlo N-Particle code was used to design the thin layer of metal for energy compensation, which was used to flatten energy response The portable dose rate detector has been calibrated by the standard radiation field method, and it can be used alone or in combination with additional radiation detectors.

Detection of Explosive Devices using X-ray Backscatter Radiation

NASA Astrophysics Data System (ADS)

Faust, Anthony A.

2002-09-01

It is our goal to develop a coded aperture based X-ray backscatter imaging detector that will provide sufficient speed, contrast and spatial resolution to detect Antipersonnel Landmines and Improvised Explosive Devices (IED). While our final objective is to field a hand-held detector, we have currently constrained ourselves to a design that can be fielded on a small robotic platform. Coded aperture imaging has been used by the observational gamma astronomy community for a number of years. However, it has been the recent advances in the field of medical nuclear imaging which has allowed for the application of the technique to a backscatter scenario. In addition, driven by requirements in medical applications, advances in X-ray detection are continually being made, and detectors are now being produced that are faster, cheaper and lighter than those only a decade ago. With these advances, a coded aperture hand-held imaging system has only recently become a possibility. This paper will begin with an introduction to the technique, identify recent advances which have made this approach possible, present a simulated example case, and conclude with a discussion on future work.

Updated User's Guide for Sammy: Multilevel R-Matrix Fits to Neutron Data Using Bayes' Equations

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

Larson, Nancy M

2008-10-01

In 1980 the multilevel multichannel R-matrix code SAMMY was released for use in analysis of neutron-induced cross section data at the Oak Ridge Electron Linear Accelerator. Since that time, SAMMY has evolved to the point where it is now in use around the world for analysis of many different types of data. SAMMY is not limited to incident neutrons but can also be used for incident protons, alpha particles, or other charged particles; likewise, Coulomb exit hannels can be included. Corrections for a wide variety of experimental conditions are available in the code: Doppler and resolution broadening, multiple-scattering corrections formore » capture or reaction yields, normalizations and backgrounds, to name but a few. The fitting procedure is Bayes' method, and data and parameter covariance matrices are properly treated within the code. Pre- and post-processing capabilities are also available, including (but not limited to) connections with the Evaluated Nuclear Data Files. Though originally designed for use in the resolved resonance region, SAMMY also includes a treatment for data analysis in the unresolved resonance region.« less

76 FR 14699 - Southern Nuclear Operating Company; Notice of Availability of Application for a Combined License

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-17

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 52-025 and 52-026; NRC-2008-0252] Southern Nuclear... Atomic Energy Act and Title 10 of the Code of Federal Regulations (10 CFR) Part 52, ``Licenses... of 10 CFR Part 52. The information submitted by the applicant includes certain administrative...

76 FR 16645 - Southern Nuclear Operating Company; Notice of Availability of Application for a Combined License

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-24

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 52-025 and 52-026; NRC-2008-0252] Southern Nuclear... Atomic Energy Act and Title 10 of the Code of Federal Regulations (10 CFR) Part 52, ``Licenses... of 10 CFR Part 52. The information submitted by the applicant includes certain administrative...

The Science of Little Boy

ERIC Educational Resources Information Center

Askew, Jennifer; Gray, Ron

2017-01-01

Near the end of World War II, the United States dropped the first nuclear bomb ever used in warfare. The bomb was code named "Little Boy." The fission-type nuclear bomb exploded with the energy equivalent of approximately 13 kilotons of TNT. This article describes a 16 day model-based inquiry (MBI) unit on nuclear chemistry that…

Uncertainty evaluation of nuclear reaction model parameters using integral and microscopic measurements. Covariances evaluation with CONRAD code

NASA Astrophysics Data System (ADS)

de Saint Jean, C.; Habert, B.; Archier, P.; Noguere, G.; Bernard, D.; Tommasi, J.; Blaise, P.

2010-10-01

In the [eV;MeV] energy range, modelling of the neutron induced reactions are based on nuclear reaction models having parameters. Estimation of co-variances on cross sections or on nuclear reaction model parameters is a recurrent puzzle in nuclear data evaluation. Major breakthroughs were asked by nuclear reactor physicists to assess proper uncertainties to be used in applications. In this paper, mathematical methods developped in the CONRAD code[2] will be presented to explain the treatment of all type of uncertainties, including experimental ones (statistical and systematic) and propagate them to nuclear reaction model parameters or cross sections. Marginalization procedure will thus be exposed using analytical or Monte-Carlo solutions. Furthermore, one major drawback found by reactor physicist is the fact that integral or analytical experiments (reactor mock-up or simple integral experiment, e.g. ICSBEP, …) were not taken into account sufficiently soon in the evaluation process to remove discrepancies. In this paper, we will describe a mathematical framework to take into account properly this kind of information.

NDEC: A NEA platform for nuclear data testing, verification and benchmarking

NASA Astrophysics Data System (ADS)

Díez, C. J.; Michel-Sendis, F.; Cabellos, O.; Bossant, M.; Soppera, N.

2017-09-01

The selection, testing, verification and benchmarking of evaluated nuclear data consists, in practice, in putting an evaluated file through a number of checking steps where different computational codes verify that the file and the data it contains complies with different requirements. These requirements range from format compliance to good performance in application cases, while at the same time physical constraints and the agreement with experimental data are verified. At NEA, the NDEC (Nuclear Data Evaluation Cycle) platform aims at providing, in a user friendly interface, a thorough diagnose of the quality of a submitted evaluated nuclear data file. Such diagnose is based on the results of different computational codes and routines which carry out the mentioned verifications, tests and checks. NDEC also searches synergies with other existing NEA tools and databases, such as JANIS, DICE or NDaST, including them into its working scheme. Hence, this paper presents NDEC, its current development status and its usage in the JEFF nuclear data project.

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

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

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

2009-09-01

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

Advanced instrumentation and analysis methods for in-pile thermal and nuclear measurements: from out-of-pile studies to irradiation campaigns

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

Reynard-Carette, C.; Lyoussi, A.

