Vectorized Monte Carlo methods for reactor lattice analysis
Brown, F. B.
1984-01-01
Some of the new computational methods and equivalent mathematical representations of physics models used in the MCV code, a vectorized continuous-enery Monte Carlo code for use on the CYBER-205 computer are discussed. While the principal application of MCV is the neutronics analysis of repeating reactor lattices, the new methods used in MCV should be generally useful for vectorizing Monte Carlo for other applications. For background, a brief overview of the vector processing features of the CYBER-205 is included, followed by a discussion of the fundamentals of Monte Carlo vectorization. The physics models used in the MCV vectorized Monte Carlo code are then summarized. The new methods used in scattering analysis are presented along with details of several key, highly specialized computational routines. Finally, speedups relative to CDC-7600 scalar Monte Carlo are discussed.
Advanced Mesh-Enabled Monte carlo capability for Multi-Physics Reactor Analysis
Energy Technology Data Exchange (ETDEWEB)
Wilson, Paul; Evans, Thomas; Tautges, Tim
2012-12-24
This project will accumulate high-precision fluxes throughout reactor geometry on a non- orthogonal grid of cells to support multi-physics coupling, in order to more accurately calculate parameters such as reactivity coefficients and to generate multi-group cross sections. This work will be based upon recent developments to incorporate advanced geometry and mesh capability in a modular Monte Carlo toolkit with computational science technology that is in use in related reactor simulation software development. Coupling this capability with production-scale Monte Carlo radiation transport codes can provide advanced and extensible test-beds for these developments. Continuous energy Monte Carlo methods are generally considered to be the most accurate computational tool for simulating radiation transport in complex geometries, particularly neutron transport in reactors. Nevertheless, there are several limitations for their use in reactor analysis. Most significantly, there is a trade-off between the fidelity of results in phase space, statistical accuracy, and the amount of computer time required for simulation. Consequently, to achieve an acceptable level of statistical convergence in high-fidelity results required for modern coupled multi-physics analysis, the required computer time makes Monte Carlo methods prohibitive for design iterations and detailed whole-core analysis. More subtly, the statistical uncertainty is typically not uniform throughout the domain, and the simulation quality is limited by the regions with the largest statistical uncertainty. In addition, the formulation of neutron scattering laws in continuous energy Monte Carlo methods makes it difficult to calculate adjoint neutron fluxes required to properly determine important reactivity parameters. Finally, most Monte Carlo codes available for reactor analysis have relied on orthogonal hexahedral grids for tallies that do not conform to the geometric boundaries and are thus generally not well
Monte Carlo analysis of Very High Temperature gas-cooled Reactor for hydrogen production
Energy Technology Data Exchange (ETDEWEB)
Kim, J. G.; Kim, H. C.; Kim, S. Y.; Shin, C. H.; Han, C. Y.; Kim, J. C. [Hanyang Univ., Seoul (Korea, Republic of)
2006-03-15
This work has been pursued during 2 years. In the first year, the development of Monte Carlo analysis method for pebble-type VHTR core was focused with zero-power reactor. The pebble-bed cores of HTR-PROTEUS critical facility in Switzerland were selected for the benchmark model and detailed full-scope MCNP modeling was carried out. Especially, accurate and effective modeling of UO{sub 2} particles and their distributions in fuel pebble was pursed as well as the pebbles distribution within core region. After the detailed MCNP modeling of the whole facility, analyses of nuclear characteristics were carried out, and the results were compared with experiments and those of other research groups. The effective multiplication factors (k{sub eff}) were calculated for the two HTR-PROTEUS cores, and then homogenization effect of TRISO fuel on criticality investigated. Control rod and shutdown rod worths were also calculated, and the criticality calculations with different cross-section library and various reflector thickness were carried out. In the 2nd year of the research period, the Monte Carol analysis method developed in the 1st year was applied to the core with thermal power. The pebble-bed cores of HTR-10 test reactor in China were selected for the benchmark model. After the detailed full-scope MCNP modeling the Monte Carlo analysis results calculated in this work were verified with the benchmark results which have been done for first criticality state and initial core.
Monte Carlo analysis of the accelerator-driven system at Kyoto University Research Reactor Institute
Energy Technology Data Exchange (ETDEWEB)
Kim, Won Kyeong; Lee, Deok Jung [Nuclear Engineering Division, Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Lee, Hyun Chul [VHTR Technology Development Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Pyeon, Cheol Ho [Nuclear Engineering Science Division, Kyoto University Research Reactor Institute, Osaka (Japan); Shin, Ho Cheol [Core and Fuel Analysis Group, Korea Hydro and Nuclear Power Central Research Institute, Daejeon (Korea, Republic of)
2016-04-15
An accelerator-driven system consists of a subcritical reactor and a controllable external neutron source. The reactor in an accelerator-driven system can sustain fission reactions in a subcritical state using an external neutron source, which is an intrinsic safety feature of the system. The system can provide efficient transmutations of nuclear wastes such as minor actinides and long-lived fission products and generate electricity. Recently at Kyoto University Research Reactor Institute (KURRI; Kyoto, Japan), a series of reactor physics experiments was conducted with the Kyoto University Critical Assembly and a Cockcroft-Walton type accelerator, which generates the external neutron source by deuterium-tritium reactions. In this paper, neutronic analyses of a series of experiments have been re-estimated by using the latest Monte Carlo code and nuclear data libraries. This feasibility study is presented through the comparison of Monte Carlo simulation results with measurements.
Monte Carlo Analysis of the Accelerator-Driven System at Kyoto University Research Reactor Institute
Directory of Open Access Journals (Sweden)
Wonkyeong Kim
2016-04-01
Full Text Available An accelerator-driven system consists of a subcritical reactor and a controllable external neutron source. The reactor in an accelerator-driven system can sustain fission reactions in a subcritical state using an external neutron source, which is an intrinsic safety feature of the system. The system can provide efficient transmutations of nuclear wastes such as minor actinides and long-lived fission products and generate electricity. Recently at Kyoto University Research Reactor Institute (KURRI; Kyoto, Japan, a series of reactor physics experiments was conducted with the Kyoto University Critical Assembly and a Cockcroft–Walton type accelerator, which generates the external neutron source by deuterium–tritium reactions. In this paper, neutronic analyses of a series of experiments have been re-estimated by using the latest Monte Carlo code and nuclear data libraries. This feasibility study is presented through the comparison of Monte Carlo simulation results with measurements.
Nuclear reactor transient analysis via a quasi-static kinetics Monte Carlo method
Jo, YuGwon; Cho, Bumhee; Cho, Nam Zin
2015-12-01
The predictor-corrector quasi-static (PCQS) method is applied to the Monte Carlo (MC) calculation for reactor transient analysis. To solve the transient fixed-source problem of the PCQS method, fission source iteration is used and a linear approximation of fission source distributions during a macro-time step is introduced to provide delayed neutron source. The conventional particle-tracking procedure is modified to solve the transient fixed-source problem via MC calculation. The PCQS method with MC calculation is compared with the direct time-dependent method of characteristics (MOC) on a TWIGL two-group problem for verification of the computer code. Then, the results on a continuous-energy problem are presented.
Nuclear reactor transient analysis via a quasi-static kinetics Monte Carlo method
Energy Technology Data Exchange (ETDEWEB)
Jo, YuGwon; Cho, Bumhee; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr [Korea Advanced Institute of Science and Technology 291 Daehak-ro, Yuseong-gu, Daejeon, Korea 305-701 (Korea, Republic of)
2015-12-31
The predictor-corrector quasi-static (PCQS) method is applied to the Monte Carlo (MC) calculation for reactor transient analysis. To solve the transient fixed-source problem of the PCQS method, fission source iteration is used and a linear approximation of fission source distributions during a macro-time step is introduced to provide delayed neutron source. The conventional particle-tracking procedure is modified to solve the transient fixed-source problem via MC calculation. The PCQS method with MC calculation is compared with the direct time-dependent method of characteristics (MOC) on a TWIGL two-group problem for verification of the computer code. Then, the results on a continuous-energy problem are presented.
The Dynamic Monte Carlo Method for Transient Analysis of Nuclear Reactors
Sjenitzer, B.L.
2013-01-01
In this thesis a new method for the analysis of power transients in a nuclear reactor is developed, which is more accurate than the present state-of-the-art methods. Transient analysis is important tool when designing nuclear reactors, since they predict the behaviour of a reactor during changing co
Energy Technology Data Exchange (ETDEWEB)
Espinosa-Paredes, Gilberto, E-mail: gepe@xanum.uam.m [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco, 186, Col. Vicentina, Mexico D.F., 09340 (Mexico); Verma, Surendra P. [Centro de Investigacion en Energia, Universidad Nacional Autonoma de Mexico, Priv. Xochicalco s/no., Col Centro, Apartado Postal 34, Temixco 62580 (Mexico); Vazquez-Rodriguez, Alejandro [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco, 186, Col. Vicentina, Mexico D.F., 09340 (Mexico); Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Doctor Barragan 779, Col. Narvarte, Mexico D.F. 03020 (Mexico)
2010-05-15
Our aim was to evaluate the sensitivity and uncertainty of mass flow rate in the core on the performance of natural circulation boiling water reactor (NCBWR). This analysis was carried out through Monte Carlo simulations of sizes up to 40,000, and the size, i.e., repetition of 25,000 was considered as valid for routine applications. A simplified boiling water reactor (SBWR) was used as an application example of Monte Carlo method. The numerical code to simulate the SBWR performance considers a one-dimensional thermo-hydraulics model along with non-equilibrium thermodynamics and non-homogeneous flow approximation, one-dimensional fuel rod heat transfer. The neutron processes were simulated with a point reactor kinetics model with six groups of delayed neutrons. The sensitivity was evaluated in terms of 99% confidence intervals of the mean to understand the range of mean values that may represent the entire statistical population of performance variables. The regression analysis with mass flow rate as the predictor variable showed statistically valid linear correlations for both neutron flux and fuel temperature and quadratic relationship for the void fraction. No statistically valid correlation was observed for the total heat flux as a function of the mass flow rate although heat flux at individual nodes was positively correlated with this variable. These correlations are useful for the study, analysis and design of any NCBWR. The uncertainties were propagated as follows: for 10% change in the mass flow rate in the core, the responses for neutron power, total heat flux, average fuel temperature and average void fraction changed by 8.74%, 7.77%, 2.74% and 0.58%, respectively.
Monte Carlo modelling of TRIGA research reactor
El Bakkari, B.; Nacir, B.; El Bardouni, T.; El Younoussi, C.; Merroun, O.; Htet, A.; Boulaich, Y.; Zoubair, M.; Boukhal, H.; Chakir, M.
2010-10-01
The Moroccan 2 MW TRIGA MARK II research reactor at Centre des Etudes Nucléaires de la Maâmora (CENM) achieved initial criticality on May 2, 2007. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes for their use in agriculture, industry, and medicine. This study deals with the neutronic analysis of the 2-MW TRIGA MARK II research reactor at CENM and validation of the results by comparisons with the experimental, operational, and available final safety analysis report (FSAR) values. The study was prepared in collaboration between the Laboratory of Radiation and Nuclear Systems (ERSN-LMR) from Faculty of Sciences of Tetuan (Morocco) and CENM. The 3-D continuous energy Monte Carlo code MCNP (version 5) was used to develop a versatile and accurate full model of the TRIGA core. The model represents in detailed all components of the core with literally no physical approximation. Continuous energy cross-section data from the more recent nuclear data evaluations (ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1, and JENDL-3.3) as well as S( α, β) thermal neutron scattering functions distributed with the MCNP code were used. The cross-section libraries were generated by using the NJOY99 system updated to its more recent patch file "up259". The consistency and accuracy of both the Monte Carlo simulation and neutron transport physics were established by benchmarking the TRIGA experiments. Core excess reactivity, total and integral control rods worth as well as power peaking factors were used in the validation process. Results of calculations are analysed and discussed.
Monte Carlo based radial shield design of typical PWR reactor
Energy Technology Data Exchange (ETDEWEB)
Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.
2016-11-15
Neutron and gamma flux and dose equivalent rate distribution are analysed in radial and shields of a typical PWR type reactor based on the Monte Carlo radiation transport computer code MCNP5. The ENDF/B-VI continuous energy cross-section library has been employed for the criticality and shielding analysis. The computed results are in good agreement with the reference results (maximum difference is less than 56 %). It implies that MCNP5 a good tool for accurate prediction of neutron and gamma flux and dose rates in radial shield around the core of PWR type reactors.
Analysis of dpa Rates in the HFIR Reactor Vessel using a Hybrid Monte Carlo/Deterministic Method*
Directory of Open Access Journals (Sweden)
Risner J.M.
2016-01-01
Full Text Available The Oak Ridge High Flux Isotope Reactor (HFIR, which began full-power operation in 1966, provides one of the highest steady-state neutron flux levels of any research reactor in the world. An ongoing vessel integrity analysis program to assess radiation-induced embrittlement of the HFIR reactor vessel requires the calculation of neutron and gamma displacements per atom (dpa, particularly at locations near the beam tube nozzles, where radiation streaming effects are most pronounced. In this study we apply the Forward-Weighted Consistent Adjoint Driven Importance Sampling (FW-CADIS technique in the ADVANTG code to develop variance reduction parameters for use in the MCNP radiation transport code. We initially evaluated dpa rates for dosimetry capsule locations, regions in the vicinity of the HB-2 beamline, and the vessel beltline region. We then extended the study to provide dpa rate maps using three-dimensional cylindrical mesh tallies that extend from approximately 12 in. below to approximately 12 in. above the height of the core. The mesh tally structures contain over 15,000 mesh cells, providing a detailed spatial map of neutron and photon dpa rates at all locations of interest. Relative errors in the mesh tally cells are typically less than 1%.
Analysis of dpa rates in the HFIR reactor vessel using a hybrid Monte Carlo/deterministic method
Energy Technology Data Exchange (ETDEWEB)
Blakeman, Edward [Retired
2016-01-01
The Oak Ridge High Flux Isotope Reactor (HFIR), which began full-power operation in 1966, provides one of the highest steady-state neutron flux levels of any research reactor in the world. An ongoing vessel integrity analysis program to assess radiation-induced embrittlement of the HFIR reactor vessel requires the calculation of neutron and gamma displacements per atom (dpa), particularly at locations near the beam tube nozzles, where radiation streaming effects are most pronounced. In this study we apply the Forward-Weighted Consistent Adjoint Driven Importance Sampling (FW-CADIS) technique in the ADVANTG code to develop variance reduction parameters for use in the MCNP radiation transport code. We initially evaluated dpa rates for dosimetry capsule locations, regions in the vicinity of the HB-2 beamline, and the vessel beltline region. We then extended the study to provide dpa rate maps using three-dimensional cylindrical mesh tallies that extend from approximately 12 below to approximately 12 above the axial extent of the core. The mesh tally structures contain over 15,000 mesh cells, providing a detailed spatial map of neutron and photon dpa rates at all locations of interest. Relative errors in the mesh tally cells are typically less than 1%.
Energy Technology Data Exchange (ETDEWEB)
Kim, Chang Hyo; Hong, In Seob; Han, Beom Seok; Jeong, Jong Seong [Seoul National University, Seoul (Korea)
2002-03-01
The objective of this project is to verify neutronics characteristics of the SMART core design as to compare computational results of the MCNAP code with those of the MASTER code. To achieve this goal, we will analyze neutronics characteristics of the SMART core using the MCNAP code and compare these results with results of the MASTER code. We improved parallel computing module and developed error analysis module of the MCNAP code. We analyzed mechanism of the error propagation through depletion computation and developed a calculation module for quantifying these errors. We performed depletion analysis for fuel pins and assemblies of the SMART core. We modeled a 3-D structure of the SMART core and considered a variation of material compositions by control rods operation and performed depletion analysis for the SMART core. We computed control-rod worths of assemblies and a reactor core for operation of individual control-rod groups. We computed core reactivity coefficients-MTC, FTC and compared these results with computational results of the MASTER code. To verify error analysis module of the MCNAP code, we analyzed error propagation through depletion of the SMART B-type assembly. 18 refs., 102 figs., 36 tabs. (Author)
Monte Carlo Modeling Electronuclear Processes in Cascade Subcritical Reactor
Bznuni, S A; Zhamkochyan, V M; Polyanskii, A A; Sosnin, A N; Khudaverdian, A G
2000-01-01
Accelerator driven subcritical cascade reactor composed of the main thermal neutron reactor constructed analogous to the core of the VVER-1000 reactor and a booster-reactor, which is constructed similar to the core of the BN-350 fast breeder reactor, is taken as a model example. It is shown by means of Monte Carlo calculations that such system is a safe energy source (k_{eff}=0.94-0.98) and it is capable of transmuting produced radioactive wastes (neutron flux density in the thermal zone is PHI^{max} (r,z)=10^{14} n/(cm^{-2} s^{-1}), neutron flux in the fast zone is respectively equal PHI^{max} (r,z)=2.25 cdot 10^{15} n/(cm^{-2} s^{-1}) if the beam current of the proton accelerator is k_{eff}=0.98 and I=5.3 mA). Suggested configuration of the "cascade" reactor system essentially reduces the requirements on the proton accelerator current.
MONTE CARLO SIMULATIONS OF PERIODIC PULSED REACTOR WITH MOVING GEOMETRY PARTS
Energy Technology Data Exchange (ETDEWEB)
Cao, Yan; Gohar, Yousry
2015-11-01
In a periodic pulsed reactor, the reactor state varies periodically from slightly subcritical to slightly prompt supercritical for producing periodic power pulses. Such periodic state change is accomplished by a periodic movement of specific reactor parts, such as control rods or reflector sections. The analysis of such reactor is difficult to perform with the current reactor physics computer programs. Based on past experience, the utilization of the point kinetics approximations gives considerable errors in predicting the magnitude and the shape of the power pulse if the reactor has significantly different neutron life times in different zones. To accurately simulate the dynamics of this type of reactor, a Monte Carlo procedure using the transfer function TRCL/TR of the MCNP/MCNPX computer programs is utilized to model the movable reactor parts. In this paper, two algorithms simulating the geometry part movements during a neutron history tracking have been developed. Several test cases have been developed to evaluate these procedures. The numerical test cases have shown that the developed algorithms can be utilized to simulate the reactor dynamics with movable geometry parts.
Eigenvalue analysis using a full-core Monte Carlo method
Energy Technology Data Exchange (ETDEWEB)
Okafor, K.C.; Zino, J.F. (Westinghouse Savannah River Co., Aiken, SC (United States))
1992-01-01
The reactor physics codes used at the Savannah River Site (SRS) to predict reactor behavior have been continually benchmarked against experimental and operational data. A particular benchmark variable is the observed initial critical control rod position. Historically, there has been some difficulty predicting this position because of the difficulties inherent in using computer codes to model experimental or operational data. The Monte Carlo method is applied in this paper to study the initial critical control rod positions for the SRS K Reactor. A three-dimensional, full-core MCNP model of the reactor was developed for this analysis.
Monte Carlo modeling of Lead-Cooled Fast Reactor in adiabatic equilibrium state
Energy Technology Data Exchange (ETDEWEB)
Stanisz, Przemysław, E-mail: pstanisz@agh.edu.pl; Oettingen, Mikołaj, E-mail: moettin@agh.edu.pl; Cetnar, Jerzy, E-mail: cetnar@mail.ftj.agh.edu.pl
2016-05-15
Graphical abstract: - Highlights: • We present the Monte Carlo modeling of the LFR in the adiabatic equilibrium state. • We assess the adiabatic equilibrium fuel composition using the MCB code. • We define the self-adjusting process of breeding gain by the control rod operation. • The designed LFR can work in the adiabatic cycle with zero fuel breeding. - Abstract: Nuclear power would appear to be the only energy source able to satisfy the global energy demand while also achieving a significant reduction of greenhouse gas emissions. Moreover, it can provide a stable and secure source of electricity, and plays an important role in many European countries. However, nuclear power generation from its birth has been doomed by the legacy of radioactive nuclear waste. In addition, the looming decrease in the available resources of fissile U235 may influence the future sustainability of nuclear energy. The integrated solution to both problems is not trivial, and postulates the introduction of a closed-fuel cycle strategy based on breeder reactors. The perfect choice of a novel reactor system fulfilling both requirements is the Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state. In such a state, the reactor converts depleted or natural uranium into plutonium while consuming any self-generated minor actinides and transferring only fission products as waste. We present the preliminary design of a Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state with the Monte Carlo Continuous Energy Burnup Code – MCB. As a reference reactor model we apply the core design developed initially under the framework of the European Lead-cooled SYstem (ELSY) project and refined in the follow-up Lead-cooled European Advanced DEmonstration Reactor (LEADER) project. The major objective of the study is to show to what extent the constraints of the adiabatic cycle are maintained and to indicate the phase space for further improvements. The analysis
Monte Carlo simulation of NSE at reactor and spallation sources
Energy Technology Data Exchange (ETDEWEB)
Zsigmond, G.; Wechsler, D.; Mezei, F. [Hahn-Meitner-Institut Berlin, Berlin (Germany)
2001-03-01
A MC (Monte Carlo) computation study of NSE (Neutron Spin Echo) has been performed by means of VITESS investigating the classic and TOF-NSE options at spallation sources. The use of white beams in TOF-NSE makes the flipper efficiency in function of the neutron wavelength an important issue. The emphasis was put on exact evaluation of flipper efficiencies for wide wavelength-band instruments. (author)
Alhassan, Erwin; Sjöstrand, Henrik; Duan, Junfeng; Gustavsson, Cecilia; Koning, Arjan; Pomp, Stephan; Rochman, Dimitri; Österlund, Michael
2014-01-01
Analyses are carried out to assess the impact of nuclear data uncertainties on some reactor safety parameters for the European Lead Cooled Training Reactor (ELECTRA) using the Total Monte Carlo method. A large number of Pu-239 random ENDF-format libraries, generated using the TALYS based system were processed into ACE format with NJOY99.336 code and used as input into the Serpent Monte Carlo code to obtain distribution in reactor safety parameters. The distribution in keff obtained was compar...
On the Calculation of Reactor Time Constants Using the Monte Carlo Method
Energy Technology Data Exchange (ETDEWEB)
Leppaenen, Jaakko [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland)
2008-07-01
Full-core reactor dynamics calculation involves the coupled modelling of thermal hydraulics and the time-dependent behaviour of core neutronics. The reactor time constants include prompt neutron lifetimes, neutron reproduction times, effective delayed neutron fractions and the corresponding decay constants, typically divided into six or eight precursor groups. The calculation of these parameters is traditionally carried out using deterministic lattice transport codes, which also produce the homogenised few-group constants needed for resolving the spatial dependence of neutron flux. In recent years, there has been a growing interest in the production of simulator input parameters using the stochastic Monte Carlo method, which has several advantages over deterministic transport calculation. This paper reviews the methodology used for the calculation of reactor time constants. The calculation techniques are put to practice using two codes, the PSG continuous-energy Monte Carlo reactor physics code and MORA, a new full-core Monte Carlo neutron transport code entirely based on homogenisation. Both codes are being developed at the VTT Technical Research Centre of Finland. The results are compared to other codes and experimental reference data in the CROCUS reactor kinetics benchmark calculation. (author)
Energy Technology Data Exchange (ETDEWEB)
Guerra, Bruno T.; Pereira, Claubia, E-mail: brunoteixeiraguerra@yahoo.com.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (DEN/UFMG), Belo Horizonte, MG (Brazil). Departmento de Energia Nuclear; Soares, Alexandre L.; Menezes, Maria Angela B.C., E-mail: menezes@cdtn.br, E-mail: asleal@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)
2015-07-01
The IPR-R1 is a reactor type TRIGA, Mark-I model, manufactured by the General Atomic Company and installed at Nuclear Technology Development Centre (CDTN), Brazilian Commission for Nuclear Energy (CNEN), in Belo Horizonte, Brazil. It is a light water moderated and cooled, graphite-reflected, open-pool type research reactor and operates at 100 kW. It presents low power, low pressure, for application in research, training and radioisotopes production. The fuel is an alloy of zirconium hydride and uranium enriched at 20% in {sup 235}U. The implementation of the PGNAA (Prompt Gamma Neutron Activation Analysis) using this research reactor will significantly increase in number of chemical elements analysed and the kind of matrices. A project is underway in order to implement this technique at CDTN. The objective of this study was to contribute in feasibility analysis of implementing this technique. For this purpose, MCNP is being used. Some variance reduction tools in the methodology, that has been already developed, was introduced for calculating of the neutron flux in the neutron extractor inclined. The objective was to reduce the code error and thereby increasing the reliability of the results. With the implementation of the variance reduction tools, the results of the thermal and epithermal neutron fluxes presented a significant improvement in both calculations. (author)
Energy Technology Data Exchange (ETDEWEB)
Villafan-Vidales, H.I.; Arancibia-Bulnes, C.A.; Dehesa-Carrasco, U. [Centro de Investigacion en Energia, Universidad Nacional Autonoma de Mexico, Privada Xochicalco s/n, Col. Centro, A.P. 34, Temixco, Morelos 62580 (Mexico); Romero-Paredes, H. [Departamento de Ingenieria de Procesos e Hidraulica, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco No.186, Col. Vicentina, A.P. 55-534, Mexico D.F 09340 (Mexico)
2009-01-15
Radiative heat transfer in a solar thermochemical reactor for the thermal reduction of cerium oxide is simulated with the Monte Carlo method. The directional characteristics and the power distribution of the concentrated solar radiation that enters the cavity is obtained by carrying out a Monte Carlo ray tracing of a paraboloidal concentrator. It is considered that the reactor contains a gas/particle suspension directly exposed to concentrated solar radiation. The suspension is treated as a non-isothermal, non-gray, absorbing, emitting, and anisotropically scattering medium. The transport coefficients of the particles are obtained from Mie-scattering theory by using the optical properties of cerium oxide. From the simulations, the aperture radius and the particle concentration were optimized to match the characteristics of the considered concentrator. (author)
Energy Technology Data Exchange (ETDEWEB)
Abdel-Khalik, Hany S. [North Carolina State Univ., Raleigh, NC (United States); Zhang, Qiong [North Carolina State Univ., Raleigh, NC (United States)
2014-05-20
The development of hybrid Monte-Carlo-Deterministic (MC-DT) approaches, taking place over the past few decades, have primarily focused on shielding and detection applications where the analysis requires a small number of responses, i.e. at the detector locations(s). This work further develops a recently introduced global variance reduction approach, denoted by the SUBSPACE approach is designed to allow the use of MC simulation, currently limited to benchmarking calculations, for routine engineering calculations. By way of demonstration, the SUBSPACE approach is applied to assembly level calculations used to generate the few-group homogenized cross-sections. These models are typically expensive and need to be executed in the order of 10^{3} - 10^{5} times to properly characterize the few-group cross-sections for downstream core-wide calculations. Applicability to k-eigenvalue core-wide models is also demonstrated in this work. Given the favorable results obtained in this work, we believe the applicability of the MC method for reactor analysis calculations could be realized in the near future.
Nuclear Reactor Engineering Analysis Laboratory
Energy Technology Data Exchange (ETDEWEB)
Carlos Chavez-Mercado; Jaime B. Morales-Sandoval; Benjamin E. Zayas-Perez
1998-12-31
The Nuclear Reactor Engineering Analysis Laboratory (NREAL) is a sophisticated computer system with state-of-the-art analytical tools and technology for analysis of light water reactors. Multiple application software tools can be activated to carry out different analyses and studies such as nuclear fuel reload evaluation, safety operation margin measurement, transient and severe accident analysis, nuclear reactor instability, operator training, normal and emergency procedures optimization, and human factors engineering studies. An advanced graphic interface, driven through touch-sensitive screens, provides the means to interact with specialized software and nuclear codes. The interface allows the visualization and control of all observable variables in a nuclear power plant (NPP), as well as a selected set of nonobservable or not directly controllable variables from conventional control panels.
Directory of Open Access Journals (Sweden)
MEHMET E. KORKMAZ
2014-06-01
Full Text Available In this research, we investigated the burnup characteristics and the conversion of fertile 232Th into fissile 233U in the core of a Sodium-Cooled Fast Reactor (SFR. The SFR fuel assemblies were designed for burning 232Th fuel (fuel pin 1 and 233U fuel (fuel pin 2 and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method and TTA (Transmutation Trajectory Analysis method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff was between 0.964 and 0.954 and peaking factor is 1.88867.
Energy Technology Data Exchange (ETDEWEB)
Korkmaz, Mehmet E.; Agar, Osman [Karamanoglu Mehmetbey University, Faculty of Kamil Oezdag Science, Karaman (Turkmenistan)
2014-06-15
In this research, we investigated the burnup characteristics and the conversion of fertile {sup 232}Th into fissile {sup 233}U in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning {sup 232}Th fuel (fuel pin 1) and {sup 233}U fuel (fuel pin 2) and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method) and TTA (Transmutation Trajectory Analysis) method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff) was between 0.964 and 0.954 and peaking factor is 1.88867.
Fernandes, A C; Gonçalves, I C; Santos, J; Cardoso, J; Santos, L; Ferro Carvalho, A; Marques, J G; Kling, A; Ramalho, A J G; Osvay, M
2006-01-01
This work presents an extensive study on Monte Carlo radiation transport simulation and thermoluminescent (TL) dosimetry for characterising mixed radiation fields (neutrons and photons) occurring in nuclear reactors. The feasibility of these methods is investigated for radiation fields at various locations of the Portuguese Research Reactor (RPI). The performance of the approaches developed in this work is compared with dosimetric techniques already existing at RPI. The Monte Carlo MCNP-4C code was used for a detailed modelling of the reactor core, the fast neutron beam and the thermal column of RPI. Simulations using these models allow to reproduce the energy and spatial distributions of the neutron field very well (agreement better than 80%). In the case of the photon field, the agreement improves with decreasing intensity of the component related to fission and activation products. (7)LiF:Mg,Ti, (7)LiF:Mg,Cu,P and Al(2)O(3):Mg,Y TL detectors (TLDs) with low neutron sensitivity are able to determine photon dose and dose profiles with high spatial resolution. On the other hand, (nat)LiF:Mg,Ti TLDs with increased neutron sensitivity show a remarkable loss of sensitivity and a high supralinearity in high-intensity fields hampering their application at nuclear reactors.
Generation of XS library for the reflector of VVER reactor core using Monte Carlo code Serpent
Usheva, K. I.; Kuten, S. A.; Khruschinsky, A. A.; Babichev, L. F.
2017-01-01
A physical model of the radial and axial reflector of VVER-1200-like reactor core has been developed. Five types of radial reflector with different material composition exist for the VVER reactor core and 1D and 2D models were developed for all of them. Axial top and bottom reflectors are described by the 1D model. A two-group XS library for diffusion code DYN3D has been generated for all types of reflectors by using Serpent 2 Monte Carlo code. Power distribution in the reactor core calculated in DYN3D is flattened in the core central region to more extent in the 2D model of the radial reflector than in its 1D model.
Ma, X. B.; Qiu, R. M.; Chen, Y. X.
2017-02-01
Uncertainties regarding fission fractions are essential in understanding antineutrino flux predictions in reactor antineutrino experiments. A new Monte Carlo-based method to evaluate the covariance coefficients between isotopes is proposed. The covariance coefficients are found to vary with reactor burnup and may change from positive to negative because of balance effects in fissioning. For example, between 235U and 239Pu, the covariance coefficient changes from 0.15 to -0.13. Using the equation relating fission fraction and atomic density, consistent uncertainties in the fission fraction and covariance matrix were obtained. The antineutrino flux uncertainty is 0.55%, which does not vary with reactor burnup. The new value is about 8.3% smaller.
Energy Technology Data Exchange (ETDEWEB)
Liang, Jingang; Wang, Kan; Qiu, Yishu [Dept. of Engineering Physics, LiuQing Building, Tsinghua University, Beijing (China); Chai, Xiao Ming; Qiang, Sheng Long [Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu (China)
2016-06-15
Because of prohibitive data storage requirements in large-scale simulations, the memory problem is an obstacle for Monte Carlo (MC) codes in accomplishing pin-wise three-dimensional (3D) full-core calculations, particularly for whole-core depletion analyses. Various kinds of data are evaluated and quantificational total memory requirements are analyzed based on the Reactor Monte Carlo (RMC) code, showing that tally data, material data, and isotope densities in depletion are three major parts of memory storage. The domain decomposition method is investigated as a means of saving memory, by dividing spatial geometry into domains that are simulated separately by parallel processors. For the validity of particle tracking during transport simulations, particles need to be communicated between domains. In consideration of efficiency, an asynchronous particle communication algorithm is designed and implemented. Furthermore, we couple the domain decomposition method with MC burnup process, under a strategy of utilizing consistent domain partition in both transport and depletion modules. A numerical test of 3D full-core burnup calculations is carried out, indicating that the RMC code, with the domain decomposition method, is capable of pin-wise full-core burnup calculations with millions of depletion regions.
The Specific Bias in Dynamic Monte Carlo Simulations of Nuclear Reactor
Yamamoto, Toshihisa; Endo, Hiroshi; Ishizu, Tomoko; Tatewaki, Isao
2014-06-01
During the development of Monte-Carlo-based dynamic code system, we have encountered two major Monte-Carlo-specific problems. One is the break down due to "false super-criticality" which is caused by an accidentally large eigenvalue due to statistical error in spite of the fact that the reactor is actually not. The other problem, which is the main topic in this paper, is that the statistical error in power level using the reactivity calculated with Monte Carlo code is not symmetric about its mean but always positively biased. This signifies that the bias is accumulated as the calculation proceeds and consequently results in over-estimation of the final power level. It should be noted that the bias will not eliminated by refining time step as long as the variance is not zero. A preliminary investigation on this matter using the one-group-precursor point kinetic equations was made and it was concluded that the bias in power level is approximately proportional to the product of variance in Monte Carlo calculation and elapsed time. This conclusion was verified with some numerical experiments. This outcome is important in quantifying the required precision of the Monte-Carlo-based reactivity calculations.
An Advanced Neutronic Analysis Toolkit with Inline Monte Carlo capability for BHTR Analysis
Energy Technology Data Exchange (ETDEWEB)
William R. Martin; John C. Lee
2009-12-30
Monte Carlo capability has been combined with a production LWR lattice physics code to allow analysis of high temperature gas reactor configurations, accounting for the double heterogeneity due to the TRISO fuel. The Monte Carlo code MCNP5 has been used in conjunction with CPM3, which was the testbench lattice physics code for this project. MCNP5 is used to perform two calculations for the geometry of interest, one with homogenized fuel compacts and the other with heterogeneous fuel compacts, where the TRISO fuel kernels are resolved by MCNP5.
Shchurovskaya, M. V.; Alferov, V. P.; Geraskin, N. I.; Radaev, A. I.
2017-01-01
The results of the validation of a research reactor calculation using Monte Carlo and deterministic codes against experimental data and based on code-to-code comparison are presented. The continuous energy Monte Carlo code MCU-PTR and the nodal diffusion-based deterministic code TIGRIS were used for full 3-D calculation of the IRT MEPhI research reactor. The validation included the investigations for the reactor with existing high enriched uranium (HEU, 90 w/o) fuel and low enriched uranium (LEU, 19.7 w/o, U-9%Mo) fuel.
Reactor Physics Analysis Models for a CANDU Reactor
Energy Technology Data Exchange (ETDEWEB)
Choi, Hang Bok
2007-10-15
Canada deuterium uranium (CANDU) reactor physics analysis is typically performed in three steps. At first, macroscopic cross-sections of the reference lattice is produced by modeling the reference fuel channel. Secondly macroscopic cross-sections of reactivity devices in the reactor are generated. The macroscopic cross-sections of a reactivity device are calculated as incremental cross-sections by subtracting macroscopic cross-sections of a three-dimensional lattice without reactivity device from those of a three-dimensional lattice with a reactivity device. Using the macroscopic cross-sections of the reference lattice and incremental cross-sections of the reactivity devices, reactor physics calculations are performed. This report summarizes input data of typical CANDU reactor physics codes, which can be utilized for the future CANDU reactor physics analysis.
Reactor Physics Analysis Models for a CANDU Reactor
Energy Technology Data Exchange (ETDEWEB)
Choi, Hang Bok
2007-10-15
Canada deuterium uranium (CANDU) reactor physics analysis is typically performed in three steps. At first, macroscopic cross-sections of the reference lattice is produced by modeling the reference fuel channel. Secondly macroscopic cross-sections of reactivity devices in the reactor are generated. The macroscopic cross-sections of a reactivity device are calculated as incremental cross-sections by subtracting macroscopic cross-sections of a three-dimensional lattice without reactivity device from those of a three-dimensional lattice with a reactivity device. Using the macroscopic cross-sections of the reference lattice and incremental cross-sections of the reactivity devices, reactor physics calculations are performed. This report summarizes input data of typical CANDU reactor physics codes, which can be utilized for the future CANDU reactor physics analysis.
Monte Carlo optimisation of a BNCT facility for treating brain gliomas at the TAPIRO reactor.
Nava, E; Burn, K W; Casalini, L; Petrovich, C; Rosi, G; Sarotto, M; Tinti, R
2005-01-01
An epithermal boron neutron capture therapy facility for treating brain gliomas is currently under construction at the 5 kW fast-flux reactor TAPIRO located at ENEA, Casaccia, near Rome. In this work, the sensitivity of the results to the boron concentrations in healthy tissue and tumour is investigated and the change in beam quality on modifying the moderator thickness (within design limits) is studied. The Monte Carlo codes MCNP and MCNPX were used together with the DSA in-house variance reduction patch. Both usual free beam parameters and the in-phantom treatment planning figures-of-merit have been calculated in a realistic anthropomorphic phantom ('ADAM').
Energy Technology Data Exchange (ETDEWEB)
Amharrak, H.; Reynard-Carette, C.; Carette, M. [Aix Marseille Universite, CNRS, Universite de Toulon, IM2NP UMR 7334, 13397, Marseille (France); Lemaire, M.; Vaglio-Gaudard, C. [CEA, DEN, DER, SPRC, LPN, Cadarache, F-13108 Saint Paul Lez Durance (France); Fourmentel, D.; Lyoussi, A. [CEA, DEN, Departement d' Etudes des Reacteurs, Service de Physique Experimentale, Laboratoire Dosimetrie Capteurs Instrumentation, 13108 Saint-Paul-lez-Durance (France)
2015-07-01
The nuclear heating measurements in Material Testing Reactors (MTRs) are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material. These measurements are then used for other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices. Nuclear heating is a great deal of interest at the moment as the measurement of such heating is an important issue for MTRs reactors. This need is especially generated by the new Jules Horowitz Reactor (JHR), under construction at CEA/Cadarache 'French Alternative Energies and Atomic Energy Commission'. This new reactor, that will be operational in late 2019, is a new facility for the nuclear research on materials and fuels. Indeed the expected nuclear heating rate is about 20 W/g for nominal capacity of 100 MW. The present Monte Carlo calculation works belong to the IN-CORE (Instrumentation for Nuclear radiation and Calorimetry On line in Reactor): a joint research program between the CEA and Aix- Marseille University in 2009. One scientific aim of this program is to design and develop a multi-sensors device, called CARMEN, dedicated to the measurements of main physical parameters simultaneously encountered inside JHR's experimental channels (core and reflector) such as neutron fluxes, photon fluxes, temperature, and nuclear heating. A first prototype was already developed. This prototype includes two mock-ups dedicated respectively to neutronic measurements (CARMEN-1N) and to photonic measurements (CARMEN-1P) with in particular a specific differential calorimeter. Two irradiation campaigns were performed successfully in the periphery of OSIRIS reactor (a MTR located at Saclay, France) in 2012 for nuclear heating levels up to 2 W/g. First Monte Carlo calculations reduced to the graphite sample of the
Some new viewpoints in reactor noise analysis
Institute of Scientific and Technical Information of China (English)
罗征培; 李富; 等
1996-01-01
It is propsed that the linearity criterion and order criterion via frequency spectrum features without any limitation of the model's phase can be used in reactor noise analysis.The time constant,natural frequency as well as the recovered transfer function of reactors can bhe obtained via the analyzable model based on reactor noise.
Bznuni, S A; Zhamkochyan, V M; Polanski, A; Sosnin, A N; Khudaverdyan, A H
2001-01-01
Parameters of a subcritical cascade reactor driven by a proton accelerator and based on a primary lead-bismuth target, main reactor constructed analogously to the molten salt breeder (MSBR) reactor core and a booster-reactor analogous to the core of the BN-350 liquid metal cooled fast breeder reactor (LMFBR). It is shown by means of Monte-Carlo modeling that the reactor under study provides safe operation modes (k_{eff}=0.94-0.98), is apable to transmute effectively radioactive nuclear waste and reduces by an order of magnitude the requirements on the accelerator beam current. Calculations show that the maximal neutron flux in the thermal zone is 10^{14} cm^{12}\\cdot s^_{-1}, in the fast booster zone is 5.12\\cdot10^{15} cm^{12}\\cdot s{-1} at k_{eff}=0.98 and proton beam current I=2.1 mA.
Chiapetto, M.; Messina, L.; Becquart, C. S.; Olsson, P.; Malerba, L.
2017-02-01
This work presents a revised set of parameters to be used in an Object kinetic Monte Carlo model to simulate the microstructure evolution under neutron irradiation of reactor pressure vessel steels at the operational temperature of light water reactors (∼300 °C). Within a "grey-alloy" approach, a more physical description than in a previous work is used to translate the effect of Mn and Ni solute atoms on the defect cluster diffusivity reduction. The slowing down of self-interstitial clusters, due to the interaction between solutes and crowdions in Fe is now parameterized using binding energies from the latest DFT calculations and the solute concentration in the matrix from atom-probe experiments. The mobility of vacancy clusters in the presence of Mn and Ni solute atoms was also modified on the basis of recent DFT results, thereby removing some previous approximations. The same set of parameters was seen to predict the correct microstructure evolution for two different types of alloys, under very different irradiation conditions: an Fe-C-MnNi model alloy, neutron irradiated at a relatively high flux, and a high-Mn, high-Ni RPV steel from the Swedish Ringhals reactor surveillance program. In both cases, the predicted self-interstitial loop density matches the experimental solute cluster density, further corroborating the surmise that the MnNi-rich nanofeatures form by solute enrichment of immobilized small interstitial loops, which are invisible to the electron microscope.
Dynamic analysis of process reactors
Energy Technology Data Exchange (ETDEWEB)
Shadle, L.J.; Lawson, L.O.; Noel, S.D.
1995-06-01
The approach and methodology of conducting a dynamic analysis is presented in this poster session in order to describe how this type of analysis can be used to evaluate the operation and control of process reactors. Dynamic analysis of the PyGas{trademark} gasification process is used to illustrate the utility of this approach. PyGas{trademark} is the gasifier being developed for the Gasification Product Improvement Facility (GPIF) by Jacobs-Siffine Engineering and Riley Stoker. In the first step of the analysis, process models are used to calculate the steady-state conditions and associated sensitivities for the process. For the PyGas{trademark} gasifier, the process models are non-linear mechanistic models of the jetting fluidized-bed pyrolyzer and the fixed-bed gasifier. These process sensitivities are key input, in the form of gain parameters or transfer functions, to the dynamic engineering models.
San Carlos Apache Tribe - Energy Organizational Analysis
Energy Technology Data Exchange (ETDEWEB)
Rapp, James; Albert, Steve
2012-04-01
The San Carlos Apache Tribe (SCAT) was awarded $164,000 in late-2011 by the U.S. Department of Energy (U.S. DOE) Tribal Energy Program's "First Steps Toward Developing Renewable Energy and Energy Efficiency on Tribal Lands" Grant Program. This grant funded: The analysis and selection of preferred form(s) of tribal energy organization (this Energy Organization Analysis, hereinafter referred to as "EOA"). Start-up staffing and other costs associated with the Phase 1 SCAT energy organization. An intern program. Staff training. Tribal outreach and workshops regarding the new organization and SCAT energy programs and projects, including two annual tribal energy summits (2011 and 2012). This report documents the analysis and selection of preferred form(s) of a tribal energy organization.
Energy Technology Data Exchange (ETDEWEB)
Bastida O, G. E.; Francois L, J. L., E-mail: gbo729@yahoo.com.mx [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)
2014-10-15
In this work a neutron analysis of the fuel of some high temperature nuclear reactors is presented, studying its main features, besides some alternatives of compound fuel by uranium and plutonium, and of coolant: sodium and helium. For this study was necessary the use of a code able to carry out a reliable calculation of the main parameters of the fuel. The use of the Monte Carlo method was convenient to simulate the neutrons transport in the reactor core, which is the base of the Serpent code, with which the calculations will be made for the analysis. (Author)
Clement, S D; Choi, J R; Zamenhof, R G; Yanch, J C; Harling, O K
1990-01-01
Monte Carlo methods of coupled neutron/photon transport are being used in the design of filtered beams for Neutron Capture Therapy (NCT). This method of beam analysis provides segregation of each individual dose component, and thereby facilitates beam optimization. The Monte Carlo method is discussed in some detail in relation to NCT epithermal beam design. Ideal neutron beams (i.e., plane-wave monoenergetic neutron beams with no primary gamma-ray contamination) have been modeled both for comparison and to establish target conditions for a practical NCT epithermal beam design. Detailed models of the 5 MWt Massachusetts Institute of Technology Research Reactor (MITR-II) together with a polyethylene head phantom have been used to characterize approximately 100 beam filter and moderator configurations. Using the Monte Carlo methodology of beam design and benchmarking/calibrating our computations with measurements, has resulted in an epithermal beam design which is useful for therapy of deep-seated brain tumors. This beam is predicted to be capable of delivering a dose of 2000 RBE-cGy (cJ/kg) to a therapeutic advantage depth of 5.7 cm in polyethylene assuming 30 micrograms/g 10B in tumor with a ten-to-one tumor-to-blood ratio, and a beam diameter of 18.4 cm. The advantage ratio (AR) is predicted to be 2.2 with a total irradiation time of approximately 80 minutes. Further optimization work on the MITR-II epithermal beams is expected to improve the available beams.
McCARD for Neutronics Design and Analysis of Research Reactor Cores
Shim, Hyung Jin; Park, Ho Jin; Kwon, Soonwoo; Seo, Geon Ho; Hyo Kim, Chang
2014-06-01
McCARD is a Monte Carlo (MC) neutron-photon transport simulation code developed exclusively for the neutronics design and analysis of nuclear reactor cores. McCARD is equipped with the hierarchical modeling and scripting functions, the CAD-based geometry processing module, the adjoint-weighted kinetics parameter and source multiplication factor estimation modules as well as the burnup analysis capability for the neutronics design and analysis of both research and power reactor cores. This paper highlights applicability of McCARD for the research reactor core neutronics analysis, as demonstrated for Kyoto University Critical Assembly, HANARO, and YALINA.
Amharrak, H.; Reynard-Carette, C.; Lyoussi, A.; Carette, M.; Brun, J.; De Vita, C.; Fourmentel, D.; Villard, J.-F.; Guimbal, P.
2016-02-01
The nuclear heating measurements in Material Testing Reactors (MTRs) are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material. Then these measurements are used for other materials, other geometries, or other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices. This paper will present new simulations with MCNP Monte-Carlo transport code to determine the gamma heating profile inside the calorimeter. The whole complex geometry of the sensor has been considered. We use as an input source in the model, the photon spectra calculated in various positions of CARMEN-1 irradiation program in OSIRIS reactor. After a description of the differential calorimeter device, the MCNP modeling used for the calculations of radial profile of nuclear heating inside the calorimeter elements will be introduced. The obtained results of different simulations will be detailed and discussed in this paper. The charged particle equilibrium inside the calorimeter elements will be studied. Then we will focus on parametric studies of the various components of the calorimeter. The influence of source type will be also took into account. Moreover the influence of the material used for the sample will be described.
Directory of Open Access Journals (Sweden)
Amharrak H.
2016-01-01
Full Text Available The nuclear heating measurements in Material Testing Reactors (MTRs are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material. Then these measurements are used for other materials, other geometries, or other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices. This paper will present new simulations with MCNP Monte-Carlo transport code to determine the gamma heating profile inside the calorimeter. The whole complex geometry of the sensor has been considered. We use as an input source in the model, the photon spectra calculated in various positions of CARMEN-1 irradiation program in OSIRIS reactor. After a description of the differential calorimeter device, the MCNP modeling used for the calculations of radial profile of nuclear heating inside the calorimeter elements will be introduced. The obtained results of different simulations will be detailed and discussed in this paper. The charged particle equilibrium inside the calorimeter elements will be studied. Then we will focus on parametric studies of the various components of the calorimeter. The influence of source type will be also took into account. Moreover the influence of the material used for the sample will be described.
PCCF flow analysis -- DR Reactor
Energy Technology Data Exchange (ETDEWEB)
Calkin, J.F.
1961-04-26
This report contains an analysis of PCCF tube flow and Panellit pressure relations at DR reactor. Supply curves are presented at front header pressures from 480 to 600 psig using cold water and the standard 0.236 inch orifice with taper down stream and the pigtail valve (plug or ball) open. Demand curves are presented for slug column lengths of 200 inches to 400 inches using 1.44 inch O.D. solid poison pieces (either Al or Pb-Cd) and cold water with a rear header pressure of 50 psig. Figure 1 is a graph of Panellit pressure vs. flow with the above supply and demand curves and clearly shows the effect of front header pressure and charge length on flow.
Shoushtari, M. K.; Kakavand, T.; Sadat Kiai, S. M.; Ghaforian, H.
2010-03-01
The Monte Carlo simulation has been established for a research reactor with nominal power of 7 MW. A detailed model of the reactor core was employed including standard and control fuel elements, reflectors, irradiation channels, control rods, reactor pool and thermal column. The following physical parameters of reactor core were calculated for the present LEU core: core reactivity ( ρ), control rod (CR) worth, thermal and epithermal neutron flux distributions, shutdown margin and delayed neutron fraction. Reduction of unfavorable effects of blockage probability of control safety rod (CSR)s in their interiors because of not enough space in their sites, and lack of suitable capabilities to fabricate very thin plates for CSR cladding, is the main aim of the present study. Making the absorber rod thinner and CSR cladding thicker by introducing a better blackness absorbing material and a new stainless steel alloy, respectively, are two studied ways to reduce the effects of mentioned problems.
Thermal Analysis for Mobile Reactor
Institute of Scientific and Technical Information of China (English)
2008-01-01
<正>Mobile reactor design in the paper is consisted of two grades of thermal electric conversion. The first grade is the thermionic conversion inside the core and the second grade is thermocouple conversion
Mathematical Modeling for Simulation of Nuclear Reactor Analysis
Salah Ud-Din Khan; Shahab Ud-Din Khan
2013-01-01
In this paper, we have developed a mathematical model for the nuclear reactor analysis to be implemented in the nuclear reactor code. THEATRe is nuclear reactor analysis code which can only work for the cylindrical type fuel reactor and cannot applicable for the plate type fuel nuclear reactor. Therefore, the current studies encompasses on the modification of THEATRe code for the plate type fuel element. This mathematical model is applicable to the thermal analysis of the reactor which is ver...
Analysis of Adiabatic Batch Reactor
Directory of Open Access Journals (Sweden)
Erald Gjonaj
2016-05-01
Full Text Available A mixture of acetic anhydride is reacted with excess water in an adiabatic batch reactor to form an exothermic reaction. The concentration of acetic anhydride and the temperature inside the adiabatic batch reactor are calculated with an initial temperature of 20°C, an initial temperature of 30°C, and with a cooling jacket maintaining the temperature at a constant of 20°C. The graphs of the three different scenarios show that the highest temperatures will cause the reaction to occur faster.
International Electrotechnical Commission. Geneva
1988-01-01
This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.
An analysis of Monte Carlo tree search
CSIR Research Space (South Africa)
James, S
2017-02-01
Full Text Available Monte Carlo Tree Search (MCTS) is a family of directed search algorithms that has gained widespread attention in recent years. Despite the vast amount of research into MCTS, the effect of modifications on the algorithm, as well as the manner...
Gritzay, Olena; Kalchenko, Oleksandr; Klimova, Nataliya; Razbudey, Volodymyr; Sanzhur, Andriy; Binney, Stephen
2005-05-01
The presented results show our consecutive steps in developing a neutron source with parameters required by Boron Neutron Capture Therapy (BNCT) at the Kyiv Research Reactor (KRR). The main goal of this work was to analyze the influence of installation of different types of uranium converters close to the reactor core on neutron beam characteristics and on level of reactor safety. The general Monte Carlo radiation transport code MCNP, version 4B, has been used for these calculations.
Bayesian phylogeny analysis via stochastic approximation Monte Carlo
Cheon, Sooyoung
2009-11-01
Monte Carlo methods have received much attention in the recent literature of phylogeny analysis. However, the conventional Markov chain Monte Carlo algorithms, such as the Metropolis-Hastings algorithm, tend to get trapped in a local mode in simulating from the posterior distribution of phylogenetic trees, rendering the inference ineffective. In this paper, we apply an advanced Monte Carlo algorithm, the stochastic approximation Monte Carlo algorithm, to Bayesian phylogeny analysis. Our method is compared with two popular Bayesian phylogeny software, BAMBE and MrBayes, on simulated and real datasets. The numerical results indicate that our method outperforms BAMBE and MrBayes. Among the three methods, SAMC produces the consensus trees which have the highest similarity to the true trees, and the model parameter estimates which have the smallest mean square errors, but costs the least CPU time. © 2009 Elsevier Inc. All rights reserved.
Bayesian phylogeny analysis via stochastic approximation Monte Carlo.
Cheon, Sooyoung; Liang, Faming
2009-11-01
Monte Carlo methods have received much attention in the recent literature of phylogeny analysis. However, the conventional Markov chain Monte Carlo algorithms, such as the Metropolis-Hastings algorithm, tend to get trapped in a local mode in simulating from the posterior distribution of phylogenetic trees, rendering the inference ineffective. In this paper, we apply an advanced Monte Carlo algorithm, the stochastic approximation Monte Carlo algorithm, to Bayesian phylogeny analysis. Our method is compared with two popular Bayesian phylogeny software, BAMBE and MrBayes, on simulated and real datasets. The numerical results indicate that our method outperforms BAMBE and MrBayes. Among the three methods, SAMC produces the consensus trees which have the highest similarity to the true trees, and the model parameter estimates which have the smallest mean square errors, but costs the least CPU time.
Fuel burnup analysis for Thai research reactor by using MCNPX computer code
Sangkaew, S.; Angwongtrakool, T.; Srimok, B.
2017-06-01
This paper presents the fuel burnup analysis of the Thai research reactor (TRR-1/M1), TRIGA Mark-III, operated by Thailand Institute of Nuclear Technology (TINT) in Bangkok, Thailand. The modelling software used in this analysis is MCNPX (MCNP eXtended) version 2.6.0, a Fortran90 Monte Carlo radiation transport computer code. The analysis results will cover the core excess reactivity, neutron fluxes at the irradiation positions and neutron detector tubes, power distribution, fuel burnup, and fission products based on fuel cycle of first reactor core arrangement.
Comparative Criticality Analysis of Two Monte Carlo Codes on Centrifugal Atomizer: MCNPS and SCALE
Energy Technology Data Exchange (ETDEWEB)
Kang, H-S; Jang, M-S; Kim, S-R [NESS, Daejeon (Korea, Republic of); Park, J-M; Kim, K-N [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2015-10-15
There are two well-known Monte Carlo codes for criticality analysis, MCNP5 and SCALE. MCNP5 is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical system as a main analysis code. 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. SCALE was conceived and funded by US NRC to perform standardized computer analysis for licensing evaluation and is used widely in the world. We performed a validation test of MCNP5 and a comparative analysis of Monte Carlo codes, MCNP5 and SCALE, in terms of the critical analysis of centrifugal atomizer. In the criticality analysis using MCNP5 code, we obtained the statistically reliable results by using a large number of source histories per cycle and performing of uncertainty analysis.
Verification of HELIOS/MASTER Nuclear Analysis System for SMART Research Reactor, Rev. 1.0
Energy Technology Data Exchange (ETDEWEB)
Lee, Kyung Hoon; Kim, Kang Seog; Cho, Jin Young; Lee, Chung Chan; Zee, Sung Quun
2005-12-15
Nuclear design for the SMART reactor is performed by using the transport lattice code HELIOS and the core analysis code MASTER. HELIOS code developed by Studsvik Scandpower in Norway is a transport lattice code for the neutron and gamma behavior, and is used to generate few group constants. MASTER code is a nodal diffusion code developed by KAERI, and is used to analyze reactor physics. This nuclear design code package requires verification. Since the SMART reactor is unique, it is impossible to verify this code system through the comparison of the calculation results with the measured ones. Therefore, the uncertainties for the nuclear physics parameters calculated by HELIOS/MASTER have been evaluated indirectly. Since Monte Carlo calculation includes least approximations an assumptions to simulate a neutron behavior, HELIOS/MASTER has been verified by this one. Monte Carlo code has been verified by the Kurchatov critical experiments similar to SMART reactor, and HELIOS/MASTER code package has been verified by Monte Carlo calculations for the SMART research reactor.
Verification of HELIOS/MASTER Nuclear Analysis System for SMART Research Reactor
Energy Technology Data Exchange (ETDEWEB)
Kim, Kang Seog; Cho, Jin Young; Lee, Chung Chan; Zee, Sung Quun
2005-07-15
Nuclear design for the SMART reactor is performed by using the transport lattice code HELIOS and the core analysis code MASTER. HELIOS code developed by Studsvik Scandpower in Norway is a transport lattice code for the neutron and gamma behavior, and is used to generate few group constants. MASTER code is a nodal diffusion code developed by KAERI, and is used to analyze reactor physics. This nuclear design code package requires verification. Since the SMART reactor is unique, it is impossible to verify this code system through the comparison of the calculation results with the measured ones. Therefore, the uncertainties for the nuclear physics parameters calculated by HELIOS/MASTER have been evaluated indirectly. Since Monte Carlo calculation includes least approximations an assumptions to simulate a neutron behavior, HELIOS/MASTER has been verified by this one. Monte Carlo code has been verified by the Kurchatov critical experiments similar to SMART reactor, and HELIOS/MASTER code package has been verified by Monte Carlo calculations for the SMART research reactor.
Energy Technology Data Exchange (ETDEWEB)
Rodenas, J.; Gallardo, S.
2011-07-01
The control rods of a boiling water reactor (BWR) are subject to a neutron flux and thus become activated during their stay in the reactor core. Activation occurs especially in the stainless steel components and impurities. The activity generated results in a dose around the bar, while it le unimportant in the reactor, but to be taken into account when removed f ron it. The bars drawn are stored on hangers placed in the storage pools of spent fuel f ron the plant. Each hanger 12 accommodates control rods and are arranged so that at least three meters of water abode the heads of the control rods. The dose received by potentially exposed workers who are in the vicinity of the storage must be calculated to ensure adequate protection of the came. This dose can be decreased significantly by changing the arrangement of the bars on hangers.
Energy Technology Data Exchange (ETDEWEB)
Ghoos, K., E-mail: kristel.ghoos@kuleuven.be [KU Leuven, Department of Mechanical Engineering, Celestijnenlaan 300A, 3001 Leuven (Belgium); Dekeyser, W. [KU Leuven, Department of Mechanical Engineering, Celestijnenlaan 300A, 3001 Leuven (Belgium); Samaey, G. [KU Leuven, Department of Computer Science, Celestijnenlaan 200A, 3001 Leuven (Belgium); Börner, P. [Institute of Energy and Climate Research (IEK-4), FZ Jülich GmbH, D-52425 Jülich (Germany); Baelmans, M. [KU Leuven, Department of Mechanical Engineering, Celestijnenlaan 300A, 3001 Leuven (Belgium)
2016-10-01
The plasma and neutral transport in the plasma edge of a nuclear fusion reactor is usually simulated using coupled finite volume (FV)/Monte Carlo (MC) codes. However, under conditions of future reactors like ITER and DEMO, convergence issues become apparent. This paper examines the convergence behaviour and the numerical error contributions with a simplified FV/MC model for three coupling techniques: Correlated Sampling, Random Noise and Robbins Monro. Also, practical procedures to estimate the errors in complex codes are proposed. Moreover, first results with more complex models show that an order of magnitude speedup can be achieved without any loss in accuracy by making use of averaging in the Random Noise coupling technique.
Directory of Open Access Journals (Sweden)
Jingang Liang
2016-06-01
Full Text Available Because of prohibitive data storage requirements in large-scale simulations, the memory problem is an obstacle for Monte Carlo (MC codes in accomplishing pin-wise three-dimensional (3D full-core calculations, particularly for whole-core depletion analyses. Various kinds of data are evaluated and quantificational total memory requirements are analyzed based on the Reactor Monte Carlo (RMC code, showing that tally data, material data, and isotope densities in depletion are three major parts of memory storage. The domain decomposition method is investigated as a means of saving memory, by dividing spatial geometry into domains that are simulated separately by parallel processors. For the validity of particle tracking during transport simulations, particles need to be communicated between domains. In consideration of efficiency, an asynchronous particle communication algorithm is designed and implemented. Furthermore, we couple the domain decomposition method with MC burnup process, under a strategy of utilizing consistent domain partition in both transport and depletion modules. A numerical test of 3D full-core burnup calculations is carried out, indicating that the RMC code, with the domain decomposition method, is capable of pin-wise full-core burnup calculations with millions of depletion regions.
The PEC reactor. Safety analysis: Detailed reports
Energy Technology Data Exchange (ETDEWEB)
1988-01-01
In the safety-analysis of the PEC Brasimone reactor (Italy), attention was focused on the role of plant-incident analysis during the design stage and the conclusions reached. The analysis regarded the following: thermohydraulic incidents at full power; incidents with the reactor shut down; reactivity incidents; core local faults; analysis of fuel-handling incidents; engineered safeguards and passive safety features; coolant leakage and sodium fires; research and development studies on the seismic behaviour of the PEC fast reactor; generalized sodium fire; severe accidents, accident sequences with shudown; reference accident. Both the theoretical and experimental analyses demonstrated the adequacy of the design of the PEC fast reactor, aimed at minimizing the consequences of a hypothetical disruptive core accident with mechanical energy release. It was shown that the containment barriers were sized correctly and that the residual heat from a disassembled core would be removed. The re-evaluation of the source term emphasized the conservative nature of the hypotheses assumed in the preliminary safety analysis for calculating the risk to the public.
Griesheimer, D. P.; Gill, D. F.; Nease, B. R.; Sutton, T. M.; Stedry, M. H.; Dobreff, P. S.; Carpenter, D. C.; Trumbull, T. H.; Caro, E.; Joo, H.; Millman, D. L.
2014-06-01
MC21 is a continuous-energy Monte Carlo radiation transport code for the calculation of the steady-state spatial distributions of reaction rates in three-dimensional models. The code supports neutron and photon transport in fixed source problems, as well as iterated-fission-source (eigenvalue) neutron transport problems. MC21 has been designed and optimized to support large-scale problems in reactor physics, shielding, and criticality analysis applications. The code also supports many in-line reactor feedback effects, including depletion, thermal feedback, xenon feedback, eigenvalue search, and neutron and photon heating. MC21 uses continuous-energy neutron/nucleus interaction physics over the range from 10-5 eV to 20 MeV. The code treats all common neutron scattering mechanisms, including fast-range elastic and non-elastic scattering, and thermal- and epithermal-range scattering from molecules and crystalline materials. For photon transport, MC21 uses continuous-energy interaction physics over the energy range from 1 keV to 100 GeV. The code treats all common photon interaction mechanisms, including Compton scattering, pair production, and photoelectric interactions. All of the nuclear data required by MC21 is provided by the NDEX system of codes, which extracts and processes data from EPDL-, ENDF-, and ACE-formatted source files. For geometry representation, MC21 employs a flexible constructive solid geometry system that allows users to create spatial cells from first- and second-order surfaces. The system also allows models to be built up as hierarchical collections of previously defined spatial cells, with interior detail provided by grids and template overlays. Results are collected by a generalized tally capability which allows users to edit integral flux and reaction rate information. Results can be collected over the entire problem or within specific regions of interest through the use of phase filters that control which particles are allowed to score each
Directory of Open Access Journals (Sweden)
Lecouey Jean-Luc
2015-01-01
Full Text Available The GUINEVERE project was launched in 2006, within the 6th Euratom Framework Program IP-EUROTRANS, in order to study the feasibility of transmutation in Accelerator Driven subcritical Systems (ADS. This zero-power facility hosted at the SCK·CEN site in Mol (Belgium couples the fast subcritical lead reactor VENUS-F with an external neutron source provided by interaction of deuterons delivered by the GENEPI-3C accelerator and a tritiated target located at the reactor core center. In order to test on-line subcriticality monitoring techniques, the reactivity of all the VENUS-F configurations used must be known beforehand to serve as benchmark values. That is why the Modified Source Multiplication Method (MSM is under consideration to estimate the reactivity worth of the control rods when the reactor is largely subcritical as well as near-critical. The MSM method appears to be a technique well adapted to measure control rod worth over a large range of subcriticality levels. The MSM factors which are required to account for spatial effects in the reactor can be successfully calculated using a Monte Carlo neutron transport code.
Energy Technology Data Exchange (ETDEWEB)
Palau, J.M. [CEA Cadarache, Service de Physique des Reacteurs et du Cycle, Lab. de Projets Nucleaires, 13 - Saint-Paul-lez-Durance (France)
2005-07-01
This paper presents how Monte-Carlo calculations (French TRIPOLI4 poly-kinetic code with an appropriate pre-processing and post-processing software called OVNI) are used in the case of 3-dimensional heterogeneous benchmarks (slab reactor cores) to reduce model biases and enable a thorough and detailed analysis of the performances of deterministic methods and their associated data libraries with respect to key neutron parameters (reactivity, local power). Outstanding examples of application of these tools are presented regarding the new numerical methods implemented in the French lattice code APOLLO2 (advanced self-shielding models, new IDT characteristics method implemented within the discrete-ordinates flux solver model) and the JEFF3.1 nuclear data library (checked against JEF2.2 previous file). In particular we have pointed out, by performing multigroup/point-wise TRIPOLI4 (assembly and core) calculations, the efficiency (in terms of accuracy and computation time) of the new IDT method developed in APOLLO2. In addition, by performing 3-dimensional TRIPOLI4 calculations of the whole slab core (few millions of elementary volumes), the high quality of the new JEFF3.1 nuclear data files and revised evaluations (U{sup 235}, U{sup 238}, Hf) for reactivity prediction of slab cores critical experiments has been stressed. As a feedback of the whole validation process, improvements in terms of nuclear data (mainly Hf capture cross-sections) and numerical methods (advanced quadrature formulas accounting validation results, validation of new self-shielding models, parallelization) are suggested to improve even more the APOLLO2-CRONOS2 standard calculation route. (author)
Molten Salt Breeder Reactor Analysis Methods
Energy Technology Data Exchange (ETDEWEB)
Park, Jinsu; Jeong, Yongjin; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)
2015-05-15
Utilizing the uranium-thorium fuel cycle shows considerable potential for the possibility of MSR. The concept of MSBR should be revised because of molten salt reactor's advantage such as outstanding neutron economy, possibility of continuous online reprocessing and refueling, a high level of inherent safety, and economic benefit by keeping off the fuel fabrication process. For the development of MSR research, this paper provides the MSBR single-cell, two-cell and whole core model for computer code input, and several calculation results including depletion calculation of each models. The calculations are carried out by using MCNP6, a Monte Carlo computer code, which has CINDER90 for depletion calculation using ENDF-VII nuclear data. From the calculation results of various reactor design parameters, the temperature coefficients are all negative at the initial state and MTC becomes positive at the equilibrium state. From the results of core rod worth, the graphite control rod alone cannot makes the core subcritical at initial state. But the equilibrium state, the core can be made subcritical state only by graphite control rods. Through the comparison of the results of each models, the two-cell method can represent the MSBR core model more accurately with a little more computational resources than the single-cell method. Many of the thermal spectrum MSR have adopted a multi-region single-fluid strategy.
Reaction kinetic analysis of reactor surveillance data
Yoshiie, T.; Kinomura, A.; Nagai, Y.
2017-02-01
In the reactor pressure vessel surveillance data of a European-type pressurized water reactor (low-Cu steel), it was found that the concentration of matrix defects was very high, and a large number of precipitates existed. In this study, defect structure evolution obtained from surveillance data was simulated by reaction kinetic analysis using 15 rate equations. The saturation of precipitation and the growth of loops were simulated, but it was not possible to explain the increase in DBTT on the basis of the defect structures. The sub-grain boundary segregation of solutes was discussed for the origin of the DBTT increase.
Monte-Carlo Application for Nondestructive Nuclear Waste Analysis
Carasco, C.; Engels, R.; Frank, M.; Furletov, S.; Furletova, J.; Genreith, C.; Havenith, A.; Kemmerling, G.; Kettler, J.; Krings, T.; Ma, J.-L.; Mauerhofer, E.; Neike, D.; Payan, E.; Perot, B.; Rossbach, M.; Schitthelm, O.; Schumann, M.; Vasquez, R.
2014-06-01
Radioactive waste has to undergo a process of quality checking in order to check its conformance with national regulations prior to its transport, intermediate storage and final disposal. Within the quality checking of radioactive waste packages non-destructive assays are required to characterize their radio-toxic and chemo-toxic contents. The Institute of Energy and Climate Research - Nuclear Waste Management and Reactor Safety of the Forschungszentrum Jülich develops in the framework of cooperation nondestructive analytical techniques for the routine characterization of radioactive waste packages at industrial-scale. During the phase of research and development Monte Carlo techniques are used to simulate the transport of particle, especially photons, electrons and neutrons, through matter and to obtain the response of detection systems. The radiological characterization of low and intermediate level radioactive waste drums is performed by segmented γ-scanning (SGS). To precisely and accurately reconstruct the isotope specific activity content in waste drums by SGS measurement, an innovative method called SGSreco was developed. The Geant4 code was used to simulate the response of the collimated detection system for waste drums with different activity and matrix configurations. These simulations allow a far more detailed optimization, validation and benchmark of SGSreco, since the construction of test drums covering a broad range of activity and matrix properties is time consuming and cost intensive. The MEDINA (Multi Element Detection based on Instrumental Neutron Activation) test facility was developed to identify and quantify non-radioactive elements and substances in radioactive waste drums. MEDINA is based on prompt and delayed gamma neutron activation analysis (P&DGNAA) using a 14 MeV neutron generator. MCNP simulations were carried out to study the response of the MEDINA facility in terms of gamma spectra, time dependence of the neutron energy spectrum
Energy Technology Data Exchange (ETDEWEB)
Song, Wei [Nuclear and Radiation Safety Center, P. O. Box 8088, Beijing (China); Wu, Yuanyu [ITER Organization, Route de Vinon-sur-Verdon, 13115 Saint-Paul-lès-Durance (France); Hu, Wenjun [China Institute of Atomic Energy, P. O. Box 275(34), Beijing (China); Zuo, Jiaxu, E-mail: zuojiaxu@chinansc.cn [Nuclear and Radiation Safety Center, P. O. Box 8088, Beijing (China)
2015-11-15
Highlights: • Monte-Carlo principle coupling with fire dynamic code is adopted to perform sodium fire vulnerability assessment. • The method can be used to calculate the failure probability of sodium fire scenarios. • A calculation example and results are given to illustrate the feasibility of the methodology. • Some critical parameters and experience are shared. - Abstract: Sodium fire is a typical and distinctive hazard in sodium cooled fast reactors, which is significant for nuclear safety. In this paper, a method of sodium fire vulnerability assessment based on the Monte-Carlo principle was introduced, which could be used to calculate the probabilities of every failure mode in sodium fire scenarios. After that, the sodium fire scenario vulnerability assessment of primary cold trap room of China Experimental Fast Reactor was performed to illustrate the feasibility of the methodology. The calculation result of the example shows that the conditional failure probability of key cable is 23.6% in the sodium fire scenario which is caused by continuous sodium leakage because of the isolation device failure, but the wall temperature, the room pressure and the aerosol discharge mass are all lower than the safety limits.
PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR.
Energy Technology Data Exchange (ETDEWEB)
Cheng, L.; Diamond, D.; Xu, J.; Carew, J.; Rorer, D.
2004-03-31
Detailed reactor physics and safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analyses provide an update to the Final Safety Analysis Report (FSAR) and employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron and photon transport calculations were performed with the MCNP code to determine the safety parameters for the NBSR. The core depletion and determination of the fuel compositions were performed with MONTEBURNS. MCNP calculations were performed to determine the beginning, middle, and end-of-cycle power distributions, moderator temperature coefficient, and shim safety arm, beam tube and void reactivity worths. The calculational model included a plate-by-plate description of each fuel assembly, axial mid-plane water gap, beam tubes and the tubular geometry of the shim safety arms. The time-dependent analysis of the primary loop was determined with a RELAP5 transient analysis model that includes the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. The statistical analysis used to assure protection from critical heat flux (CHF) was performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF were determined with MCNP. Evaluations were performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element. In both the startup and maximum reactivity insertion accidents, the
Alhassan, E.; Sjöstrand, H.; Duan, J.; Gustavsson, C.; Koning, A. J.; Pomp, S.; Rochman, D.; Österlund, M.
2014-04-01
Analyses are carried out to assess the impact of nuclear data uncertainties on keff for the European Lead Cooled Training Reactor (ELECTRA) using the Total Monte Carlo method. A large number of 239Pu random ENDF-formatted libraries generated using the TALYS based system were processed into ACE format with NJOY-99.336 code and used as input into the Serpent Monte Carlo neutron transport code to obtain distribution in keff. The mean of the keff distribution obtained was compared with the major nuclear data libraries, JEFF-3.1.1, ENDF/B-VII.1 and JENDL-4.0. A method is proposed for the selection of benchmarks for specific applications using the Total Monte Carlo approach. Finally, an accept/reject criterion was investigated based on χ2 values obtained using the 239Pu Jezebel criticality benchmark. It was observed that nuclear data uncertainties in keff were reduced considerably from 748 to 443 pcm by applying a more rigid acceptance criteria for accepting random files.
Analysis of muon radiography of the Toshiba nuclear critical assembly reactor
Energy Technology Data Exchange (ETDEWEB)
Morris, C. L.; Bacon, Jeffery; Borozdin, Konstantin; Fabritius, J. M.; Perry, John; Ramsey, John [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States); Ban, Yuichiro; Izumi, Mikio; Sano, Yuji; Yoshida, Noriyuki [Toshiba Corporation, 8 Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan); Miyadera, Haruo [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States); Toshiba Corporation, 8 Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan); Mizokami, Shinya; Otsuka, Yasuyuki; Yamada, Daichi [Tokyo Electric Power Company, 1-1-3 Uchisaiwai-cho, Chiyoda-ku, Tokyo (Japan); Sugita, Tsukasa; Yoshioka, Kenichi [Toshiba Corporation, 4-1 Ukishima-cho, Kawasaki-ku, Kawasaki 210-0862 (Japan)
2014-01-13
A 1.2 × 1.2 m{sup 2} muon tracker was moved from Los Alamos to the Toshiba facility at Kawasaki, Japan, where it was used to take ∼4 weeks of data radiographing the Toshiba Critical Assembly Reactor with cosmic ray muons. In this paper, we describe the analysis procedure, show results of this experiment, and compare the results to Monte Carlo predictions. The results validate the concept of using cosmic rays to image the damaged cores of the Fukushima Daiichi reactors.
Lee, Y.-K.; Brun, E.
2014-04-01
The Sodium-cooled fast neutron reactor ASTRID is currently under design and development in France. Traditional ECCO/ERANOS fast reactor code system used for ASTRID core design calculations relies on multi-group JEFF-3.1.1 data library. To gauge the use of ENDF/B-VII.0 and JEFF-3.1.1 nuclear data libraries in the fast reactor applications, two recent OECD/NEA computational benchmarks specified by Argonne National Laboratory were calculated. Using the continuous-energy TRIPOLI-4 Monte Carlo transport code, both ABR-1000 MWth MOX core and metallic (U-Pu) core were investigated. Under two different fast neutron spectra and two data libraries, ENDF/B-VII.0 and JEFF-3.1.1, reactivity impact studies were performed. Using JEFF-3.1.1 library under the BOEC (Beginning of equilibrium cycle) condition, high reactivity effects of 808 ± 17 pcm and 1208 ± 17 pcm were observed for ABR-1000 MOX core and metallic core respectively. To analyze the causes of these differences in reactivity, several TRIPOLI-4 runs using mixed data libraries feature allow us to identify the nuclides and the nuclear data accounting for the major part of the observed reactivity discrepancies.
Structural analysis of reactor fuel elements
Energy Technology Data Exchange (ETDEWEB)
Weeks, R.W.
1977-01-01
An overview of fuel-element modeling is presented that traces the development of codes for the prediction of light-water-reactor and fast-breeder-reactor fuel-element performance. It is concluded that although the mathematical analysis is now far advanced, the development and incorporation of mechanistic constitutive equations has not kept pace. The resultant reliance on empirical correlations severely limits the physical insight that can be gained from code extrapolations. Current efforts include modeling of alternate fuel systems, analysis of local fuel-cladding interactions, and development of a predictive capability for off-normal behavior. Future work should help remedy the current constitutive deficiencies and should include the development of deterministic failure criteria for use in design.
Reactivity determination in accelerator driven reactors using reactor noise analysis
Directory of Open Access Journals (Sweden)
Kostić Ljiljana 1
2002-01-01
Full Text Available Feynman-alpha and Rossi-alpha methods are used in traditional nuclear reactors to determine the subcritical reactivity of a system. The methods are based on the measurement of the mean value, variance and the covariance of detector counts for different measurement times. Such methods attracted renewed attention recently with the advent of the so-called accelerator driven reactors (ADS proposed some time ago. The ADS systems, intended to be used either in energy generation or transuranium transmutation, will use a subcritical core with a strong spallation source. A spallation source has statistical properties that are different from those traditionally used by radioactive sources. In such reactors the monitoring of the subcritical reactivity is very important, and a statistical method, such as the Feynman-alpha method, is capable of resolving this problem.
Stress analysis of magnetically controlled reactor
Directory of Open Access Journals (Sweden)
Ben Tong
2017-05-01
Full Text Available To provide technique references for vibration reduction of magnetically controlled reactors (MCRs, stress, which is the inherent reason of vibration and noise, should be investigated. Stresses in reactor cores are produced due to the magnetostriction deformation of silicon steel and electromagnetic force between the core discs. So far, stress analysis on reactor cores was based on one-way coupled numerical method, which did not consider the influence of the stress on magnetic properties of the core material. Thus, multi-group magnetization and magnetostriction characteristics curves of silicon steel under different tensile stresses are measured firstly to support the computation. From the experiment results, it can be seen that magnetic properties of silicon steel change with stress. Then an electromagneto-mechanical two-way coupled numerical model for MCRs considering magnetostrictive effect and electromagnetic force effect is proposed. Stress distribution of MCR cores under the combination excitation of the sinusoidal wave current and different direct currents are calculated. From the computed results, it can be seen that a larger direct current has greater influence on MCRs vibration, which provides a theory basis for further analysis of vibration and noise reduction.
Stress analysis of magnetically controlled reactor
Tong, Ben; Qingxin, Yang; Rongge, Yan; Lihua, Zhu; Ling, Weng; Ying, Sun
2017-05-01
To provide technique references for vibration reduction of magnetically controlled reactors (MCRs), stress, which is the inherent reason of vibration and noise, should be investigated. Stresses in reactor cores are produced due to the magnetostriction deformation of silicon steel and electromagnetic force between the core discs. So far, stress analysis on reactor cores was based on one-way coupled numerical method, which did not consider the influence of the stress on magnetic properties of the core material. Thus, multi-group magnetization and magnetostriction characteristics curves of silicon steel under different tensile stresses are measured firstly to support the computation. From the experiment results, it can be seen that magnetic properties of silicon steel change with stress. Then an electromagneto-mechanical two-way coupled numerical model for MCRs considering magnetostrictive effect and electromagnetic force effect is proposed. Stress distribution of MCR cores under the combination excitation of the sinusoidal wave current and different direct currents are calculated. From the computed results, it can be seen that a larger direct current has greater influence on MCRs vibration, which provides a theory basis for further analysis of vibration and noise reduction.
Sherbini, S; Tamasanis, D; Sykes, J; Porter, S W
1986-12-01
A program was developed to calculate the exposure rate resulting from airborne gases inside a reactor containment building. The calculations were performed at the location of a wall-mounted area radiation monitor. The program uses Monte Carlo techniques and accounts for both the direct and scattered components of the radiation field at the detector. The scattered component was found to contribute about 30% of the total exposure rate at 50 keV and dropped to about 7% at 2000 keV. The results of the calculations were normalized to unit activity per unit volume of air in the containment. This allows the exposure rate readings of the area monitor to be used to estimate the airborne activity in containment in the early phases of an accident. Such estimates, coupled with containment leak rates, provide a method to obtain a release rate for use in offsite dose projection calculations.
Systems analysis of the CANDU 3 Reactor
Energy Technology Data Exchange (ETDEWEB)
Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H. [Oak Ridge National Lab., TN (United States)
1993-07-01
This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.