Research and development on nuclear fuel behavior under irradiations and accelerated ageing of structure materials is a key issue for sustainable nuclear energy in order to meet specific needs by keeping the best level of safety. A new Material Testing Reactor (MTR), the Jules Horowitz Reactor (JHR) currently under construction in the South of France in the CEA Cadarache research centre will offer a real opportunity to perform R and D programs and hence will crucially contribute to the selection, optimization and qualification of innovative materials and fuels. To perform such programs advanced accurate and innovative experiments, irradiation devices thatmore » contain material and fuel samples are required to be set up inside or beside the reactor core. These experiments needs beforehand in situ and on line sophisticated measurements to accurately reach specific and determining parameters such as thermal and fast neutron fluxes, nuclear heating and temperature conditions to precisely monitor and control the conducted assays. Consequently, since 2009 CEA and Aix-Marseille University collaborate in order to design and develop a new multi-sensor device which will be dedicated to measuring profiles of such conditions inside the experimental channels of the JHR. These works are performed in the framework of two complementary joint research programs called MAHRI-BETHY and INCORE. These programs couple experimental studies carried out both out-of nuclear fluxes (in laboratory) and under irradiation conditions (in OSIRIS MTR reactor in France and MARIA MTR reactor in Poland) with numerical works realized by thermal simulations (CAST3M code) and Monte Carlo simulations (MCNP code). These programs deal with three main aims. The first one corresponds to the design and/or the test of new in-pile instrumentation. The second one concerns the development of advanced calibration procedures in particular in the case of one specific sensor: a differential calorimeter used to quantify nuclear heating. The last one consists in the development of accurate measurement and analysis methods. The paper will be dedicated to a complete review of the experimental and numerical works performed since 2009 thanks to two parts. The first part will detail a new thermal approach implemented to improve nuclear heating measurements by radiometric calorimeters. New experimental tools (calorimeter prototypes and set-ups such BETHY Bench) developed to perform preliminary out-of-pile studies under suitable conditions will be presented (temperature and velocity of the external cooling fluid, heat source localization and intensity inside the calorimetric cells). Then the response of two kinds of sensors, their calibrations curves and their thermal behaviors will be compared for various parameters. Finally validated numerical thermal and Monte Carlo works will be discussed to propose new improvements. The second parts of the paper will focus on works realized in order to design, develop and test the first prototype of the multi-sensor device called CARMEN [7-9]. The two mock-ups dedicated respectively to neutron measurements and photon measurements will be detailed. The results obtained during two irradiation campaigns inside the periphery of OSIRIS reactor will be shown. The new analysis method will be discussed. (authors)« less

Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

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

Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

2012-06-06

Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spentmore » fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.« less

The skyshine benchmark experiment revisited.

PubMed

Terry, Ian R

2005-01-01

With the coming renaissance of nuclear power, heralded by new nuclear power plant construction in Finland, the issue of qualifying modern tools for calculation becomes prominent. Among the calculations required may be the determination of radiation levels outside the plant owing to skyshine. For example, knowledge of the degree of accuracy in the calculation of gamma skyshine through the turbine hall roof of a BWR plant is important. Modern survey programs which can calculate skyshine dose rates tend to be qualified only by verification with the results of Monte Carlo calculations. However, in the past, exacting experimental work has been performed in the field for gamma skyshine, notably the benchmark work in 1981 by Shultis and co-workers, which considered not just the open source case but also the effects of placing a concrete roof above the source enclosure. The latter case is a better reflection of reality as safety considerations nearly always require the source to be shielded in some way, usually by substantial walls but by a thinner roof. One of the tools developed since that time, which can both calculate skyshine radiation and accurately model the geometrical set-up of an experiment, is the code RANKERN, which is used by Framatome ANP and other organisations for general shielding design work. The following description concerns the use of this code to re-address the experimental results from 1981. This then provides a realistic gauge to validate, but also to set limits on, the program for future gamma skyshine applications within the applicable licensing procedures for all users of the code.

Evaluation of the finite element fuel rod analysis code (FRANCO)

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

Lee, K.; Feltus, M.A.

1994-12-31

Knowledge of temperature distribution in a nuclear fuel rod is required to predict the behavior of fuel elements during operating conditions. The thermal and mechanical properties and performance characteristics are strongly dependent on the temperature, which can vary greatly inside the fuel rod. A detailed model of fuel rod behavior can be described by various numerical methods, including the finite element approach. The finite element method has been successfully used in many engineering applications, including nuclear piping and reactor component analysis. However, fuel pin analysis has traditionally been carried out with finite difference codes, with the exception of Electric Powermore » Research Institute`s FREY code, which was developed for mainframe execution. This report describes FRANCO, a finite element fuel rod analysis code capable of computing temperature disrtibution and mechanical deformation of a single light water reactor fuel rod.« less

ASME Code Efforts Supporting HTGRs

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

D.K. Morton

2010-09-01

In 1999, an international collaborative initiative for the development of advanced (Generation IV) reactors was started. The idea behind this effort was to bring nuclear energy closer to the needs of sustainability, to increase proliferation resistance, and to support concepts able to produce energy (both electricity and process heat) at competitive costs. The U.S. Department of Energy has supported this effort by pursuing the development of the Next Generation Nuclear Plant, a high temperature gas-cooled reactor. This support has included research and development of pertinent data, initial regulatory discussions, and engineering support of various codes and standards development. This reportmore » discusses the various applicable American Society of Mechanical Engineers (ASME) codes and standards that are being developed to support these high temperature gascooled reactors during construction and operation. ASME is aggressively pursuing these codes and standards to support an international effort to build the next generation of advanced reactors so that all can benefit.« less

ASME Code Efforts Supporting HTGRs

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

D.K. Morton

2011-09-01

In 1999, an international collaborative initiative for the development of advanced (Generation IV) reactors was started. The idea behind this effort was to bring nuclear energy closer to the needs of sustainability, to increase proliferation resistance, and to support concepts able to produce energy (both electricity and process heat) at competitive costs. The U.S. Department of Energy has supported this effort by pursuing the development of the Next Generation Nuclear Plant, a high temperature gas-cooled reactor. This support has included research and development of pertinent data, initial regulatory discussions, and engineering support of various codes and standards development. This reportmore » discusses the various applicable American Society of Mechanical Engineers (ASME) codes and standards that are being developed to support these high temperature gascooled reactors during construction and operation. ASME is aggressively pursuing these codes and standards to support an international effort to build the next generation of advanced reactors so that all can benefit.« less

ASME Code Efforts Supporting HTGRs

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

D.K. Morton

2012-09-01

In 1999, an international collaborative initiative for the development of advanced (Generation IV) reactors was started. The idea behind this effort was to bring nuclear energy closer to the needs of sustainability, to increase proliferation resistance, and to support concepts able to produce energy (both electricity and process heat) at competitive costs. The U.S. Department of Energy has supported this effort by pursuing the development of the Next Generation Nuclear Plant, a high temperature gas-cooled reactor. This support has included research and development of pertinent data, initial regulatory discussions, and engineering support of various codes and standards development. This reportmore » discusses the various applicable American Society of Mechanical Engineers (ASME) codes and standards that are being developed to support these high temperature gascooled reactors during construction and operation. ASME is aggressively pursuing these codes and standards to support an international effort to build the next generation of advanced reactors so that all can benefit.« less

Multi-scale modeling of irradiation effects in spallation neutron source materials

NASA Astrophysics Data System (ADS)

Yoshiie, T.; Ito, T.; Iwase, H.; Kaneko, Y.; Kawai, M.; Kishida, I.; Kunieda, S.; Sato, K.; Shimakawa, S.; Shimizu, F.; Hashimoto, S.; Hashimoto, N.; Fukahori, T.; Watanabe, Y.; Xu, Q.; Ishino, S.