Energy Technology Data Exchange (ETDEWEB)
Messina, Luca; Olsson, Paer [KTH Royal Institute of Technology, Stockholm (Sweden); Chiapetto, Monica [SCK - CEN, Nuclear Materials Science Institute, Mol (Belgium); Unite Materiaux et Transformations (UMET), UMR 8207, Universite de Lille 1, ENSCL, Villeneuve d' Ascq (France); Becquart, Charlotte S. [Unite Materiaux et Transformations (UMET), UMR 8207, Universite de Lille 1, ENSCL, Villeneuve d' Ascq (France); Malerba, Lorenzo [SCK - CEN, Nuclear Materials Science Institute, Mol (Belgium)
2016-11-15
This work presents a full object kinetic Monte Carlo framework for the simulation of the microstructure evolution of reactor pressure vessel (RPV) steels. The model pursues a ''gray-alloy'' approach, where the effect of solute atoms is seen exclusively as a reduction of the mobility of defect clusters. The same set of parameters yields a satisfactory evolution for two different types of alloys, in very different irradiation conditions: an Fe-C-MnNi model alloy (high flux) and a high-Mn, high-Ni RPV steel (low flux). A satisfactory match with the experimental characterizations is obtained only if assuming a substantial immobilization of vacancy clusters due to solute atoms, which is here verified by means of independent atomistic kinetic Monte Carlo simulations. The microstructure evolution of the two alloys is strongly affected by the dose rate; a predominance of single defects and small defect clusters is observed at low dose rates, whereas larger defect clusters appear at high dose rates. In both cases, the predicted density of interstitial loops matches the experimental solute-cluster density, suggesting that the MnNi-rich nanofeatures might form as a consequence of solute enrichment on immobilized small interstitial loops, which are invisible to the electron microscope. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)
Lázaro, Ignacio; Ródenas, José; Marques, José G.; Gallardo, Sergio
2014-06-01
Materials in a nuclear reactor are activated by neutron irradiation. When they are withdrawn from the reactor and placed in some storage, the potential dose received by workers in the surrounding area must be taken into account. In previous papers, activation of control rods in a NPP with BWR and dose rates around the storage pool have been estimated using the MCNP5 code based on the Monte Carlo method. Models were validated comparing simulation results with experimental measurements. As the activation is mostly produced in stainless steel components of control rods the activation model can be also validated by means of experimental measurements on a stainless steel sample after being irradiated in a reactor. This has been done in the Portuguese Research Reactor at Instituto Tecnológico e Nuclear. The neutron activation has been calculated by two different methods, Monte Carlo and CINDER'90, and results have been compared. After irradiation, dose rates at the water surface of the reactor pool were measured, with the irradiated stainless steel sample submerged at different positions under water. Experimental measurements have been compared with simulation results using Monte Carlo. The comparison shows a good agreement confirming the validation of models.
TRIGA IPR-R1 reactor simulation using Monte Carlo transport methods
Hugo Moura Dalle
2005-01-01
Resumo: A utilização do método Monte Carlo na simulação do transporte de partículas em reatores nucleares é crescente e constitui uma tendência mundial. O maior inconveniente dessa técnica, a grande exigência de capacidade de processamento, vem sendo superado pelo contínuo desenvolvimento de processadores cada vez mais rápidos. Esse contexto permitiu o desenvolvimento de metodologias de cálculo neutrônico de reatores nas quais se acopla a parte do transporte de partículas, feita com um código...
Yamamoto, Takahisa; MITACHI, Koshi; Suzuki, Takashi
2005-01-01
The Molten Salt Reactor (MSR) is a thermal neutron reactor with graphite moderation and operates on the thorium-uranium fuel cycle. The feature of the MSR is that fuel salt flows inside the reactor during the nuclear fission reaction. In the previous study, the authors developed numerical model with which to simulate the effects of fuel salt flow on the reactor characteristics. In this study, we apply the model to the steady-state analysis of a small MSR system and estimate the effects of fue...
A Preliminary Analysis of Reactor Performance Test (LOEP) for a Research Reactor
Energy Technology Data Exchange (ETDEWEB)
Kim, Hyeonil; Park, Su-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2015-10-15
The final phase of commissioning is reactor performance test, which is to prove the integrated performance and safety of the research reactor at full power with fuel loaded such as neutron power calibration, Control Absorber Rod/Second Shutdown Rod drop time, InC function test, Criticality, Rod worth, Core heat removal with natural mechanism, and so forth. The last test will be safety-related one to assure the result of the safety analysis of the research reactor is marginal enough to be sure about the nuclear safety by showing the reactor satisfies the acceptance criteria of the safety functions such as for reactivity control, maintenance of auxiliaries, reactor pool water inventory control, core heat removal, and confinement isolation. After all, the fuel integrity will be ensured by verifying there is no meaningful change in the radiation levels. To confirm the performance of safety equipment, loss of normal electric power (LOEP), possibly categorized as Anticipated Operational Occurrence (AOO), is selected as a key experiment to figure out how safe the research reactor is before turning over the research reactor to the owner. This paper presents a preliminary analysis of the reactor performance test (LOEP) for a research reactor. The results showed how different the transient between conservative estimate and best estimate will look. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters.
Human Reliability Analysis for Small Modular Reactors
Energy Technology Data Exchange (ETDEWEB)
Ronald L. Boring; David I. Gertman
2012-06-01
Because no human reliability analysis (HRA) method was specifically developed for small modular reactors (SMRs), the application of any current HRA method to SMRs represents tradeoffs. A first- generation HRA method like THERP provides clearly defined activity types, but these activity types do not map to the human-system interface or concept of operations confronting SMR operators. A second- generation HRA method like ATHEANA is flexible enough to be used for SMR applications, but there is currently insufficient guidance for the analyst, requiring considerably more first-of-a-kind analyses and extensive SMR expertise in order to complete a quality HRA. Although no current HRA method is optimized to SMRs, it is possible to use existing HRA methods to identify errors, incorporate them as human failure events in the probabilistic risk assessment (PRA), and quantify them. In this paper, we provided preliminary guidance to assist the human reliability analyst and reviewer in understanding how to apply current HRA methods to the domain of SMRs. While it is possible to perform a satisfactory HRA using existing HRA methods, ultimately it is desirable to formally incorporate SMR considerations into the methods. This may require the development of new HRA methods. More practicably, existing methods need to be adapted to incorporate SMRs. Such adaptations may take the form of guidance on the complex mapping between conventional light water reactors and small modular reactors. While many behaviors and activities are shared between current plants and SMRs, the methods must adapt if they are to perform a valid and accurate analysis of plant personnel performance in SMRs.
Energy Technology Data Exchange (ETDEWEB)
Campioni, Guillaume; Mounier, Claude [Commissariat a l' Energie Atomique, CEA, 31-33, rue de la Federation, 75752 Paris cedex (France)
2006-07-01
The main goal of the thesis about studies of cold neutrons sources (CNS) in research reactors was to create a complete set of tools to design efficiently CNS. The work raises the problem to run accurate simulations of experimental devices inside reactor reflector valid for parametric studies. On one hand, deterministic codes have reasonable computation times but introduce problems for geometrical description. On the other hand, Monte Carlo codes give the possibility to compute on precise geometry, but need computation times so important that parametric studies are impossible. To decrease this computation time, several developments were made in the Monte Carlo code TRIPOLI-4.4. An uncoupling technique is used to isolate a study zone in the complete reactor geometry. By recording boundary conditions (incoming flux), further simulations can be launched for parametric studies with a computation time reduced by a factor 60 (case of the cold neutron source of the Orphee reactor). The short response time allows to lead parametric studies using Monte Carlo code. Moreover, using biasing methods, the flux can be recorded on the surface of neutrons guides entries (low solid angle) with a further gain of running time. Finally, the implementation of a coupling module between TRIPOLI- 4.4 and the Monte Carlo code McStas for research in condensed matter field gives the possibility to obtain fluxes after transmission through neutrons guides, thus to have the neutron flux received by samples studied by scientists of condensed matter. This set of developments, involving TRIPOLI-4.4 and McStas, represent a complete computation scheme for research reactors: from nuclear core, where neutrons are created, to the exit of neutrons guides, on samples of matter. This complete calculation scheme is tested against ILL4 measurements of flux in cold neutron guides. (authors)
Integrated systems analysis of the PIUS reactor
Energy Technology Data Exchange (ETDEWEB)
Fullwood, F.; Kroeger, P.; Higgins, J. [Brookhaven National Lab., Upton, NY (United States)] [and others
1993-11-01
Results are presented of a systems failure analysis of the PIUS plant systems that are used during normal reactor operation and postulated accidents. This study was performed to provide the NRC with an understanding of the behavior of the plant. The study applied two diverse failure identification methods, Failure Modes Effects & Criticality Analysis (FMECA) and Hazards & Operability (HAZOP) to the plant systems, supported by several deterministic analyses. Conventional PRA methods were also used along with a scheme for classifying events by initiator frequency and combinations of failures. Principal results of this study are: (a) an extensive listing of potential event sequences, grouped in categories that can be used by the NRC, (b) identification of support systems that are important to safety, and (c) identification of key operator actions.
Heat-Flux Analysis of Solar Furnace Using the Monte Carlo Ray-Tracing Method
Energy Technology Data Exchange (ETDEWEB)
Lee, Hyun Jin; Kim, Jong Kyu; Lee, Sang Nam; Kang, Yong Heack [Korea Institute of Energy Research, Daejeon (Korea, Republic of)
2011-10-15
An understanding of the concentrated solar flux is critical for the analysis and design of solar-energy-utilization systems. The current work focuses on the development of an algorithm that uses the Monte Carlo ray-tracing method with excellent flexibility and expandability; this method considers both solar limb darkening and the surface slope error of reflectors, thereby analyzing the solar flux. A comparison of the modeling results with measurements at the solar furnace in Korea Institute of Energy Research (KIER) show good agreement within a measurement uncertainty of 10%. The model evaluates the concentration performance of the KIER solar furnace with a tracking accuracy of 2 mrad and a maximum attainable concentration ratio of 4400 sun. Flux variations according to measurement position and flux distributions depending on acceptance angles provide detailed information for the design of chemical reactors or secondary concentrators.
Directory of Open Access Journals (Sweden)
A. Rais
2015-01-01
Full Text Available In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor.
Energy Technology Data Exchange (ETDEWEB)
Mandelli, Diego; Smith, Curtis; Prescott, Steven; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua; Kinoshita, Robert
2014-08-01
The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impacts of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization project aims to provide insights to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This paper focuses on the impacts of power uprate on the safety margin of a boiling water reactor for a flooding induced station black-out event. Analysis is performed by using a combination of thermal-hydraulic codes and a stochastic analysis tool currently under development at the Idaho National Laboratory, i.e. RAVEN. We employed both classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. Results obtained give a detailed investigation of the issues associated with a plant power uprate including the effects of station black-out accident scenarios. We were able to quantify how the timing of specific events was impacted by a higher nominal reactor core power. Such safety insights can provide useful information to the decision makers to perform risk informed margins management.
The impact of Monte Carlo simulation: a scientometric analysis of scholarly literature
Pia, Maria Grazia; Bell, Zane W; Dressendorfer, Paul V
2010-01-01
A scientometric analysis of Monte Carlo simulation and Monte Carlo codes has been performed over a set of representative scholarly journals related to radiation physics. The results of this study are reported and discussed. They document and quantitatively appraise the role of Monte Carlo methods and codes in scientific research and engineering applications.
Light water reactor lower head failure analysis
Energy Technology Data Exchange (ETDEWEB)
Rempe, J.L.; Chavez, S.A.; Thinnes, G.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)] [and others
1993-10-01
This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.
Energy Technology Data Exchange (ETDEWEB)
Yamamoto, Toshihiro; Sakurai, Kiyoshi; Tonoike, Kotaro; Miyoshi, Yoshinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2001-01-01
Reactor noise analysis methods using Monte Carlo technique have been proposed and developed in the field of nuclear criticality safety. The Monte Carlo simulation for noise analysis can be made by simulating physical phenomena in the course of neutron transport in a nuclear fuel as practically as possible. MCNP-DSP was developed by T. Valentine of ORNL for this purpose and it is a modified version of MCNP-4A. The authors applied this code to frequency analysis measurements performed in light-water critical assembly TCA. Prompt neutron generation times for critical and subcritical cores were measured by doing the frequency analysis of detector signals. The Monte Carlo simulations for these experiments were carried out using MCNP-DSP, and prompt neutron generation times were calculated. (author)
Energy Technology Data Exchange (ETDEWEB)
Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I. [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)
2016-09-15
A burn-up calculation of large systems by Monte-Carlo code (MCU) is complex process and it requires large computational costs. Previously prepared isotopic compositions are proposed to be used for the Monte-Carlo code calculations of different system states with burnt fuel. Isotopic compositions are calculated by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by the engineering codes (TVS-M, BIPR-7A and PERMAK-A). The multiplication factors and power distributions of FAs from a 3-D reactor core are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The separate conditions of the burnt core are observed. The results of MCU calculations were compared with those that were obtained by engineering codes.
Mohammadi, A; Hassanzadeh, M; Gharib, M
2016-02-01
In this study, shielding calculation and criticality safety analysis were carried out for general material testing reactor (MTR) research reactors interim storage and relevant transportation cask. During these processes, three major terms were considered: source term, shielding, and criticality calculations. The Monte Carlo transport code MCNP5 was used for shielding calculation and criticality safety analysis and ORIGEN2.1 code for source term calculation. According to the results obtained, a cylindrical cask with body, top, and bottom thicknesses of 18, 13, and 13 cm, respectively, was accepted as the dual-purpose cask. Furthermore, it is shown that the total dose rates are below the normal transport criteria that meet the standards specified.
Improving PWR core simulations by Monte Carlo uncertainty analysis and Bayesian inference
Castro, Emilio; Buss, Oliver; Garcia-Herranz, Nuria; Hoefer, Axel; Porsch, Dieter
2016-01-01
A Monte Carlo-based Bayesian inference model is applied to the prediction of reactor operation parameters of a PWR nuclear power plant. In this non-perturbative framework, high-dimensional covariance information describing the uncertainty of microscopic nuclear data is combined with measured reactor operation data in order to provide statistically sound, well founded uncertainty estimates of integral parameters, such as the boron letdown curve and the burnup-dependent reactor power distribution. The performance of this methodology is assessed in a blind test approach, where we use measurements of a given reactor cycle to improve the prediction of the subsequent cycle. As it turns out, the resulting improvement of the prediction quality is impressive. In particular, the prediction uncertainty of the boron letdown curve, which is of utmost importance for the planning of the reactor cycle length, can be reduced by one order of magnitude by including the boron concentration measurement information of the previous...
Criticality accident detector coverage analysis using the Monte Carlo Method
Energy Technology Data Exchange (ETDEWEB)
Zino, J.F.; Okafor, K.C.
1993-12-31
As a result of the need for a more accurate computational methodology, the Los Alamos developed Monte Carlo code MCNP is used to show the implementation of a more advanced and accurate methodology in criticality accident detector analysis. This paper will detail the application of MCNP for the analysis of the areas of coverage of a criticality accident alarm detector located inside a concrete storage vault at the Savannah River Site. The paper will discuss; (1) the generation of fixed-source representations of various criticality fission sources (for spherical geometries); (2) the normalization of these sources to the ``minimum criticality of concern`` as defined by ANS 8.3; (3) the optimization process used to determine which source produces the lowest total detector response for a given set of conditions; and (4) the use of this minimum source for the analysis of the areas of coverage of the criticality accident alarm detector.
Energy Technology Data Exchange (ETDEWEB)
Biondo, Elliott D [ORNL; Ibrahim, Ahmad M [ORNL; Mosher, Scott W [ORNL; Grove, Robert E [ORNL
2015-01-01
Detailed radiation transport calculations are necessary for many aspects of the design of fusion energy systems (FES) such as ensuring occupational safety, assessing the activation of system components for waste disposal, and maintaining cryogenic temperatures within superconducting magnets. Hybrid Monte Carlo (MC)/deterministic techniques are necessary for this analysis because FES are large, heavily shielded, and contain streaming paths that can only be resolved with MC. The tremendous complexity of FES necessitates the use of CAD geometry for design and analysis. Previous ITER analysis has required the translation of CAD geometry to MCNP5 form in order to use the AutomateD VAriaNce reducTion Generator (ADVANTG) for hybrid MC/deterministic transport. In this work, ADVANTG was modified to support CAD geometry, allowing hybrid (MC)/deterministic transport to be done automatically and eliminating the need for this translation step. This was done by adding a new ray tracing routine to ADVANTG for CAD geometries using the Direct Accelerated Geometry Monte Carlo (DAGMC) software library. This new capability is demonstrated with a prompt dose rate calculation for an ITER computational benchmark problem using both the Consistent Adjoint Driven Importance Sampling (CADIS) method an the Forward Weighted (FW)-CADIS method. The variance reduction parameters produced by ADVANTG are shown to be the same using CAD geometry and standard MCNP5 geometry. Significant speedups were observed for both neutrons (as high as a factor of 7.1) and photons (as high as a factor of 59.6).
Ion beam analysis of materials in the PBMR reactor
Malherbe, Johan B.; Friedland, E.; van der Berg, N. G.
2008-04-01
South Africa is developing a new type of high temperature nuclear reactor, the so-called pebble bed modular reactor (PBMR). The planned reactor outlet temperature of this gas-cooled reactor is approximately 900 °C. This high temperature places some severe restrictions on materials, which can be used. The name of the reactor is derived from the form of the fuel elements, which are in the form of pebbles, each with a diameter of 60 mm. Each pebble is composed of several thousands of coated fuel particles. The coated particle consists of a nucleus of UO2 surrounded by several layers of different carbons and SiC. The diameter of the fuel particles is 0.92 mm. A brief review will be given of the advantages of this nuclear reactor, of the materials in the fuel elements and their analysis using ion beam techniques.
Perturbation analysis of the TRIGA Mark II reactor Vienna
Energy Technology Data Exchange (ETDEWEB)
Khan, R. [Pakistan Institute of Engineering and Applied Sciences (PIEAS), Islamabad (Pakistan); Villa, M.; Stummer, T.; Boeck, H. [Vienna Univ. of Technology (Austria). Atominstitut; Saeedbadshah [International Islamic Univ., Islamabad (Pakistan)
2013-04-15
The safety design of a nuclear reactor needs to maintain the steady state operation at desired power level. The safe and reliable reactor operation demands the complete knowledge of the core multiplication and its changes during the reactor operation. Therefore it is frequently of interest to compute the changes in core multiplication caused by small disturbances in the field of reactor physics. These disturbances can be created either by geometry or composition changes of the core. Fortunately if these changes (or perturbations) are very small, one does not have to repeat the reactivity calculations. This article focuses the study of small perturbations created in the Central Irradiation Channel (CIC) of the TRIGA mark II core to investigate their reactivity influences on the core reactivity. For this purpose, 3 different kinds of perturbations are created by inserting 3 different samples in the CIC. The cylindrical void (air), heavy water (D2O) and Cadmium (Cd) samples are inserted into the CIC separately to determine their neutronics behavior along the length of the core. The Monte Carlo N-Particle radiation transport code (MCNP) is applied to simulate these perturbations in the CIC. The MCNP theoretical predictions are verified by the experiments performed on the current reactor core. The behavior of void in the whole core and its dependence on position and water fraction is also presented in this article. (orig.)
Development of fault diagnostic technique using reactor noise analysis
Energy Technology Data Exchange (ETDEWEB)
Park, Jin Ho; Kim, J. S.; Oh, I. S.; Ryu, J. S.; Joo, Y. S.; Choi, S.; Yoon, D. B
1999-04-01
The ultimate goal of this project is to establish the analysis technique to diagnose the integrity of reactor internals using reactor noise. The reactor noise analyses techniques for the PWR and CANDU NPP(Nuclear Power Plants) were established by which the dynamic characteristics of reactor internals and SPND instrumentations could be identified, and the noise database corresponding to each plant(both Korean and foreign one) was constructed and compared. Also the change of dynamic characteristics of the Ulchin 1 and 2 reactor internals were simulated under presumed fault conditions. Additionally portable reactor noise analysis system was developed so that real time noise analysis could directly be able to be performed at plant site. The reactor noise analyses techniques developed and the database obtained from the fault simulation, can be used to establish a knowledge based expert system to diagnose the NPP's abnormal conditions. And the portable reactor noise analysis system may be utilized as a substitute for plant IVMS(Internal Vibration Monitoring System). (author)
Stratified source-sampling techniques for Monte Carlo eigenvalue analysis.
Energy Technology Data Exchange (ETDEWEB)
Mohamed, A.
1998-07-10
In 1995, at a conference on criticality safety, a special session was devoted to the Monte Carlo ''Eigenvalue of the World'' problem. Argonne presented a paper, at that session, in which the anomalies originally observed in that problem were reproduced in a much simplified model-problem configuration, and removed by a version of stratified source-sampling. In this paper, stratified source-sampling techniques are generalized and applied to three different Eigenvalue of the World configurations which take into account real-world statistical noise sources not included in the model problem, but which differ in the amount of neutronic coupling among the constituents of each configuration. It is concluded that, in Monte Carlo eigenvalue analysis of loosely-coupled arrays, the use of stratified source-sampling reduces the probability of encountering an anomalous result over that if conventional source-sampling methods are used. However, this gain in reliability is substantially less than that observed in the model-problem results.
Assessing Pretreatment Reactor Scaling Through Empirical Analysis
Energy Technology Data Exchange (ETDEWEB)
Lischeske, James J.; Crawford, Nathan C.; Kuhn, Erik; Nagle, Nicholas J.; Schell, Daniel J.; Tucker, Melvin P.; McMillan, James D.; Wolfrum, Edward J.
2016-12-01
Pretreatment is a critical step in the biochemical conversion of lignocellulosic biomass to fuels and chemicals. Due to the complexity of the physicochemical transformations involved, predictively scaling up technology from bench- to pilot-scale is difficult. This study examines how pretreatment effectiveness under nominally similar reaction conditions is influenced by pretreatment reactor design and scale using four different pretreatment reaction systems ranging from a 3 g batch reactor to a 10 dry-ton/d continuous reactor. The reactor systems examined were an Automated Solvent Extractor (ASE), Steam Explosion Reactor (SER), ZipperClave(R) reactor (ZCR), and Large Continuous Horizontal-Screw Reactor (LHR). To our knowledge, this is the first such study performed on pretreatment reactors across a range of reaction conditions (time and temperature) and at different reactor scales. The comparative pretreatment performance results obtained for each reactor system were used to develop response surface models for total xylose yield after pretreatment and total sugar yield after pretreatment followed by enzymatic hydrolysis. Near- and very-near-optimal regions were defined as the set of conditions that the model identified as producing yields within one and two standard deviations of the optimum yield. Optimal conditions identified in the smallest-scale system (the ASE) were within the near-optimal region of the largest scale reactor system evaluated. A reaction severity factor modeling approach was shown to inadequately describe the optimal conditions in the ASE, incorrectly identifying a large set of sub-optimal conditions (as defined by the RSM) as optimal. The maximum total sugar yields for the ASE and LHR were 95%, while 89% was the optimum observed in the ZipperClave. The optimum condition identified using the automated and less costly to operate ASE system was within the very-near-optimal space for the total xylose yield of both the ZCR and the LHR, and was
Hygro-Thermo-Mechanical Analysis of a Reactor Vessel
Directory of Open Access Journals (Sweden)
Jaroslav Kruis
2012-01-01
Full Text Available Determining the durability of a reactor vessel requires a hygro-thermo-mechanical analysis of the vessel throughout its service life. Damage, prestress losses, distribution of heat and moisture and some other quantities are needed for a durability assessment. A coupled analysis was performed on a two-level model because of the huge demands on computer hardware. This paper deals with a hygro-thermo-mechanical analysis of a reactor vessel made of prestressed concrete with a steel inner liner. The reactor vessel is located in Temelín, Czech Republic.
Energy Technology Data Exchange (ETDEWEB)
Lopez S, R. C.; Francois L, J. L., E-mail: rcarlos.lope@gmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)
2015-09-15
The fast reactors cooled by sodium are one of the options considered in the Generation IV. Since most of the reactors of Fourth Generation are still in development stage, is necessary to have efficient and reliable computational tools, this in order to obtain accurate results in reasonable computational times. In this paper is introduced and describes the deterministic code KANEXT (KArlsruhe Neutronic EXtended Tool) and is compared against a Monte Carlo code of more diffusion: Serpent. KANEXT, being a modular code requires the interaction of different modules to perform a job, this interaction of modules is described in this article. The parameters to be compared are the results of the neutron multiplication effective factor and the evolution of isotopes during the burning. The mentioned comparison is carried out for a fast reactor cooled by sodium of relatively small size compared to commercial size reactors. In this paper the particularities of the reactor are described, important for the analysis such as geometry, enrichments, reflector, etc. The considerations in the implementation in both codes are also described, as are simplifications, length of the burning steps, possible solutions of the Bateman equations for the burning fuel in Serpent and the solution options for transport (P3) and diffusion (P1) in KANEXT. The results show good correspondence between Serpent and KANEXT, which give confidence to continue using KANEXT as the main tool. Respect to computation time, time saving is evident with the use of deterministic codes instead of Monte Carlo codes, in this particular case, the time savings using KANEXT is about 98.5% of the time used by Serpent. (Author)
Energy Technology Data Exchange (ETDEWEB)
Carbajo, Juan (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin, Madison, WI); Schmidt, Rodney Cannon; Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Ludewig, Hans (Brookhaven National Laboratory, Upton, NY); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache %3CU%2B2013%3E CEA, France)
2011-06-01
This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the
Sensitivity analysis for oblique incidence reflectometry using Monte Carlo simulations
DEFF Research Database (Denmark)
Kamran, Faisal; Andersen, Peter E.
2015-01-01
Oblique incidence reflectometry has developed into an effective, noncontact, and noninvasive measurement technology for the quantification of both the reduced scattering and absorption coefficients of a sample. The optical properties are deduced by analyzing only the shape of the reflectance...... profiles. This article presents a sensitivity analysis of the technique in turbid media. Monte Carlo simulations are used to investigate the technique and its potential to distinguish the small changes between different levels of scattering. We present various regions of the dynamic range of optical...... properties in which system demands vary to be able to detect subtle changes in the structure of the medium, translated as measured optical properties. Effects of variation in anisotropy are discussed and results presented. Finally, experimental data of milk products with different fat content are considered...
Nagaya, Yasunobu
2014-06-01
The methods to calculate the kinetics parameters of βeff and Λ with the differential operator sampling have been reviewed. The comparison of the results obtained with the differential operator sampling and iterated fission probability approaches has been performed. It is shown that the differential operator sampling approach gives the same results as the iterated fission probability approach within the statistical uncertainty. In addition, the prediction accuracy of the evaluated nuclear data library JENDL-4.0 for the measured βeff/Λ and βeff values is also examined. It is shown that JENDL-4.0 gives a good prediction except for the uranium-233 systems. The present results imply the need for revisiting the uranium-233 nuclear data evaluation and performing the detailed sensitivity analysis.
Implementation and analysis of an adaptive multilevel Monte Carlo algorithm
Hoel, Hakon
2014-01-01
We present an adaptive multilevel Monte Carlo (MLMC) method for weak approximations of solutions to Itô stochastic dierential equations (SDE). The work [11] proposed and analyzed an MLMC method based on a hierarchy of uniform time discretizations and control variates to reduce the computational effort required by a single level Euler-Maruyama Monte Carlo method from O(TOL-3) to O(TOL-2 log(TOL-1)2) for a mean square error of O(TOL2). Later, the work [17] presented an MLMC method using a hierarchy of adaptively re ned, non-uniform time discretizations, and, as such, it may be considered a generalization of the uniform time discretizationMLMC method. This work improves the adaptiveMLMC algorithms presented in [17] and it also provides mathematical analysis of the improved algorithms. In particular, we show that under some assumptions our adaptive MLMC algorithms are asymptotically accurate and essentially have the correct complexity but with improved control of the complexity constant factor in the asymptotic analysis. Numerical tests include one case with singular drift and one with stopped diusion, where the complexity of a uniform single level method is O(TOL-4). For both these cases the results con rm the theory, exhibiting savings in the computational cost for achieving the accuracy O(TOL) from O(TOL-3) for the adaptive single level algorithm to essentially O(TOL-2 log(TOL-1)2) for the adaptive MLMC algorithm. © 2014 by Walter de Gruyter Berlin/Boston 2014.
Multiscale Analysis of Pebble Bed Reactors
Energy Technology Data Exchange (ETDEWEB)
Hans Gougar; Woo Yoon; Abderrafi Ougouag
2010-10-01
– The PEBBED code was developed at the Idaho National Laboratory for design and analysis of pebble-bed high temperature reactors. The diffusion-depletion-pebble-mixing algorithm of the original PEBBED code was enhanced through coupling with the THERMIX-KONVEK code for thermal fluid analysis and by the COMBINE code for online cross section generation. The COMBINE code solves the B-1 or B-3 approximations to the transport equation for neutron slowing down and resonance interactions in a homogeneous medium with simple corrections for shadowing and thermal self-shielding. The number densities of materials within specified regions of the core are averaged and transferred to COMBINE from PEBBED for updating during the burnup iteration. The simple treatment of self-shielding in previous versions of COMBINE led to inaccurate results for cross sections and unsatisfactory core performance calculations. A new version of COMBINE has been developed that treats all levels of heterogeneity using the 1D transport code ANISN. In a 3-stage calculation, slowing down is performed in 167 groups for each homogeneous subregion (kernel, particle layers, graphite shell, control rod absorber annulus, etc.) Particles in a local average pebble are homogenized using ANISN then passed to the next (pebble) stage. A 1D transport solution is again performed over the pebble geometry and the homogenized pebble cross sections are passed to a 1-d radial model of a wedge of the pebble bed core. This wedge may also include homogeneous reflector regions and a control rod region composed of annuli of different absorbing regions. Radial leakage effects are therefore captured with discrete ordinates transport while axial and azimuthal effects are captured with a transverse buckling term. In this paper, results of various PBR models will be compared with comparable models from literature. Performance of the code will be assessed.
Adaptive Nodal Transport Methods for Reactor Transient Analysis
Energy Technology Data Exchange (ETDEWEB)
Thomas Downar; E. Lewis
2005-08-31
Develop methods for adaptively treating the angular, spatial, and time dependence of the neutron flux in reactor transient analysis. These methods were demonstrated in the DOE transport nodal code VARIANT and the US NRC spatial kinetics code, PARCS.
Monte Carlo analysis of radiative transport in oceanographic lidar measurements
Energy Technology Data Exchange (ETDEWEB)
Cupini, E.; Ferro, G. [ENEA, Divisione Fisica Applicata, Centro Ricerche Ezio Clementel, Bologna (Italy); Ferrari, N. [Bologna Univ., Bologna (Italy). Dipt. Ingegneria Energetica, Nucleare e del Controllo Ambientale
2001-07-01
The analysis of oceanographic lidar systems measurements is often carried out with semi-empirical methods, since there is only a rough understanding of the effects of many environmental variables. The development of techniques for interpreting the accuracy of lidar measurements is needed to evaluate the effects of various environmental situations, as well as of different experimental geometric configurations and boundary conditions. A Monte Carlo simulation model represents a tool that is particularly well suited for answering these important questions. The PREMAR-2F Monte Carlo code has been developed taking into account the main molecular and non-molecular components of the marine environment. The laser radiation interaction processes of diffusion, re-emission, refraction and absorption are treated. In particular are considered: the Rayleigh elastic scattering, produced by atoms and molecules with small dimensions with respect to the laser emission wavelength (i.e. water molecules), the Mie elastic scattering, arising from atoms or molecules with dimensions comparable to the laser wavelength (hydrosols), the Raman inelastic scattering, typical of water, the absorption of water, inorganic (sediments) and organic (phytoplankton and CDOM) hydrosols, the fluorescence re-emission of chlorophyll and yellow substances. PREMAR-2F is an extension of a code for the simulation of the radiative transport in atmospheric environments (PREMAR-2). The approach followed in PREMAR-2 was to combine conventional Monte Carlo techniques with analytical estimates of the probability of the receiver to have a contribution from photons coming back after an interaction in the field of view of the lidar fluorosensor collecting apparatus. This offers an effective mean for modelling a lidar system with realistic geometric constraints. The retrieved semianalytic Monte Carlo radiative transfer model has been developed in the frame of the Italian Research Program for Antarctica (PNRA) and it is
Thermal-hydraulic analysis of nuclear reactors
Zohuri, Bahman
2015-01-01
This text covers the fundamentals of thermodynamics required to understand electrical power generation systems and the application of these principles to nuclear reactor power plant systems. It is not a traditional general thermodynamics text, per se, but a practical thermodynamics volume intended to explain the fundamentals and apply them to the challenges facing actual nuclear power plants systems, where thermal hydraulics comes to play. Written in a lucid, straight-forward style while retaining scientific rigor, the content is accessible to upper division undergraduate students and aimed at practicing engineers in nuclear power facilities and engineering scientists and technicians in industry, academic research groups, and national laboratories. The book is also a valuable resource for students and faculty in various engineering programs concerned with nuclear reactors. This book also: Provides extensive coverage of thermal hydraulics with thermodynamics in nuclear reactors, beginning with fundamental ...
Further experience in Bayesian analysis using Monte Carlo Integration
H.K. van Dijk (Herman); T. Kloek (Teun)
1980-01-01
textabstractAn earlier paper [Kloek and Van Dijk (1978)] is extended in three ways. First, Monte Carlo integration is performed in a nine-dimensional parameter space of Klein's model I [Klein (1950)]. Second, Monte Carlo is used as a tool for the elicitation of a uniform prior on a finite region by
Performance Analysis of Korean Liquid metal type TBM based on Monte Carlo code
Energy Technology Data Exchange (ETDEWEB)
Kim, C. H.; Han, B. S.; Park, H. J.; Park, D. K. [Seoul National Univ., Seoul (Korea, Republic of)
2007-01-15
The objective of this project is to analyze a nuclear performance of the Korean HCML(Helium Cooled Molten Lithium) TBM(Test Blanket Module) which will be installed in ITER(International Thermonuclear Experimental Reactor). This project is intended to analyze a neutronic design and nuclear performances of the Korean HCML ITER TBM through the transport calculation of MCCARD. In detail, we will conduct numerical experiments for analyzing the neutronic design of the Korean HCML TBM and the DEMO fusion blanket, and improving the nuclear performances. The results of the numerical experiments performed in this project will be utilized further for a design optimization of the Korean HCML TBM. In this project, Monte Carlo transport calculations for evaluating TBR (Tritium Breeding Ratio) and EMF (Energy Multiplication factor) were conducted to analyze a nuclear performance of the Korean HCML TBM. The activation characteristics and shielding performances for the Korean HCML TBM were analyzed using ORIGEN and MCCARD. We proposed the neutronic methodologies for analyzing the nuclear characteristics of the fusion blanket, which was applied to the blanket analysis of a DEMO fusion reactor. In the results, the TBR of the Korean HCML ITER TBM is 0.1352 and the EMF is 1.362. Taking into account a limitation for the Li amount in ITER TBM, it is expected that tritium self-sufficiency condition can be satisfied through a change of the Li quantity and enrichment. In the results of activation and shielding analysis, the activity drops to 1.5% of the initial value and the decay heat drops to 0.02% of the initial amount after 10 years from plasma shutdown.
NUSAR: N Reactor Updated Safety Analysis Report, Amendment 21
Energy Technology Data Exchange (ETDEWEB)
Smith, G L
1989-12-01
The enclosed pages are Amendment 21 of the N Reactor Updated Safety Analysis Report (NUSAR). NUSAR, formerly UNI-M-90, was revised by 18 amendments that were issued by UNC Nuclear Industries, the contractor previously responsible for N Reactor operations. As of June 1987, Westinghouse Hanford Company (WHC) acquired the operations and engineering contract for N Reactor and other facilities at Hanford. The document number for NUSAR then became WHC-SP-0297. The first revision was issued by WHC as Amendment 19, prepared originally by UNC. Summaries of each of the amendments are included in NUSAR Section 1.1.
Energy Technology Data Exchange (ETDEWEB)
Park, Jae Phil; Bahn, Chi Bum [Pusan National University, Busan (Korea, Republic of)
2015-10-15
Probabilistic Fracture Mechanics (PFM) analysis was generally used to consider the scatter and uncertainty of parameters in complex phenomenon. Weld defects could be present in weld regions of Pressurized Water Reactors (PWRs), which cannot be considered by the typical fracture mechanics analysis. It is necessary to evaluate the effects of the pre-existing cracks in welds for the integrity of the welds. In this paper, PFM analysis for pre-existing cracks on Alloy 182 weld in PWR primary water environment was carried out using a Monte Carlo simulation. PFM analysis for pre-existing cracks on Alloy 182 weld in PWR primary water environment was carried out. It was shown that inspection decreases the gradient of the failure probability. And failure probability caused by the pre-existing cracks was stabilized after 15 years of operation time in this input condition.
Seismic Base Isolation Analysis for PASCAR Liquid Metal Reactor
Energy Technology Data Exchange (ETDEWEB)
Lee, Kuk Hee; Yoo, Bong; Kim, Yun Jae [Korea Univ., Seoul (Korea, Republic of)
2008-10-15
This paper presents a study for developing a seismic isolation system for the PASCAR (Proliferation resistant, Accident-tolerant, Self-supported, Capsular and Assured Reactor) liquid metal reactor design. PASCAR use lead-bismuth eutectic (LBE) as coolant. Because the density (10,000kg/m{sup 3}) of LBE coolant is very heavier than sodium coolant and water, this presents a challenge to designers of the seismic isolation systems that will be used with these heavy liquid metal reactors. Finite element analysis is adapted to determine the characteristics of the isolator device. Results are presented from a study on the use of three-dimensional seismic isolation devices to the full-scale reactor. The seismic analysis responses of the two-dimensional and the three-dimensional isolation systems for the PASCAR are compared with that of the conventional fixed base system.
Parameter analysis calculation on characteristics of portable FAST reactor
Energy Technology Data Exchange (ETDEWEB)
Otsubo, Akira; Kowata, Yasuki [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center
1998-06-01
In this report, we performed a parameter survey analysis by using the analysis program code STEDFAST (Space, TErrestrial and Deep sea FAST reactor-gas turbine system). Concerning the deep sea fast reactor-gas turbine system, calculations with many variable parameters were performed on the base case of a NaK cooled reactor of 40 kWe. We aimed at total equipment weight and surface area necessary to remove heat from the system as important values of the characteristics of the system. Electric generation power and the material of a pressure hull were specially influential for the weight. The electric generation power, reactor outlet/inlet temperatures, a natural convection heat transfer coefficient of sea water were specially influential for the area. Concerning the space reactor-gas turbine system, the calculations with the variable parameters of compressor inlet temperature, reactor outlet/inlet temperatures and turbine inlet pressure were performed on the base case of a Na cooled reactor of 40 kWe. The first and the second variable parameters were influential for the total equipment weight of the important characteristic of the system. Concerning the terrestrial fast reactor-gas turbine system, the calculations with the variable parameters of heat transferred pipe number in a heat exchanger to produce hot water of 100degC for cogeneration, compressor stage number and the kind of primary coolant material were performed on the base case of a Pb cooled reactor of 100 MWt. In the comparison of calculational results for Pb and Na of primary coolant material, the primary coolant weight flow rate was naturally large for the former case compared with for the latter case because density is very different between them. (J.P.N.)
Thermal hydraulics analysis of the Advanced High Temperature Reactor
Energy Technology Data Exchange (ETDEWEB)
Wang, Dean, E-mail: Dean_Wang@uml.edu [University of Massachusetts Lowell, One University Avenue, Lowell, MA 01854 (United States); Yoder, Graydon L.; Pointer, David W.; Holcomb, David E. [Oak Ridge National Laboratory, 1 Bethel Valley RD #6167, Oak Ridge, TN 37831 (United States)
2015-12-01
Highlights: • The TRACE AHTR model was developed and used to define and size the DRACS and the PHX. • A LOFF transient was simulated to evaluate the reactor performance during the transient. • Some recommendations for modifying FHR reactor system component designs are discussed. - Abstract: The Advanced High Temperature Reactor (AHTR) is a liquid salt-cooled nuclear reactor design concept, featuring low-pressure molten fluoride salt coolant, a carbon composite fuel form with embedded coated particle fuel, passively triggered negative reactivity insertion mechanisms, and fully passive decay heat rejection. This paper describes an AHTR system model developed using the Nuclear Regulatory Commission (NRC) thermal hydraulic transient code TRAC/RELAP Advanced Computational Engine (TRACE). The TRACE model includes all of the primary components: the core, downcomer, hot legs, cold legs, pumps, direct reactor auxiliary cooling system (DRACS), the primary heat exchangers (PHXs), etc. The TRACE model was used to help define and size systems such as the DRACS and the PHX. A loss of flow transient was also simulated to evaluate the performance of the reactor during an anticipated transient event. Some initial recommendations for modifying system component designs are also discussed. The TRACE model will be used as the basis for developing more detailed designs and ultimately will be used to perform transient safety analysis for the reactor.
Versatile in situ gas analysis apparatus for nanomaterials reactors.
Meysami, Seyyed Shayan; Snoek, Lavina C; Grobert, Nicole
2014-09-02
We report a newly developed technique for the in situ real-time gas analysis of reactors commonly used for the production of nanomaterials, by showing case-study results obtained using a dedicated apparatus for measuring the gas composition in reactors operating at high temperature (nanomaterials with tailored properties. Our studies demonstrate that the composition of the precursors dynamically changes as they travel inside of the reactor, causing a nonuniform growth of nanomaterials. Moreover, mapping of the nanomaterials reactor using quantitative gas analysis revealed the actual contribution of thermocatalytic cracking and a quantification of individual precursor fragments. This information is particularly important for quality control of the produced nanomaterials and for the recycling of exhaust residues, ultimately leading toward a more cost-effective continuous production of nanomaterials in large quantities. Our case study of multiwall carbon nanotube synthesis was conducted using the probe in conjunction with chemical vapor deposition (CVD) techniques. Given the similarities of this particular CVD setup to other CVD reactors and high-temperature setups generally used for nanomaterials synthesis, the concept and methodology of in situ gas analysis presented here does also apply to other systems, making it a versatile and widely applicable method across a wide range of materials/manufacturing methods, catalysis, as well as reactor design and engineering.
The Monte Carlo Simulation Method for System Reliability and Risk Analysis
Zio, Enrico
2013-01-01
Monte Carlo simulation is one of the best tools for performing realistic analysis of complex systems as it allows most of the limiting assumptions on system behavior to be relaxed. The Monte Carlo Simulation Method for System Reliability and Risk Analysis comprehensively illustrates the Monte Carlo simulation method and its application to reliability and system engineering. Readers are given a sound understanding of the fundamentals of Monte Carlo sampling and simulation and its application for realistic system modeling. Whilst many of the topics rely on a high-level understanding of calculus, probability and statistics, simple academic examples will be provided in support to the explanation of the theoretical foundations to facilitate comprehension of the subject matter. Case studies will be introduced to provide the practical value of the most advanced techniques. This detailed approach makes The Monte Carlo Simulation Method for System Reliability and Risk Analysis a key reference for senior undergra...
Energy Technology Data Exchange (ETDEWEB)
D. E. Shropshire
2009-01-01
The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.
Monte Carlo Simulations for Likelihood Analysis of the PEN experiment
Glaser, Charles; PEN Collaboration
2017-01-01
The PEN collaboration performed a precision measurement of the π+ ->e+νe(γ) branching ratio with the goal of obtaining a relative uncertainty of 5 ×10-4 or better at the Paul Scherrer Institute. A precision measurement of the branching ratio Γ(π -> e ν (γ)) / Γ(π -> μ ν (γ)) can be used to give mass bounds on ``new'', or non V -A, particles and interactions. This ratio also proves to be one of the most sensitive tests for lepton universality. The PEN detector consists of beam counters, an active target, a mini-time projection chamber, multi-wire proportional chamber, a plastic scintillating hodoscope, and a CsI electromagnetic calorimeter. The Geant4 Monte Carlo simulation is used to construct ultra-realistic events by digitizing energies and times, creating synthetic target waveforms, and fully accounting for photo-electron statistics. We focus on the detailed detector response to specific decay and background processes in order to sharpen the discrimination between them in the data analysis. Work supported by NSF grants PHY-0970013, 1307328, and others.
Monte Carlo Alpha Iteration Algorithm for a Subcritical System Analysis
Directory of Open Access Journals (Sweden)
Hyung Jin Shim
2015-01-01
Full Text Available The α-k iteration method which searches the fundamental mode alpha-eigenvalue via iterative updates of the fission source distribution has been successfully used for the Monte Carlo (MC alpha-static calculations of supercritical systems. However, the α-k iteration method for the deep subcritical system analysis suffers from a gigantic number of neutron generations or a huge neutron weight, which leads to an abnormal termination of the MC calculations. In order to stably estimate the prompt neutron decay constant (α of prompt subcritical systems regardless of subcriticality, we propose a new MC alpha-static calculation method named as the α iteration algorithm. The new method is derived by directly applying the power method for the α-mode eigenvalue equation and its calculation stability is achieved by controlling the number of time source neutrons which are generated in proportion to α divided by neutron speed in MC neutron transport simulations. The effectiveness of the α iteration algorithm is demonstrated for two-group homogeneous problems with varying the subcriticality by comparisons with analytic solutions. The applicability of the proposed method is evaluated for an experimental benchmark of the thorium-loaded accelerator-driven system.
Analysis of Credible Accidents for Argonaut Reactors
Energy Technology Data Exchange (ETDEWEB)
Hawley, S. C.; Kathern, R. L.; Robkin, M. A.
1981-04-01
Five areas of potential accidents have been evaluated for the Argonaut-UTR reactors. They are: • insertion of excess reactivity • catastrophic rearrangement of the core • explosive chemical reaction • graphite fire • fuel-handling accident. A nuclear excursion resulting from the rapid insertion of the maximum available excess reactivity would produce only 12 MWs which is insufficient to cause fuel melting even with conservative assumptions. Although precise structural rearrangement of the core would create a potential hazard, it is simply not credible to assume that such an arrangement would result from the forces of an earthquake or other catastrophic event. Even damage to the fuel from falling debris or other objects is unlikely given the normal reactor structure. An explosion from a metal-water reaction could not occur because there is no credible source of sufficient energy to initiate the reaction. A graphite fire could conceivably create some damage to the reactor but not enough to melt any fuel or initiate a metal-water reaction. The only credible accident involving offsite doses was determined to be a fuel-handling accident which, given highly conservative assumptions, would produce a whole-body dose equivalent of 2 rem from noble gas immersion and a lifetime dose equivalent commitment to the thyroid of 43 rem from radioiodines.
Energy Technology Data Exchange (ETDEWEB)
Park, Goon Cherl; Park, Ik Gyu; Kim, Jae Hak; Lee, Sang Min; Kim, Tae Wan [Seoul National University, Seoul (Korea)
1999-04-01
The objective of this study is to understand the natural circulation characteristics of integral type reactors and to develope the natural circulation analysis code for integral type reactors. This study is focused on the asymmetric 3-dimensional flow during natural circulation such as 1/4 steam generator section isolation and the inclination of the reactor systems. Natural circulation experiments were done using small-scale facilities of integral reactor SMART (System-Integrated Modular Advanced ReacTor). CFX4 code was used to investigate the flow patterns and thermal mixing phenomena in upper pressure header and downcomer. Differences between normal operation of all steam generators and the 1/4 section isolation conditions were observed and the results were used as the data 1/4 section isolation conditions were observed and the results were used as the data for RETRAN-03/INT code validation. RETRAN-03 code was modified for the development of natural circulation analysis code for integral type reactors, which was development of natural circulation analysis code for integral type reactors, which was named as RETRAN-03/INT. 3-dimensional analysis models for asymmetric flow in integral type reactors were developed using vector momentum equations in RETRAN-03. Analysis results using RETRAN-03/INT were compared with experimental and CFX4 analysis results and showed good agreements. The natural circulation characteristics obtained in this study will provide the important and fundamental design features for the future small and medium integral reactors. (author). 29 refs., 75 figs., 18 tabs.
Energy Technology Data Exchange (ETDEWEB)
Park, Goon Cherl; Park, Ik Gyu; Kim, Jae Hak; Lee, Sang Min; Kim, Tae Wan [Seoul National University, Seoul (Korea)
1999-04-01
The objective of this study is to understand the natural circulation characteristics of integral type reactors and to develope the natural circulation analysis code for integral type reactors. This study is focused on the asymmetric 3-dimensional flow during natural circulation such as 1/4 steam generator section isolation and the inclination of the reactor systems. Natural circulation experiments were done using small-scale facilities of integral reactor SMART (System-Integrated Modular Advanced ReacTor). CFX4 code was used to investigate the flow patterns and thermal mixing phenomena in upper pressure header and downcomer. Differences between normal operation of all steam generators and the 1/4 section isolation conditions were observed and the results were used as the data 1/4 section isolation conditions were observed and the results were used as the data for RETRAN-03/INT code validation. RETRAN-03 code was modified for the development of natural circulation analysis code for integral type reactors, which was development of natural circulation analysis code for integral type reactors, which was named as RETRAN-03/INT. 3-dimensional analysis models for asymmetric flow in integral type reactors were developed using vector momentum equations in RETRAN-03. Analysis results using RETRAN-03/INT were compared with experimental and CFX4 analysis results and showed good agreements. The natural circulation characteristics obtained in this study will provide the important and fundamental design features for the future small and medium integral reactors. (author). 29 refs., 75 figs., 18 tabs.
A spectral analysis of the domain decomposed Monte Carlo method for linear systems
Energy Technology Data Exchange (ETDEWEB)
Slattery, Stuart R., E-mail: slatterysr@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831 (United States); Evans, Thomas M., E-mail: evanstm@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831 (United States); Wilson, Paul P.H., E-mail: wilsonp@engr.wisc.edu [University of Wisconsin - Madison, 1500 Engineering Dr., Madison, WI 53706 (United States)
2015-12-15
The domain decomposed behavior of the adjoint Neumann-Ulam Monte Carlo method for solving linear systems is analyzed using the spectral properties of the linear operator. Relationships for the average length of the adjoint random walks, a measure of convergence speed and serial performance, are made with respect to the eigenvalues of the linear operator. In addition, relationships for the effective optical thickness of a domain in the decomposition are presented based on the spectral analysis and diffusion theory. Using the effective optical thickness, the Wigner rational approximation and the mean chord approximation are applied to estimate the leakage fraction of random walks from a domain in the decomposition as a measure of parallel performance and potential communication costs. The one-speed, two-dimensional neutron diffusion equation is used as a model problem in numerical experiments to test the models for symmetric operators with spectral qualities similar to light water reactor problems. In general, the derived approximations show good agreement with random walk lengths and leakage fractions computed by the numerical experiments.
Energy Technology Data Exchange (ETDEWEB)
Ahn, Chang Kee; Shim, Joo Sup [Shinwa Technology Information, Seoul (Korea)
2001-04-01
The objective of this study is to deduce the dynamic correlation between the fuel assembly and the reactor structure. Dynamic characteristics analyses for reactor structure related with vibration of HANARO fuel assembly have been performed For the dynamic characteristic analysis, the in-air models of the round and hexagonal flow tubes, 18-element and 36-element fuel assemblies, and reactor structure were developed. By calculating the hydrodynamic mass and distributing it on the in-air models, the in-water models of the flow tubes, the fuel assemblies, and the reactor structure were developed. Then, modal analyses for developed in-air and in-water models have been performed. Especially, two 18-element fuel assemblies and three 36-element fuel assemblies were included in the in-water reactor models. For the verification of the modal analysis results, the natural frequencies and the mode shapes of the fuel assembly were compared with those obtained from the experiment. Finally the analysis results of the reactor structure were compared with them performed by AECL Based on the reactor model without PCS piping, the in-water reactor model including the fuel assemblies was developed, and its modal analysis was performed. The analysis results demonstrate that there are no resonance between the fuel assembly and the reactor structures. 26 refs., 419 figs., 85 tabs. (Author)
Perkó, Z.
2015-01-01
This thesis presents novel adjoint and spectral methods for the sensitivity and uncertainty (S&U) analysis of multi-physics problems encountered in the field of reactor physics. The first part focuses on the steady state of reactors and extends the adjoint sensitivity analysis methods well establish
Perkó, Z.
2015-01-01
This thesis presents novel adjoint and spectral methods for the sensitivity and uncertainty (S&U) analysis of multi-physics problems encountered in the field of reactor physics. The first part focuses on the steady state of reactors and extends the adjoint sensitivity analysis methods well
Energy Technology Data Exchange (ETDEWEB)
Kim, Young In; Hahn, Do Hee; Won, Byung Chool; Lee, Dong Uk
2008-01-15
Using the DANESS code newly employed for future scenario analysis, reactor deployment scenarios with the introduction of sodium cooled fast reactors(SFRs) having different conversion ratios in the existing PWRs dominant nuclear fleet have been analyzed to find the SFR deployment strategy for replacing PWRs with the view of a spent fuel reduction and an efficient uranium utilization through its reuse in a closed nuclear fuel cycle. Descriptions of the DANESS code and how to use are briefly given from the viewpoint of its first application. The use of SFRs and recycling of TRUs by reusing PWR spent fuel leads to the substantial reduction of the amount of PWR spent fuel and environmental burden by decreasing radiotoxicity of high level waste, and a significant improvement on the natural uranium resources utilization. A continuous deployment of burners effectively decreases the amount of PWR spent fuel accumulation, thus lightening the burden for PWR spent fuel management. An introduction of breakeven reactors effectively reduces the uranium demand through producing excess TRU during the operation, thus contributing to a sustainable nuclear power development. With SFR introduction starting in 2040, PWRs will remain as a main power reactor type till 2100 and SFRs will be in support of waste minimization and fuel utilization.
Hydrogen analysis depth calibration by CORTEO Monte-Carlo simulation
Energy Technology Data Exchange (ETDEWEB)
Moser, M., E-mail: marcus.moser@unibw.de [Universität der Bundeswehr München, Institut für Angewandte Physik und Messtechnik LRT2, Fakultät für Luft- und Raumfahrttechnik, 85577 Neubiberg (Germany); Reichart, P.; Bergmaier, A.; Greubel, C. [Universität der Bundeswehr München, Institut für Angewandte Physik und Messtechnik LRT2, Fakultät für Luft- und Raumfahrttechnik, 85577 Neubiberg (Germany); Schiettekatte, F. [Université de Montréal, Département de Physique, Montréal, QC H3C 3J7 (Canada); Dollinger, G., E-mail: guenther.dollinger@unibw.de [Universität der Bundeswehr München, Institut für Angewandte Physik und Messtechnik LRT2, Fakultät für Luft- und Raumfahrttechnik, 85577 Neubiberg (Germany)
2016-03-15
Hydrogen imaging with sub-μm lateral resolution and sub-ppm sensitivity has become possible with coincident proton–proton (pp) scattering analysis (Reichart et al., 2004). Depth information is evaluated from the energy sum signal with respect to energy loss of both protons on their path through the sample. In first order, there is no angular dependence due to elastic scattering. In second order, a path length effect due to different energy loss on the paths of the protons causes an angular dependence of the energy sum. Therefore, the energy sum signal has to be de-convoluted depending on the matrix composition, i.e. mainly the atomic number Z, in order to get a depth calibrated hydrogen profile. Although the path effect can be calculated analytically in first order, multiple scattering effects lead to significant deviations in the depth profile. Hence, in our new approach, we use the CORTEO Monte-Carlo code (Schiettekatte, 2008) in order to calculate the depth of a coincidence event depending on the scattering angle. The code takes individual detector geometry into account. In this paper we show, that the code correctly reproduces measured pp-scattering energy spectra with roughness effects considered. With more than 100 μm thick Mylar-sandwich targets (Si, Fe, Ge) we demonstrate the deconvolution of the energy spectra on our current multistrip detector at the microprobe SNAKE at the Munich tandem accelerator lab. As a result, hydrogen profiles can be evaluated with an accuracy in depth of about 1% of the sample thickness.
Accuracy Analysis of Assembly Success Rate with Monte Carlo Simulations
Institute of Scientific and Technical Information of China (English)
仲昕; 杨汝清; 周兵
2003-01-01
Monte Carlo simulation was applied to Assembly Success Rate (ASR) analyses.ASR of two peg-in-hole robot assemblies was used as an example by taking component parts' sizes,manufacturing tolerances and robot repeatability into account.A statistic arithmetic expression was proposed and deduced in this paper,which offers an alternative method of estimating the accuracy of ASR,without having to repeat the simulations.This statistic method also helps to choose a suitable sample size,if error reduction is desired.Monte Carlo simulation results demonstrated the feasibility of the method.
Energy Technology Data Exchange (ETDEWEB)
Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1998-12-31
The neutronics analysis of a liquid metal reactor for burning minor actinides has shown that uncertainties in the nuclear data of several key minor actinide isotopes can introduce large uncertainties in the predicted performance of the core. A comprehensive sensitivity and uncertainty analysis was performed on a 1200 MWth actinide burner designed for a low burnup reactivity swing, negative doppler coefficient, and low sodium void worth. Sensitivities were generated using depletion perturbation methods for the equilibrium cycle of the reactor and covariance data was taken ENDF-B/V and other published sources. The relative uncertainties in the burnup swing, doppler coefficient, and void worth were conservatively estimated to be 180%, 97%, and 46%, respectively. 5 refs., 1 fig., 3 tabs. (Author)
Nonlinear analysis and prediction of time series in multiphase reactors
Liu, Mingyan
2014-01-01
This book reports on important nonlinear aspects or deterministic chaos issues in the systems of multi-phase reactors. The reactors treated in the book include gas-liquid bubble columns, gas-liquid-solid fluidized beds and gas-liquid-solid magnetized fluidized beds. The authors take pressure fluctuations in the bubble columns as time series for nonlinear analysis, modeling and forecasting. They present qualitative and quantitative non-linear analysis tools which include attractor phase plane plot, correlation dimension, Kolmogorov entropy and largest Lyapunov exponent calculations and local non-linear short-term prediction.
Energy Technology Data Exchange (ETDEWEB)
Dan Maljovec; Bei Wang; Valerio Pascucci; Peer-Timo Bremer; Michael Pernice; Robert Nourgaliev
2013-05-01
The next generation of methodologies for nuclear reactor Probabilistic Risk Assessment (PRA) explicitly accounts for the time element in modeling the probabilistic system evolution and uses numerical simulation tools to account for possible dependencies between failure events. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. A challenge of dynamic PRA algorithms is the large amount of data they produce which may be difficult to visualize and analyze in order to extract useful information. We present a software tool that is designed to address these goals. We model a large-scale nuclear simulation dataset as a high-dimensional scalar function defined over a discrete sample of the domain. First, we provide structural analysis of such a function at multiple scales and provide insight into the relationship between the input parameters and the output. Second, we enable exploratory analysis for users, where we help the users to differentiate features from noise through multi-scale analysis on an interactive platform, based on domain knowledge and data characterization. Our analysis is performed by exploiting the topological and geometric properties of the domain, building statistical models based on its topological segmentations and providing interactive visual interfaces to facilitate such explorations. We provide a user’s guide to our software tool by highlighting its analysis and visualization capabilities, along with a use case involving dataset from a nuclear reactor safety simulation.
Energy Technology Data Exchange (ETDEWEB)
Dan Maljovec; Bei Wang; Valerio Pascucci; Peer-Timo Bremer; Michael Pernice; Robert Nourgaliev
2013-05-01
The next generation of methodologies for nuclear reactor Probabilistic Risk Assessment (PRA) explicitly accounts for the time element in modeling the probabilistic system evolution and uses numerical simulation tools to account for possible dependencies between failure events. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. A challenge of dynamic PRA algorithms is the large amount of data they produce which may be difficult to visualize and analyze in order to extract useful information. We present a software tool that is designed to address these goals. We model a large-scale nuclear simulation dataset as a high-dimensional scalar function defined over a discrete sample of the domain. First, we provide structural analysis of such a function at multiple scales and provide insight into the relationship between the input parameters and the output. Second, we enable exploratory analysis for users, where we help the users to differentiate features from noise through multi-scale analysis on an interactive platform, based on domain knowledge and data characterization. Our analysis is performed by exploiting the topological and geometric properties of the domain, building statistical models based on its topological segmentations and providing interactive visual interfaces to facilitate such explorations. We provide a user’s guide to our software tool by highlighting its analysis and visualization capabilities, along with a use case involving dataset from a nuclear reactor safety simulation.
Conversion Preliminary Safety Analysis Report for the NIST Research Reactor
Energy Technology Data Exchange (ETDEWEB)
Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Baek, J. S. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hanson, A. L. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, L-Y [Brookhaven National Lab. (BNL), Upton, NY (United States); Brown, N. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cuadra, A. [Brookhaven National Lab. (BNL), Upton, NY (United States)
2015-01-30
The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.
A probabilistic safety analysis of incidents in nuclear research reactors.
Lopes, Valdir Maciel; Agostinho Angelo Sordi, Gian Maria; Moralles, Mauricio; Filho, Tufic Madi
2012-06-01
This work aims to evaluate the potential risks of incidents in nuclear research reactors. For its development, two databases of the International Atomic Energy Agency (IAEA) were used: the Research Reactor Data Base (RRDB) and the Incident Report System for Research Reactor (IRSRR). For this study, the probabilistic safety analysis (PSA) was used. To obtain the result of the probability calculations for PSA, the theory and equations in the paper IAEA TECDOC-636 were used. A specific program to analyse the probabilities was developed within the main program, Scilab 5.1.1. for two distributions, Fischer and chi-square, both with the confidence level of 90 %. Using Sordi equations, the maximum admissible doses to compare with the risk limits established by the International Commission on Radiological Protection (ICRP) were obtained. All results achieved with this probability analysis led to the conclusion that the incidents which occurred had radiation doses within the stochastic effects reference interval established by the ICRP-64.
Algebraic Monte Carlo precedure reduces statistical analysis time and cost factors
Africano, R. C.; Logsdon, T. S.
1967-01-01
Algebraic Monte Carlo procedure statistically analyzes performance parameters in large, complex systems. The individual effects of input variables can be isolated and individual input statistics can be changed without having to repeat the entire analysis.
Accident analysis of heavy water cooled thorium breeder reactor
Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki
2015-04-01
Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition
Development of safety analysis technology for integral reactor
Energy Technology Data Exchange (ETDEWEB)
Sim, Suk K.; Song, J. H.; Chung, Y. J. and others
1999-03-01
Inherent safety features and safety system characteristics of the SMART integral reactor are investigated in this study. Performance and safety of the SMART conceptual design have been evaluated and confirmed through the performance and safety analyses using safety analysis system codes as well as a preliminary performance and safety analysis methodology. SMART design base events and their acceptance criteria are identified to develop a preliminary PIRT for the SMART integral reactor. Using the preliminary PIRT, a set of experimental program for the thermal hydraulic separate effect tests and the integral effect tests was developed for the thermal hydraulic model development and the system code validation. Safety characteristics as well as the safety issues of the integral reactor has been identified during the study, which will be used to resolve the safety issues and guide the regulatory criteria for the integral reactor. The results of the performance and safety analyses performed during the study were used to feedback for the SMART conceptual design. The performance and safety analysis code systems as well as the preliminary safety analysis methodology developed in this study will be validated as the SMART design evolves. The performance and safety analysis technology developed during the study will be utilized for the SMART basic design development. (author)
Utility guidelines for reactor noise analysis: Final report
Energy Technology Data Exchange (ETDEWEB)
Sweeney, F.J.
1987-02-01
Noise analysis techniques have been extensively utilized to monitor the health and performance of nuclear power plant systems. However, few utilities have adequate programs to effectively utilize these techniques. These programs usually provide low-quality data, which can lead to misinterpretation and false alarms. The objective of this work is to provide utilities and noise analysts with guidelines for data acquisition, data analysis, and interpretation of noise analysis results for surveillance and diagnosis of reactor systems.
Improved analysis of bias in Monte Carlo criticality safety
Haley, Thomas C.
2000-08-01
Criticality safety, the prevention of nuclear chain reactions, depends on Monte Carlo computer codes for most commercial applications. One major shortcoming of these codes is the limited accuracy of the atomic and nuclear data files they depend on. In order to apply a code and its data files to a given criticality safety problem, the code must first be benchmarked against similar problems for which the answer is known. The difference between a code prediction and the known solution is termed the "bias" of the code. Traditional calculations of the bias for application to commercial criticality problems are generally full of assumptions and lead to large uncertainties which must be conservatively factored into the bias as statistical tolerances. Recent trends in storing commercial nuclear fuel---narrowed regulatory margins of safety, degradation of neutron absorbers, the desire to use higher enrichment fuel, etc.---push the envelope of criticality safety. They make it desirable to minimize uncertainty in the bias to accommodate these changes, and they make it vital to understand what assumptions are safe to make under what conditions. A set of improved procedures is proposed for (1) developing multivariate regression bias models, and (2) applying multivariate regression bias models. These improved procedures lead to more accurate estimates of the bias and much smaller uncertainties about this estimate, while also generally providing more conservative results. The drawback is that the procedures are not trivial and are highly labor intensive to implement. The payback in savings in margin to criticality and conservatism for calculations near regulatory and safety limits may be worth this cost. To develop these procedures, a bias model using the statistical technique of weighted least squares multivariate regression is developed in detail. Problems that can occur from a weak statistical analysis are highlighted, and a solid statistical method for developing the bias
Development of Burnup Calculation Function in Reactor Monte Carlo Code RMC%堆用蒙卡程序燃耗计算功能开发
Institute of Scientific and Technical Information of China (English)
佘顶; 王侃; 余纲林
2012-01-01
This paper presents the burnup calculation capability of RMC, which is a new Monte Carlo (MC) neutron transport code developed by Reactor Engineering Analysis Laboratory (REAL) in Tsinghua university of China. Unlike most of existing MC depletion codes which explicitly couple the depletion module, RMC incorporates ORIGEN 2.1 in an implicit way. Different burn step strategies, including the middle-of-step approximation and the predictor-corrector method, are adopted by RMC to assure the accuracy under large burnup step size. RMC employs a spectrum-based method of tallying one-group cross section, which can considerably saves computational time with negligible accuracy loss. According to the validation results of benchmarks and examples, it is proved that the burnup function of RMC performs quite well in accuracy and efficiency.%堆用蒙卡程序(RMC)是由清华大学工程物理系REAL实验室自主开发的用于反应堆物理分析的中子输运蒙卡程序,本文主要介绍其燃耗计算功能的开发与验证.RMC的燃耗计算功能具有的特点:内部耦合ORIGEN,相比于外耦合方式,更加灵活和高效；使用基于能谱的单群截面统计方法,可在保证精度的前提下,显著提高计算效率；采取预估修正和中点近似等多种燃耗步策略,减小大燃耗步长时的计算误差.通过计算压水堆栅元、沸水堆组件、快堆等一系列基准题和算例,验证了RMC燃耗计算的正确性和速度优势.
Ramos, A.; Filtvedt, W. O.; Lindholm, D.; Ramachandran, P. A.; Rodríguez, A.; del Cañizo, C.
2015-12-01
Polysilicon production costs contribute approximately to 25-33% of the overall cost of the solar panels and a similar fraction of the total energy invested in their fabrication. Understanding the energy losses and the behaviour of process temperature is an essential requirement as one moves forward to design and build large scale polysilicon manufacturing plants. In this paper we present thermal models for two processes for poly production, viz., the Siemens process using trichlorosilane (TCS) as precursor and the fluid bed process using silane (monosilane, MS). We validate the models with some experimental measurements on prototype laboratory reactors relating the temperature profiles to product quality. A model sensitivity analysis is also performed, and the effects of some key parameters such as reactor wall emissivity and gas distributor temperature, on temperature distribution and product quality are examined. The information presented in this paper is useful for further understanding of the strengths and weaknesses of both deposition technologies, and will help in optimal temperature profiling of these systems aiming at lowering production costs without compromising the solar cell quality.
Evans, Robert M.
1976-10-05
1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.
A Monte Carlo based spent fuel analysis safeguards strategy assessment
Energy Technology Data Exchange (ETDEWEB)
Fensin, Michael L [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Sandoval, Nathan P [Los Alamos National Laboratory
2009-01-01
assessment process, the techniques employed to automate the coupled facets of the assessment process, and the standard burnup/enrichment/cooling time dependent spent fuel assembly library. We also clearly define the diversion scenarios that will be analyzed during the standardized assessments. Though this study is currently limited to generic PWR assemblies, it is expected that the results of the assessment will yield an adequate spent fuel analysis strategy knowledge that will help the down-select process for other reactor types.
Deterministic sensitivity analysis for first-order Monte Carlo simulations: a technical note.
Geisler, Benjamin P; Siebert, Uwe; Gazelle, G Scott; Cohen, David J; Göhler, Alexander
2009-01-01
Monte Carlo microsimulations have gained increasing popularity in decision-analytic modeling because they can incorporate discrete events. Although deterministic sensitivity analyses are essential for interpretation of results, it remains difficult to combine these alongside Monte Carlo simulations in standard modeling packages without enormous time investment. Our purpose was to facilitate one-way deterministic sensitivity analysis of TreeAge Markov state-transition models requiring first-order Monte Carlo simulations. Using TreeAge Pro Suite 2007 and Microsoft Visual Basic for EXCEL, we constructed a generic script that enables one to perform automated deterministic one-way sensitivity analyses in EXCEL employing microsimulation models. In addition, we constructed a generic EXCEL-worksheet that allows for use of the script with little programming knowledge. Linking TreeAge Pro Suite 2007 and Visual Basic enables the performance of deterministic sensitivity analyses of first-order Monte Carlo simulations. There are other potentially interesting applications for automated analysis.
Institute of Scientific and Technical Information of China (English)
彭钢
2012-01-01
采用三维堆芯连续能量蒙特卡罗程序( MCNP)对高通量工程试验堆(HFETR)零功率物理实验进行计算分析.从计算结果可以看出,在零功率反应堆上,径向铍反射层应当考虑金属铍中的杂质和密度修正,同时需要考虑控制棒过渡段的10B含量修正；而HFETR上的铍块则可以认为是纯金属铍,控制棒过渡段中10B含量也需另作考虑.从计算结果来看,多数参数(堆芯中子有效增殖系数keff、中子通量密度相对分布、γ剂量率相对分布以及多数部件反应性价值)均较好满足误差要求,而个别小反应性部件计算误差较大可能与MCNP程序模型有关.%Three-dimensional, continuous energy Monte Carlo code (MCNP) is adopted to carry out the analysis of zero power experiment of HFETR. From the results, the impurity, density of Beryllium block and the B concentration in control rod transition part should be carefully determined in the analysis of zero power experiment. While in the experiment of HFETR, the Beryllium block is considered as pure metal and the 10B concentration in control rod transition part is different from that of zero power experiment. From the calculation results, these parameters (effective neutron multiplication factor Keff relative distribution of neutron flux density, y dose rate distribution and component reactivity) are quite fit with the experiment. The difference of small reactivity between calculation and experiment is quite large, and may be related to the deficiency of MCNP model.
The MC21 Monte Carlo Transport Code
Energy Technology Data Exchange (ETDEWEB)
Sutton TM, Donovan TJ, Trumbull TH, Dobreff PS, Caro E, Griesheimer DP, Tyburski LJ, Carpenter DC, Joo H
2007-01-09
MC21 is a new Monte Carlo neutron and photon transport code currently under joint development at the Knolls Atomic Power Laboratory and the Bettis Atomic Power Laboratory. MC21 is the Monte Carlo transport kernel of the broader Common Monte Carlo Design Tool (CMCDT), which is also currently under development. The vision for CMCDT is to provide an automated, computer-aided modeling and post-processing environment integrated with a Monte Carlo solver that is optimized for reactor analysis. CMCDT represents a strategy to push the Monte Carlo method beyond its traditional role as a benchmarking tool or ''tool of last resort'' and into a dominant design role. This paper describes various aspects of the code, including the neutron physics and nuclear data treatments, the geometry representation, and the tally and depletion capabilities.
Accelerator-driven transmutation reactor analysis code system (ATRAS)
Energy Technology Data Exchange (ETDEWEB)
Sasa, Toshinobu; Tsujimoto, Kazufumi; Takizuka, Takakazu; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1999-03-01
JAERI is proceeding a design study of the hybrid type minor actinide transmutation system which mainly consist of an intense proton accelerator and a fast subcritical core. Neutronics and burnup characteristics of the accelerator-driven system is important from a view point of the maintenance of subcriticality and energy balance during the system operation. To determine those characteristics accurately, it is necessary to involve reactions at high-energy region, which are not treated on ordinary reactor analysis codes. The authors developed a code system named ATRAS to analyze the neutronics and burnup characteristics of accelerator-driven subcritical reactor systems. ATRAS has a function of burnup analysis taking account of the effect of spallation neutron source. ATRAS consists of a spallation analysis code, a neutron transport codes and a burnup analysis code. Utility programs for fuel exchange, pre-processing and post-processing are also incorporated. (author)
Energy Technology Data Exchange (ETDEWEB)
Motalab, Mohammad Abdul; Kim, Woosong; Kim, Yonghee, E-mail: yongheekim@kaist.ac.kr
2015-12-15
Highlights: • The PCR of the CANDU6 reactor is slightly negative at low power, e.g. <80% P. • Doppler broadening of scattering resonances improves noticeably the FTC and make the PCR more negative or less positive in CANDU6. • The elevated inlet coolant condition can worsen significantly the PCR of CANDU6. • Improved design tools are needed for the safety evaluation of CANDU6 reactor. - Abstract: The power coefficient of reactivity (PCR) is a very important parameter for inherent safety and stability of nuclear reactors. The combined effect of a relatively less negative fuel temperature coefficient and a positive coolant temperature coefficient make the CANDU6 (CANada Deuterium Uranium) PCR very close to zero. In the original CANDU6 design, the PCR was calculated to be clearly negative. However, the latest physics design tools predict that the PCR is slightly positive for a wide operational range of reactor power. It is upon this contradictory observation that the CANDU6 PCR is re-evaluated in this work. In our previous study, the CANDU6 PCR was evaluated through a standard lattice analysis at mid-burnup and was found to be negative at low power. In this paper, the study was extended to a detailed 3-D CANDU6 whole-core model using the Monte Carlo code Serpent2. The Doppler broadening rejection correction (DBRC) method was implemented in the Serpent2 code in order to take into account thermal motion of the heavy uranium nucleus in the neutron-U scattering reactions. Time-average equilibrium core was considered for the evaluation of the representative PCR of CANDU6. Two thermal hydraulic models were considered in this work: one at design condition and the other at operating condition. Bundle-wise distributions of the coolant properties are modeled and the bundle-wise fuel temperature is also considered in this study. The evaluated nuclear data library ENDF/B-VII.0 was used throughout this Serpent2 evaluation. In these Monte Carlo calculations, a large number
Directory of Open Access Journals (Sweden)
Malouch Fadhel
2016-01-01
Full Text Available An irradiation program DV50 was carried out from 2002 to 2006 in the OSIRIS material testing reactor (CEA-Saclay center to assess the pressure vessel steel toughness curve for a fast neutron fluence (E > 1 MeV equivalent to a French 900-MWe PWR lifetime of 50 years. This program allowed the irradiation of 120 specimens out of vessel steel, subdivided in two successive irradiations DV50 n∘1 and DV50 n∘2. To measure the fast neutron fluence (E > 1 MeV received by specimens after each irradiation, sample holders were equipped with activation foils that were withdrawn at the end of irradiation for activity counting and processing. The fast effective cross-sections used in the dosimeter processing were determined with a specific calculation scheme based on the Monte-Carlo code TRIPOLI-3 (and the nuclear data ENDF/B-VI and IRDF-90. In order to put vessel-steel experiments at the same standard, a new dosimetric interpretation of the DV50 experiment has been performed by using the Monte-Carlo code TRIPOLI-4 and more recent nuclear data (JEFF3.1.1 and IRDF-2002. This paper presents a comparison of previous and recent calculations performed for the DV50 vessel-steel experiment to assess the impact on the dosimetric interpretation.
Multiscale Methods for Nuclear Reactor Analysis
Collins, Benjamin S.
The ability to accurately predict local pin powers in nuclear reactors is necessary to understand the mechanisms that cause fuel pin failure during steady state and transient operation. In the research presented here, methods are developed to improve the local solution using high order methods with boundary conditions from a low order global solution. Several different core configurations were tested to determine the improvement in the local pin powers compared to the standard techniques, that use diffusion theory and pin power reconstruction (PPR). Two different multiscale methods were developed and analyzed; the post-refinement multiscale method and the embedded multiscale method. The post-refinement multiscale methods use the global solution to determine boundary conditions for the local solution. The local solution is solved using either a fixed boundary source or an albedo boundary condition; this solution is "post-refinement" and thus has no impact on the global solution. The embedded multiscale method allows the local solver to change the global solution to provide an improved global and local solution. The post-refinement multiscale method is assessed using three core designs. When the local solution has more energy groups, the fixed source method has some difficulties near the interface: however the albedo method works well for all cases. In order to remedy the issue with boundary condition errors for the fixed source method, a buffer region is used to act as a filter, which decreases the sensitivity of the solution to the boundary condition. Both the albedo and fixed source methods benefit from the use of a buffer region. Unlike the post-refinement method, the embedded multiscale method alters the global solution. The ability to change the global solution allows for refinement in areas where the errors in the few group nodal diffusion are typically large. The embedded method is shown to improve the global solution when it is applied to a MOX/LEU assembly
Gauge Potts model with generalized action: A Monte Carlo analysis
Energy Technology Data Exchange (ETDEWEB)
Fanchiotti, H.; Canal, C.A.G.; Sciutto, S.J.
1985-08-15
Results of a Monte Carlo calculation on the q-state gauge Potts model in d dimensions with a generalized action involving planar 1 x 1, plaquette, and 2 x 1, fenetre, loop interactions are reported. For d = 3 and q = 2, first- and second-order phase transitions are detected. The phase diagram for q = 3 presents only first-order phase transitions. For d = 2, a comparison with analytical results is made. Here also, the behavior of the numerical simulation in the vicinity of a second-order transition is analyzed.
Non-linear analysis in Light Water Reactor design
Energy Technology Data Exchange (ETDEWEB)
Rashid, Y.R.; Sharabi, M.N.; Nickell, R.E.; Esztergar, E.P.; Jones, J.W.
1980-03-01
The results obtained from a scoping study sponsored by the US Department of Energy (DOE) under the Light Water Reactor (LWR) Safety Technology Program at Sandia National Laboratories are presented. Basically, this project calls for the examination of the hypothesis that the use of nonlinear analysis methods in the design of LWR systems and components of interest include such items as: the reactor vessel, vessel internals, nozzles and penetrations, component support structures, and containment structures. Piping systems are excluded because they are being addressed by a separate study. Essentially, the findings were that nonlinear analysis methods are beneficial to LWR design from a technical point of view. However, the costs needed to implement these methods are the roadblock to readily adopting them. In this sense, a cost-benefit type of analysis must be made on the various topics identified by these studies and priorities must be established. This document is the complete report by ANATECH International Corporation.
Monte Carlo Criticality Methods and Analysis Capabilities in SCALE
Energy Technology Data Exchange (ETDEWEB)
Goluoglu, Sedat [ORNL; Petrie Jr, Lester M [ORNL; Dunn, Michael E [ORNL; Hollenbach, Daniel F [ORNL; Rearden, Bradley T [ORNL
2011-01-01
This paper describes the Monte Carlo codes KENO V.a and KENO-VI in SCALE that are primarily used to calculate multiplication factors and flux distributions of fissile systems. Both codes allow explicit geometric representation of the target systems and are used internationally for safety analyses involving fissile materials. KENO V.a has limiting geometric rules such as no intersections and no rotations. These limitations make KENO V.a execute very efficiently and run very fast. On the other hand, KENO-VI allows very complex geometric modeling. Both KENO codes can utilize either continuous-energy or multigroup cross-section data and have been thoroughly verified and validated with ENDF libraries through ENDF/B-VII.0, which has been first distributed with SCALE 6. Development of the Monte Carlo solution technique and solution methodology as applied in both KENO codes is explained in this paper. Available options and proper application of the options and techniques are also discussed. Finally, performance of the codes is demonstrated using published benchmark problems.
Energy Technology Data Exchange (ETDEWEB)
Papazoglou, I A; Gyftopoulos, E P
1978-09-01
A methodology for the assessment of the uncertainties about the reliability of nuclear reactor systems described by Markov models is developed, and the uncertainties about the probability of loss of coolable core geometry (LCG) of the Clinch River Breeder Reactor (CRBR) due to shutdown system failures, are assessed. Uncertainties are expressed by assuming the failure rates, the repair rates and all other input variables of reliability analysis as random variables, distributed according to known probability density functions (pdf). The pdf of the reliability is then calculated by the moment matching technique. Two methods have been employed for the determination of the moments of the reliability: the Monte Carlo simulation; and the Taylor-series expansion. These methods are adopted to Markovian problems and compared for accuracy and efficiency.