2011-07-01

Changes in mechanical property of Ni under irradiation by 3 GeV protons were estimated by multi-scale modeling. The code consisted of four parts. The first part was based on the Particle and Heavy-Ion Transport code System (PHITS) code for nuclear reactions, and modeled the interactions between high energy protons and nuclei in the target. The second part covered atomic collisions by particles without nuclear reactions. Because the energy of the particles was high, subcascade analysis was employed. The direct formation of clusters and the number of mobile defects were estimated using molecular dynamics (MD) and kinetic Monte-Carlo (kMC) methods in each subcascade. The third part considered damage structural evolutions estimated by reaction kinetic analysis. The fourth part involved the estimation of mechanical property change using three-dimensional discrete dislocation dynamics (DDD). Using the above four part code, stress-strain curves for high energy proton irradiated Ni were obtained.

Probabilistic approach for decay heat uncertainty estimation using URANIE platform and MENDEL depletion code

NASA Astrophysics Data System (ADS)

Tsilanizara, A.; Gilardi, N.; Huynh, T. D.; Jouanne, C.; Lahaye, S.; Martinez, J. M.; Diop, C. M.

2014-06-01

The knowledge of the decay heat quantity and the associated uncertainties are important issues for the safety of nuclear facilities. Many codes are available to estimate the decay heat. ORIGEN, FISPACT, DARWIN/PEPIN2 are part of them. MENDEL is a new depletion code developed at CEA, with new software architecture, devoted to the calculation of physical quantities related to fuel cycle studies, in particular decay heat. The purpose of this paper is to present a probabilistic approach to assess decay heat uncertainty due to the decay data uncertainties from nuclear data evaluation like JEFF-3.1.1 or ENDF/B-VII.1. This probabilistic approach is based both on MENDEL code and URANIE software which is a CEA uncertainty analysis platform. As preliminary applications, single thermal fission of uranium 235, plutonium 239 and PWR UOx spent fuel cell are investigated.

Observations Regarding Use of Advanced CFD Analysis, Sensitivity Analysis, and Design Codes in MDO

NASA Technical Reports Server (NTRS)

Newman, Perry A.; Hou, Gene J. W.; Taylor, Arthur C., III

1996-01-01

Observations regarding the use of advanced computational fluid dynamics (CFD) analysis, sensitivity analysis (SA), and design codes in gradient-based multidisciplinary design optimization (MDO) reflect our perception of the interactions required of CFD and our experience in recent aerodynamic design optimization studies using CFD. Sample results from these latter studies are summarized for conventional optimization (analysis - SA codes) and simultaneous analysis and design optimization (design code) using both Euler and Navier-Stokes flow approximations. The amount of computational resources required for aerodynamic design using CFD via analysis - SA codes is greater than that required for design codes. Thus, an MDO formulation that utilizes the more efficient design codes where possible is desired. However, in the aerovehicle MDO problem, the various disciplines that are involved have different design points in the flight envelope; therefore, CFD analysis - SA codes are required at the aerodynamic 'off design' points. The suggested MDO formulation is a hybrid multilevel optimization procedure that consists of both multipoint CFD analysis - SA codes and multipoint CFD design codes that perform suboptimizations.

Dosimetric evaluation of nanotargeted (188)Re-liposome with the MIRDOSE3 and OLINDA/EXM programs.

PubMed

Chang, Chih-Hsien; Chang, Ya-Jen; Lee, Te-Wei; Ting, Gann; Chang, Kwo-Ping

2012-06-01

The OLINDA/EXM computer code was created as a replacement for the widely used MIRDOSE3 code for radiation dosimetry in nuclear medicine. A dosimetric analysis with these codes was performed to evaluate nanoliposomes as carriers of radionuclides ((188)Re-liposomes) in colon carcinoma-bearing mice. Pharmacokinetic data for (188)Re-N, N-bis (2-mercaptoethyl)-N',N'-diethylethylenediamine ((188)Re-BMEDA) and (188)Re-liposome were obtained for estimation of absorbed doses in normal organs. Radiation dose estimates for normal tissues were calculated using the MIRDOSE3 and OLINDA/EXM programs for a colon carcinoma solid tumor mouse model. Mean absorbed doses derived from(188)Re-BMEDA and (188)Re-liposome in normal tissues were generally similar as calculated by MIRDOSE3 and OLINDA/EXM programs. One notable exception to this was red marrow, wherein MIRDOSE3 resulted in higher absorbed doses than OLINDA/EXM (1.53- and 1.60-fold for (188)Re-BMEDA and (188)Re-liposome, respectively). MIRDOSE3 and OLINDA have very similar residence times and organ doses. Bone marrow doses were estimated by designating cortical bone rather than bone marrow as a source organ. The bone marrow doses calculated by MIRDOSE3 are higher than those by OLINDA. If the bone marrow is designated as a source organ, the doses estimated by MIRDOSE3 and OLINDA programs will be very similar.

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

Steiner, J.L.; Lime, J.F.; Elson, J.S.