Energy Technology Data Exchange (ETDEWEB)
Woo, Tae Ho [Seoul National Univ. (Korea, Republic of). Dept. of Nuclear Engineering
2012-07-15
The power stabilization of the nuclear power plants (NPPs) is investigated in the aspect of the liquid metal coolant. The quantification of the risk analysis is performed by the system dynamics (SD) method which is processed by the feedback and accumulation complex algorithms. The Vensim software package is used for the simulations, which is supported by the Monte-Carlo method. There are 2 kinds of considerations as the economic and safety properties. The result shows the stability of the operations when the power can be decided. This shows the higher efficiency of the reactor. The failure frequency is 16/60 = 27%. In the event of Power Stabilized, the failure event is in the quite lower frequency rate. The commercial use of the reactor is important in the operations. (orig.)
Qualitative diagnosis for transients analysis on nuclear reactors
Energy Technology Data Exchange (ETDEWEB)
Lorre, J.P.; Dorlet, E.; Evrard, J.M.
1995-12-31
One of the major aims of an intelligent monitoring system, is the supervision task which assist the operator in understanding what occurs on a process. Failures hypotheses must be located and the inferring process must be explained. This paper demonstrate a second generation expert system (SEXTANT) decided to the transients analysis on PWR nuclear reactors. This system detects failures by simulating the process with a numerical model. A diagnosis module uses an even graph built from a causal graph model of the plant to generate hypotheses, and a numerical model to validate these hypotheses. Hypotheses are stored into scenarios which are concurrent possible interpretations of the process evolution. The approach is illustrated by an application for the analysis of the house load operation on a pressurized water reactor. (authors). 9 refs., 10 figs.
Neutronics shielding analysis for the end plug of a tandem mirror fusion reactor
Ragheb, Magdi M. H.; Maynard, Charles W.
1981-10-01
A neutronics analysis using the Monte Carlo method is carried out for the end-plug penetration and magnet system of a tandem mirror fusion reactor. Detailed penetration and the magnets' three-dimensional configurations are modeled. A method of position dependent angular source biasing is developed to adequately sample the DT fusion source in the central cell region and obtain flux contributions at the penetration components. To assure cryogenic stability, the barrier cylindrical solenoid is identified as needing substantial shielding of about 1 m of a steel-lead-boron-carbide-water mixture. Heating rates there would require a thermal-hydraulic design similar to that in the central cell blanket region. The transition coils, however, need a minimal 0.2 m thickness shield. The leakage neutron flux at the direct converters is estimated at 1.3×1015 n/(m2·s), two orders of magnitude lower than that reported at the neutral beam injectors for tokamaks around 1017 n/(m2·s) for a 1 MW/m2 14 MeV neutron wall loading. This result is obtained through a coupling between the nuclear and plasma physics designs in which hydrogen ions rather than deuterium atoms are used for energy injection at the end plug, to avoid creating a neutron source there. This lower and controllable radiation leakage problem is perceived as a potential major advantage of tandem mirrors compared to tokamaks and laser reactor systems.
pyNSMC: A Python Module for Null-Space Monte Carlo Uncertainty Analysis
White, J.; Brakefield, L. K.
2015-12-01
The null-space monte carlo technique is a non-linear uncertainty analyses technique that is well-suited to high-dimensional inverse problems. While the technique is powerful, the existing workflow for completing null-space monte carlo is cumbersome, requiring the use of multiple commandline utilities, several sets of intermediate files and even a text editor. pyNSMC is an open-source python module that automates the workflow of null-space monte carlo uncertainty analyses. The module is fully compatible with the PEST and PEST++ software suites and leverages existing functionality of pyEMU, a python framework for linear-based uncertainty analyses. pyNSMC greatly simplifies the existing workflow for null-space monte carlo by taking advantage of object oriented design facilities in python. The core of pyNSMC is the ensemble class, which draws and stores realized random vectors and also provides functionality for exporting and visualizing results. By relieving users of the tedium associated with file handling and command line utility execution, pyNSMC instead focuses the user on the important steps and assumptions of null-space monte carlo analysis. Furthermore, pyNSMC facilitates learning through flow charts and results visualization, which are available at many points in the algorithm. The ease-of-use of the pyNSMC workflow is compared to the existing workflow for null-space monte carlo for a synthetic groundwater model with hundreds of estimable parameters.
Example Work Domain Analysis for a Reference Sodium Fast Reactor
Energy Technology Data Exchange (ETDEWEB)
Hugo, Jacques [Idaho National Lab. (INL), Idaho Falls, ID (United States); Oxstrand, Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States)
2015-01-01
The nuclear industry is currently designing and building a new generation of reactors that will include different structural, functional, and environmental aspects, all of which are likely to have a significant impact on the way these plants are operated. In order to meet economic and safety objectives, these new reactors will all use advanced technologies to some extent, including new materials and advanced digital instrumentation and control systems. New technologies will affect not only operational strategies, but will also require a new approach to how functions are allocated to humans or machines to ensure optimal performance. Uncertainty about the effect of large scale changes in plant design will remain until sound technical bases are developed for new operational concepts and strategies. Up-to-date models and guidance are required for the development of operational concepts for complex socio-technical systems. This report describes how the classical Work Domain Analysis method was adapted to develop operational concept frameworks for new plants. This adaptation of the method is better able to deal with the uncertainty and incomplete information typical of first-of-a-kind designs. Practical examples are provided of the systematic application of the method in the operational analysis of sodium-cooled reactors. Insights from this application and its utility are reviewed and arguments for the formal adoption of Work Domain Analysis as a value-added part of the Systems Engineering process are presented.
Maucec, M.; Rigollet, C.
2004-01-01
The performance of a detection system based on the pulsed fast/thermal neutron analysis technique was assessed using Monte Carlo simulations. The aim was to develop and implement simulation methods, to support and advance the data analysis techniques of the characteristic gamma-ray spectra, potentia
Recent Developments in Real and Harmonic Analysis In Honor of Carlos Segovia
Cabrelli, Carlos A
2008-01-01
Featuring a collection of invited chapters dedicated to Carlos Segovia, this volume examines the developments in real and harmonic analysis. It includes topics such as: Vector-valued singular integral equations; Harmonic analysis related to Hermite expansions; Gas flow in porous media; and, Global well-posedness of the KPI Equation
Dumenci, Levent; Windle, Michael
2001-01-01
Used Monte Carlo methods to evaluate the adequacy of cluster analysis to recover group membership based on simulated latent growth curve (LCG) models. Cluster analysis failed to recover growth subtypes adequately when the difference between growth curves was shape only. Discusses circumstances under which it was more successful. (SLD)
A Markov Chain Monte Carlo Approach to Confirmatory Item Factor Analysis
Edwards, Michael C.
2010-01-01
Item factor analysis has a rich tradition in both the structural equation modeling and item response theory frameworks. The goal of this paper is to demonstrate a novel combination of various Markov chain Monte Carlo (MCMC) estimation routines to estimate parameters of a wide variety of confirmatory item factor analysis models. Further, I show…
Contributions to the neutronic analysis of a gas-cooled fast reactor
Energy Technology Data Exchange (ETDEWEB)
Martin-del-Campo, Cecilia, E-mail: cecilia.martin.del.campo@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Reyes-Ramirez, Ricardo, E-mail: ricarera@yahoo.com.mx [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Francois, Juan-Luis, E-mail: juan.luis.francois@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Reinking-Cejudo, Arturo G., E-mail: reinking@servidor.unam.mx [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico)
2011-06-15
Highlights: > Differences on reactivity with MCNPX and TRIPOLI-4 are negligible. > Fuel lattice and core criticality calculations were done. > A higher Doppler coefficient than coolant density coefficient. > Zirconium carbide is a better reflector than silicon carbide. > Adequate active height, radial size and reflector thickness were obtained. - Abstract: In this work the Monte Carlo codes MCNPX and TRIPOLI-4 were used to perform the criticality calculations of the fuel assembly and the core configuration of a gas-cooled fast reactor (GFR) concept, currently in development. The objective is to make contributions to the neutronic analysis of a gas-cooled fast reactor. In this study the fuel assembly is based on a hexagonal lattice of fuel-pins. The materials used are uranium and plutonium carbide as fuel, silicon carbide as cladding, and helium gas as coolant. Criticality calculations were done for a fuel assembly where the axial reflector thickness was varied in order to find the optimal thickness. In order to determine the best material to be used as a reflector, in the reactor core with neutrons of high energy spectrum, criticality calculations were done for three reflector materials: zirconium carbide, silicon carbide and natural uranium. It was found that the zirconium carbide provides the best neutron reflection. Criticality calculations using different active heights were done to determine the optimal height, and the reflector thickness was adjusted. Core criticality calculations were performed with different radius sizes to determine the active radial dimension of the core. A negative temperature coefficient of reactivity was verified for the fuel. The effect on reactivity produced by changes in the coolant density was also evaluated. We present the main neutronic characteristics of a preliminary fuel and core designs for the GFR concept. ENDF-VI cross-sections libraries were used in both the MCNPX and TRIPOLI-4 codes, and we verified that the obtained
Advanced High Temperature Reactor Systems and Economic Analysis
Energy Technology Data Exchange (ETDEWEB)
Holcomb, David Eugene [ORNL; Peretz, Fred J [ORNL; Qualls, A L [ORNL
2011-09-01
The Advanced High Temperature Reactor (AHTR) is a design concept for a large-output [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR's large thermal output enables direct comparison of its performance and requirements with other high output reactor concepts. As high-temperature plants, FHRs can support either high-efficiency electricity generation or industrial process heat production. The AHTR analysis presented in this report is limited to the electricity generation mission. FHRs, in principle, have the potential to be low-cost electricity producers while maintaining full passive safety. However, no FHR has been built, and no FHR design has reached the stage of maturity where realistic economic analysis can be performed. The system design effort described in this report represents early steps along the design path toward being able to predict the cost and performance characteristics of the AHTR as well as toward being able to identify the technology developments necessary to build an FHR power plant. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High temperature gas-cooled reactors provide experience with coated particle fuel and graphite components. Light water reactors (LWRs) show the potentials of transparent, high-heat capacity coolants with low chemical reactivity. Modern coal-fired power plants provide design experience
Design and analysis of multicavity prestressed concrete reactor vessels. [HTGR
Energy Technology Data Exchange (ETDEWEB)
Goodpasture, D.W.; Burdette, E.G.; Callahan, J.P.
1977-01-01
During the past 25 years, a rather rapid evolution has taken place in the design and use of prestressed concrete reactor vessels (PCRVs). Initially the concrete vessel served as a one-to-one replacement for its steel counterpart. This was followed by the development of the integral design which led eventually to the more recent multicavity vessel concept. Although this evolution has seen problems in construction and operation, a state-of-the-art review which was recently conducted by the Oak Ridge National Laboratory indicated that the PCRV has proven to be a satisfactory and inherently safe type of vessel for containment of gas-cooled reactors from a purely functional standpoint. However, functionalism is not the only consideration in a demanding and highly competitive industry. A summary is presented of the important considerations in the design and analysis of multicavity PCRVs together with overall conclusions concerning the state of the art of these vessels.
Energy Technology Data Exchange (ETDEWEB)
Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Mueller, Donald E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division
2016-08-31
Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 ^{7}LiF-BeF_{2}) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.
Accuracy Analysis for 6-DOF PKM with Sobol Sequence Based Quasi Monte Carlo Method
Institute of Scientific and Technical Information of China (English)
Jianguang Li; Jian Ding; Lijie Guo; Yingxue Yao; Zhaohong Yi; Huaijing Jing; Honggen Fang
2015-01-01
To improve the precisions of pose error analysis for 6⁃dof parallel kinematic mechanism ( PKM) during assembly quality control, a Sobol sequence based on Quasi Monte Carlo ( QMC) method is introduced and implemented in pose accuracy analysis for the PKM in this paper. The Sobol sequence based on Quasi Monte Carlo with the regularity and uniformity of samples in high dimensions, can prevail traditional Monte Carlo method with up to 98�59% and 98�25% enhancement for computational precision of pose error statistics. Then a PKM tolerance design system integrating this method is developed and with it pose error distributions of the PKM within a prescribed workspace are finally obtained and analyzed.
Application of CFD Codes in Nuclear Reactor Safety Analysis
Directory of Open Access Journals (Sweden)
T. Höhne
2010-01-01
Full Text Available Computational Fluid Dynamics (CFD is increasingly being used in nuclear reactor safety (NRS analyses as a tool that enables safety relevant phenomena occurring in the reactor coolant system to be described in more detail. Numerical investigations on single phase coolant mixing in Pressurised Water Reactors (PWR have been performed at the FZD for almost a decade. The work is aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity. For the experimental investigation of horizontal two phase flows, different non pressurized channels and the TOPFLOW Hot Leg model in a pressure chamber was build and simulated with ANSYS CFX. In a common project between the University of Applied Sciences Zittau/Görlitz and FZD the behaviour of insulation material released by a LOCA released into the containment and might compromise the long term emergency cooling systems is investigated. Moreover, the actual capability of CFD is shown to contribute to fuel rod bundle design with a good CHF performance.
Directory of Open Access Journals (Sweden)
Héctor Armando Durán Peralta
2010-04-01
Full Text Available The stability of reactors having encompassing concentration and temperature parameters, such as continuous flow stirred tank reactors (CSTR, has been widely explored in the literature; however, there are few papers about the stability of tubular reactor having distributed spatial concentration and temperature parameters such as the plow flow tubular reactor (PFTR. This paper analyses the stability of isothermal and non-isothermal PFTR reactors using the Lyapunov functional method. The first order kinetic reaction was selected because one of this paper’s oblectives was to apply Lyapunov functionals to stability analysis of distributed parameter reactors (technique used in electrical engineering systems’ stability analysis. The stability analysis revealed asymptotically stable tempe- rature and concentration profiles for isothermal PFTR, non-isothermal PFTR with kinetic constant independent of temperature and adiabatic non-isothermal PFTR. Analysis revealed an asymptotically stability region for the heat exchange reactor and an uncertain region where it may have oscillations.
Preliminary Core Analysis of a Micro Modular Reactor
Energy Technology Data Exchange (ETDEWEB)
Jo, Chang Keun; Chang, Jongwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Venneri, Francesco [Ultra Safe Nuclear Corporation, Los Alamos (United States); Hawari, Ayman [NC State Univ., Raleigh (United States)
2014-05-15
The Micro Modular Reactor (MMR) will be 'melt-down proof'(MDP) under all circumstances, including the complete loss of coolant, and will be easily transportable and retrievable, and suitable for use with very little site preparation and Balance of Plant (BOP) requirements for a variety of applications, from power generation and process heat applications in remote areas to grid-unattached locations, including ship propulsion. The Micro Modular Reactor design proposed in this paper has 3 meter diameter core (2 meter active core) which is suitable for 'factory manufactured' and has few tens year of service life for remote deployment. We confirmed the feasibility of long term service life by a preliminary neutronic analysis in terms of the excess reactivity, the temperature feedback coefficient, and the control margins. We are able to achieve a reasonably long core life time of 5 ∼ 10 years under typical thermal hydraulic condition of a helium cooled reactor. However, on a situation where longer service period and safety is important, we can reduce the power density to the level of typical pebble bed reactor. In this case we can design 10 MWt MMR with core diameter for 10 ∼ 40 years core life time without much loss in the economics. Several burnable poisons are studied and it is found that erbia mixed in the compact matrix seems reasonably good poison. The temperature feedback coefficients were remaining negative during lifetime. Drum type control rods at reflector region and few control rods inside core region are sufficient to control the reactivity during operation and to achieve safe cold shutdown state.
Ex-vessel Steam Explosion Analysis for Pressurized Water Reactor and Boiling Water Reactor
Matjaž Leskovar; Mitja Uršič
2016-01-01
A steam explosion may occur during a severe accident, when the molten core comes into contact with water. The pressurized water reactor and boiling water reactor ex-vessel steam explosion study, which was carried out with the multicomponent three-dimensional Eulerian fuel–coolant interaction code under the conditions of the Organisation for Economic Co-operation and Development (OECD) Steam Explosion Resolution for Nuclear Applications project reactor exercise, is presented and discussed. In ...
Reliability analysis of tunnel surrounding rock stability by Monte-Carlo method
Institute of Scientific and Technical Information of China (English)
XI Jia-mi; YANG Geng-she
2008-01-01
Discussed advantages of improved Monte-Carlo method and feasibility aboutproposed approach applying in reliability analysis for tunnel surrounding rock stability. Onthe basis of deterministic parsing for tunnel surrounding rock, reliability computing methodof surrounding rock stability was derived from improved Monte-Carlo method. The com-puting method considered random of related parameters, and therefore satisfies relativityamong parameters. The proposed method can reasonably determine reliability of sur-rounding rock stability. Calculation results show that this method is a scientific method indiscriminating and checking surrounding rock stability.
Experimental assessment of computer codes used for safety analysis of integral reactors
Energy Technology Data Exchange (ETDEWEB)
Falkov, A.A.; Kuul, V.S.; Samoilov, O.B. [OKB Mechanical Engineering, Nizhny Novgorod (Russian Federation)
1995-09-01
Peculiarities of integral reactor thermohydraulics in accidents are associated with presence of noncondensable gas in built-in pressurizer, absence of pumped ECCS, use of guard vessel for LOCAs localisation and passive RHRS through in-reactor HX`s. These features defined the main trends in experimental investigations and verification efforts for computer codes applied. The paper reviews briefly the performed experimental investigation of thermohydraulics of AST-500, VPBER600-type integral reactors. The characteristic of UROVEN/MB-3 code for LOCAs analysis in integral reactors and results of its verification are given. The assessment of RELAP5/mod3 applicability for accident analysis in integral reactor is presented.
Neutronic analysis of absorbing materials for the control rod system in reactor ALLEGRO
Energy Technology Data Exchange (ETDEWEB)
Cajko, Frantisek; Secansky, Michal; Chrebet, Tomas; Zajac, Radoslav; Darilek, Petr [VUJE, a.s., Trnava (Slovakia)
2016-09-15
Experimental reactor ALLEGRO is a gas cooled fast reactor in the design stage. The current design of its reactivity control system is based on control rods filled with boron carbide as the absorber. Because of disadvantages connected to high boron enrichment a possibility of using other absorbent materials was explored to lower the boron enrichment and increase the worth of the control rods. The results of neutronic Monte-Carlo analyses in a computational supercell are presented in this paper. Three absorbent materials most suitable for a use in reactor ALLEGRO (B{sub 4}C, EuB{sub 6} and ReB{sub 2}) have been analysed also in a full core model. A possible benefit of a neutron trap concept is explored as well but materials with satisfactory neutronic properties proved to be not suitable for expected high temperatures in the reactor.
SAFETY ANALYSIS METHODOLOGY FOR AGED CANDU® 6 NUCLEAR REACTORS
Directory of Open Access Journals (Sweden)
WOLFGANG HARTMANN
2013-10-01
Full Text Available This paper deals with the Safety Analysis for CANDU® 6 nuclear reactors as affected by main Heat Transport System (HTS aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermalhydraulic analytic models. Safety analyses ranging from the assessment of safety limits associated with the prevention of intermittent fuel sheath dryout for a slow Loss of Regulation (LOR analysis and fission gas release after a fuel failure are summarized. Specifically for fission gas release, the thermalhydraulic analysis for a fresh core and an 11 Effective Full Power Years (EFPY aged core was summarized, leading to the most severe stagnation break sizes for the inlet feeder break and the channel failure time. Associated coolant conditions provide the input data for fuel analyses. Based on the thermalhydraulic data, the fission product inventory under normal operating conditions may be calculated for both fresh and aged cores, and the fission gas release may be evaluated during the transient. This analysis plays a major role in determining possible radiation doses to the public after postulated accidents have occurred.
Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2
Energy Technology Data Exchange (ETDEWEB)
1992-04-01
This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)
Computation system for nuclear reactor core analysis. [LMFBR
Energy Technology Data Exchange (ETDEWEB)
Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.
1977-04-01
This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals.
Ainscow, E K; Brand, M D
1998-09-21
The errors associated with experimental application of metabolic control analysis are difficult to assess. In this paper, we give examples where Monte-Carlo simulations of published experimental data are used in error analysis. Data was simulated according to the mean and error obtained from experimental measurements and the simulated data was used to calculate control coefficients. Repeating the simulation 500 times allowed an estimate to be made of the error implicit in the calculated control coefficients. In the first example, state 4 respiration of isolated mitochondria, Monte-Carlo simulations based on the system elasticities were performed. The simulations gave error estimates similar to the values reported within the original paper and those derived from a sensitivity analysis of the elasticities. This demonstrated the validity of the method. In the second example, state 3 respiration of isolated mitochondria, Monte-Carlo simulations were based on measurements of intermediates and fluxes. A key feature of this simulation was that the distribution of the simulated control coefficients did not follow a normal distribution, despite simulation of the original data being based on normal distributions. Consequently, the error calculated using simulation was greater and more realistic than the error calculated directly by averaging the original results. The Monte-Carlo simulations are also demonstrated to be useful in experimental design. The individual data points that should be repeated in order to reduce the error in the control coefficients can be highlighted.
Energy Technology Data Exchange (ETDEWEB)
Shi, P.; Mok, D.H.B. [AMEC NSS, Toronto, Ontario (Canada)
2011-07-01
Even though pressure tubes are major components of a CANDU reactor, only small proportions of pressure tubes are sampled for in-service inspections due to execution cost, outage duration, and site cumulative radiation exposure limits. In general, a realistic core assessment was not carried out based on all known information related to in-service degradation mechanisms. Recently, a hybrid deterministic and probabilistic core assessment (HDPCA) has been introduced to address the uncertainties associated with uninspected pressure tubes and diverse degradation mechanisms. In the present paper, the HDPCA was carried out for a CANDU unit based on cumulative operating experience and history in order to satisfy the requirements of Clause 7 of CSA Standard N285.8 by considering the uncertainties associated with the estimated distribution parameters, the limited inspected data, and pressure tube properties. The HDPCA is composed of two parts: a simulation part and a deterministic evaluation part. The outcome of the core assessment is the expected pressure tube failure frequency due to pressure tube flaws. In the simulations, pressure tube material properties were sampled from distributions derived from material surveillance and testing programs. The flaw dimensions and intensities were sampled from distributions fitted to in-service inspection data. The pressure tubes were then populated with flaws. Each simulated flaw was evaluated for DHC initiation under constant loading conditions. When Delayed Hydride Cracking initiation from a flaw was predicted, the pressure tube was evaluated for rupture in the Leak-Before-Break evaluation. Based on all the predicted pressure tube ruptures from simulations, the failure frequency was calculated on an annual basis. The largest expected mean and the 95% upper bound of the mean failure frequencies for any evaluation subinterval to the end of pressure tube design life of 210,000 EFPH are significantly below the allowable failure frequency
Two Step Procedure Using a 1-D Slab Spectral Geometry in a Pebble Bed Reactor Core Analysis
Energy Technology Data Exchange (ETDEWEB)
Lee, Hyun Chul; Kim, Kang Seog; Noh, Jae Man; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
2005-07-01
A strong spectral interaction between the core and the reflector has been one of the main concerns in the analysis of pebble bed reactor cores. To resolve this problem, VSOP adopted iteration between the spectrum calculation in a spectral zone and the global core calculation. In VSOP, the whole problem domain is divided into many spectral zones in which the fine group spectrum is calculated using bucklings for fast groups and albedos for thermal groups from the global core calculation. The resulting spectrum in each spectral zone is used to generate broad group cross sections of the spectral zone for the global core calculation. In this paper, we demonstrate a two step procedure in a pebble bed reactor core analysis. In the first step, we generate equivalent cross sections from a 1-D slab spectral geometry model with the help of the equivalence theory. The equivalent cross sections generated in this way include the effect of the spectral interaction between the core and the reflector. In the second step, we perform a diffusion calculation using the equivalent cross sections generated in the first step. A simple benchmark problem derived from the PMBR-400 Reactor was introduced to verify this approach. We compared the two step solutions with the Monte Carlo (MC) solutions for the problem.
Dynamic analysis on methanation reactor using a double-input-multi-output linearized model☆
Institute of Scientific and Technical Information of China (English)
Xingxing Li; Jiageng Li; Bolun Yang; Yong Zhang
2015-01-01
A double-input–multi-output linearized system is developed using the state-space method for dynamic analysis of methanation process of coke oven gas. The stability of reactor alone and reactor with feed-effluent heat ex-changer is compared through the dominant poles of the system transfer functions. With single or double distur-bance of temperature and CO concentration at the reactor inlet, typical dynamic behavior in the reactor, including fast concentration response, slow temperature response and inverse response, is revealed for further under-standing of the counteraction and synergy effects caused by simultaneous variation of concentration and temper-ature. Analysis results show that the stability of the reactor loop is more sensitive than that of reactor alone due to the positive heat feedback. Remarkably, with the decrease of heat exchange efficiency, the reactor system may display limit cycle behavior for a pair of complex conjugate poles across the imaginary axis.
Ex-vessel Steam Explosion Analysis for Pressurized Water Reactor and Boiling Water Reactor
Directory of Open Access Journals (Sweden)
Matjaž Leskovar
2016-02-01
Full Text Available A steam explosion may occur during a severe accident, when the molten core comes into contact with water. The pressurized water reactor and boiling water reactor ex-vessel steam explosion study, which was carried out with the multicomponent three-dimensional Eulerian fuel–coolant interaction code under the conditions of the Organisation for Economic Co-operation and Development (OECD Steam Explosion Resolution for Nuclear Applications project reactor exercise, is presented and discussed. In reactor calculations, the largest uncertainties in the prediction of the steam explosion strength are expected to be caused by the large uncertainties related to the jet breakup. To obtain some insight into these uncertainties, premixing simulations were performed with both available jet breakup models, i.e., the global and the local models. The simulations revealed that weaker explosions are predicted by the local model, compared to the global model, due to the predicted smaller melt droplet size, resulting in increased melt solidification and increased void buildup, both reducing the explosion strength. Despite the lower active melt mass predicted for the pressurized water reactor case, pressure loads at the cavity walls are typically higher than that for the boiling water reactor case. This is because of the significantly larger boiling water reactor cavity, where the explosion pressure wave originating from the premixture in the center of the cavity has already been significantly weakened on reaching the distant cavity wall.
Stability analysis on natural circulation boiling water reactors
Energy Technology Data Exchange (ETDEWEB)
Metz, Peter
1999-05-01
The purpose of the study is a stability analysis of the simplified boiling water reactor concept. A fluid dynamics code, DYNOS, was developed and successfully validated against FRIGG and DESIRE data and a stability benchmark on the Ringhals 1 forced circulation BWR. Three simplified desings were considered in the analysis: The SWRIOOO by Siemens and the SBWR and ESBWR from the General Electric Co. For all three design operational characteristics, i.e. power versus flow rate maps, were calculated. The effects which different geometric and operational parameters, such as the riser height, inlet subcooling etc., have on the characteristics have been investigated. Dynamic simulations on the three simplified design revealed the geysering and the natural circulation oscillations modes only. They were, however, only encountered at pressure below 0.6 MPa. Stability maps for all tree simplified BWRs were calculated and plotted. The study concluded that a fast pressurisation of the reactor vessel is necessary to eliminate the possibility of geysering or natural circulation oscillations mode instability. (au) 26 tabs., 88 ills.
An Economic Analysis of Generation IV Small Modular Reactors
Energy Technology Data Exchange (ETDEWEB)
Stewart, J S; Lamont, A D; Rothwell, G S; Smith, C F; Greenspan, E; Brown, N; Barak, A
2002-03-01
This report examines some conditions necessary for Generation IV Small Modular Reactors (SMRs) to be competitive in the world energy market. The key areas that make nuclear reactors an attractive choice for investors are reviewed, and a cost model based on the ideal conditions is developed. Recommendations are then made based on the output of the cost model and on conditions and tactics that have proven successful in other industries. The Encapsulated Nuclear Heat Source (ENHS), a specific SMR design concept, is used to develop the cost model and complete the analysis because information about the ENHS design is readily available from the University of California at Berkeley Nuclear Engineering Department. However, the cost model can be used to analyze any of the current SMR designs being considered. On the basis of our analysis, we determined that the nuclear power industry can benefit from and SMRs can become competitive in the world energy market if a combination of standardization and simplification of orders, configuration, and production are implemented. This would require wholesale changes in the way SMRs are produced, manufactured and regulated, but nothing that other industries have not implemented and proven successful.
Full core analysis of IRIS reactor by using MCNPX.
Amin, E A; Bashter, I I; Hassan, Nabil M; Mustafa, S S
2016-07-01
This paper describes neutronic analysis for fresh fuelled IRIS (International Reactor Innovative and Secure) reactor by MCNPX code. The analysis included criticality calculations, radial power and axial power distribution, nuclear peaking factor and axial offset percent at the beginning of fuel cycle. The effective multiplication factor obtained by MCNPX code is compared with previous calculations by HELIOS/NESTLE, CASMO/SIMULATE, modified CORD-2 nodal calculations and SAS2H/KENO-V code systems. It is found that k-eff value obtained by MCNPX is closer to CORD-2 value. The radial and axial powers are compared with other published results carried out using SAS2H/KENO-V code. Moreover, the WIMS-D5 code is used for studying the effect of enriched boron in form of ZrB2 on the effective multiplication factor (K-eff) of the fuel pin. In this part of calculation, K-eff is calculated at different concentrations of Boron-10 in mg/cm at different stages of burnup of unit cell. The results of this part are compared with published results performed by HELIOS code.
Energy Technology Data Exchange (ETDEWEB)
Morzinski, J.A.; Gilmore, W.; Hahn, H.A.
1998-07-10
A job task and functional analysis was recently completed for the positions that make up the regional Divisions of Reactor Projects. Among the conclusions of that analysis was a recommendation to clarify roles and responsibilities among site, regional, and headquarters personnel. As that analysis did not cover headquarters personnel, a similar analysis was undertaken of three headquarters positions within the Division of Reactor Projects: Licensing Assistants, Project Managers, and Project Directors. The goals of this analysis were to systematically evaluate the tasks performed by these headquarters personnel to determine job training requirements, to account for variations due to division/regional assignment or differences in several experience categories, and to determine how, and by which positions, certain functions are best performed. The results of this analysis include recommendations for training and for job design. Data to support this analysis was collected by a survey instrument and through several sets of focus group meetings with representatives from each position.
Energy Technology Data Exchange (ETDEWEB)
Kobori, Hikaru [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hiwatari, Ryoji [Central Research Institute of Electric Power Industry, Tokyo (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)
2016-11-01
Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO{sub 2} emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.
Mission Command Analysis Using Monte Carlo Tree Search
2013-06-14
Luther King Drive, White Sands Missile Range, NM 88002-5502, 28 September 2011. REF-1 ... lessons learned: • When implementing MCTS into partially observable games, we must be able to produce a fully-specified state based on available information...Ohman. COMBATXXI, defined. COMBAT XXI online documentation, Training and Doctrine Command Analysis Center—White Sands Missile Range (TRAC-WSMR), Martin
Innovative fusion reactor design analysis: Annual performance report, May 15, 1988--January 31, 1989
Energy Technology Data Exchange (ETDEWEB)
Klein, A.C.
1989-01-31
This report discusses the following topics on fusion reactor component design: FLiBe intermediate heat exchanger design analysis; FLiBe properties; design methodology; FLiBe system steam generator freezeup; FLiBe reactor systems studies; tritium breeding ratio control; analysis of original objectives; and budget analysis. 15 refs., 13 figs., 3 tabs. (LSP)
Leak rate analysis of the Westinghouse Reactor Coolant Pump
Energy Technology Data Exchange (ETDEWEB)
Boardman, T.; Jeanmougin, N.; Lofaro, R.; Prevost, J.
1985-07-01
An independent analysis was performed by ETEC to determine what the seal leakage rates would be for the Westinghouse Reactor Coolant Pump (RCP) during a postulated station blackout resulting from loss of ac electric power. The object of the study was to determine leakage rates for the following conditions: Case 1: All three seals function. Case 2: No. 1 seal fails open while Nos. 2 and 3 seals function. Case 3: All three seals fail open. The ETEC analysis confirmed Westinghouse calculations on RCP seal performance for the conditions investigated. The leak rates predicted by ETEC were slightly lower than those predicted by Westinghouse for each of the three cases as summarized below. Case 1: ETEC predicted 19.6 gpm, Westinghouse predicted 21.1 gpm. Case 2: ETEC predicted 64.7 gpm, Westinghouse predicted 75.6 gpm. Case 3: ETEC predicted 422 gpm, Westinghouse predicted 480 gpm. 3 refs., 22 figs., 6 tabs.
Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS
Energy Technology Data Exchange (ETDEWEB)
Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States)
2015-11-30
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 support 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.
Development of a system model for advanced small modular reactors.
Energy Technology Data Exchange (ETDEWEB)
Lewis, Tom Goslee,; Holschuh, Thomas Vernon,
2014-01-01
This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandias concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.
Multiphysics methods development for high temperature gas reactor analysis
Seker, Volkan
Multiphysics computational methods were developed to perform design and safety analysis of the next generation Pebble Bed High Temperature Gas Cooled Reactors. A suite of code modules was developed to solve the coupled thermal-hydraulics and neutronics field equations. The thermal-hydraulics module is based on the three dimensional solution of the mass, momentum and energy equations in cylindrical coordinates within the framework of the porous media method. The neutronics module is a part of the PARCS (Purdue Advanced Reactor Core Simulator) code and provides a fine mesh finite difference solution of the neutron diffusion equation in three dimensional cylindrical coordinates. Coupling of the two modules was performed by mapping the solution variables from one module to the other. Mapping is performed automatically in the code system by the use of a common material mesh in both modules. The standalone validation of the thermal-hydraulics module was performed with several cases of the SANA experiment and the standalone thermal-hydraulics exercise of the PBMR-400 benchmark problem. The standalone neutronics module was validated by performing the relevant exercises of the PBMR-268 and PBMR-400 benchmark problems. Additionally, the validation of the coupled code system was performed by analyzing several steady state and transient cases of the OECD/NEA PBMR-400 benchmark problem.
Directory of Open Access Journals (Sweden)
Wonkyeong Kim
2015-01-01
Full Text Available A high-leakage core has been known to be a challenging problem not only for a two-step homogenization approach but also for a direct heterogeneous approach. In this paper the DIMPLE S06 core, which is a small high-leakage core, has been analyzed by a direct heterogeneous modeling approach and by a two-step homogenization modeling approach, using contemporary code systems developed for reactor core analysis. The focus of this work is a comprehensive comparative analysis of the conventional approaches and codes with a small core design, DIMPLE S06 critical experiment. The calculation procedure for the two approaches is explicitly presented in this paper. Comprehensive comparative analysis is performed by neutronics parameters: multiplication factor and assembly power distribution. Comparison of two-group homogenized cross sections from each lattice physics codes shows that the generated transport cross section has significant difference according to the transport approximation to treat anisotropic scattering effect. The necessity of the ADF to correct the discontinuity at the assembly interfaces is clearly presented by the flux distributions and the result of two-step approach. Finally, the two approaches show consistent results for all codes, while the comparison with the reference generated by MCNP shows significant error except for another Monte Carlo code, SERPENT2.
Further analysis of multilevel Monte Carlo methods for elliptic PDEs with random coefficients
Teckentrup, A. L.; Scheichl, R.; Giles, M. B.; Ullmann, E
2012-01-01
We consider the application of multilevel Monte Carlo methods to elliptic PDEs with random coefficients. We focus on models of the random coefficient that lack uniform ellipticity and boundedness with respect to the random parameter, and that only have limited spatial regularity. We extend the finite element error analysis for this type of equation, carried out recently by Charrier, Scheichl and Teckentrup, to more difficult problems, posed on non--smooth domains and with discontinuities in t...
SAFETY ANALYSIS AND RISK ASSESSMENT FOR BRIDGES HEALTH MONITORING WITH MONTE CARLO METHODS
2016-01-01
With the increasing requirements of building safety in the past few decades, healthy monitoring and risk assessment of structures is of more and more importance. Especially since traffic loads are heavier, risk Assessment for bridges are essential. In this paper we take advantage of Monte Carlo Methods to analysis the safety of bridge and monitoring the destructive risk. One main goal of health monitoring is to reduce the risk of unexpected damage of artificial objects
Architecture dependent availability analysis of RPS for Research Reactor Applications
Energy Technology Data Exchange (ETDEWEB)
Rahman, Khalilur; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Son, Hanseong [Joongbu Univ., Geumsan (Korea, Republic of); Kim, Youngki; Park, Jaekwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2013-05-15
The research reactors are categorized into two broad categories, Low power research reactors and medium to high power research reactors. According to IAEA TECDOC-1234, Research reactors with 0.250- 2.0 MW power rating or 2.5-10 Χ 10{sup 11} η/cm{sup 2}. s flux are termed low power reactor whereas research reactors ranging from 2-10 MW power rating or 0.1-10 Χ 10{sup 13} η/cm{sup 2}. s are considered as Medium to High power research reactors. Some other standards (IAEA NP-T-5.1) define multipurpose research reactor ranging from power few hundred KW to 10 MW as low power research reactor. The aim of this research, in this article, was to identify a configuration of architecture which gives highest availability with maintaining low cost of manufacturing. In this regard, two configurations of a single channel of RPS are formulated in the current article and their fault trees were developed using AIMS PSA software to get the unavailability. This is a starting point of attempt towards the standardization of I and C architecture for low and medium power research reactors.
Thermal stability analysis of the liquid phase methanol synthesis reactor
Energy Technology Data Exchange (ETDEWEB)
Gogate, M.R.; Desirazu, S.; Berty, J.M.; Lee, S. (Akron University, Akron, OH (USA). Dept. of Chemical Engineering)
1992-01-01
The effect of addition of an inert liquid phase on the rate of heat generation in the catalytic synthesis of methanol from syngas has been studied. Gas compositions typical of product gases from Lurgi and Koppers-Totzek gasifiers, represented by H[sub 2]-rich and CO-rich syngas respectively, were used to experimentally verify the 'slope' and 'dynamic' criteria in a three-phase fixed bed recycle reactor. The liquid medium, Witco-40 oil, has been effective in controlling the rate of heat generation and in preventing catalyst overheating, signifying that the liquid phase synthesis is thermally far more stable than the vapour phase synthesis. The experimental thermal stability study provides crucial and valuable information in commercializing the liquid phase methanol synthesis process. The current approach of thermal stability analysis does not require any a priori assumption or predetermined reaction kinetics. 22 refs., 6 figs., 7 tabs.
Energy Technology Data Exchange (ETDEWEB)
Ait Abderrahim, A
2001-04-01
The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.
Taheriyoun, Masoud; Moradinejad, Saber
2015-01-01
The reliability of a wastewater treatment plant is a critical issue when the effluent is reused or discharged to water resources. Main factors affecting the performance of the wastewater treatment plant are the variation of the influent, inherent variability in the treatment processes, deficiencies in design, mechanical equipment, and operational failures. Thus, meeting the established reuse/discharge criteria requires assessment of plant reliability. Among many techniques developed in system reliability analysis, fault tree analysis (FTA) is one of the popular and efficient methods. FTA is a top down, deductive failure analysis in which an undesired state of a system is analyzed. In this study, the problem of reliability was studied on Tehran West Town wastewater treatment plant. This plant is a conventional activated sludge process, and the effluent is reused in landscape irrigation. The fault tree diagram was established with the violation of allowable effluent BOD as the top event in the diagram, and the deficiencies of the system were identified based on the developed model. Some basic events are operator's mistake, physical damage, and design problems. The analytical method is minimal cut sets (based on numerical probability) and Monte Carlo simulation. Basic event probabilities were calculated according to available data and experts' opinions. The results showed that human factors, especially human error had a great effect on top event occurrence. The mechanical, climate, and sewer system factors were in subsequent tier. Literature shows applying FTA has been seldom used in the past wastewater treatment plant (WWTP) risk analysis studies. Thus, the developed FTA model in this study considerably improves the insight into causal failure analysis of a WWTP. It provides an efficient tool for WWTP operators and decision makers to achieve the standard limits in wastewater reuse and discharge to the environment.