One dimensional TRAC transient calculations of the process inherent ultimate safety (PIUS) advanced reactor design were performed for a pump-trip SCRAM. The TRAC calculations showed that the reactor power response and shutdown were in qualitative agreement with the one-dimensional analyses presented in the PIUS Preliminary Safety Information Document (PSID) submitted by Asea Brown Boveri (ABB) to the US Nuclear Regulatory Commission for preapplication safety review. The PSID analyses were performed with the ABB-developed RIGEL code. The TRAC-calculated phenomena and trends were also similar to those calculated with another one-dimensional PIUS model, the Brookhaven National Laboratory developed PIPA code. A TRACmore » pump-trip SCRAM transient has also been calculated with a TRAC model containing a multi-dimensional representation of the PIUS intemal flow structures and core region. The results obtained using the TRAC fully one-dimensional PIUS model are compared to the RIGEL, PIPA, and TRAC multi-dimensional results.« less

Effects of Cluster Porosity on the Tensile Properties of Butt-Weldments in T-1 Steel

DTIC Science & Technology

1974-11-01

i 12 Boiler and Pressure Vessel Code .19 In this code, the algebraic difference between the largest and smallest principal stresses is defined...Report U1LU- HN(J 7l-2()24 (University ot Illinois. 1971). "Nuclear Power Components.’* ASME Boiler and Pressure Vessel Code . Section HI. Subsections

Elbow stress indices using finite element analysis

NASA Astrophysics Data System (ADS)

Yu, Lixin

Section III of the ASME Boiler and Pressure Vessel Code (the Code) specifies rules for the design of nuclear power plant components. NB-3600 of the Code presents a simplified design method using stress indices---Scalar Coefficients used the modify straight pipe stress equations so that they can be applied to elbows, tees and other piping components. The stress indices of piping components are allowed to be determined both analytically and experimentally. This study concentrates on the determination of B2 stress indices for elbow components using finite element analysis (FEA). First, the previous theoretical, numerical and experimental investigations on elbow behavior were comprehensively reviewed, as was the philosophy behind the use of stress indices. The areas of further research was defined. Then, a comprehensive investigation was carried out to determine how the finite element method should be used to correctly simulate an elbow's structural behavior. This investigation included choice of element type, convergence of mesh density, use of boundary restraint and a reconciliation study between FEA and laboratory experiments or other theoretical formulations in both elastic and elasto-plastic domain. Results from different computer programs were also compared. Reasonably good reconciliation was obtained. Appendix II of the Code describes the experimental method to determine B2 stress indices based on load-deflection curves. This procedure was used to compute the B2 stress indices for various loading modes on one particular elbow configuration. The B2 stress indices thus determined were found to be about half of the value calculated from the Code equation. Then the effect on B2 stress indices of those factors such as internal pressure and flange attachments were studied. Finally, the investigation was extended to other configurations of elbow components. A parametric study was conducted on different elbow sizes and schedules. Regression analysis was then used to obtain a modified coefficient and exponent for the Code equation used to calculate B2 index for elbows.

The behaviour of transuranic mixed oxide fuel in a Candu-900 reactor

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

Morreale, A. C.; Ball, M. R.; Novog, D. R.

2012-07-01

The production of transuranic actinide fuels for use in current thermal reactors provides a useful intermediary step in closing the nuclear fuel cycle. Extraction of actinides reduces the longevity, radiation and heat loads of spent material. The burning of transuranic fuels in current reactors for a limited amount of cycles reduces the infrastructure demand for fast reactors and provides an effective synergy that can result in a reduction of as much as 95% of spent fuel waste while reducing the fast reactor infrastructure needed by a factor of almost 13.5 [1]. This paper examines the features of actinide mixed oxidemore » fuel, TRUMOX, in a CANDU{sup R}* nuclear reactor. The actinide concentrations used were based on extraction from 30 year cooled spent fuel and mixed with natural uranium in 3.1 wt% actinide MOX fuel. Full lattice cell modeling was performed using the WIMS-AECL code, super-cell calculations were analyzed in DRAGON and full core analysis was executed in the RFSP 2-group diffusion code. A time-average full core model was produced and analyzed for reactor coefficients, reactivity device worth and online fuelling impacts. The standard CANDU operational limits were maintained throughout operations. The TRUMOX fuel design achieved a burnup of 27.36 MWd/kg HE. A full TRUMOX fuelled CANDU was shown to operate within acceptable limits and provided a viable intermediary step for burning actinides. The recycling, reprocessing and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle. (authors)« less

Advances in Monte-Carlo code TRIPOLI-4®'s treatment of the electromagnetic cascade

NASA Astrophysics Data System (ADS)

Mancusi, Davide; Bonin, Alice; Hugot, François-Xavier; Malouch, Fadhel

2018-01-01

TRIPOLI-4® is a Monte-Carlo particle-transport code developed at CEA-Saclay (France) that is employed in the domains of nuclear-reactor physics, criticality-safety, shielding/radiation protection and nuclear instrumentation. The goal of this paper is to report on current developments, validation and verification made in TRIPOLI-4 in the electron/positron/photon sector. The new capabilities and improvements concern refinements to the electron transport algorithm, the introduction of a charge-deposition score, the new thick-target bremsstrahlung option, the upgrade of the bremsstrahlung model and the improvement of electron angular straggling at low energy. The importance of each of the developments above is illustrated by comparisons with calculations performed with other codes and with experimental data.

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

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

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

2011-09-01

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

Extension of the energy range of the experimental activation cross-sections data of longer-lived products of proton induced nuclear reactions on dysprosium up to 65MeV.

PubMed

Tárkányi, F; Ditrói, F; Takács, S; Hermanne, A; Ignatyuk, A V

2015-04-01

Activation cross-sections data of longer-lived products of proton induced nuclear reactions on dysprosium were extended up to 65MeV by using stacked foil irradiation and gamma spectrometry experimental methods. Experimental cross-sections data for the formation of the radionuclides (159)Dy, (157)Dy, (155)Dy, (161)Tb, (160)Tb, (156)Tb, (155)Tb, (154m2)Tb, (154m1)Tb, (154g)Tb, (153)Tb, (152)Tb and (151)Tb are reported in the 36-65MeV energy range, and compared with an old dataset from 1964. The experimental data were also compared with the results of cross section calculations of the ALICE and EMPIRE nuclear model codes and of the TALYS nuclear reaction model code as listed in the latest on-line libraries TENDL 2013. Copyright © 2015. Published by Elsevier Ltd.

Assessment of WWMCCS Performance in a Post-Nuclear Attack Environment. Sanitized

DTIC Science & Technology

1975-11-01

34cONTRACT No. 001-75-C-0217 * 1 THIS WORK SPONSOQRED BY THE DEFENSE NUCLEAR AGENC" UNDER RDT&E RMSS CODE X36075669 Q94OAXDDO01O1 H2590D. Pr p r d 1 rTHIS DOCUM...8217 < : " ! , . .I/• I V...... .... . . . . . . 1 "- 7 . . . • .tEPORT DOCJMENTATION PAGE "rAr, . Irt’. ±~ATSSIS9.C.7NF VMC EPPINEI 21 Jan 75-31 Oc- 75 IOST-I.CLLEA...U .. 4II4 .i %Iy•11 .4 @$ . .~ *󈧯.I . 1 . S $P.IL%-V ka"I 1 "- This work sponsored by the Defe.xse Nuclear Agency under %DT&E WISS Code X360075469 Q9

Coulomb effects in low-energy nuclear fragmentation

NASA Technical Reports Server (NTRS)

Wilson, John W.; Chun, Sang Y.; Badavi, Francis F.; John, Sarah

1993-01-01

Early versions of the Langley nuclear fragmentation code NUCFRAG (and a publicly released version called HZEFRG1) assumed straight-line trajectories throughout the interaction. As a consequence, NUCFRAG and HZEFRG1 give unrealistic cross sections for large mass removal from the projectile and target at low energies. A correction for the distortion of the trajectory by the nuclear Coulomb fields is used to derive fragmentation cross sections. A simple energy-loss term is applied to estimate the energy downshifts that greatly alter the Coulomb trajectory at low energy. The results, which are far more realistic than prior versions of the code, should provide the data base for future transport calculations. The systematic behavior of charge-removal cross sections compares favorably with results from low-energy experiments.