Energy Technology Data Exchange (ETDEWEB)
Kim, Hee Chul; Kim, Woong Sik; Lee, J. H. [Korea Institute of Nuclear Safety, Taejeon (Korea)
2002-03-01
The Nuclear Desalination Plant (NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in this study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current light water reactor and advanced reactor designs, and user requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified and discussed. They include the use of proven technology for new safety features, systematic event classification and selection, strengthening containment function, and the safety impacts on desalination-related systems. The study presents the general safety requirements applicable to licensing of an integral reactor and suggests additional regulatory requirements, which need to be developed, based on the direction to resolution of the safety concerns. The efforts to identify and technically resolve the safety concerns in the design stage will provide the early confidence of SMART safety and the technical basis to evaluate the safety to designers and reviewers in the future. Suggestion on the development of additional regulatory requirements will contribute for the regulator to taking actions for licensing of an integral reactor. 66 refs., 5 figs., 24 tabs. (Author)
The application of research reactor Maria for analysis of thorium use in nuclear power plant
Energy Technology Data Exchange (ETDEWEB)
Chwaszczewski, S.; Andrzejewski, K.; Myslek-Laurikainen, B.; Pytel, B.; Szczurek, J. [Dep. Thorium Project, Institute of Atomic Energy POLATOM, 05-400 Otwock-Swierk (Poland); Polkowska-Motrenko, H. [Institute of Nuclear Chemistry and Technology, ul.Dorodna 16 03-195 Warszawa (Poland)
2010-07-01
The MARIA reactor, pool-type light-water cooled and beryllium moderated nuclear research reactor was used to evaluate the {sup 233}U breeding during the experimental irradiation of the thorium samples. The level of impurities concentrations was determined using ICP-MS method. The associated development of computer programs for analysis of application of thorium in EPR reactor consist of PC version of CORD-2/GNOMER system are presented. (authors)
Russell, Charles R
1962-01-01
Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor
Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)
Clement, J. D.; Rust, J. H.
1977-01-01
Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.
ANALYSIS OF UNCERTAINTY QUANTIFICATION METHOD BY COMPARING MONTE-CARLO METHOD AND WILKS’ FORMULA
Directory of Open Access Journals (Sweden)
SEUNG WOOK LEE
2014-08-01
Full Text Available An analysis of the uncertainty quantification related to LBLOCA using the Monte-Carlo calculation has been performed and compared with the tolerance level determined by the Wilks’ formula. The uncertainty range and distribution of each input parameter associated with the LOCA phenomena were determined based on previous PIRT results and documentation during the BEMUSE project. Calulations were conducted on 3,500 cases within a 2-week CPU time on a 14-PC cluster system. The Monte-Carlo exercise shows that the 95% upper limit PCT value can be obtained well, with a 95% confidence level using the Wilks’ formula, although we have to endure a 5% risk of PCT under-prediction. The results also show that the statistical fluctuation of the limit value using Wilks’ first-order is as large as the uncertainty value itself. It is therefore desirable to increase the order of the Wilks’ formula to be higher than the second-order to estimate the reliable safety margin of the design features. It is also shown that, with its ever increasing computational capability, the Monte-Carlo method is accessible for a nuclear power plant safety analysis within a realistic time frame.
Number of iterations needed in Monte Carlo Simulation using reliability analysis for tunnel supports
Directory of Open Access Journals (Sweden)
E. Bukaçi
2016-06-01
Full Text Available There are many methods in geotechnical engineering which could take advantage of Monte Carlo Simulation to establish probability of failure, since closed form solutions are almost impossible to use in most cases. The problem that arises with using Monte Carlo Simulation is the number of iterations needed for a particular simulation.This article will show why it’s important to calculate number of iterations needed for Monte Carlo Simulation used in reliability analysis for tunnel supports using convergence – confinement method. Number if iterations needed will be calculated with two methods. In the first method, the analyst has to accept a distribution function for the performance function. The other method suggested by this article is to calculate number of iterations based on the convergence of the factor the analyst is interested in the calculation. Reliability analysis will be performed for the diversion tunnel in Rrëshen, Albania, by using both methods mentioned and results will be confronted
Monte Carlo Calculation for Landmine Detection using Prompt Gamma Neutron Activation Analysis
Energy Technology Data Exchange (ETDEWEB)
Park, Seungil; Kim, Seong Bong; Yoo, Suk Jae [Plasma Technology Research Center, Gunsan (Korea, Republic of); Shin, Sung Gyun; Cho, Moohyun [POSTECH, Pohang (Korea, Republic of); Han, Seunghoon; Lim, Byeongok [Samsung Thales, Yongin (Korea, Republic of)
2014-05-15
Identification and demining of landmines are a very important issue for the safety of the people and the economic development. To solve the issue, several methods have been proposed in the past. In Korea, National Fusion Research Institute (NFRI) is developing a landmine detector using prompt gamma neutron activation analysis (PGNAA) as a part of the complex sensor-based landmine detection system. In this paper, the Monte Carlo calculation results for this system are presented. Monte Carlo calculation was carried out for the design of the landmine detector using PGNAA. To consider the soil effect, average soil composition is analyzed and applied to the calculation. This results has been used to determine the specification of the landmine detector.
Stolarski, R. S.; Butler, D. M.; Rundel, R. D.
1977-01-01
A concise stratospheric model was used in a Monte-Carlo analysis of the propagation of reaction rate uncertainties through the calculation of an ozone perturbation due to the addition of chlorine. Two thousand Monte-Carlo cases were run with 55 reaction rates being varied. Excellent convergence was obtained in the output distributions because the model is sensitive to the uncertainties in only about 10 reactions. For a 1 ppby chlorine perturbation added to a 1.5 ppby chlorine background, the resultant 1 sigma uncertainty on the ozone perturbation is a factor of 1.69 on the high side and 1.80 on the low side. The corresponding 2 sigma factors are 2.86 and 3.23. Results are also given for the uncertainties, due to reaction rates, in the ambient concentrations of stratospheric species.
Directory of Open Access Journals (Sweden)
Xisheng Yu
2014-01-01
Full Text Available The paper by Liu (2010 introduces a method termed the canonical least-squares Monte Carlo (CLM which combines a martingale-constrained entropy model and a least-squares Monte Carlo algorithm to price American options. In this paper, we first provide the convergence results of CLM and numerically examine the convergence properties. Then, the comparative analysis is empirically conducted using a large sample of the S&P 100 Index (OEX puts and IBM puts. The results on the convergence show that choosing the shifted Legendre polynomials with four regressors is more appropriate considering the pricing accuracy and the computational cost. With this choice, CLM method is empirically demonstrated to be superior to the benchmark methods of binominal tree and finite difference with historical volatilities.
Monte Carlo Analysis as a Trajectory Design Driver for the TESS Mission
Nickel, Craig; Lebois, Ryan; Lutz, Stephen; Dichmann, Donald; Parker, Joel
2016-01-01
The Transiting Exoplanet Survey Satellite (TESS) will be injected into a highly eccentric Earth orbit and fly 3.5 phasing loops followed by a lunar flyby to enter a mission orbit with lunar 2:1 resonance. Through the phasing loops and mission orbit, the trajectory is significantly affected by lunar and solar gravity. We have developed a trajectory design to achieve the mission orbit and meet mission constraints, including eclipse avoidance and a 30-year geostationary orbit avoidance requirement. A parallelized Monte Carlo simulation was performed to validate the trajectory after injecting common perturbations, including launch dispersions, orbit determination errors, and maneuver execution errors. The Monte Carlo analysis helped identify mission risks and is used in the trajectory selection process.
Nickel, Craig; Parker, Joel; Dichmann, Don; Lebois, Ryan; Lutz, Stephen
2016-01-01
The Transiting Exoplanet Survey Satellite (TESS) will be injected into a highly eccentric Earth orbit and fly 3.5 phasing loops followed by a lunar flyby to enter a mission orbit with lunar 2:1 resonance. Through the phasing loops and mission orbit, the trajectory is significantly affected by lunar and solar gravity. We have developed a trajectory design to achieve the mission orbit and meet mission constraints, including eclipse avoidance and a 30-year geostationary orbit avoidance requirement. A parallelized Monte Carlo simulation was performed to validate the trajectory after injecting common perturbations, including launch dispersions, orbit determination errors, and maneuver execution errors. The Monte Carlo analysis helped identify mission risks and is used in the trajectory selection process.
Modified divergence theorem for analysis and optimization of wall reflecting cylindrical UV reactor
Directory of Open Access Journals (Sweden)
Milanović Đurđe R.
2011-01-01
Full Text Available In this paper, Modified Divergence Theorem (MDT, known in earlier literature as Gauss-Ostrogradsky theorem, was formulated and proposed as a general approach to electromagnetic (EM radiation, especially ultraviolet (UV radiation reactor modeling. Formulated mathematical model, based on MDT, for multilamp UV reactor was applied to all sources in a reactor in order to obtain intensity profiles at chosen surfaces inside reactor. Applied modification of MDT means that intensity at a real opaque or transparent surface or through a virtual surface, opened or closed, from different sides of the surface are added and not subtracted as in some other areas of physics. Derived model is applied to an example of the multiple UV sources reactor, where sources are arranged inside a cylindrical reactor at the coaxial virtual cylinder, having the radius smaller than the radius of the reactor. In this work, optimization of a reactor means maximum transfer of EM energy sources into the fluid for given fluid absorbance and fluid flow-dose product. Obtained results, for in advanced known water quality, gives unique solution for an optimized model of a multilamp reactor geometry. As everyone can easily verify, MDT is very good starting point for every reactor modeling and analysis.
Energy Technology Data Exchange (ETDEWEB)
Kim, Tae Ryong; Jeong, Seong Ho; Park, Jin Ho; Park, Jin Suk [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1994-08-01
The unfavorable phenomena, such as flow induced vibration and aging process in reactor internals, cause degradation of structural integrity and may result in loosing some mechanical binding components which might impact other equipments and components or cause flow blockage. Since these malfunctions and potential failures change reactor noise signal, it is necessary to analyze reactor noise signal for early fault diagnosis in the point of few of safety and plant economics. The objectives of this study are to establish fault diagnostic and TS(thermal shield), and to develop a data acquisition and signal processing software system. In the first year of this study, an analysis technique for the reactor internal vibration using the reactor noise was proposed. With the technique proposed and the reactor noise signals (ex-core neutron and acceleration), the dynamic characteristics of Ulchin-1 reactor internals were obtained, and compared with those of Tricastin-1 which is the prototype of Ulchin-1. In the second year, a PC-based expert system for reactor internals fault diagnosis is developed, which included data acquisition, signal processing, feature extraction function, and represented diagnostic knowledge by the IF-THEN rule. To know the effect of the faults, the reactor internals of Ulchin-1 is modeled using FEM and simulated with an artificial defect given in the hold-down spring. Trend in the dynamic characteristics of reactor internals is also observed during one fuel cycle to know the effect of boron concentration. 100 figs, 7 tabs, 18 refs. (Author).
REACTOR ANALYSIS AND VIRTUAL CONTROL ENVIRONMENT (RAVEN) FY12 REPORT
Energy Technology Data Exchange (ETDEWEB)
Cristian Rabiti; Andrea Alfonsi; Joshua Cogliati; Diego Mandelli; Robert Kinoshita
2012-09-01
RAVEN is a complex software tool that will have tasks spanning from being the RELAP-7 user interface, to using RELAP-7 to perform Risk Informed Safety Characterization (RISMC), and to controlling RELAP-7 calculation execution. The goal of this document is to: 1. Highlight the functional requirements of the different tasks of RAVEN 2. Identify shared functions that could be aggregate in modules so to obtain a minimal software redundancy and maximize software utilization. RAVEN is in fact a software framework that will allow exploiting the following functionalities: • Derive and actuate the control logic required to: o Simulate the plant control system o Simulate the operator (procedure guided) actions o Perform Monte Carlo sampling of random distributed events o Perform event three based analysis • Provide a GUI to: o Input a plant description to RELAP-7 (component, control variable, control parameters) o Concurrent monitoring of Control Parameters o Concurrent alteration of control parameters • Provide Post Processing data mining capability based on o Dimensionality reduction o Cardinality reduction In this document it will be shown how an appropriate mathematical formulation of the control logic and probabilistic analysis leads to have most of the software infrastructure leveraged between the two main tasks. Further, this document will go through the development accomplished this year, including simulation results, and priorities for the next years development
Energy Technology Data Exchange (ETDEWEB)
Jaradat, S.Q., E-mail: sqjxv3@mst.edu; Alajo, A.B., E-mail: alajoa@mst.edu
2017-04-01
Highlights: • The verification for FUJI-U3-(0)—a molten salt reactor—was performed. • The MCNP6 was used to study the reactor physics characteristics for FUJI-U3 type. • The results from the MCNP6 were comparable with the ones obtained from literature. - Abstract: The verification for FUJI-U3-(0)—a molten salt reactor—was performed. The reactor used LiF-BeF2-ThF4-UF4 as the mixed liquid fuel salt, and the core was graphite moderated. The MCNP6 code was used to study the reactor physics characteristics for the FUJI-U3-(0) reactor. Results for reactor physics characteristic of the FUJI-U3-(0) exist in literature, which were used as reference. The reference results were obtained using SRAC95 (a reactor analysis code) coupled with ORIGEN2 (a depletion code). Some modifications were made in the reconstruction of the FUJI-U3-(0) reactor in MCNP due to unavailability of more detailed description of the reactor core. The assumptions resulted in two representative models of the reactor. The results from the MCNP6 models were compared with the reference results obtained from literature. The results were comparable with each other, but with some notable differences. The differences are because of the approximations that were done on the SRAC95 model of the FUJI-U3 to simplify the simulation. Based on the results, it is concluded that MCNP6 code predicts well the overall simulation of neutronics analysis to the previous simulation works using SRAC95 code.
Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion
Energy Technology Data Exchange (ETDEWEB)
Baek J.; Diamond D.; Cuadra, A.; Hanson, A.L.; Cheng, L-Y.; Brown, N.R.
2012-09-30
Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a model of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.
Introduction to Chemical Engineering Reactor Analysis: A Web-Based Reactor Design Game
Orbey, Nese; Clay, Molly; Russell, T.W. Fraser
2014-01-01
An approach to explain chemical engineering through a Web-based interactive game design was developed and used with college freshman and junior/senior high school students. The goal of this approach was to demonstrate how to model a lab-scale experiment, and use the results to design and operate a chemical reactor. The game incorporates both…
Introduction to Chemical Engineering Reactor Analysis: A Web-Based Reactor Design Game
Orbey, Nese; Clay, Molly; Russell, T.W. Fraser
2014-01-01
An approach to explain chemical engineering through a Web-based interactive game design was developed and used with college freshman and junior/senior high school students. The goal of this approach was to demonstrate how to model a lab-scale experiment, and use the results to design and operate a chemical reactor. The game incorporates both…
Energy Technology Data Exchange (ETDEWEB)
Kim, W. S.; Kim, W. K.; Yun, Y. G.; Ahn, H. J.; Lee, J. S.; Lee, S. G.; Sin, A. D. [Korea Atomic Energy Research Institute, Taejeon (Korea)
2000-03-01
The Nuclear Desalination Plant (NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in a present study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current light water reactor and advanced reactor designs, and user requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified and discussed. They includes the use of proven technology for new safety features, systematic event classification and selection, strengthening containment function, and the safety impacts on desalination-related systems. These efforts to identify and technically resolve the safety concerns in the design stage will provide the early confidence of SMART safety and the technical basis to evaluate the safety to designers and reviewers in the future. 62 refs., 3 figs., 21 tabs. (Author)
Development of analysis system and analysis on reactor physics for CANDU advanced fuel
Energy Technology Data Exchange (ETDEWEB)
Kim, Bong Gi; Bae, Chang Joon; Kwon, Oh Sun [Korea Power Electric Corporation, Taejon (Korea, Republic of)
1997-07-01
characteristics of reactor physics for CANFLEX-NU fuel core were calculated using final fuel design data. The results of analysis showed that there was no impact on reactor operations and safety. The above results of calculations and analysis were described in the physics design for CANFLEX-NU= fuel core. Various fuel models were evaluated for selecting high burnup fuel using recovered uranium. It is judged to be worse effects for reactor safety Hence, the use of graphite within fuel was proposed and its results showed to be better. The analysis system of reactor physics for design and analysis of high burnup fuel was evaluated. Lattice codes and core code were reviewed. From the results, the probability of WIMS-AECL and HELIOS is known to be high for analysis of high burnup fuel. For the core code, RFSP, it was evaluated that the simplified 2 group equation should be replaced by explicit 2 group equation. (Author) 32 refs., 25 tabs., 79 figs.
System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors
Energy Technology Data Exchange (ETDEWEB)
Moiseyev, A.V. [SSC RF - IPPE, 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation)
2008-07-01
There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k{sub eff}, control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)
Comparative microbial analysis before and after foaming incidents in biogas reactors
DEFF Research Database (Denmark)
Kougias, Panagiotis; De Francisci, Davide; Treu, Laura
2014-01-01
retention time (HRT) of all reactors was kept constant at 15 days. The whole experiment was divided into two periods. During the first period, the reactors were fed only with cattle manure. Once steady state conditions were reached, liquid sample from all reactors was obtained for DNA extraction...... foam was steady, samples were taken again for DNA extraction and metagenomic analysis. Results from the present study revealed significant variations in the microbiology of the manure-based biogas reactors after foam initiation. A number of genera could be linked to foaming as they produce...
Monte-Carlo Analysis of the Flavour Changing Neutral Current B \\to Gamma at Babar
Energy Technology Data Exchange (ETDEWEB)
Smith, D. [Imperial College, London (United Kingdom)
2001-09-01
The main theme of this thesis is a Monte-Carlo analysis of the rare Flavour Changing Neutral Current (FCNC) decay b→sγ. The analysis develops techniques that could be applied to real data, to discriminate between signal and background events in order to make a measurement of the branching ratio of this rare decay using the BaBar detector. Also included in this thesis is a description of the BaBar detector and the work I have undertaken in the development of the electronic data acquisition system for the Electromagnetic calorimeter (EMC), a subsystem of the BaBar detector.
First Monte Carlo analysis of fragmentation functions from single-inclusive $e^+ e^-$ annihilation
Sato, N; Melnitchouk, W; Hirai, M; Kumano, S; Accardi, A
2016-01-01
We perform the first iterative Monte Carlo (IMC) analysis of fragmentation functions constrained by all available data from single-inclusive $e^+ e^-$ annihilation into pions and kaons. The IMC method eliminates potential bias in traditional analyses based on single fits introduced by fixing parameters not well contrained by the data and provides a statistically rigorous determination of uncertainties. Our analysis reveals specific features of fragmentation functions using the new IMC methodology and those obtained from previous analyses, especially for light quarks and for strange quark fragmentation to kaons.
First Monte Carlo analysis of fragmentation functions from single-inclusive e+e- annihilation
Sato, Nobuo; Ethier, J. J.; Melnitchouk, W.; Hirai, M.; Kumano, S.; Accardi, A.; Jefferson Lab Angular Momentum Collaboration
2016-12-01
We perform the first iterative Monte Carlo (IMC) analysis of fragmentation functions constrained by all available data from single-inclusive e+e- annihilation into pions and kaons. The IMC method eliminates potential bias in traditional analyses based on single fits introduced by fixing parameters not well constrained by the data and provides a statistically rigorous determination of uncertainties. Our analysis reveals specific features of fragmentation functions using the new IMC methodology and those obtained from previous analyses, especially for light quarks and for strange quark fragmentation to kaons.
Uncertainty analysis of fission fraction for reactor antineutrino experiments
Ma, X. B.; Lu, F.; Wang, L. Z.; Chen, Y. X.; Zhong, W. L.; An, F. P.
2016-06-01
Reactor simulation is an important source of uncertainties for a reactor neutrino experiment. Therefore, how to evaluate the antineutrino flux uncertainty results from reactor simulation is an important question. In this study, a method of the antineutrino flux uncertainty result from reactor simulation was proposed by considering the correlation coefficient. In order to use this method in the Daya Bay antineutrino experiment, the open source code DRAGON was improved and used for obtaining the fission fraction and correlation coefficient. The average fission fraction between DRAGON and SCIENCE code was compared and the difference was less than 5% for all the four isotopes. The uncertainty of fission fraction was evaluated by comparing simulation atomic density of four main isotopes with Takahama-3 experiment measurement. After that, the uncertainty of the antineutrino flux results from reactor simulation was evaluated as 0.6% per core for Daya Bay antineutrino experiment.
Fuel burnup analysis of the TRIGA Mark II Reactor at the University of Pavia
Chiesa, Davide; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica; Alloni, Daniele; Magrotti, Giovanni; Manera, Sergio; Prata, Michele; Salvini, Andrea; Cammi, Antonio; Zanetti, Matteo; Sartori, Alberto
2015-01-01
A time evolution model was developed to study fuel burnup for the TRIGA Mark II reactor at the University of Pavia. The results were used to predict the effects of a complete core reconfiguration and the accuracy of this prediction was tested experimentally. We used the Monte Carlo code MCNP5 to reproduce system neutronics in different operating conditions and to analyse neutron fluxes in the reactor core. The software that took care of time evolution, completely designed in-house, used the neutron fluxes obtained by MCNP5 to evaluate fuel consumption. This software was developed specifically to keep into account some features that differentiate experimental reactors from power ones, such as the daily ON/OFF cycle and the long fuel lifetime. These effects can not be neglected to properly account for neutron poison accumulation. We evaluated the effect of 48 years of reactor operation and predicted a possible new configuration for the reactor core: the objective was to remove some of the fuel elements from the...
Energy Technology Data Exchange (ETDEWEB)
Ahn, Sang Kyu [Korea Institute of Nuclear Safety, 19 Kusong-dong, Yuseong-gu, Daejeon 305-338 (Korea, Republic of); Kim, Inn Seock, E-mail: innseockkim@gmail.co [ISSA Technology, 21318 Seneca Crossing Drive, Germantown, MD 20876 (United States); Oh, Kyu Myung [Korea Institute of Nuclear Safety, 19 Kusong-dong, Yuseong-gu, Daejeon 305-338 (Korea, Republic of)
2010-05-15
The objective of this paper and a companion paper in this issue (part II, risk-informed approaches) is to derive technical insights from a critical review of deterministic and risk-informed safety analysis approaches that have been applied to develop licensing requirements for water-cooled reactors, or proposed for safety verification of the advanced reactor design. To this end, a review was made of a number of safety analysis approaches including those specified in regulatory guides and industry standards, as well as novel methodologies proposed for licensing of advanced reactors. This paper and the companion paper present the review insights on the deterministic and risk-informed safety analysis approaches, respectively. These insights could be used in making a safety case or developing a new licensing review infrastructure for advanced reactors including Generation IV reactors.
Discussion about modeling the effects of neutron flux exposure for nuclear reactor core analysis
Energy Technology Data Exchange (ETDEWEB)
Vondy, D.R.
1986-04-01
Methods used to calculate the effects of exposure to a neutron flux are described. The modeling of the nuclear-reactor core history presents an analysis challenge. The nuclide chain equations must be solved, and some of the methods in use for this are described. Techniques for treating reactor-core histories are discussed and evaluated.
Energy efficiency analysis of reactor for torrefaction of biomass with direct heating
Kuzmina, J. S.; Director, L. B.; Shevchenko, A. L.; Zaichenko, V. M.
2016-11-01
Paper presents energy analysis of reactor for torrefaction with direct heating of granulated biomass by exhaust gases. Various schemes of gas flow through the reactor zones are presented. Performed is a comparative evaluation of the specific energy consumption for the considered schemes. It has been shown that one of the most expensive processes of torrefaction technology is recycling of pyrolysis gases.
Degraded core analysis for the pressurized-water reactor
Energy Technology Data Exchange (ETDEWEB)
Gittus, J.H.
1987-02-09
An analysis of the likelihood and the consequences of 'degraded-core accidents' has been undertaken for the proposed Sizewell B PWR. In such accidents, degradation of the core geometry occurs as a result of overheating. Radionuclides are released and may enter the environment, causing harmful effects. The analysis concludes that degraded-core accidents are highly improbable, the plant having been designed to reduce the frequency of such accidents to a value of order 10/sup -6/ per year. Tbe building containing the reactor would only fail in a small proportion of degraded-core accidents. In the great majority of cases the containment would remain intact and the release of radioactivity to the environment would be small. The risk to individuals have been calculated for both immediate and long term effects. Although the estimates of risk are approximate, studies to investigate the uncertainties, and sensitivities to different assumptions, show that potential errors are small compared with the very large 'margin of safety' between the risks estimated for Sizewell B and those that already exist in society.
2013-01-01
The Handbook of Nuclear Engineering is an authoritative compilation of information regarding methods and data used in all phases of nuclear engineering. Addressing nuclear engineers and scientists at all academic levels, this five volume set provides the latest findings in nuclear data and experimental techniques, reactor physics, kinetics, dynamics and control. Readers will also find a detailed description of data assimilation, model validation and calibration, sensitivity and uncertainty analysis, fuel management and cycles, nuclear reactor types and radiation shielding. A discussion of radioactive waste disposal, safeguards and non-proliferation, and fuel processing with partitioning and transmutation is also included. As nuclear technology becomes an important resource of non-polluting sustainable energy in the future, The Handbook of Nuclear Engineering is an excellent reference for practicing engineers, researchers and professionals.
REACTOR PRESSURE VESSEL TEMPERATURE ANALYSIS OF CANDIDATE VERY HIGH TEMPERATURE REACTOR DESIGNS
Energy Technology Data Exchange (ETDEWEB)
Hans D. Gougar; Cliff B. Davis; George Hayner; Kevan Weaver
2006-10-01
Analyses were performed to determine maximum temperatures in the reactor pressure vessel for two potential Very-High Temperature Reactor (VHTR) designs during normal operation and during a depressurized conduction cooldown accident. The purpose of the analyses was to aid in the determination of appropriate reactor vessel materials for the VHTR. The designs evaluated utilized both prismatic and pebble-bed cores that generated 600 MW of thermal power. Calculations were performed for fluid outlet temperatures of 900 and 950 °C, corresponding to the expected range for the VHTR. The analyses were performed using the RELAP5-3D and PEBBED-THERMIX computer codes. Results of the calculations were compared with preliminary temperature limits derived from the ASME pressure vessel code. Because PEBBED-THERMIX has not been extensively validated, confirmatory calculations were also performed with RELAP5-3D for the pebble-bed design. During normal operation, the predicted axial profiles in reactor vessel temperature were similar with both codes and the predicted maximum values were within 2 °C. The trends of the calculated vessel temperatures were similar during the depressurized conduction cooldown accident. The maximum value predicted with RELAP5-3D during the depressurized conduction cooldown accident was about 40 °C higher than that predicted with PEBBED. This agreement is considered reasonable based on the expected uncertainty in either calculation. The differences between the PEBBED and RELAP5-3D calculations were not large enough to affect conclusions concerning comparisons between calculated and allowed maximum temperatures during normal operation and the depressurized conduction cooldown accident.
Current status of neutron activation analysis in HANARO Research Reactor
Energy Technology Data Exchange (ETDEWEB)
Chung, Yong Sam; Moon, Jong Hwa; Sohn, Jae Min [Korea Atomic Energy Research Institute, Daejeon (Korea)
2003-03-01
The facilities for neutron activation analysis in the HANARO (Hi-flux Advanced Neutron Application Research Reactor) are described and the main applications of NAA (Neutron Activation Analysis) are reviewed. The sample irradiation tube, automatic and manual pneumatic transfer system were installed at three irradiation holes of HANARO at the end of 1995. The performance of the NAA facility was examined to identify the characteristics of the tube transfer system, irradiation sites and custom-made polyethylene irradiation capsule. The available thermal neutron fluxes at irradiation sites are in the range of 3 x 10{sup 13} - 1 x 10{sup 14} n/cm{sup 2}{center_dot}s and cadmium ratios are in 15 - 250. For an automatic sample changer for gamma-ray counting, a domestic product was designed and manufactured. An integrated computer program (Labview) to analyse the content was developed. In 2001, PGNAA (Prompt Gamma Neutron Activation Analysis) facility has been installed using a diffracted neutron beam of ST1. NAA has been applied in the trace component analysis of nuclear, geological, biological, environmental and high purity materials, and various polymers for research and development. The improvement of analytical procedures and establishment of an analytical quality control and assurance system were studied. Applied research and development for the environment, industry and human health by NAA and its standardization were carried out. For the application of the KOLAS (Korea Laboratory Accreditation Scheme), evaluation of measurement uncertainty and proficiency testing of reference materials were performed. Also to verify the reliability and to validate analytical results, intercomparison studies between laboratories were carried out. (author)
System Requirements Analysis for a Computer-based Procedure in a Research Reactor Facility
Energy Technology Data Exchange (ETDEWEB)
Park, Jaek Wan; Jang, Gwi Sook; Seo, Sang Moon; Shin, Sung Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2014-10-15
This can address many of the routine problems related to human error in the use of conventional, hard-copy operating procedures. An operating supporting system is also required in a research reactor. A well-made CBP can address the staffing issues of a research reactor and reduce the human errors by minimizing the operator's routine tasks. A CBP for a research reactor has not been proposed yet. Also, CBPs developed for nuclear power plants have powerful and various technical functions to cover complicated plant operation situations. However, many of the functions may not be required for a research reactor. Thus, it is not reasonable to apply the CBP to a research reactor directly. Also, customizing of the CBP is not cost-effective. Therefore, a compact CBP should be developed for a research reactor. This paper introduces high level requirements derived by the system requirements analysis activity as the first stage of system implementation. Operation support tools are under consideration for application to research reactors. In particular, as a full digitalization of the main control room, application of a computer-based procedure system has been required as a part of man-machine interface system because it makes an impact on the operating staffing and human errors of a research reactor. To establish computer-based system requirements for a research reactor, this paper addressed international standards and previous practices on nuclear plants.
Full-Band Monte Carlo Analysis of Hot-Carrier Light Emission in GaAs
Ferretti, I.; Abramo, A.; Brunetti, R.; Jacobini, C.
1997-11-01
A computational analysis of light emission from hot carriers in GaAs due to direct intraband conduction-conduction (c-c) transitions is presented. The emission rates have been evaluated by means of a Full-Band Monte-Carlo simulator (FBMC). Results have been obtained for the emission rate as a function of the photon energy, for the emitted and absorbed light polarization along and perpendicular to the electric field direction. Comparison has been made with available experimental data in MESFETs.
Passive gamma analysis of the boiling-water-reactor assemblies
Vo, D.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S.; Trellue, H.; Vaccaro, S.
2016-09-01
This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative-Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.
Reactor coolant pump testing using motor current signatures analysis
Energy Technology Data Exchange (ETDEWEB)
Burstein, N.; Bellamy, J.
1996-12-01
This paper describes reactor coolant pump motor testing carried out at Florida Power Corporation`s Crystal River plant using Framatome Technologies` new EMPATH (Electric Motor Performance Analysis and Trending Hardware) system. EMPATH{trademark} uses an improved form of Motor Current Signature Analysis (MCSA), technology, originally developed at Oak Ridge National Laboratories, for detecting deterioration in the rotors of AC induction motors. Motor Current Signature Analysis (MCSA) is a monitoring tool for motor driven equipment that provides a non-intrusive means for detecting the presence of mechanical and electrical abnormalities in the motor and the driven equipment. The base technology was developed at the Oak Ridge National Laboratory as a means for determining the affects of aging and service wear specifically on motor-operated valves used in nuclear power plant safety systems, but it is applicable to a broad range of electric machinery. MCSA is based on the recognition that an electric motor (ac or dc) driving a mechanical load acts as an efficient and permanently available transducer by sensing mechanical load variations, large and small, long-term and rapid, and converting them into variations in the induced current generated in the motor windings. The motor current variations, resulting from changes in load caused by gears, pulleys, friction, bearings, and other conditions that may change over the life of the motor, are carried by the electrical cables powering the motor and are extracted at any convenient location along the motor lead. These variations modulate the 60 Hz carrier frequency and appear as sidebands in the spectral plot.
Passive gamma analysis of the boiling-water-reactor assemblies
Energy Technology Data Exchange (ETDEWEB)
Vo, D., E-mail: ducvo@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Atomic Energy Community (EURATOM), Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Atomic Energy Community (EURATOM), Luxemburg (Luxembourg)
2016-09-11
This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative–Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: {sup 137}Cs, {sup 154}Eu, {sup 134}Cs, and to a lesser extent, {sup 106}Ru and {sup 144}Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.
Model Based Cyber Security Analysis for Research Reactor Protection System
Energy Technology Data Exchange (ETDEWEB)
Sho, Jinsoo; Rahman, Khalil Ur; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Son, Hanseong [Joongbu Univ., Geumsan (Korea, Republic of)
2013-07-01
The study on the qualitative risk due to cyber-attacks into research reactors was performed using bayesian Network (BN). This was motivated to solve the issues of cyber security raised due to digitalization of instrumentation and control (I and C) system. As a demonstrative example, we chose the reactor protection system (RPS) of research reactors. Two scenarios of cyber-attacks on RPS were analyzed to develop mitigation measures against vulnerabilities. The one is the 'insertion of reactor trip' and the other is the 'scram halt'. The six mitigation measures are developed for five vulnerability for these scenarios by getting the risk information from BN.
Impact of inflow transport approximation on light water reactor analysis
Choi, Sooyoung; Smith, Kord; Lee, Hyun Chul; Lee, Deokjung
2015-10-01
The impact of the inflow transport approximation on light water reactor analysis is investigated, and it is verified that the inflow transport approximation significantly improves the accuracy of the transport and transport/diffusion solutions. A methodology for an inflow transport approximation is implemented in order to generate an accurate transport cross section. The inflow transport approximation is compared to the conventional methods, which are the consistent-PN and the outflow transport approximations. The three transport approximations are implemented in the lattice physics code STREAM, and verification is performed for various verification problems in order to investigate their effects and accuracy. From the verification, it is noted that the consistent-PN and the outflow transport approximations cause significant error in calculating the eigenvalue and the power distribution. The inflow transport approximation shows very accurate and precise results for the verification problems. The inflow transport approximation shows significant improvements not only for the high leakage problem but also for practical large core problem analyses.
Non normal modal analysis of oscillations in boiling water reactors
Energy Technology Data Exchange (ETDEWEB)
Suarez-Antola, Roberto, E-mail: roberto.suarez@miem.gub.uy [Ministerio de Industria, Energia y Mineria (MIEM), Montevideo (Uruguay); Flores-Godoy, Jose-Job, E-mail: job.flores@ibero.mx [Universidad Iberoamericana (UIA), Mexico, DF (Mexico). Dept. de Fisica Y Matematicas
2013-07-01
The first objective of the present work is to construct a simple reduced order model for BWR stability analysis, combining a two nodes nodal model of the thermal hydraulics with a two modes modal model of the neutronics. Two coupled non-linear integral-differential equations are obtained, in terms of one global (in phase) and one local (out of phase) power amplitude, with direct and cross feedback reactivities given as functions of thermal hydraulics core variables (void fractions and temperatures). The second objective is to apply the effective life time approximation to further simplify the nonlinear equations. Linear approximations for the equations of the amplitudes of the global and regional modes are derived. The linearized equation for the amplitude of the global mode corresponds to a decoupled and damped harmonic oscillator. An analytical closed form formula for the damping coefficient, as a function of the parameters space of the BWR, is obtained. The coefficient changes its sign (with the corresponding modification in the decay ratio) when a stability boundary is crossed. This produces a supercritical Hopf bifurcation, with the steady state power of the reactor as the bifurcation parameter. However, the linearized equation for the amplitude of the regional mode corresponds always to an over-damped and always coupled (with the amplitude of the global mode) harmonic oscillator, for every set of possible values of core parameters (including the steady state power of the reactor) in the framework of the present mathematical model. The equation for the above mentioned over damped linear oscillator is closely connected with a non-normal operator. Due to this connection, there could be a significant transient growth of some solutions of the linear equation. This behavior allows a significant shrinking of the basin of attraction of the equilibrium state. The third objective is to apply the above approach to partially study the stability of the regional mode and
Models and Stability Analysis of Boiling Water Reactors
Energy Technology Data Exchange (ETDEWEB)
John Dorning
2002-04-15
We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model that includes: space-time modal neutron kinetics based on spatial w-modes; single- and two-phase flow in parallel boiling channels; fuel rod heat conduction dynamics; and a simple model of the recirculation loop. The BR model is represented by a set of time-dependent nonlinear ordinary differential equations, and is studied as a dynamical system using the modern bifurcation theory and nonlinear dynamical systems analysis. We first determine the stability boundary (SB) - or Hopf bifurcation set- in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value; then we transform the SB to the practical power-flow map used by BWR operating engineers and regulatory agencies. Using this SB, we show that the normal operating point at 100% power is very stable, that stability of points on the 100% rod line decreases as the flow rate is reduced, and that operating points in the low-flow/high-power region are least stable. We also determine the SB that results when the modal kinetics is replaced by simple point reactor kinetics, and we thereby show that the first harmonic mode does not have a significant effect on the SB. However, we later show that it nevertheless has a significant effect on stability because it affects the basin of attraction of stable operating points. Using numerical simulations we show that, in the important low-flow/high-power region, the Hopf bifurcation that occurs as the SB is crossed is subcritical; hence, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line at points in the low-flow/high-power region. Numerical simulations are also performed to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is
Institute of Scientific and Technical Information of China (English)
张志宏; 夏晓彬; 蔡军; 王建华; 李长园; 葛良全; 张庆贤
2015-01-01
The Chinese Academy of Science has launched a thorium-based molten-salt reactor (TMSR) research project with a mission to research and develop a fission energy system of the fourth generation. The TMSR project intends to construct a liquid fuel molten-salt reactor (TMSR-LF), which uses fluoride salt as both the fuel and coolant, and a solid fuel molten-salt reactor (TMSR-SF), which uses fluoride salt as coolant and TRISO fuel. An optimized 2 MWth TMSR-LF has been designed to solve major technological challenges in the Th-U fuel cycle. Preliminary conceptual shielding design has also been performed to develop bulk shielding. In this study, the radiation dose and temperature distribution of the shielding bulk due to the core were simulated and analyzed by performing Monte Carlo simulations and computational fluid dynamics (CFD) analysis. The MCNP calculated dose rate and neutron and gamma spectra indicate that the total dose rate due to the core at the external surface of the concrete wall was 1.91 µSv/h in the radial direction, 1.16 µSv/h above and 1.33 µSv/h below the bulk shielding. All the radiation dose rates due to the core were below the design criteria. Thermal analysis results show that the temperature at the outermost surface of the bulk shielding was 333.86 K, which was below the required limit value. The results indicate that the designed bulk shielding satisfies the radiation shielding requirements for the 2 MWth TMSR-LF.
A study on naphtha catalytic reforming reactor simulation and analysis
Liang, Ke-min; Guo, Hai-Yan; Pan, Shi-Wei
2005-01-01
A naphtha catalytic reforming unit with four reactors in series is analyzed. A physical model is proposed to describe the catalytic reforming radial flow reactor. Kinetics and thermodynamics equations are selected to describe the naphtha catalytic reforming reactions characteristics based on idealizing the complex naphtha mixture by representing the paraffin, naphthene, and aromatic groups by single compounds. The simulation results based above models agree very well with actual operation uni...
A study on naphtha catalytic reforming reactor simulation and analysis.
Liang, Ke-min; Guo, Hai-yan; Pan, Shi-wei
2005-06-01
A naphtha catalytic reforming unit with four reactors in series is analyzed. A physical model is proposed to describe the catalytic reforming radial flow reactor. Kinetics and thermodynamics equations are selected to describe the naphtha catalytic reforming reactions characteristics based on idealizing the complex naphtha mixture by representing the paraffin, naphthene, and aromatic groups by single compounds. The simulation results based above models agree very well with actual operation unit data.
Study and Analysis on Naphtha Catalytic Reforming Reactor Simulation
Institute of Scientific and Technical Information of China (English)
Liang Ke min; Song Yongji; Pan Shiwei
2004-01-01
A naphtha catalytic reforming unit with four reactors connected in series is analyzed. A physical model is proposed to describe the catalytic reforming radial flow reactor. Kinetics and thermodynamics equations are selected to describe the naphtha catalytic reforming reaction characteristics based on idealizing the complex naphtha mixture to represent the paraffin, naphthene, and aromatic groups with individual compounds. The simulation results based on above models agree very well with actual operating data of process unit.
A study on naphtha catalytic reforming reactor simulation and analysis
Institute of Scientific and Technical Information of China (English)
LIANG Ke-min; GUO Hai-yan; PAN Shi-wei
2005-01-01
A naphtha catalytic reforming unit with four reactors in series is analyzed. A physical model is proposed to describe the catalytic reforming radial flow reactor. Kinetics and thermodynamics equations are selected to describe the naphtha catalytic reforming reactions characteristics based on idealizing the complex naphtha mixture by representing the paraffin, naphthene, and aromatic groups by single compounds. The simulation results based above models agree very well with actual operation unit data.
Analysis of N-16 concentration in primary cooling system of AP1000 power reactor
Energy Technology Data Exchange (ETDEWEB)
Rohanda, Anis [Center for Reactor Technology and Nuclear Safety – BATAN Kawasan PUSPIPTEK Gd. No. 80 Serpong, Tangerang Selatan 15310 (Indonesia); Waris, Abdul [Physics Department of ITB, Indonesia anis-rohanda@yahoo.com (Indonesia)
2015-04-16
Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on {sup 16}O(n,p){sup 16}N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.