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

Joint design of QC-LDPC codes for coded cooperation system with joint iterative decoding

NASA Astrophysics Data System (ADS)

Zhang, Shunwai; Yang, Fengfan; Tang, Lei; Ejaz, Saqib; Luo, Lin; Maharaj, B. T.

2016-03-01

In this paper, we investigate joint design of quasi-cyclic low-density-parity-check (QC-LDPC) codes for coded cooperation system with joint iterative decoding in the destination. First, QC-LDPC codes based on the base matrix and exponent matrix are introduced, and then we describe two types of girth-4 cycles in QC-LDPC codes employed by the source and relay. In the equivalent parity-check matrix corresponding to the jointly designed QC-LDPC codes employed by the source and relay, all girth-4 cycles including both type I and type II are cancelled. Theoretical analysis and numerical simulations show that the jointly designed QC-LDPC coded cooperation well combines cooperation gain and channel coding gain, and outperforms the coded non-cooperation under the same conditions. Furthermore, the bit error rate performance of the coded cooperation employing jointly designed QC-LDPC codes is better than those of random LDPC codes and separately designed QC-LDPC codes over AWGN channels.

Cooperative Threat Reduction: Cooperation Threat Reduction Program Liquid Propellant Disposition Project

NASA Astrophysics Data System (ADS)

2002-09-01

This audit is one in a series of audits the Deputy Secretary of Defense requested. As part of the Cooperative Threat Reduction (CTR) Program, DoD agreed to assist the Russian Federation in disposing of its liquid rocket propellant. Public Law 102-228 (section 2551 NOTE, title 22, United States Code), the Soviet Nuclear Threat Reduction Act of 1991 designates DoD as the executive agent for the CTR Program. Specific objectives of the act are to destroy chemical, nuclear, and other weapons; transport, store, disable, and safeguard weapons in connection with their destruction; and establish verifiable safeguards against proliferation of weapons of mass destruction. The Office of the Assistant Secretary of Defense (International Security Policy), under the Office of the Under Secretary of Defense for Policy, develops, coordinates, and oversees implementation of policy for the CTR Program. The CTR Directorate, Defense Threat Reduction Agency operates the program.

A qualitative analysis of the neutron population in fresh and spent fuel assemblies during simulated interrogation using the differential die-away technique

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

Tobin, Stephen J.; Lundkvist, Niklas; Goodsell, Alison V.

In this study, Monte Carlo simulations were performed for the differential die-away (DDA) technique to analyse the time-dependent behaviour of the neutron population in fresh and spent nuclear fuel assemblies as part of the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Simulations were performed to investigate both a possibly portable as well as a permanent DDA instrument. Taking advantage of a custom made modification to the MCNPX code, the variation in the neutron population, simultaneously in time and space, was examined. The motivation for this research was to improve the design of the DDA instrument, as it is bemore » ing considered for possible deployment at the Central Storage of Spent Nuclear Fuel and Encapsulation Plant in Sweden (Clab), as well as to assist in the interpretation of the both simulated and measured signals.« less

A qualitative analysis of the neutron population in fresh and spent fuel assemblies during simulated interrogation using the differential die-away technique

DOE PAGES

Tobin, Stephen J.; Lundkvist, Niklas; Goodsell, Alison V.; ...

2015-12-01

In this study, Monte Carlo simulations were performed for the differential die-away (DDA) technique to analyse the time-dependent behaviour of the neutron population in fresh and spent nuclear fuel assemblies as part of the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Simulations were performed to investigate both a possibly portable as well as a permanent DDA instrument. Taking advantage of a custom made modification to the MCNPX code, the variation in the neutron population, simultaneously in time and space, was examined. The motivation for this research was to improve the design of the DDA instrument, as it is bemore » ing considered for possible deployment at the Central Storage of Spent Nuclear Fuel and Encapsulation Plant in Sweden (Clab), as well as to assist in the interpretation of the both simulated and measured signals.« less

Microstructured silicon neutron detectors for security applications

NASA Astrophysics Data System (ADS)

Esteban, S.; Fleta, C.; Guardiola, C.; Jumilla, C.; Pellegrini, G.; Quirion, D.; Rodriguez, J.; Lozano, M.

2014-12-01

In this paper we present the design and performance of a perforated thermal neutron silicon detector with a 6LiF neutron converter. This device was manufactured within the REWARD project workplace whose aim is to develop and enhance technologies for the detection of nuclear and radiological materials. The sensor perforated structure results in a higher efficiency than that obtained with an equivalent planar sensor. The detectors were tested in a thermal neutron beam at the nuclear reactor at the Instituto Superior Técnico in Lisbon and the intrinsic detection efficiency for thermal neutrons and the gamma sensitivity were obtained. The Geant4 Monte Carlo code was used to simulate the experimental conditions, i.e. thermal neutron beam and the whole detector geometry. An intrinsic thermal neutron detection efficiency of 8.6%±0.4% with a discrimination setting of 450 keV was measured.

Monte Carlo capabilities of the SCALE code system

DOE PAGES

Rearden, Bradley T.; Petrie, Jr., Lester M.; Peplow, Douglas E.; ...

2014-09-12

SCALE is a broadly used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a “plug-and-play” framework that includes three deterministic and three Monte Carlo radiation transport solvers that can be selected based on the desired solution, including hybrid deterministic/Monte Carlo simulations. SCALE includes the latest nuclear data libraries for continuous-energy and multigroup radiation transport asmore » well as activation, depletion, and decay calculations. SCALE’s graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.2 will provide several new capabilities and significant improvements in many existing features, especially with expanded continuous-energy Monte Carlo capabilities for criticality safety, shielding, depletion, and sensitivity and uncertainty analysis. Finally, an overview of the Monte Carlo capabilities of SCALE is provided here, with emphasis on new features for SCALE 6.2.« less

Annual Report on the Activities and Publications of the DHS-DNDO-NTNFC Sponsored Post-doctoral Fellow at Los Alamos National Laboratory

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

Rim, Jung Ho; Tandon, Lav

This report is a summary of the projects Jung Rim is working on as a DHS postdoctoral fellow at Los Alamos National Laboratory. These research projects are designed to explore different radioanalytical methods to support nuclear forensics applications. The current projects discussed here include development of alpha spectroscopy method for 240/239Pu Isotopic ratio measurement, non-destructive uranium assay method using gamma spectroscopy, and 236U non-destructive uranium analysis using FRAM code. This report documents the work that has been performed since the start of the postdoctoral appointment.

Abstracts for student symposium

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

Goldman, B.

Lawrence Livermore National Laboratory Science and Engineering Research Semester (SERS) students are participants in a national program sponsored by the DOE Office of Energy Research. Presented topics from Fall 1993 include: Laser glass, wiring codes, lead in food and food containers, chromium removal from ground water, fiber optic sensors for ph measurement, CFC replacement, predator/prey simulation, detection of micronuclei in germ cells, DNA conformation, stimulated brillouin scattering, DNA sequencing, evaluation of education programs, neural network analysis of nuclear glass, lithium ion batteries, Indonesian snails, optical switching systems, and photoreceiver design. Individual papers are indexed separately on the Energy Data Base.

The new interactive CESAR

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

Fox, P.B.; Yatabe, M.

1987-01-01

In this report the Nuclear Criticality Safety Analytical Methods Resource Center describes a new interactive version of CESAR, a critical experiments storage and retrieval program available on the Nuclear Criticality Information System (NCIS) database at Lawrence Livermore National Laboratory. The original version of CESAR did not include interactive search capabilities. The CESAR database was developed to provide a convenient, readily accessible means of storing and retrieving code input data for the SCALE Criticality Safety Analytical Sequences and the codes comprising those sequences. The database includes data for both cross section preparation and criticality safety calculations. 3 refs., 1 tab.

New interactive CESAR

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

Fox, P.B.; Yatabe, M.

1987-01-01

The Nuclear Criticality Safety Analytical Methods Resource Center announces the availability of a new interactive version of CESAR, a critical experiments storage and retrieval program available on the Nuclear Criticality Information System (NCIS) data base at Lawrence Livermore National Laboratory. The original version of CESAR did not include interactive search capabilities. The CESAR data base was developed to provide a convenient, readily accessible means of storing and retrieving code input data for the SCALE criticality safety analytical sequences and the codes comprising those sequences. The data base includes data for both cross-section preparation and criticality safety calculations.

Comparison of MELCOR and SCDAP/RELAP5 results for a low-pressure, short-term station blackout at Browns Ferry

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

Carbajo, J.J.

1995-12-31

This study compares results obtained with two U.S. Nuclear Regulatory Commission (NRC)-sponsored codes, MELCOR version 1.8.3 (1.8PQ) and SCDAP/RELAP5 Mod3.1 release C, for the same transient - a low-pressure, short-term station blackout accident at the Browns Ferry nuclear plant. This work is part of MELCOR assessment activities to compare core damage progression calculations of MELCOR against SCDAP/RELAP5 since the two codes model core damage progression very differently.

78 FR 15009 - Consideration of Withdrawal From Commercial Production and Distribution of the Radioisotope...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-08

... may be addressed to: Dr. Marc Garland, Program Manager, Office of Nuclear Physics, Office of Science... Management Division, Office of Nuclear Physics, Office of Science, U.S. Department of Energy, Germantown..., Office of Nuclear Physics, Office of Science. [FR Doc. 2013-05444 Filed 3-7-13; 8:45 am] BILLING CODE...

75 FR 57535 - Northern States Power Company-Minnesota Notice of Issuance of Amendments to Facility Operating...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-21

... issuance. FOR FURTHER INFORMATION CONTACT: Thomas J. Wengert, Office of Nuclear Reactor Regulation, U.S... Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2010-23516 Filed 9-20-10; 8:45 am] BILLING CODE... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-282 and 50-306; NRC-2010-0290] Northern States Power...

SkyNet: A Modular Nuclear Reaction Network Library

NASA Astrophysics Data System (ADS)

Lippuner, Jonas; Roberts, Luke F.

2017-12-01

Almost all of the elements heavier than hydrogen that are present in our solar system were produced by nuclear burning processes either in the early universe or at some point in the life cycle of stars. In all of these environments, there are dozens to thousands of nuclear species that interact with each other to produce successively heavier elements. In this paper, we present SkyNet, a new general-purpose nuclear reaction network that evolves the abundances of nuclear species under the influence of nuclear reactions. SkyNet can be used to compute the nucleosynthesis evolution in all astrophysical scenarios where nucleosynthesis occurs. SkyNet is free and open source, and aims to be easy to use and flexible. Any list of isotopes can be evolved, and SkyNet supports different types of nuclear reactions. SkyNet is modular so that new or existing physics, like nuclear reactions or equations of state, can easily be added or modified. Here, we present in detail the physics implemented in SkyNet with a focus on a self-consistent transition to and from nuclear statistical equilibrium to non-equilibrium nuclear burning, our implementation of electron screening, and coupling of the network to an equation of state. We also present comprehensive code tests and comparisons with existing nuclear reaction networks. We find that SkyNet agrees with published results and other codes to an accuracy of a few percent. Discrepancies, where they exist, can be traced to differences in the physics implementations.

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)

Comparative Study on Various Geometrical Core Design of 300 MWth Gas Cooled Fast Reactor with UN-PuN Fuel Longlife without Refuelling

NASA Astrophysics Data System (ADS)

Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

2017-07-01

Nuclear power has progressive improvement in the operating performance of exiting reactors and ensuring economic competitiveness of nuclear electricity around the world. The GFR use gas coolant and fast neutron spectrum. This research use helium coolant which has low neutron moderation, chemical inert and single phase. Comparative study on various geometrical core design for modular GFR with UN-PuN fuel long life without refuelling has been done. The calculation use SRAC2006 code both PIJ calculation and CITATION calculation. The data libraries use JENDL 4.0. The variation of fuel fraction is 40% until 65%. In this research, we varied the geometry of core reactor to find the optimum geometry design. The variation of the geometry design is balance cylinder; it means that the diameter active core (D) same with height active core (H). Second, pancake cylinder (D>H) and third, tall cylinder (D

A journey from nuclear criticality methods to high energy density radflow experiments