Linear stability analysis of a nuclear reactor using the lumped model
Directory of Open Access Journals (Sweden)
Kale Vivek A.
2016-01-01
Full Text Available The stability analysis of a nuclear reactor is an important aspect in the design and operation of the reactor. A stable neutronic response to perturbations is essential from the safety point of view. In this paper, a general methodology has been developed for the linear stability analysis of nuclear reactors using the lumped reactor model. The reactor kinetics has been modelled using the point kinetics equations and the reactivity feedbacks from fuel, coolant and xenon have been modelled through the appropriate time dependent equations. These governing equations are linearized considering small perturbations in the reactor state around a steady operating point. The characteristic equation of the system is used to establish the stability zone of the reactor considering the reactivity coefficients as parameters. This methodology has been used to identify the stability region of a typical pressurized heavy water reactor. It is shown that the positive reactivity feedback from xenon narrows down the stability region. Further, it is observed that the neutron kinetics parameters (such as the number of delayed neutron precursor groups considered, the neutron generation time, the delayed neutron fractions, etc. do not have a significant influence on the location of the stability boundary. The stability boundary is largely influenced by the parameters governing the evolution of the fuel and coolant temperature and xenon concentration.
A Monte Carlo Uncertainty Analysis of Ozone Trend Predictions in a Two Dimensional Model. Revision
Considine, D. B.; Stolarski, R. S.; Hollandsworth, S. M.; Jackman, C. H.; Fleming, E. L.
1998-01-01
We use Monte Carlo analysis to estimate the uncertainty in predictions of total O3 trends between 1979 and 1995 made by the Goddard Space Flight Center (GSFC) two-dimensional (2D) model of stratospheric photochemistry and dynamics. The uncertainty is caused by gas-phase chemical reaction rates, photolysis coefficients, and heterogeneous reaction parameters which are model inputs. The uncertainty represents a lower bound to the total model uncertainty assuming the input parameter uncertainties are characterized correctly. Each of the Monte Carlo runs was initialized in 1970 and integrated for 26 model years through the end of 1995. This was repeated 419 times using input parameter sets generated by Latin Hypercube Sampling. The standard deviation (a) of the Monte Carlo ensemble of total 03 trend predictions is used to quantify the model uncertainty. The 34% difference between the model trend in globally and annually averaged total O3 using nominal inputs and atmospheric trends calculated from Nimbus 7 and Meteor 3 total ozone mapping spectrometer (TOMS) version 7 data is less than the 46% calculated 1 (sigma), model uncertainty, so there is no significant difference between the modeled and observed trends. In the northern hemisphere midlatitude spring the modeled and observed total 03 trends differ by more than 1(sigma) but less than 2(sigma), which we refer to as marginal significance. We perform a multiple linear regression analysis of the runs which suggests that only a few of the model reactions contribute significantly to the variance in the model predictions. The lack of significance in these comparisons suggests that they are of questionable use as guides for continuing model development. Large model/measurement differences which are many multiples of the input parameter uncertainty are seen in the meridional gradients of the trend and the peak-to-peak variations in the trends over an annual cycle. These discrepancies unambiguously indicate model formulation
Detailed analysis for a control rod worth of the gas turbine high temperature reactor (GTHTR300)
Energy Technology Data Exchange (ETDEWEB)
Nakata, Tetsuo; Katanishi, Shoji; Takada, Shoji; Yan, Xing; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment
2002-11-01
GTHTR300 is composed of a simplified and economical power plant based on an inherent safe 600 MWt reactor and a nearly 50% high efficiency gas turbine power conversion cycle. GTHTR300 core consist of annular fuel region, center and outer side reflectors because of cooling it effectively in depressurized accident conditions, and all control rods are located in both side reflectors of annular core. As a thermal neutron spectrum is strongly distorted in reflector regions, an accurate calculation is especially required for the control rod worth evaluation. In this study, we applied the detailed Monte Carlo calculations of a full core model, and confirmed that our design method has enough accuracy. (author)
Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Fast Salt Reactor (MFSR)
Laureau, A.; Rubiolo, P. R.; Heuer, D.; Merle-Lucotte, E.; Brovchenko, M.
2014-06-01
Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor are presented. These preliminary numerical simulations are carried-out using the Monte Carlo code MCNP and the Computation Fluid Dynamic code OpenFOAM. The main objectives of this analysis performed at steady-reactor conditions are to confirm the acceptability of the current neutronic and thermalhydraulic designs of the reactor, to study the effects of the reactor operating conditions on some of the key MSFR design parameters such as the temperature peaking factor. The effects of the precursor's motion on the reactor safety parameters such as the effective fraction of delayed neutrons have been evaluated.
Vishnuganth, M A; Remya, Neelancherry; Kumar, Mathava; Selvaraju, N
2017-02-22
Carbofuran (CBF) removal in a continuous-flow photocatalytic reactor with granular activated carbon supported titanium dioxide (GAC-TiO2) catalyst was investigated. The effects of feed flow rate, TiO2 concentration and addition of supplementary oxidants on CBF removal were investigated. The central composite design (CCD) was used to design the experiments and to estimate the effects of feed flow rate and TiO2 concentration on CBF removal. The outcome of CCD experiments demonstrated that reactor performance was influenced mainly by feed flow rate compared to TiO2 concentration. A second-order polynomial model developed based on CCD experiments fitted the experimental data with good correlation (R(2) ∼ 0.964). The addition of 1 mL min(-1) hydrogen peroxide has shown complete CBF degradation and 76% chemical oxygen demand removal under the following operating conditions of CBF ∼50 mg L(-1), TiO2 ∼5 mg L(-1) and feed flow rate ∼82.5 mL min(-1). Rate constant of the photodegradation process was also calculated by applying the kinetic data in pseudo-first-order kinetics. Four major degradation intermediates of CBF were identified using GC-MS analysis. As a whole, the reactor system and GAC-TiO2 catalyst used could be constructive in cost-effective CBF removal with no impact to receiving environment through getaway of photocatalyst.
Hoffmann, Max J; Matera, Sebastian
2016-01-01
Lattice kinetic Monte Carlo simulations have become a vital tool for predictive quality atomistic understanding of complex surface chemical reaction kinetics over a wide range of reaction conditions. In order to expand their practical value in terms of giving guidelines for atomic level design of catalytic systems, it is very desirable to readily evaluate a sensitivity analysis for a given model. The result of such a sensitivity analysis quantitatively expresses the dependency of the turnover frequency, being the main output variable, on the rate constants entering the model. In the past the application of sensitivity analysis, such as Degree of Rate Control, has been hampered by its exuberant computational effort required to accurately sample numerical derivatives of a property that is obtained from a stochastic simulation method. In this study we present an efficient and robust three stage approach that is capable of reliably evaluating the sensitivity measures for stiff microkinetic models as we demonstrat...
Use of Monte Carlo simulations for cultural heritage X-ray fluorescence analysis
Energy Technology Data Exchange (ETDEWEB)
Brunetti, Antonio, E-mail: brunetti@uniss.it [Polcoming Department, University of Sassari (Italy); Golosio, Bruno [Polcoming Department, University of Sassari (Italy); Schoonjans, Tom; Oliva, Piernicola [Chemical and Pharmaceutical Department, University of Sassari (Italy)
2015-06-01
The analytical study of Cultural Heritage objects often requires merely a qualitative determination of composition and manufacturing technology. However, sometimes a qualitative estimate is not sufficient, for example when dealing with multilayered metallic objects. Under such circumstances a quantitative estimate of the chemical contents of each layer is sometimes required in order to determine the technology that was used to produce the object. A quantitative analysis is often complicated by the surface state: roughness, corrosion, incrustations that remain even after restoration, due to efforts to preserve the patina. Furthermore, restorers will often add a protective layer on the surface. In all these cases standard quantitative methods such as the fundamental parameter based approaches are generally not applicable. An alternative approach is presented based on the use of Monte Carlo simulations for quantitative estimation. - Highlights: • We present an application of fast Monte Carlo codes for Cultural Heritage artifact analysis. • We show applications to complex multilayer structures. • The methods allow estimating both the composition and the thickness of multilayer, such as bronze with patina. • The performance in terms of accuracy and uncertainty is described for the bronze samples.
Directory of Open Access Journals (Sweden)
Shchekoturova S. D.
2015-04-01
Full Text Available The article presents an analysis of an innovative activity of four Russian metallurgical enterprises: "Ruspolimet", JSC "Ural Smithy", JSC "Stupino Metallurgical Company", JSC "VSMPO" via mathematical modeling using Monte Carlo method. The results of the assessment of innovative activity of Russian metallurgical companies were identified in five years dynamics. An assessment of the current innovative activity was made by the calculation of an integral index of the innovative activity. The calculation was based on such six indicators as the proportion of staff employed in R & D; the level of development of new technology; the degree of development of new products; share of material resources for R & D; degree of security of enterprise intellectual property; the share of investment in innovative projects and it was analyzed from 2007 to 2011. On the basis of this data the integral indicator of the innovative activity of metallurgical companies was calculated by well-known method of weighting coefficients. The comparative analysis of integral indicators of the innovative activity of considered companies made it possible to range their level of the innovative activity and to characterize the current state of their business. Based on Monte Carlo method a variation interval of the integral indicator was obtained and detailed instructions to choose the strategy of the innovative development of metallurgical enterprises were given as well
Seismic fragility analysis of buried steel piping at P, L, and K reactors
Energy Technology Data Exchange (ETDEWEB)
Wingo, H.E.
1989-10-01
Analysis of seismic strength of buried cooling water piping in reactor areas is necessary to evaluate the risk of reactor operation because seismic events could damage these buried pipes and cause loss of coolant accidents. This report documents analysis of the ability of this piping to withstand the combined effects of the propagation of seismic waves, the possibility that the piping may not behave in a completely ductile fashion, and the distortions caused by relative displacements of structures connected to the piping.
Empirical Risk Analysis of Severe Reactor Accidents in Nuclear Power Plants after Fukushima
Directory of Open Access Journals (Sweden)
Jan Christian Kaiser
2012-01-01
Full Text Available Many countries are reexamining the risks connected with nuclear power generation after the Fukushima accidents. To provide updated information for the corresponding discussion a simple empirical approach is applied for risk quantification of severe reactor accidents with International Nuclear and Radiological Event Scale (INES level ≥5. The analysis is based on worldwide data of commercial nuclear facilities. An empirical hazard of 21 (95% confidence intervals (CI 4; 62 severe accidents among the world’s reactors in 100,000 years of operation has been estimated. This result is compatible with the frequency estimate of a probabilistic safety assessment for a typical pressurised power reactor in Germany. It is used in scenario calculations concerning the development in numbers of reactors in the next twenty years. For the base scenario with constant reactor numbers the time to the next accident among the world's 441 reactors, which were connected to the grid in 2010, is estimated to 11 (95% CI 3.7; 52 years. In two other scenarios a moderate increase or decrease in reactor numbers have negligible influence on the results. The time to the next accident can be extended well above the lifetime of reactors by retiring a sizeable number of less secure ones and by safety improvements for the rest.
Simple analysis of an External Vessel Cooling Thermosyphon for a Sodium-cooled Fast Reactor
Energy Technology Data Exchange (ETDEWEB)
Choi, Jae Young; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Song, Sub Lee [Handong Global University, Pohang (Korea, Republic of)
2015-05-15
KALIMER has three different DHR systems: two non-safety grade systems and one safety grade system. The non-safety grade systems are an IRACS (Intermediate Reactor Auxiliary Cooling System) and a steam/feedwater system. The safety grade system is a PDRC (Passive Decay Heat Removal Circuit). In case of the foreign reactor designs, ABTR (Advanced Burner Test Reactor) has a DRACS (Direct Reactor Auxiliary Cooling System), a PFBR (Indian Prototype Fast Breeder Reactor) has an SGDHRS (Safety Grade Decay Heat Removal System), and an EFR (European Fast Reactor) has DRC (Direct Reactor Cooling). Those designs have advantage on relatively high decay heat removal capacity. However, larger vessel size due to subsidiary in-vessel structure and possible accident propagation to reactor induced by sodium fire. In this paper, an ex-vessel thermosyphon design was proposed for the removal of decay heat for an iSFR. The proposed ex-vessel thermosyphon was designed to remove decay heat in both transient cases and BDBA cases, such as vessel failure. Proper working fluid was selected based on thermodynamic properties and chemical stability. Mercury was chosen as the working fluid, and SUS 314 was used for the corresponding structure material. Possible chemical reactions and adverse effects from using the thermosyphon were inherently eliminated by the system layout. A model for a high-temperature thermosyphon and numerical algorithms were used for the analysis. As a result of the simulation, the thermosyphon design was optimized, and it showed sufficient DHR performance to maintain core integrity.
Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme
Widiawati, Nina; Su'ud, Zaki
2015-09-01
Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uranium fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from -0.6695443 % at BOC to -0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.
Energy Technology Data Exchange (ETDEWEB)
Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer
2010-09-01
The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450
Analysis of barium hydroxide and calcium hydroxide slurry carbonation reactors
Energy Technology Data Exchange (ETDEWEB)
Patch, K.D.; Hart, R.P.; Schumacher, W.A.
1980-05-01
The removal of CO/sub 2/ from air was investigated by using a continuous-agitated-slurry carbonation reactor containing either barium hydroxide (Ba(OH)/sub 2/) or calcium hydroxide (Ca(OH)/sub 2/). Such a process would be applied to scrub /sup 14/CO/sub 2/ from stack gases at nuclear-fuel reprocessing plants. Decontamination factors were characterized for reactor conditions which could alter hydrodynamic behavior. An attempt was made to characterize reactor performance with models assuming both plug flow and various degrees of backmixing in the gas phase. The Ba(OH)/sub 2/ slurry enabled increased conversion, but apparently the process was controlled under some conditions by phenomena differing from those observed for carbonation by Ca(OH)/sub 2/. Overall reaction mechanisms are postulated.
Configurational analysis of an EBT reactor in various magnetic geometries
Energy Technology Data Exchange (ETDEWEB)
Owen, L.W.; Uckan, N.A.
1980-01-01
Optimization of vacuum field particle confinement in an ELMO Bumpy Torus (EBT) reactor has been considered. Several methods of improving the efficient utilization of magnetic fields and the particle confinement characteristics of a reactor have been analyzed. These include the use of (1) magnets with a large mirror ratio, (2) high field Nb/sub 3/Sn or Nb/sub 3/Sn/NbTi hybrid mirror coils, (3) split-wedge mirror coils, (4) aspect ratio enhancement (ARE) coils, and (5) recently developed field symmetrizing (SYM) coils. Of these, particle drift orbits and three-dimensional tensor pressure equilibrium calculations have shown that the ARE and SYM coils used in conjunction with high field magnets offer the most promise of good plasma performance in a smaller size (up to 50%) EBT reactor. The relative merits of each magnetic configuration are discussed, and the design characteristics are given.
Total loss of AC power analysis for EPR reactor
Energy Technology Data Exchange (ETDEWEB)
Darnowski, Piotr, E-mail: piotr.darnowski@itc.pw.edu.pl [Warsaw University of Technology, Institute of Heat Engineering, Nowowiejska 21/25, 00-665 Warsaw (Poland); Skrzypek, Eleonora, E-mail: eleonora.skrzypek@ncbj.gov.pl [Warsaw University of Technology, Institute of Heat Engineering, Nowowiejska 21/25, 00-665 Warsaw (Poland); National Centre for Nuclear Research (NCBJ), A. Sołtana 7, 05-400 Otwock-Świerk (Poland); Mazgaj, Piotr, E-mail: piotr.mazgaj@itc.pw.edu.pl [Warsaw University of Technology, Institute of Heat Engineering, Nowowiejska 21/25, 00-665 Warsaw (Poland); Świrski, Konrad [Warsaw University of Technology, Institute of Heat Engineering, Nowowiejska 21/25, 00-665 Warsaw (Poland); Gandrille, Pascal [AREVA NP SAS, Tour AREVA, 1 place Jean Millier, 92084 Paris La Défense (France)
2015-08-15
Highlights: • Total loss of AC power (Station Blackout) was simulated for the EPR reactor model. • In-vessel phase of the accident is under consideration. • Comparison of MELCOR and MAAP results is presented. • MELCOR and MAAP results are comparable. - Abstract: In this paper the results of severe accident simulations for the EPR reactor in the case of loss of offsite power combined with total failure of all diesel generators (total loss of AC power) are presented. Calculations were performed with MELCOR 2.1 computer code for in-vessel phase of the accident. In this scenario, the unavailability of all offsite and onsite power sources and the lack of cooling leads directly to core degradation, material relocation to the lower plenum and rupture of the reactor pressure vessel. MELCOR results were compared qualitatively and quantitatively with MAAP4 code results and show a good agreement.
Plasma engineering analysis of a small torsatron reactor
Energy Technology Data Exchange (ETDEWEB)
Lacatski, J.T.; Houlberg, W.A.; Uckan, N.A.
1985-10-01
This study examines the plasma physics and reactor engineering feasibility of a small, medium aspect ratio, high-beta, l = 2, D-T torsatron power reactor, based on the magnetic configuration of the Advanced Toroidal Facility, Oak Ridge National Laboratory. Plasma analyses are performed to assess whether confinement in a small, average radius plasma is sufficient to yield an ignited or high-Q driven device. Much of the physics assessment focuses on an evaluation of the radial electric field created by the nonambipolar particle flux. Detailed transport simulations are done with both fixed and self-consistent evolution of the radial electric field. Basic reactor engineering considerations taken into account are neutron wall loading, maximum magnetic field at the helical coils, coil shield thickness, and tritium breeding blanket-shield thickness.
Energy Technology Data Exchange (ETDEWEB)
Martin del Campo, C.; Nelson, P.F.; Francois, J.L. [Departamento de Sistemas Energeticos, Facultad de Ingenieria, UNAM, Paseo Cuauhnahuac 8532, Jiutepec, 62550 Morelos (Mexico)]. e-mail: cmcm@fi-b.unam.mx
2008-07-01
In this work the decision making method known as hierarchical analysis process for the selection of a new reactor in Mexico was applied. The main objective of the process it is to select the nuclear reactor technology more appropriate for Mexico, to begin the bid process inside one or two years to begin their operation in 2016. The options were restricted to four reactors that fulfill the following ones approaches: 1) its are advanced reactors, from the technological point of view, with regard to the reactors that at the moment operate in the Laguna Verde Power Station, 2) its are reactors that have the totally finished design, 3) its are reactors that already have the certification on the part of the regulator organism of the origin country or that they are in an advanced state of the certification process and 4) its are reactors offered by the companies that they have designed and built the greater number of reactors that are at the moment in operation at world level. Taking into account these restrictions it was decided to consider as alternative at the reactors: Advanced Boiling Water Reactor (A BWR), European Reactor of Pressurized Water (EPR), Water at Pressure reactor (AP1000) and Simplified Economic Reactor of Boiling Water (ESBWR). The evaluation approaches include economic and of safety indicators, qualitative some of them and other quantitative ones. Another grade of complexity in the solution of the problem is that there are actors that can be involved in the definition of the evaluation approaches and in the definition of the relative importance among them, according to each actor's interests. To simplify the problem its were only considered two actors or groups of interest that can influence in more significant way and that are the Federal Commission of Electricity and the National Commission of Nuclear Safety and Safeguards. The qualifications for each reactor in function of the evaluation approaches were obtained, being the A BWR the best
Core-scale solute transport model selection using Monte Carlo analysis
Malama, Bwalya; James, Scott C
2013-01-01
Model applicability to core-scale solute transport is evaluated using breakthrough data from column experiments conducted with conservative tracers tritium (H-3) and sodium-22, and the retarding solute uranium-232. The three models considered are single-porosity, double-porosity with single-rate mobile-immobile mass-exchange, and the multirate model, which is a deterministic model that admits the statistics of a random mobile-immobile mass-exchange rate coefficient. The experiments were conducted on intact Culebra Dolomite core samples. Previously, data were analyzed using single- and double-porosity models although the Culebra Dolomite is known to possess multiple types and scales of porosity, and to exhibit multirate mobile-immobile-domain mass transfer characteristics at field scale. The data are reanalyzed here and null-space Monte Carlo analysis is used to facilitate objective model selection. Prediction (or residual) bias is adopted as a measure of the model structural error. The analysis clearly shows ...
CAD-Based Monte Carlo Neutron Transport KSTAR Analysis for KSTAR
Seo, Geon Ho; Choi, Sung Hoon; Shim, Hyung Jin
2017-09-01
The Monte Carlo (MC) neutron transport analysis for a complex nuclear system such as fusion facility may require accurate modeling of its complicated geometry. In order to take advantage of modeling capability of the computer aided design (CAD) system for the MC neutronics analysis, the Seoul National University MC code, McCARD, has been augmented with a CAD-based geometry processing module by imbedding the OpenCASCADE CAD kernel. In the developed module, the CAD geometry data are internally converted to the constructive solid geometry model with help of the CAD kernel. An efficient cell-searching algorithm is devised for the void space treatment. The performance of the CAD-based McCARD calculations are tested for the Korea Superconducting Tokamak Advanced Research device by comparing with results of the conventional MC calculations using a text-based geometry input.
Monte Carlo based statistical power analysis for mediation models: methods and software.
Zhang, Zhiyong
2014-12-01
The existing literature on statistical power analysis for mediation models often assumes data normality and is based on a less powerful Sobel test instead of the more powerful bootstrap test. This study proposes to estimate statistical power to detect mediation effects on the basis of the bootstrap method through Monte Carlo simulation. Nonnormal data with excessive skewness and kurtosis are allowed in the proposed method. A free R package called bmem is developed to conduct the power analysis discussed in this study. Four examples, including a simple mediation model, a multiple-mediator model with a latent mediator, a multiple-group mediation model, and a longitudinal mediation model, are provided to illustrate the proposed method.
An approach to model reactor core nodalization for deterministic safety analysis
Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat @ Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd
2016-01-01
Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.
An approach to model reactor core nodalization for deterministic safety analysis
Energy Technology Data Exchange (ETDEWEB)
Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my; Samsudin, Mohd Rafie, E-mail: rafies@tnb.com.my [Nuclear Energy Department, Regulatory Economics & Planning Division, Tenaga Nasional Berhad (Malaysia); Mamat Ibrahim, Mohd Rizal, E-mail: m-rizal@nuclearmalaysia.gov.my [Prototypes & Plant Development Center, Malaysian Nuclear Agency (Malaysia); Roslan, Ridha, E-mail: ridha@aelb.gov.my; Sadri, Abd Aziz [Nuclear Installation Divisions, Atomic Energy Licensing Board (Malaysia); Farid, Mohd Fairus Abd [Reactor Technology Center, Malaysian Nuclear Agency (Malaysia)
2016-01-22
Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH{sub 1.6}, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D{sup ®} computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.
Gulik, Volodymyr; Tkaczyk, Alan Henry
2014-06-01
An optimization study of a subcritical two-zone homogeneous reactor was carried out, taking into consideration geometry, material, and economic parameters. The advantage of a two-zone subcritical system over a single-zone system is demonstrated. The study investigated the optimal volume ratio for the inner and outer zones of the subcritical reactor, in terms of the neutron-physical parameters as well as fuel cost. Optimal geometrical parameters of the system are suggested for different material compositions.
Aygun, Bünyamin; Korkut, Turgay; Karabulut, Abdulhalik
2016-05-01
Despite the possibility of depletion of fossil fuels increasing energy needs the use of radiation tends to increase. Recently the security-focused debate about planned nuclear power plants still continues. The objective of this thesis is to prevent the radiation spread from nuclear reactors into the environment. In order to do this, we produced higher performanced of new shielding materials which are high radiation holders in reactors operation. Some additives used in new shielding materials; some of iron (Fe), rhenium (Re), nickel (Ni), chromium (Cr), boron (B), copper (Cu), tungsten (W), tantalum (Ta), boron carbide (B4C). The results of this experiments indicated that these materials are good shields against gamma and neutrons. The powder metallurgy technique was used to produce new shielding materials. CERN - FLUKA Geant4 Monte Carlo simulation code and WinXCom were used for determination of the percentages of high temperature resistant and high-level fast neutron and gamma shielding materials participated components. Super alloys was produced and then the experimental fast neutron dose equivalent measurements and gamma radiation absorpsion of the new shielding materials were carried out. The produced products to be used safely reactors not only in nuclear medicine, in the treatment room, for the storage of nuclear waste, nuclear research laboratories, against cosmic radiation in space vehicles and has the qualities.
Monte Carlo simulation for slip rate sensitivity analysis in Cimandiri fault area
Energy Technology Data Exchange (ETDEWEB)
Pratama, Cecep, E-mail: great.pratama@gmail.com [Graduate Program of Earth Science, Faculty of Earth Science and Technology, ITB, JalanGanesa no. 10, Bandung 40132 (Indonesia); Meilano, Irwan [Geodesy Research Division, Faculty of Earth Science and Technology, ITB, JalanGanesa no. 10, Bandung 40132 (Indonesia); Nugraha, Andri Dian [Global Geophysical Group, Faculty of Mining and Petroleum Engineering, ITB, JalanGanesa no. 10, Bandung 40132 (Indonesia)
2015-04-24
Slip rate is used to estimate earthquake recurrence relationship which is the most influence for hazard level. We examine slip rate contribution of Peak Ground Acceleration (PGA), in probabilistic seismic hazard maps (10% probability of exceedance in 50 years or 500 years return period). Hazard curve of PGA have been investigated for Sukabumi using a PSHA (Probabilistic Seismic Hazard Analysis). We observe that the most influence in the hazard estimate is crustal fault. Monte Carlo approach has been developed to assess the sensitivity. Then, Monte Carlo simulations properties have been assessed. Uncertainty and coefficient of variation from slip rate for Cimandiri Fault area has been calculated. We observe that seismic hazard estimates is sensitive to fault slip rate with seismic hazard uncertainty result about 0.25 g. For specific site, we found seismic hazard estimate for Sukabumi is between 0.4904 – 0.8465 g with uncertainty between 0.0847 – 0.2389 g and COV between 17.7% – 29.8%.
Monte Carlo simulation for slip rate sensitivity analysis in Cimandiri fault area
Pratama, Cecep; Meilano, Irwan; Nugraha, Andri Dian
2015-04-01
Slip rate is used to estimate earthquake recurrence relationship which is the most influence for hazard level. We examine slip rate contribution of Peak Ground Acceleration (PGA), in probabilistic seismic hazard maps (10% probability of exceedance in 50 years or 500 years return period). Hazard curve of PGA have been investigated for Sukabumi using a PSHA (Probabilistic Seismic Hazard Analysis). We observe that the most influence in the hazard estimate is crustal fault. Monte Carlo approach has been developed to assess the sensitivity. Then, Monte Carlo simulations properties have been assessed. Uncertainty and coefficient of variation from slip rate for Cimandiri Fault area has been calculated. We observe that seismic hazard estimates is sensitive to fault slip rate with seismic hazard uncertainty result about 0.25 g. For specific site, we found seismic hazard estimate for Sukabumi is between 0.4904 - 0.8465 g with uncertainty between 0.0847 - 0.2389 g and COV between 17.7% - 29.8%.
Gil, Victor A; Lecina, Daniel; Grebner, Christoph; Guallar, Victor
2016-10-15
Normal mode methods are becoming a popular alternative to sample the conformational landscape of proteins. In this study, we describe the implementation of an internal coordinate normal mode analysis method and its application in exploring protein flexibility by using the Monte Carlo method PELE. This new method alternates two different stages, a perturbation of the backbone through the application of torsional normal modes, and a resampling of the side chains. We have evaluated the new approach using two test systems, ubiquitin and c-Src kinase, and the differences to the original ANM method are assessed by comparing both results to reference molecular dynamics simulations. The results suggest that the sampled phase space in the internal coordinate approach is closer to the molecular dynamics phase space than the one coming from a Cartesian coordinate anisotropic network model. In addition, the new method shows a great speedup (∼5-7×), making it a good candidate for future normal mode implementations in Monte Carlo methods. Copyright © 2016 Elsevier Ltd. All rights reserved.
Energy Technology Data Exchange (ETDEWEB)
Salome, Jean A.D.; Cardoso, Fabiano; Faria, Rochkhudson B.; Pereira, Claubia, E-mail: jadsalome@yahoo.com.br, E-mail: fabinuclear@yahoo.com.br, E-mail: rockdefaria@yahoo.com.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear
2015-07-01
The IPEN/MB-01 reactor, located in the city of Sao Paulo - Brazil, reached its first criticality on the year of 1988. The reactor is characterized by a low output power of 100 W only, even because its purpose is to produce knowledge about nuclear power plants on a smaller geometric scale without the requirement of an extremely complex cooling system. The use of devices such as this it is very interesting because it achieves the demands of nuclear engineering about the neutronic parameters needed in the design of large nuclear plants through relatively simple and inexpensive methods. In this paper, the computational mathematical code MCNP5 is used to perform the calculation of the thermal neutron flux in the core of the IPEN/MB-01 reactor. To do this is used an experiment from the LEU-COMP-THERM-077 benchmark that represents the standard rectangular configuration of the IPEN/MB-01 reactor. The thermal neutron flux is calculated at some axial planes of different heights and, after that, axial profiles of the thermal neutron flux are done and compared to experimental results issued previously. The experimental values used as reference refer to a cylindrical configuration of the core of the reactor. Finally, the pertinence and relevance of the results are checked. With this work is expected to produce more knowledge about the dynamics of neutron flux in the core of the IPEN/MB-01 reactor. (author)
Goal-oriented sensitivity analysis for lattice kinetic Monte Carlo simulations.
Arampatzis, Georgios; Katsoulakis, Markos A
2014-03-28
In this paper we propose a new class of coupling methods for the sensitivity analysis of high dimensional stochastic systems and in particular for lattice Kinetic Monte Carlo (KMC). Sensitivity analysis for stochastic systems is typically based on approximating continuous derivatives with respect to model parameters by the mean value of samples from a finite difference scheme. Instead of using independent samples the proposed algorithm reduces the variance of the estimator by developing a strongly correlated-"coupled"- stochastic process for both the perturbed and unperturbed stochastic processes, defined in a common state space. The novelty of our construction is that the new coupled process depends on the targeted observables, e.g., coverage, Hamiltonian, spatial correlations, surface roughness, etc., hence we refer to the proposed method as goal-oriented sensitivity analysis. In particular, the rates of the coupled Continuous Time Markov Chain are obtained as solutions to a goal-oriented optimization problem, depending on the observable of interest, by considering the minimization functional of the corresponding variance. We show that this functional can be used as a diagnostic tool for the design and evaluation of different classes of couplings. Furthermore, the resulting KMC sensitivity algorithm has an easy implementation that is based on the Bortz-Kalos-Lebowitz algorithm's philosophy, where events are divided in classes depending on level sets of the observable of interest. Finally, we demonstrate in several examples including adsorption, desorption, and diffusion Kinetic Monte Carlo that for the same confidence interval and observable, the proposed goal-oriented algorithm can be two orders of magnitude faster than existing coupling algorithms for spatial KMC such as the Common Random Number approach. We also provide a complete implementation of the proposed sensitivity analysis algorithms, including various spatial KMC examples, in a supplementary MATLAB
Goal-oriented sensitivity analysis for lattice kinetic Monte Carlo simulations
Arampatzis, Georgios; Katsoulakis, Markos A.
2014-03-01
In this paper we propose a new class of coupling methods for the sensitivity analysis of high dimensional stochastic systems and in particular for lattice Kinetic Monte Carlo (KMC). Sensitivity analysis for stochastic systems is typically based on approximating continuous derivatives with respect to model parameters by the mean value of samples from a finite difference scheme. Instead of using independent samples the proposed algorithm reduces the variance of the estimator by developing a strongly correlated-"coupled"- stochastic process for both the perturbed and unperturbed stochastic processes, defined in a common state space. The novelty of our construction is that the new coupled process depends on the targeted observables, e.g., coverage, Hamiltonian, spatial correlations, surface roughness, etc., hence we refer to the proposed method as goal-oriented sensitivity analysis. In particular, the rates of the coupled Continuous Time Markov Chain are obtained as solutions to a goal-oriented optimization problem, depending on the observable of interest, by considering the minimization functional of the corresponding variance. We show that this functional can be used as a diagnostic tool for the design and evaluation of different classes of couplings. Furthermore, the resulting KMC sensitivity algorithm has an easy implementation that is based on the Bortz-Kalos-Lebowitz algorithm's philosophy, where events are divided in classes depending on level sets of the observable of interest. Finally, we demonstrate in several examples including adsorption, desorption, and diffusion Kinetic Monte Carlo that for the same confidence interval and observable, the proposed goal-oriented algorithm can be two orders of magnitude faster than existing coupling algorithms for spatial KMC such as the Common Random Number approach. We also provide a complete implementation of the proposed sensitivity analysis algorithms, including various spatial KMC examples, in a supplementary MATLAB
Subcritical Noise Analysis Measurements with Fresh and Spent Research Reactor Fuels Elements
Energy Technology Data Exchange (ETDEWEB)
Valentine, T.E.; Mihalczo, J.T.; Kryter, R.C.; Miller, V.C.
1999-02-01
The verification of the subcriticality is of utmost importance for the safe transportation and storage of nuclear reactor fuels. Transportation containers and storage facilities are designed such that nuclear fuels remain in a subcritical state. Such designs often involve excess conservatism because of the lack of relevant experimental data to verify the accuracy of Monte Carlo codes used in nuclear criticality safety analyses. A joint experimental research program between Oak Ridge National Laboratory, Westinghouse Safety Management Solutions, Inc., and the University of Missouri was initiated to obtain measured quantities that could be directly related to the subcriticality of simple arrays of Missouri University Research Reactor (MURR) fuel elements. A series of measurement were performed to assess the reactivity of materials such as BORAL, stainless steel, aluminum, and lead that are typically used in the construction of shipping casks. These materials were positioned between the fuel elements. In addition, a limited number of measurements were performed with configurations of fresh and spent (irradiated) fuel elements to ascertain the reactivity of the spent fuel elements. In these experiments, fresh fuel elements were replaced by spent fuel elements such that the subcritical reactivity change could be measured. The results of these measurements were used by Westinghouse Safety Management Solutions to determine the subcriticality of MURR fuel elements isolated by absorbing materials. The measurements were interpreted using the MCNP-DSP Monte Carlo code to obtain the subcritical neutron multiplication factor k(sub eff), and the bias in K(sub eff) that are used in criticality safety analyses.
A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes
Energy Technology Data Exchange (ETDEWEB)
Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L. [Whiteshell Labs., Pinawa (Canada)] [and others
1997-04-01
A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis.
Common-Cause Failure Analysis for Reactor Protection System Reliability Studies
Energy Technology Data Exchange (ETDEWEB)
Gentillon, C.; Rasmuson, D.; Eide, S.; Wierman, T.
1999-08-01
Analyses were performed of the safety-related performance of the reactor protection system (RPS) at U.S. Westinghouse and General Electric commercial reactors during the period 1984 through 1995. RPS operational data from these reactors were collected from the Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Reports (LER). The common-cause failure (CCF) modeling in the fault trees developed for these studies and the analysis and use of common-cause failure data were sophisticated, state-of-the-art efforts. The overall CCF effort helped to test and expand the limits of the U.S. Nuclear Regulatory Commission's CCF methodology.
Nuclear data and reactor physics activities in Indonesia
Energy Technology Data Exchange (ETDEWEB)
Liem, P.H. [National Atomic Energy Agency, Tangerang (Indonesia). Center for Multipurpose Reactor
1998-03-01
The nuclear data and reactor physics activities in Indonesia, especially, in the National Atomic Energy Agency are presented. In the nuclear data field, the Agency is now taking the position of a user of the main nuclear data libraries such as JENDL and ENDF/B. These nuclear data libraries become the main sources for producing problem dependent cross section sets that are needed by cell calculation codes or transport codes for design, analysis and safety evaluation of research reactors. In the reactor physics field, besides utilising the existing core analysis codes obtained from bilateral and international co-operation, the Agency is putting much effort to self-develop Batan`s codes for reactor physics calculations, in particular, for research reactor and high temperature reactor design, analysis and fuel management. Under the collaboration with JAERI, Monte Carlo criticality calculations on the first criticality of RSG GAS (MPR-30) first core were done using JAERI continuous energy, vectorized Monte Carlo code, MVP, with JENDL-3.1 and JENDL-3.2 nuclear data libraries. The results were then compared with the experiment data collected during the commissioning phase. Monte Carlo calculations with both JENDL-3.1 and -3.2 libraries produced k{sub eff} values with excellent agreement with experiment data, however, systematically, JENDL-3.2 library showed slightly higher k{sub eff} values than JENDL-3.1 library. (author)
Human reliability analysis of the Tehran research reactor using the SPAR-H method
Directory of Open Access Journals (Sweden)
Barati Ramin
2012-01-01
Full Text Available The purpose of this paper is to cover human reliability analysis of the Tehran research reactor using an appropriate method for the representation of human failure probabilities. In the present work, the technique for human error rate prediction and standardized plant analysis risk-human reliability methods have been utilized to quantify different categories of human errors, applied extensively to nuclear power plants. Human reliability analysis is, indeed, an integral and significant part of probabilistic safety analysis studies, without it probabilistic safety analysis would not be a systematic and complete representation of actual plant risks. In addition, possible human errors in research reactors constitute a significant part of the associated risk of such installations and including them in a probabilistic safety analysis for such facilities is a complicated issue. Standardized plant analysis risk-human can be used to address these concerns; it is a well-documented and systematic human reliability analysis system with tables for human performance choices prepared in consultation with experts in the domain. In this method, performance shaping factors are selected via tables, human action dependencies are accounted for, and the method is well designed for the intended use. In this study, in consultations with reactor operators, human errors are identified and adequate performance shaping factors are assigned to produce proper human failure probabilities. Our importance analysis has revealed that human action contained in the possibility of an external object falling on the reactor core are the most significant human errors concerning the Tehran research reactor to be considered in reactor emergency operating procedures and operator training programs aimed at improving reactor safety.
Preliminary accident analysis of Flexblue® underwater reactor
Directory of Open Access Journals (Sweden)
Haratyk Geoffrey
2015-01-01
Full Text Available Flexblue® is a subsea-based, transportable, small modular reactor delivering 160 MWe. Immersion provides the reactor with an infinite heat sink – the ocean – around the metallic hull. The reference design includes a loop-type PWR with two horizontal steam generators. The safety systems are designed to operate passively; safety functions are fulfilled without operator action and external electrical input. Residual heat is removed through four natural circulation loops: two primary heat exchangers immersed in safety tanks cooled by seawater and two emergency condensers immersed in seawater. In case of a primary piping break, a two-train safety injection system is actuated. Each train includes a core makeup tank, an accumulator and a safety tank at low pressure. To assess the capability of these features to remove residual heat, the reactor and its safety systems have been modelled using thermal-hydraulics code ATHLET with conservative assumptions. The results of simulated transients for three typical PWR accidents are presented: a turbine trip with station blackout, a large break loss of coolant accident and a small break loss of coolant accident. The analyses show that the safety criteria are respected and that the reactor quickly reaches a safe shutdown state without operator action and external power.
Limiting photocurrent analysis of a wide channel photoelectrochemical flow reactor
Davis, Jonathan T.; Esposito, Daniel V.
2017-03-01
The development of efficient and scalable photoelectrochemical (PEC) reactors is of great importance for the eventual commercialization of solar fuels technology. In this study, we systematically explore the influence of convective mass transport and light intensity on the performance of a 3D-printed PEC flow cell reactor based on a wide channel, parallel plate geometry. Using this design, the limiting current density generated from the hydrogen evolution reaction at a p-Si metal–insulator–semiconductor (MIS) photocathode was investigated under varied reactant concentration, fluid velocity, and light intensity. Additionally, a simple model is introduced to predict the range of operating conditions (reactant concentration, light intensity, fluid velocity) for which the photocurrent generated in a parallel plate PEC flow cell is limited by light absorption or mass transport. This model can serve as a useful guide for the design and operation of wide-channel PEC flow reactors. The results of this study have important implications for PEC reactors operating in electrolytes with dilute reactant concentrations and/or under high light intensities where high fluid velocities are required in order to avoid operation in the mass transport-limited regime.