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

Urbatsch, Todd James

Los Alamos National Laboratory is a nuclear weapons laboratory supporting our nation's defense. In support of this mission is a high energy-density physics program in which we design and execute experiments to study radiationhydrodynamics phenomena and improve the predictive capability of our largescale multi-physics software codes on our big-iron computers. The Radflow project’s main experimental effort now is to understand why we haven't been able to predict opacities on Sandia National Laboratory's Z-machine. We are modeling an increasing fraction of the Z-machine's dynamic hohlraum to find multi-physics explanations for the experimental results. Further, we are building an entirely different opacitymore » platform on Lawrence Livermore National Laboratory's National Ignition Facility (NIF), which is set to get results early 2017. Will the results match our predictions, match the Z-machine, or give us something entirely different? The new platform brings new challenges such as designing hohlraums and spectrometers. The speaker will recount his history, starting with one-dimensional Monte Carlo nuclear criticality methods in graduate school, radiative transfer methods research and software development for his first 16 years at LANL, and, now, radflow technology and experiments. Who knew that the real world was more than just radiation transport? Experiments aren't easy and they are as saturated with politics as a presidential election, but they sure are fun.« less

The Nuclear Energy Knowledge and Validation Center – Summary of Activities Conducted in FY15

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

Gougar, Hans David; Hong, Bonnie Colleen

2016-05-01

The Nuclear Energy Knowledge and Validation Center (NEKVaC) is a new initiative by the Department of Energy and the Idaho National Laboratory to coordinate and focus the resources and expertise that exist with the DOE Complex toward solving issues in modern nuclear code validation. In time, code owners, users, and developers will view the Center as a partner and essential resource for acquiring the best practices and latest techniques for validating codes, for guidance in planning and executing experiments, for facilitating access to, and maximizing the usefulness of, existing data, and for preserving knowledge for continual use by nuclear professionalsmore » and organizations for their own validation needs. The scope of the center covers many inter-related activities which will need to be cultivated carefully in the near-term and managed properly once the Center is fully functional. Three areas comprise the principal mission: 1) identification and prioritization of projects that extend the field of validation science and its application to modern codes, 2) adapt or develop best practices and guidelines for high fidelity multiphysics/multiscale analysis code development and associated experiment design, and 3) define protocols for data acquisition and knowledge preservation and provide a portal for access to databases currently scattered among numerous organizations. These mission areas, while each having a unique focus, are inter-dependent and complementary. Likewise, all activities supported by the NEKVaC, both near-term and long-term), must possess elements supporting all three. This cross-cutting nature is essential to ensuring that activities and supporting personnel do not become ‘stove-piped’, i.e. focused so much on a specific function that the activity itself becomes the objective rather than the achieving the larger vision. Achieving the broader vision will require a healthy and accountable level of activity in each of the areas. This will take time and significant DOE support. Growing too fast (budget-wise) will not allow ideas to mature, lessons to be learned, and taxpayer money to be spent responsibly. The process should be initiated with a small set of tasks, executed over a short but reasonable term, that will exercise most if not all aspects of the Center’s potential operation. The initial activities described in this report have a high potential for near-term success in demonstrating Center objectives but also to work out some of the issues in task execution, communication between functional elements, and the ability to raise awareness of the Center and cement stakeholder buy-in. This report begins with a description of the Mission areas; specifically the role played by each and the types of activities for which they are responsible. It then lists and describes the proposed near-term tasks upon which future efforts can build.« less

KAOS/LIB-V: A library of nuclear response functions generated by KAOS-V code from ENDF/B-V and other data files

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

Farawila, Y.; Gohar, Y.; Maynard, C.

1989-04-01

KAOS/LIB-V: A library of processed nuclear responses for neutronics analyses of nuclear systems has been generated. The library was prepared using the KAOS-V code and nuclear data from ENDF/B-V. The library includes kerma (kinetic energy released in materials) factors and other nuclear response functions for all materials presently of interest in fusion and fission applications for 43 nonfissionable and 15 fissionable isotopes and elements. The nuclear response functions include gas production and tritium-breeding functions, and all important reaction cross sections. KAOS/LIB-V employs the VITAMIN-E weighting function and energy group structure of 174 neutron groups. Auxiliary nuclear data bases, e.g., themore » Japanese evaluated nuclear data library JENDL-2 were used as a source of isotopic cross sections when these data are not provided in ENDF/B-V files for a natural element. These are needed mainly to estimate average quantities such as effective Q-values for the natural element. This analysis of local energy deposition was instrumental in detecting and understanding energy balance deficiencies and other problems in the ENDF/B-V data. Pertinent information about the library and a graphical display of the main nuclear response functions for all materials in the library are given. 35 refs.« less

Kinetic: A system code for analyzing nuclear thermal propulsion rocket engine transients

NASA Astrophysics Data System (ADS)

Schmidt, Eldon; Lazareth, Otto; Ludewig, Hans

The topics are presented in viewgraph form and include the following: outline of kinetic code; a kinetic information flow diagram; kinetic neutronic equations; turbopump/nozzle algorithm; kinetic heat transfer equations per node; and test problem diagram.

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

DOE PAGES

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

2015-05-28

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

CEM2k and LAQGSM Codes as Event-Generators for Space Radiation Shield and Cosmic Rays Propagation Applications

NASA Technical Reports Server (NTRS)

Mashnik, S. G.; Gudima, K. K.; Sierk, A. J.; Moskalenko, I. V.

2002-01-01

Space radiation shield applications and studies of cosmic ray propagation in the Galaxy require reliable cross sections to calculate spectra of secondary particles and yields of the isotopes produced in nuclear reactions induced both by particles and nuclei at energies from threshold to hundreds of GeV per nucleon. Since the data often exist in a very limited energy range or sometimes not at all, the only way to obtain an estimate of the production cross sections is to use theoretical models and codes. Recently, we have developed improved versions of the Cascade-Exciton Model (CEM) of nuclear reactions: the codes CEM97 and CEM2k for description of particle-nucleus reactions at energies up to about 5 GeV. In addition, we have developed a LANL version of the Quark-Gluon String Model (LAQGSM) to describe reactions induced both by particles and nuclei at energies up to hundreds of GeVhucleon. We have tested and benchmarked the CEM and LAQGSM codes against a large variety of experimental data and have compared their results with predictions by other currently available models and codes. Our benchmarks show that CEM and LAQGSM codes have predictive powers no worse than other currently used codes and describe many reactions better than other codes; therefore both our codes can be used as reliable event-generators for space radiation shield and cosmic ray propagation applications. The CEM2k code is being incorporated into the transport code MCNPX (and several other transport codes), and we plan to incorporate LAQGSM into MCNPX in the near future. Here, we present the current status of the CEM2k and LAQGSM codes, and show results and applications to studies of cosmic ray propagation in the Galaxy.

i-PI: A Python interface for ab initio path integral molecular dynamics simulations

NASA Astrophysics Data System (ADS)

Ceriotti, Michele; More, Joshua; Manolopoulos, David E.

2014-03-01

Recent developments in path integral methodology have significantly reduced the computational expense of including quantum mechanical effects in the nuclear motion in ab initio molecular dynamics simulations. However, the implementation of these developments requires a considerable programming effort, which has hindered their adoption. Here we describe i-PI, an interface written in Python that has been designed to minimise the effort required to bring state-of-the-art path integral techniques to an electronic structure program. While it is best suited to first principles calculations and path integral molecular dynamics, i-PI can also be used to perform classical molecular dynamics simulations, and can just as easily be interfaced with an empirical forcefield code. To give just one example of the many potential applications of the interface, we use it in conjunction with the CP2K electronic structure package to showcase the importance of nuclear quantum effects in high-pressure water. Catalogue identifier: AERN_v1_0 Program summary URL: http://cpc.cs.qub.ac.uk/summaries/AERN_v1_0.html Program obtainable from: CPC Program Library, Queen’s University, Belfast, N. Ireland Licensing provisions: GNU General Public License, version 3 No. of lines in distributed program, including test data, etc.: 138626 No. of bytes in distributed program, including test data, etc.: 3128618 Distribution format: tar.gz Programming language: Python. Computer: Multiple architectures. Operating system: Linux, Mac OSX, Windows. RAM: Less than 256 Mb Classification: 7.7. External routines: NumPy Nature of problem: Bringing the latest developments in the modelling of nuclear quantum effects with path integral molecular dynamics to ab initio electronic structure programs with minimal implementational effort. Solution method: State-of-the-art path integral molecular dynamics techniques are implemented in a Python interface. Any electronic structure code can be patched to receive the atomic coordinates from the Python interface, and to return the forces and energy that are used to integrate the equations of motion. Restrictions: This code only deals with distinguishable particles. It does not include fermonic or bosonic exchanges between equivalent nuclei, which can become important at very low temperatures. Running time: Depends dramatically on the nature of the simulation being performed. A few minutes for short tests with empirical force fields, up to several weeks for production calculations with ab initio forces. The examples provided with the code run in less than an hour.

Method and apparatus for reading lased bar codes on shiny-finished fuel rod cladding tubes

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

Goldenfield, M.P.; Lambert, D.V.

1990-10-02

This patent describes, in a nuclear fuel rod identification system, a method of reading a bar code etched directly on a surface of a nuclear fuel rod. It comprises: defining a pair of light diffuser surfaces adjacent one another but in oppositely inclined relation to a beam of light emitted from a light reader; positioning a fuel rod, having a cylindrical surface portion with a bar code etched directly thereon, relative to the light diffuser surfaces such that the surfaces are disposed adjacent to and in oppositely inclined relation along opposite sides of the fuel rod surface portion and themore » fuel rod surface portion is aligned with the beam of light emitted from the light reader; directing the beam of light on the bar code on fuel rod cylindrical surface portion such that the light is reflected therefrom onto one of the light diffuser surfaces; and receiving and reading the reflected light from the bar code via the one of the light diffuser surfaces to the light reader.« less

Probabilistic accident consequence uncertainty analysis -- Late health effects uncertain assessment. Volume 2: Appendices

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

Little, M.P.; Muirhead, C.R.; Goossens, L.H.J.

1997-12-01

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library ofmore » uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA late health effects models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the expert panel on late health effects, (4) short biographies of the experts, and (5) the aggregated results of their responses.« less

Probabilistic accident consequence uncertainty analysis -- Uncertainty assessment for internal dosimetry. Volume 2: Appendices

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

Goossens, L.H.J.; Kraan, B.C.P.; Cooke, R.M.

1998-04-01

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. In 1991, the US Nuclear Regulatory Commission and the Commission of the European Communities began cosponsoring a joint uncertainty analysis of the two codes. The ultimate objective of this joint effort was to systematically develop credible and traceable uncertainty distributions for the respective code input variables. A formal expert judgment elicitation and evaluation process was identified as the best technology available for developing a library ofmore » uncertainty distributions for these consequence parameters. This report focuses on the results of the study to develop distribution for variables related to the MACCS and COSYMA internal dosimetry models. This volume contains appendices that include (1) a summary of the MACCS and COSYMA consequence codes, (2) the elicitation questionnaires and case structures, (3) the rationales and results for the panel on internal dosimetry, (4) short biographies of the experts, and (5) the aggregated results of their responses.« less

A Freirean Approach to Peacemaking.

ERIC Educational Resources Information Center

Moriarty, Pia

1989-01-01

The Nuclear Disarmament Project involving Catholic churches and schools in San Francisco used Freirean codifications, with photographs as codes, to develop discussions on the moral issues of nuclear arms. Group discussions led to concrete action in the cause of peace and social justice. (SK)

Pebble bed modular reactor safeguards: developing new approaches and implementing safeguards by design

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

Beyer, Brian David; Beddingfield, David H; Durst, Philip

2010-01-01

The design of the Pebble Bed Modular Reactor (PBMR) does not fit or seem appropriate to the IAEA safeguards approach under the categories of light water reactor (LWR), on-load refueled reactor (OLR, i.e. CANDU), or Other (prismatic HTGR) because the fuel is in a bulk form, rather than discrete items. Because the nuclear fuel is a collection of nuclear material inserted in tennis-ball sized spheres containing structural and moderating material and a PBMR core will contain a bulk load on the order of 500,000 spheres, it could be classified as a 'Bulk-Fuel Reactor.' Hence, the IAEA should develop unique safeguardsmore » criteria. In a multi-lab DOE study, it was found that an optimized blend of: (i) developing techniques to verify the plutonium content in spent fuel pebbles, (ii) improving burn-up computer codes for PBMR spent fuel to provide better understanding of the core and spent fuel makeup, and (iii) utilizing bulk verification techniques for PBMR spent fuel storage bins should be combined with the historic IAEA and South African approaches of containment and surveillance to verify and maintain continuity of knowledge of PBMR fuel. For all of these techniques to work the design of the reactor will need to accommodate safeguards and material accountancy measures to a far greater extent than has thus far been the case. The implementation of Safeguards-by-Design as the PBMR design progresses provides an approach to meets these safeguards and accountancy needs.« less