Portfolio Assessment on Chemical Reactor Analysis and Process Design Courses
Alha, Katariina
2004-01-01
Assessment determines what students regard as important: if a teacher wants to change students' learning, he/she should change the methods of assessment. This article describes the use of portfolio assessment on five courses dealing with chemical reactor and process design during the years 1999-2001. Although the use of portfolio was a new…
Development of Safety Analysis Technology for Integral Reactor
Energy Technology Data Exchange (ETDEWEB)
Sim, S. K. [Korea Atomic Energy Research Institute, Taejeon (Korea); Seul, K. W.; Kim, W. S.; Kim, W. K.; Yun, Y. G.; Ahn, H. J.; Lee, J. S.; Sin, A. D. [Korea Institute of Nuclear Safety, Taejeon (Korea)
2000-03-01
The Nuclear Desalination Plant(NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in a present study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated based on the design of foreign and domestic integral reactors. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current and advanced reactor designs, and use requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified. They includes the use of proven technology for new safety systems, the systematic classification and selection of design basis accidents, and the safety assurance of desalination-related systems. These efforts to identify and resolve the safety concerns in the design stage will provide the early confidence of SMART safety to designers, and the technical basis to evaluate the safety to reviewers in the future. 8 refs., 20 figs., 4 tabs. (Author)
Computational Fluid Dynamics Analysis of Canadian Supercritical Water Reactor (SCWR)
Movassat, Mohammad; Bailey, Joanne; Yetisir, Metin
2015-11-01
A Computational Fluid Dynamics (CFD) simulation was performed on the proposed design for the Canadian SuperCritical Water Reactor (SCWR). The proposed Canadian SCWR is a 1200 MW(e) supercritical light-water cooled nuclear reactor with pressurized fuel channels. The reactor concept uses an inlet plenum that all fuel channels are attached to and an outlet header nested inside the inlet plenum. The coolant enters the inlet plenum at 350 C and exits the outlet header at 625 C. The operating pressure is approximately 26 MPa. The high pressure and high temperature outlet conditions result in a higher electric conversion efficiency as compared to existing light water reactors. In this work, CFD simulations were performed to model fluid flow and heat transfer in the inlet plenum, outlet header, and various parts of the fuel assembly. The ANSYS Fluent solver was used for simulations. Results showed that mass flow rate distribution in fuel channels varies radially and the inner channels achieve higher outlet temperatures. At the outlet header, zones with rotational flow were formed as the fluid from 336 fuel channels merged. Results also suggested that insulation of the outlet header should be considered to reduce the thermal stresses caused by the large temperature gradients.
Microbial analysis in biogas reactors suffering by foaming incidents
DEFF Research Database (Denmark)
Kougias, Panagiotis; De Francisci, Davide; Treu, Laura
2014-01-01
Foam formation can lead to total failure of digestion process in biogas plants. In the present study, possible correlation between foaming and the presence of specific microorganisms in biogas reactors was elucidated. The microbial ecology of continuous fed digesters overloaded with proteins...
Final safeguards analysis, high temperature lattice test reactor. Revision 1
Energy Technology Data Exchange (ETDEWEB)
Hanthorn, H.E.; Brown, W.W.; Clark, R.G.; Heineman, R.E.; Humes, R.M.
1966-01-01
The PMACS `reactor-normal` signal signifies that important process variables do not exceed their set points, that various interlocks are properly set, that functional tests of the computer operation are satisfactory, and that the reactor flux level and period derived from two additional, independent, and dissimilar channels are within set limits. This safety circuit combines the features of redundancy, dissimilar components, and frequent testing which are required for best reliability. The experimental equipment auxiliary to the reactor includes two oscillator mechanisms, one to move the test cell or the adjoining cell into and out of position, the other to move small specimens in the test cell or adjoining cells. They have cooling chambers for the removal of specimens from the test cell without the necessity of cooling the reactor. A neutron chopper and time-of-flight spectrometer are provided; the neutron detectors, at the end of a 25-meter flight tube, are in an adjoining small building. Test cores may be assembled on a core dolly have a load capacity of 14,000 lb. Two wire traverse mechanisms are provided for measurements of flux distribution.
Temperature Fluctuation Characteristics Analysis for Steam Generator of Fast Reactor
Institute of Scientific and Technical Information of China (English)
ZHU; Li-na; WU; Zhi-guang
2015-01-01
In the case of boiling heat transfer deterioration,temperature fluctuating may accelerate the corrosion of heat transfer tubes and can also lead to thermal stress on the tubes.In this paper,dryout-induced temperature fluctuation for the fast reactor steam generator is investigated.The impacts of water flow rate,sodium inlet temperature and the outlet steam
MCNP-REN a Monte Carlo tool for neutron detector design
Abhold, M E
2002-01-01
The development of neutron detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model fails to accurately predict detector response in common applications. For this reason, the general Monte Carlo code developed at Los Alamos National Laboratory, Monte Carlo N-Particle (MCNP), was modified to simulate the pulse streams that would be generated by a neutron detector and normally analyzed by a shift register. This modified code, MCNP-Random Exponentially Distributed Neutron Source (MCNP-REN), along with the Time Analysis Program, predicts neutron detector response without using the point reactor model, making it unnecessary for the user to decide whether or not the assumptions of the point model are met for their application. MCNP-REN is capable of simulating standard neutron coincidence counting as well as neutron multiplicity counting. Measurements of mixed oxide fresh fuel w...
Directory of Open Access Journals (Sweden)
Tsu-Ming Yeh
2013-10-01
Full Text Available Measurements are required to maintain the consistent quality of all finished and semi-finished products in a production line. Many firms in the automobile and general precision industries apply the TS 16949:2009 Technical Specifications and Measurement System Analysis (MSA manual to establish measurement systems. This work is undertaken to evaluate gauge repeatability and reproducibility (GR&R to verify the measuring ability and quality of the measurement frame, as well as to continuously improve and maintain the verification process. Nevertheless, the implementation of GR&R requires considerable time and manpower, and is likely to affect production adversely. In addition, the evaluation value for GR&R is always different owing to the sum of man-made and machine-made variations. Using a Monte Carlo simulation and the prediction of the repeatability and reproducibility of the measurement system analysis, this study aims to determine the distribution of %GR&R and the related number of distinct categories (ndc. This study uses two case studies of an automobile parts manufacturer and the combination of a Monte Carlo simulation, statistical bases, and the prediction of the repeatability and reproducibility of the measurement system analysis to determine the probability density function, the distribution of %GR&R, and the related number of distinct categories (ndc. The method used in this study could evaluate effectively the possible range of the GR&R of the measurement capability, in order to establish a prediction model for the evaluation of the measurement capacity of a measurement system.
Energy Technology Data Exchange (ETDEWEB)
Azorin V, C.G.; Rivera M, T. [CICATA-IPN, Legaria, Mexico D.F. (Mexico)]. e-mail: claudiaazorin@yahoo.com.mx
2007-07-01
Full text: At the present time many computer programs that simulate the radiation interaction with the matter using the Monte Carlo method. Presently work is carried out the comparative analysis of four of these codes (MCNPX, EGS4, GEANT, PENELOPE) for their later one use in the development of a simple algorithm that simulates the energy deposit when passing through the matter in patients subjected to radiotherapy. The results of the analysis show that the studied simulators model the interaction of almost all type of particles with the matter, although they have their variations among those the energy intervals that manage, the programming language in which are programmed, as well as the platform under which they are executed can be mentioned. (Author)
PREMOR: a point reactor exposure model computer code for survey analysis of power plant performance
Energy Technology Data Exchange (ETDEWEB)
Vondy, D.R.
1979-10-01
The PREMOR computer code was written to exploit a simple, two-group point nuclear reactor power plant model for survey analysis. Up to thirteen actinides, fourteen fission products, and one lumped absorber nuclide density are followed over a reactor history. Successive feed batches are accounted for with provision for from one to twenty batches resident. The effect of exposure of each of the batches to the same neutron flux is determined.
Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors
Galvez, Cristhian
2011-01-01
The Pebble Bed Advanced High Temperature Reactor (PB-AHTR) is a pebble fueled, liquid salt cooled, high temperature nuclear reactor design that can be used for electricity generation or other applications requiring the availability of heat at elevated temperatures. A stage in the design evolution of this plant requires the analysis of the plant during a variety of potential transients to understand the primary and safety cooling system response. This study focuses on the performance of the pa...
Energy Technology Data Exchange (ETDEWEB)
Na, Jang-Hwan; Jeon, Ho-Jun; Hwang, Seok-Won [KHNP Central Research Institute, Daejeon (Korea, Republic of)
2015-10-15
In this paper, we focus on risk insights of Westinghouse typed reactors. We identified that Reactor Coolant Pump (RCP) seal integrity is the most important contributor to Core Damage Frequency (CDF). As we reflected the latest technical report; WCAP-15603(Rev. 1-A), 'WOG2000 RCP Seal Leakage Model for Westinghouse PWRs' instead of the old version, RCP seal integrity became more important to Westinghouse typed reactors. After Fukushima accidents, Korea Hydro and Nuclear Power (KHNP) decided to develop Low Power and Shutdown (LPSD) Probabilistic Safety Assessment (PSA) models and upgrade full power PSA models of all operating Nuclear Power Plants (NPPs). As for upgrading full power PSA models, we have tried to standardize the methodology of CCF (Common Cause Failure) and HRA (Human Reliability Analysis), which are the most influential factors to risk measures of NPPs. Also, we have reviewed and reflected the latest operating experiences, reliability data sources and technical methods to improve the quality of PSA models. KHNP has operating various types of reactors; Optimized Pressurized Reactor (OPR) 1000, CANDU, Framatome and Westinghouse. So, one of the most challengeable missions is to keep the balance of risk contributors of all types of reactors. This paper presents the method of new RCP seal leakage model and the sensitivity analysis results from applying the detailed method to PSA models of Westinghouse typed reference reactors. To perform the sensitivity analysis on LOCCW of the reference Westinghouse typed reactors, we reviewed WOG2000 RCP seal leakage model and developed the detailed event tree of LOCCW considering all scenarios of RCP seal failures. Also, we performed HRA based on the T/H analysis by using the leakage rates for each scenario. We could recognize that HRA was the sensitive contributor to CDF, and the RCP seal failure scenario of 182gpm leakage rate was estimated as the most important scenario.
Energy Technology Data Exchange (ETDEWEB)
Lee, Goany Up; Lee, Hae Cho; Kim, Bong Seok; Kim, Jong Soo; Choi, Pyung Kyu [Dept. of Emergency Preparedness, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2016-12-15
To find out the leak characteristic of research reactor 'HANARO' building in a typhoon condition MELCOR code which normally is used to simulate severe accident behavior in a nuclear power plant was used to simulate the leak rate of air and fission products from reactor hall after the shutdown of the ventilation system of HANARO reactor building. For the simulation, HANARO building was designed by MELCOR code and typhoon condition passed through Daejeon in 2012 was applied. It was found that the leak rate is 0.1%·day{sup -1} of air, 0.004%·day{sup -1} of noble gas and 3.7×10{sup -5}%·day{sup -1} of aerosol during typhoon passing. The air leak rate of 0.1%·day can be converted into 1.36 m{sup 3}·hr{sup -1} , but the design leak rate in HANARO safety analysis report was considered as 600 m3·hr{sup -1} under the condition of 20 m·sec{sup -1} wind speed outside of the building by typhoon. Most of fission products during the maximum hypothesis accident at HANARO reactor will be contained in the reactor hall, so the direct radiation by remained fission products in the reactor hall will be the most important factor in designing emergency preparedness for HANARO reactor.
100-KE REACTOR CORE REMOVAL PROJECT ALTERNATIVE ANALYSIS WORKSHOP REPORT
Energy Technology Data Exchange (ETDEWEB)
HARRINGTON RA
2010-01-15
On December 15-16, 2009, a 100-KE Reactor Core Removal Project Alternative Analysis Workshop was conducted at the Washington State University Consolidated Information Center, Room 214. Colburn Kennedy, Project Director, CH2M HILL Plateau Remediation Company (CHPRC) requested the workshop and Richard Harrington provided facilitation. The purpose of the session was to select the preferred Bio Shield Alternative, for integration with the Thermal Shield and Core Removal and develop the path forward to proceed with project delivery. Prior to this workshop, the S.A. Robotics (SAR) Obstruction Removal Alternatives Analysis (565-DLV-062) report was issued, for use prior to and throughout the session, to all the team members. The multidisciplinary team consisted ofrepresentatives from 100-KE Project Management, Engineering, Radcon, Nuclear Safety, Fire Protection, Crane/Rigging, SAR Project Engineering, the Department of Energy Richland Field Office, Environmental Protection Agency, Washington State Department of Ecology, Defense Nuclear Facility Safety Board, and Deactivation and Decommission subject matter experts from corporate CH2M HILL and Lucas. Appendix D contains the workshop agenda, guidelines and expectations, opening remarks, and attendance roster going into followed throughout the workshop. The team was successful in selecting the preferred alternative and developing an eight-point path forward action plan to proceed with conceptual design. Conventional Demolition was selected as the preferred alternative over two other alternatives: Diamond Wire with Options, and Harmonic Delamination with Conventional Demolition. The teams preferred alternative aligned with the SAR Obstruction Removal Alternative Analysis report conclusion. However, the team identified several Path Forward actions, in Appendix A, which upon completion will solidify and potentially enhance the Conventional Demolition alternative with multiple options and approaches to achieve project delivery
Analysis of Far-Field Radiation from Apertures Using Monte Carlo Integration Technique
Directory of Open Access Journals (Sweden)
Mohammad Mehdi Fakharian
2014-12-01
Full Text Available An integration technique based on the use of Monte Carlo Integration (MCI is proposed for the analysis of the electromagnetic radiation from apertures. The technique that can be applied to the calculation of the aperture antenna radiation patterns is the equivalence principle followed by physical optics, which can then be used to compute far-field antenna radiation patterns. However, this technique is often complex mathematically, because it requires integration over the closed surface. This paper presents an extremely simple formulation to calculate the far-fields from some types of aperture radiators by using MCI technique. The accuracy and effectiveness of this technique are demonstrated in three cases of radiation from the apertures and results are compared with the solutions using FE simulation and Gaussian quadrature rules.
Outlier detection in near-infrared spectroscopic analysis by using Monte Carlo cross-validation
Institute of Scientific and Technical Information of China (English)
LIU ZhiChao; CAI WenSheng; SHAO XueGuang
2008-01-01
An outlier detection method is proposed for near-infrared spectral analysis. The underlying philosophy of the method is that, in random test (Monte Carlo) cross-validation, the probability of outliers pre-senting in good models with smaller prediction residual error sum of squares (PRESS) or in bad mod-els with larger PRESS should be obviously different from normal samples. The method builds a large number of PLS models by using random test cross-validation at first, then the models are sorted by the PRESS, and at last the outliers are recognized according to the accumulative probability of each sam-ple in the sorted models. For validation of the proposed method, four data sets, including three pub-lished data sets and a large data set of tobacco lamina, were investigated. The proposed method was proved to be highly efficient and veracious compared with the conventional leave-one-out (LOO) cross validation method.
Dubecký, Matúš; Jurečka, Petr; Mitas, Lubos; Hobza, Pavel; Otyepka, Michal
2014-01-01
Reliable theoretical predictions of noncovalent interaction energies, which are important e.g. in drug-design and hydrogen-storage applications, belong to longstanding challenges of contemporary quantum chemistry. In this respect, the fixed-node diffusion Monte Carlo (FN-DMC) is a promising alternative to the commonly used ``gold standard'' coupled-cluster CCSD(T)/CBS method for its benchmark accuracy and favourable scaling, in contrast to other correlated wave function approaches. This work is focused on the analysis of protocols and possible tradeoffs for FN-DMC estimations of noncovalent interaction energies and proposes a significantly more efficient yet accurate computational protocol using simplified explicit correlation terms. Its performance is illustrated on a number of weakly bound complexes, including water dimer, benzene/hydrogen, T-shape benzene dimer and stacked adenine-thymine DNA base pair complex. The proposed protocol achieves excellent agreement ($\\sim$0.2 kcal/mol) with respect to the reli...
System Level Numerical Analysis of a Monte Carlo Simulation of the E. Coli Chemotaxis
Siettos, Constantinos I
2010-01-01
Over the past few years it has been demonstrated that "coarse timesteppers" establish a link between traditional numerical analysis and microscopic/ stochastic simulation. The underlying assumption of the associated lift-run-restrict-estimate procedure is that macroscopic models exist and close in terms of a few governing moments of microscopically evolving distributions, but they are unavailable in closed form. This leads to a system identification based computational approach that sidesteps the necessity of deriving explicit closures. Two-level codes are constructed; the outer code performs macroscopic, continuum level numerical tasks, while the inner code estimates -through appropriately initialized bursts of microscopic simulation- the quantities required for continuum numerics. Such quantities include residuals, time derivatives, and the action of coarse slow Jacobians. We demonstrate how these coarse timesteppers can be applied to perform equation-free computations of a kinetic Monte Carlo simulation of...
ANALYSIS OF NEIGHBORHOOD IMPACTS ARISING FROM IMPLEMENTATION OF SUPERMARKETS IN CITY OF SÃO CARLOS
Directory of Open Access Journals (Sweden)
Pedro Silveira Gonçalves Neto
2010-12-01
Full Text Available The study included supermarkets of different sizes (small, medium and large - defined based on the area occupied by the project and volume of activity located in São Carlos (São Paulo state, Brazil to evaluate the influence of the size of the project impacts neighborhood generated by these supermarkets. It was considered the influence of factors like the location of enterprises, size of the building, and areas of influence contribute to the increased population density and change of use of buildings since it was post-deployment analysis. The relationship between the variables of the spatial impacts was made possible by the use of geographic information system. It was noted that the legislation does not have suitable conditions to guide the studies of urban impacts due to the complex integration between the urban and impacting components.
Energy Technology Data Exchange (ETDEWEB)
Davis JE, Eddy MJ, Sutton TM, Altomari TJ
2007-03-01
Solid modeling computer software systems provide for the design of three-dimensional solid models used in the design and analysis of physical components. The current state-of-the-art in solid modeling representation uses a boundary representation format in which geometry and topology are used to form three-dimensional boundaries of the solid. The geometry representation used in these systems is cubic B-spline curves and surfaces--a network of cubic B-spline functions in three-dimensional Cartesian coordinate space. Many Monte Carlo codes, however, use a geometry representation in which geometry units are specified by intersections and unions of half-spaces. This paper describes an algorithm for converting from a boundary representation to a half-space representation.
Marathon: An Open Source Software Library for the Analysis of Markov-Chain Monte Carlo Algorithms.
Rechner, Steffen; Berger, Annabell
2016-01-01
We present the software library marathon, which is designed to support the analysis of sampling algorithms that are based on the Markov-Chain Monte Carlo principle. The main application of this library is the computation of properties of so-called state graphs, which represent the structure of Markov chains. We demonstrate applications and the usefulness of marathon by investigating the quality of several bounding methods on four well-known Markov chains for sampling perfect matchings and bipartite graphs. In a set of experiments, we compute the total mixing time and several of its bounds for a large number of input instances. We find that the upper bound gained by the famous canonical path method is often several magnitudes larger than the total mixing time and deteriorates with growing input size. In contrast, the spectral bound is found to be a precise approximation of the total mixing time.
Outlier detection in near-infrared spectroscopic analysis by using Monte Carlo cross-validation
Institute of Scientific and Technical Information of China (English)
2008-01-01
An outlier detection method is proposed for near-infrared spectral analysis. The underlying philosophy of the method is that,in random test(Monte Carlo) cross-validation,the probability of outliers presenting in good models with smaller prediction residual error sum of squares(PRESS) or in bad models with larger PRESS should be obviously different from normal samples. The method builds a large number of PLS models by using random test cross-validation at first,then the models are sorted by the PRESS,and at last the outliers are recognized according to the accumulative probability of each sample in the sorted models. For validation of the proposed method,four data sets,including three published data sets and a large data set of tobacco lamina,were investigated. The proposed method was proved to be highly efficient and veracious compared with the conventional leave-one-out(LOO) cross validation method.
Ligand-receptor binding kinetics in surface plasmon resonance cells: A Monte Carlo analysis
Carroll, Jacob; Forsten-Williams, Kimberly; Täuber, Uwe C
2016-01-01
Surface plasmon resonance (SPR) chips are widely used to measure association and dissociation rates for the binding kinetics between two species of chemicals, e.g., cell receptors and ligands. It is commonly assumed that ligands are spatially well mixed in the SPR region, and hence a mean-field rate equation description is appropriate. This approximation however ignores the spatial fluctuations as well as temporal correlations induced by multiple local rebinding events, which become prominent for slow diffusion rates and high binding affinities. We report detailed Monte Carlo simulations of ligand binding kinetics in an SPR cell subject to laminar flow. We extract the binding and dissociation rates by means of the techniques frequently employed in experimental analysis that are motivated by the mean-field approximation. We find major discrepancies in a wide parameter regime between the thus extracted rates and the known input simulation values. These results underscore the crucial quantitative importance of s...
Quasi-Monte Carlo Simulation-Based SFEM for Slope Reliability Analysis
Institute of Scientific and Technical Information of China (English)
Yu Yuzhen; Xie Liquan; Zhang Bingyin
2005-01-01
Considering the stochastic spatial variation of geotechnical parameters over the slope, a Stochastic Finite Element Method (SFEM) is established based on the combination of the Shear Strength Reduction (SSR) concept and quasi-Monte Carlo simulation. The shear strength reduction FEM is superior to the slice method based on the limit equilibrium theory in many ways, so it will be more powerful to assess the reliability of global slope stability when combined with probability theory. To illustrate the performance of the proposed method, it is applied to an example of simple slope. The results of simulation show that the proposed method is effective to perform the reliability analysis of global slope stability without presupposing a potential slip surface.
Techno-economic and Monte Carlo probabilistic analysis of microalgae biofuel production system.
Batan, Liaw Y; Graff, Gregory D; Bradley, Thomas H
2016-11-01
This study focuses on the characterization of the technical and economic feasibility of an enclosed photobioreactor microalgae system with annual production of 37.85 million liters (10 million gallons) of biofuel. The analysis characterizes and breaks down the capital investment and operating costs and the production cost of unit of algal diesel. The economic modelling shows total cost of production of algal raw oil and diesel of $3.46 and $3.69 per liter, respectively. Additionally, the effects of co-products' credit and their impact in the economic performance of algal-to-biofuel system are discussed. The Monte Carlo methodology is used to address price and cost projections and to simulate scenarios with probabilities of financial performance and profits of the analyzed model. Different markets for allocation of co-products have shown significant shifts for economic viability of algal biofuel system.
Energy Technology Data Exchange (ETDEWEB)
Nasimi, Elnara; Gabbar, Hossam A., E-mail: Hossam.gabbar@uoit.ca
2014-04-01
Highlights: • Diagnosis of neutron overpower protection (NOP) in CANDU reactors. • Accurate reactor detector modeling. • NOP detectors response analysis. • Statistical methods for quantitative analysis of NOP detector behavior. - Abstract: An accurate fault modeling and troubleshooting methodology is required to aid in making risk-informed decisions related to design and operational activities of current and future generation of CANDU{sup ®} designs. This paper attempts to develop an explanation for the unanticipated detector response and overall behavior phenomena using statistical methods to compliment traditional engineering analysis techniques. Principal component analysis (PCA) methodology is used for pattern recognition using a case study of Bruce B zone-control level oscillations.
3D TRANSIENT COUPLED THERMO-ELASTIC-PLASTIC CONTACT SEALING ANALYSIS OF REACTOR PRESSURE VESSEL
Institute of Scientific and Technical Information of China (English)
Du Xuesong; Li Runfang; Lin Tengjiao
2005-01-01
Sealing analysis of sealing system in reactor pressure vessels is relevant with multiple nonlinear coupled-field effects, so even large-scale commercial finite element software cannot finish the complicated analysis. A fmite element method of 3D transient coupled thermo-elastic-plastic contact sealing analysis for reactor pressure vessels is presented, in which the surface nonlinearity,material nonlinearity, transient heat transfer nonlinearity and multiple coupled effect are taken into account and the sealing equation is coupling solved in iterative procedure. At the same time, a computational analysis program is developed, which is applied in the sealing analysis of experimental reactor pressure vessel, and the numerical results are in good coincidence with the experimental results. This program is also successful in analyzing the practical problem in engineering.
Activation analysis with the Michigan Reactor: application to Michigan problems
Energy Technology Data Exchange (ETDEWEB)
Nass, H. W.
1963-10-15
A review is presented on activation analyses performed by the Michigan reactor. Typical determinations include: sodium deposited on gold leaf for nuclear physics experiments; cobalt and silver in high purity copper used in electron diffraction studies; sodium, manganese and chlorine in experimental reactor grade terphenyl; zirconium in an experimental tungsten - chromium alloy; argon, vanadium, manganese, chlorine, scandium, and iron in spectroscopic grade carbon rod; vanadium and aluminum in synthetically grown sapphires used for x-ray fluorescence studies; hafnium in uranyl nitrate; aluminum in silicon catalyst; vanadium in synthetic cadmium--telluride crystals used in electron spin resonance studies; and sodium and potassium in human tissue from internal organs, Analyses on oil samples and cadmium rod have also been performed. (P.C.H.)
Exposure calculation code module for reactor core analysis: BURNER
Energy Technology Data Exchange (ETDEWEB)
Vondy, D.R.; Cunningham, G.W.
1979-02-01
The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.
Temperature measuring analysis of the nuclear reactor fuel assembly
F., Urban; Ä½., Kučák; Bereznai, J.; Závodný, Z.; Muškát, P.
2014-08-01
Study was based on rapid changes of measured temperature values from the thermocouple in the VVER 440 nuclear reactor fuel assembly. Task was to determine origin of fluctuations of the temperature values by experiments on physical model of the fuel assembly. During an experiment, heated water was circulating in the system and cold water inlet through central tube to record sensitivity of the temperature sensor. Two positions of the sensor was used. First, just above the central tube in the physical model fuel assembly axis and second at the position of the thermocouple in the VVER 440 nuclear reactor fuel assembly. Dependency of the temperature values on time are presented in the diagram form in the paper.
Microbial analysis in biogas reactors suffering by foaming incidents.
Kougias, Panagiotis G; De Francisci, Davide; Treu, Laura; Campanaro, Stefano; Angelidaki, Irini
2014-09-01
Foam formation can lead to total failure of digestion process in biogas plants. In the present study, possible correlation between foaming and the presence of specific microorganisms in biogas reactors was elucidated. The microbial ecology of continuous fed digesters overloaded with proteins, lipids and carbohydrates before and after foaming incidents was characterized using 16S rRNA gene sequencing. Moreover, the microbial diversity between the liquid and foaming layer was assessed. A number of genera that are known to produce biosurfactants, contain mycolic acid in their cell wall, or decrease the surface tension of the media, increased their relative abundance after foam formation. Finally, a microorganism similar to widely known foaming bacteria (Nocardia and Desulfotomaculum) was found to increase its relative abundance in all reactors once foam was observed, regardless of the used substrate. These findings suggest that foaming and specific microorganisms might have direct association which requires to be further investigated.
Exposure calculation code module for reactor core analysis: BURNER
Energy Technology Data Exchange (ETDEWEB)
Vondy, D.R.; Cunningham, G.W.
1979-02-01
The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.
Energy Technology Data Exchange (ETDEWEB)
Uddin, M.N. [Department of Physics, Jahangirnagar University, Dhaka (Bangladesh); Sarker, M.M. [Reactor Physics and Engineering Division, Institute of Nuclear Science and Technology, Atomic Energy Research Establishment, Savar, GPO Box 3787, Dhaka 1000 (Bangladesh); Khan, M.J.H. [Reactor Physics and Engineering Division, Institute of Nuclear Science and Technology, Atomic Energy Research Establishment, Savar, GPO Box 3787, Dhaka 1000 (Bangladesh)], E-mail: jahirulkhan@yahoo.com; Islam, S.M.A. [Department of Physics, Jahangirnagar University, Dhaka (Bangladesh)
2009-10-15
The aim of this paper is to present the validation of evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 through the analysis of the integral parameters of TRX and BAPL benchmark lattices of thermal reactors for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. In this process, the 69-group cross-section library for lattice code WIMS was generated using the basic evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 with the help of nuclear data processing code NJOY99.0. Integral measurements on the thermal reactor lattices TRX-1, TRX-2, BAPL-UO{sub 2}-1, BAPL-UO{sub 2}-2 and BAPL-UO{sub 2}-3 served as standard benchmarks for testing nuclear data files and have also been selected for this analysis. The integral parameters of the said lattices were calculated using the lattice transport code WIMSD-5B based on the generated 69-group cross-section library. The calculated integral parameters were compared to the measured values as well as the results of Monte Carlo Code MCNP. It was found that in most cases, the values of integral parameters show a good agreement with the experiment and MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation of evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 through benchmarking the integral parameters of TRX and BAPL lattices and can also be essential to implement further neutronic analysis of TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh.
HELB Analysis for ESBWR Reactor Building and Main Steam Tunnel
Energy Technology Data Exchange (ETDEWEB)
Noguera Oliva, O.
2011-07-01
The Reactor Building compartments and tbe Main Steam Tunnel are modeled using GOTHIC 7.2a. These models are based on Control Volumes (Rooms/Compartments/Regions), Flow Paths (junctions such as vent path or any opening) and Boundary Conditions (Mass and energy releases and outside conditions). Due to the different break locations, four models are built to analyze the short-term pressurization response. Are shown the cases analyzed, the results obtained and the models used for this purpose.
Application of damage function analysis to reactor coolant circuits
Energy Technology Data Exchange (ETDEWEB)
MacDonald, D.D. [Center for Electrochemical Science and Technology, Pennsylvania State Univ., University Park, PA (United States)
2002-07-01
The application of deterministic models for simulating stress corrosion cracking phenomena in Boiling Water Reactor primary coolant circuits is described. The first generation code, DAMAGE-PREDICTOR, has been used to model the radiolysis of the coolant, to estimate the electrochemical corrosion potential (ECP), and to calculate the crack growth rate (CGR) at fixed state points during reactor operation in about a dozen plants worldwide. This code has been validated in ''double-blind'' comparisons between the calculated and measured hydrogen concentration, oxygen concentration, and ECP in the recirculation system of the Leibstadt BWR in Switzerland, as well as through less formal comparisons with data from other plants. Second generation codes have now been developed, including REMAIN for simulating BWRs with internal coolant pumps and the ALERT series for modeling reactors with external pumps. One of this series, ALERT, yields the integrated damage function (IDF), which is the crack length versus time, on a component-by-component basis for a specified future operating scenario. This code therefore allows one to explore proposed future operating protocols, with the objective of identifying those that are most cost-effective and which minimizes the risk of failure of components in the coolant circuit by stress corrosion cracking. The application of this code is illustrated by exploring the benefits of partial hydrogen water chemistry (HWC) for an actual reactor, in which hydrogen is added to the feedwater over only limited periods during operation. The simulations show that the benefits, in terms of reduction in the IDFs for various components, are sensitive to when HWC was initiated in the plant life and to the length of time over which it is applied. (author)
Thermal analysis of IRT-T reactor fuel elements
Naymushin, Artem Georgievich; Chertkov, Yuri Borisovich; Lebedev, Ivan Igorevich; Anikin, Mikhail Nikolaevich
2015-01-01
The article describes the method and results of thermo-physical calculations of IRT-T reactor core. Heat fluxes, temperatures of cladding, fuel meat and coolant were calculated for height of core, azimuth directions of FA and each fuel elements in FA. Average calculated values of uniformity factor of energy release distribution for height of fuel assemblies were shown in this research. Onset nucleate boiling temperature and ONB-ratio were calculated. Shows that temperature regimes of fuel ele...
Energy Technology Data Exchange (ETDEWEB)
Brown, F.B.; Sutton, T.M.
1996-02-01
This report is composed of the lecture notes from the first half of a 32-hour graduate-level course on Monte Carlo methods offered at KAPL. These notes, prepared by two of the principle developers of KAPL`s RACER Monte Carlo code, cover the fundamental theory, concepts, and practices for Monte Carlo analysis. In particular, a thorough grounding in the basic fundamentals of Monte Carlo methods is presented, including random number generation, random sampling, the Monte Carlo approach to solving transport problems, computational geometry, collision physics, tallies, and eigenvalue calculations. Furthermore, modern computational algorithms for vector and parallel approaches to Monte Carlo calculations are covered in detail, including fundamental parallel and vector concepts, the event-based algorithm, master/slave schemes, parallel scaling laws, and portability issues.
Microbial Community Analysis of Anaerobic Reactors Treating Soft Drink Wastewater
Narihiro, Takashi; Kim, Na-Kyung; Mei, Ran; Nobu, Masaru K.; Liu, Wen-Tso
2015-01-01
The anaerobic packed-bed (AP) and hybrid packed-bed (HP) reactors containing methanogenic microbial consortia were applied to treat synthetic soft drink wastewater, which contains polyethylene glycol (PEG) and fructose as the primary constituents. The AP and HP reactors achieved high COD removal efficiency (>95%) after 80 and 33 days of the operation, respectively, and operated stably over 2 years. 16S rRNA gene pyrotag analyses on a total of 25 biofilm samples generated 98,057 reads, which were clustered into 2,882 operational taxonomic units (OTUs). Both AP and HP communities were predominated by Bacteroidetes, Chloroflexi, Firmicutes, and candidate phylum KSB3 that may degrade organic compound in wastewater treatment processes. Other OTUs related to uncharacterized Geobacter and Spirochaetes clades and candidate phylum GN04 were also detected at high abundance; however, their relationship to wastewater treatment has remained unclear. In particular, KSB3, GN04, Bacteroidetes, and Chloroflexi are consistently associated with the organic loading rate (OLR) increase to 1.5 g COD/L-d. Interestingly, KSB3 and GN04 dramatically decrease in both reactors after further OLR increase to 2.0 g COD/L-d. These results indicate that OLR strongly influences microbial community composition. This suggests that specific uncultivated taxa may take central roles in COD removal from soft drink wastewater depending on OLR. PMID:25748027
Microbial community analysis of anaerobic reactors treating soft drink wastewater.
Directory of Open Access Journals (Sweden)
Takashi Narihiro
Full Text Available The anaerobic packed-bed (AP and hybrid packed-bed (HP reactors containing methanogenic microbial consortia were applied to treat synthetic soft drink wastewater, which contains polyethylene glycol (PEG and fructose as the primary constituents. The AP and HP reactors achieved high COD removal efficiency (>95% after 80 and 33 days of the operation, respectively, and operated stably over 2 years. 16S rRNA gene pyrotag analyses on a total of 25 biofilm samples generated 98,057 reads, which were clustered into 2,882 operational taxonomic units (OTUs. Both AP and HP communities were predominated by Bacteroidetes, Chloroflexi, Firmicutes, and candidate phylum KSB3 that may degrade organic compound in wastewater treatment processes. Other OTUs related to uncharacterized Geobacter and Spirochaetes clades and candidate phylum GN04 were also detected at high abundance; however, their relationship to wastewater treatment has remained unclear. In particular, KSB3, GN04, Bacteroidetes, and Chloroflexi are consistently associated with the organic loading rate (OLR increase to 1.5 g COD/L-d. Interestingly, KSB3 and GN04 dramatically decrease in both reactors after further OLR increase to 2.0 g COD/L-d. These results indicate that OLR strongly influences microbial community composition. This suggests that specific uncultivated taxa may take central roles in COD removal from soft drink wastewater depending on OLR.
Reactor scram experience for shutdown system reliability analysis. [BWR; PWR
Energy Technology Data Exchange (ETDEWEB)
Edison, G.E.; Pugliese, S.L.; Sacramo, R.F.
1976-06-01
Scram experience in a number of operating light water reactors has been reviewed. The date and reactor power of each scram was compiled from monthly operating reports and personal communications with the operating plant personnel. The average scram frequency from ''significant'' power (defined as P/sub trip//P/sub max greater than/ approximately 20 percent) was determined as a function of operating life. This relationship was then used to estimate the total number of reactor trips from above approximately 20 percent of full power expected to occur during the life of a nuclear power plant. The shape of the scram frequency vs. operating life curve resembles a typical reliability bathtub curve (failure rate vs. time), but without a rising ''wearout'' phase due to the lack of operating data near the end of plant design life. For this case the failures are represented by ''bugs'' in the plant system design, construction, and operation which lead to scram. The number of scrams would appear to level out at an average of around three per year; the standard deviations from the mean value indicate an uncertainty of about 50 percent. The total number of scrams from significant power that could be expected in a plant designed for a 40-year life would be about 130 if no wearout phase develops near the end of life.
Directory of Open Access Journals (Sweden)
Hammam Oktajianto
2014-12-01
Full Text Available Gas-cooled nuclear reactor is a Generation IV reactor which has been receiving significant attention due to many desired characteristics such as inherent safety, modularity, relatively low cost, short construction period, and easy financing. High temperature reactor (HTR pebble-bed as one of type of gas-cooled reactor concept is getting attention. In HTR pebble-bed design, radius and enrichment of the fuel kernel are the key parameter that can be chosen freely to determine the desired value of criticality. This paper models HTR pebble-bed 10 MW and determines an effective of enrichment and radius of the fuel (Kernel to get criticality value of reactor. The TRISO particle coated fuel particle which was modelled explicitly and distributed in the fuelled region of the fuel pebbles using a Simple-Cubic (SC lattice. The pebble-bed balls and moderator balls distributed in the core zone using a Body-Centred Cubic lattice with assumption of a fresh fuel by the fuel enrichment was 7-17% at 1% range and the size of the fuel radius was 175-300 µm at 25 µm ranges. The geometrical model of the full reactor is obtained by using lattice and universe facilities provided by MCNP4C. The details of model are discussed with necessary simplifications. Criticality calculations were conducted by Monte Carlo transport code MCNP4C and continuous energy nuclear data library ENDF/B-VI. From calculation results can be concluded that an effective of enrichment and radius of fuel (Kernel to achieve a critical condition was the enrichment of 15-17% at a radius of 200 µm, the enrichment of 13-17% at a radius of 225 µm, the enrichments of 12-15% at radius of 250 µm, the enrichments of 11-14% at a radius of 275 µm and the enrichment of 10-13% at a radius of 300 µm, so that the effective of enrichments and radii of fuel (Kernel can be considered in the HTR 10 MW. Keywords—MCNP4C, HTR, enrichment, radius, criticality
Analysis of Pressurized Water Reactor Primary Coolant Leak Events Caused by Thermal Fatigue
Energy Technology Data Exchange (ETDEWEB)
Atwood, Corwin Lee; Shah, Vikram Naginbhai; Galyean, William Jospeh
1999-09-01
We present statistical analyses of pressurized water reactor (PWR) primary coolant leak events caused by thermal fatigue, and discuss their safety significance. Our worldwide data contain 13 leak events (through-wall cracking) in 3509 reactor-years, all in stainless steel piping with diameter less than 25 cm. Several types of data analysis show that the frequency of leak events (events per reactor-year) is increasing with plant age, and the increase is statistically significant. When an exponential trend model is assumed, the leak frequency is estimated to double every 8 years of reactor age, although this result should not be extrapolated to plants much older than 25 years. Difficulties in arresting this increase include lack of quantitative understanding of the phenomena causing thermal fatigue, lack of understanding of crack growth, and difficulty in detecting existing cracks.
Partial Safety Analysis for a Reduced Uranium Enrichment Core for the High Flux Isotope Reactor
Energy Technology Data Exchange (ETDEWEB)
Primm, Trent [ORNL; Gehin, Jess C [ORNL
2009-04-01
A computational model of the reactor core of the High Flux Isotope Rector (HFIR) was developed in order to analyze non-destructive accidents caused by transients during reactor operation. The reactor model was built for the latest version of the nuclear analysis software package called Program for the Analysis of Reactor Transients (PARET). Analyses performed with the model constructed were compared with previous data obtained with other tools in order to benchmark the code. Finally, the model was used to analyze the behavior of the reactor under transients using a different nuclear fuel with lower enrichment of uranium (LEU) than the fuel currently used, which has a high enrichment of uranium (HEU). The study shows that the presence of fertile isotopes in LEU fuel, which increases the neutron resonance absorption, reduces the impact of transients on the fuel and enhances the negative reactivity feedback, thus, within the limitations of this study, making LEU fuel appear to be a safe alternative fuel for the reactor core.
Energy Technology Data Exchange (ETDEWEB)
Suhwan, JI; Shirahama, H.; Koshizuka, S.; Oka, Y. [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.
2001-07-01
The purpose of this study is to evaluate the thermal-hydraulic and the thermal-nuclear coupled stabilities of a supercritical pressure light water-cooled reactor. A stability analysis code at supercritical pressure is developed. Using this code, stabilities of full and partial-power reactor operating at supercritical pressure are investigated by the frequency-domain analysis. Two types of SCRs are analyzed; a supercritical light water reactor (SCLWR) and a supercritical water-cooled fast reactor (SCFR). The same stability criteria as Boiling Water Reactor are applied. The thermal-hydraulic stability of SCLWR and SCFR satisfies the criteria with a reasonable orifice loss coefficient. The decay ratio of the thermal-nuclear coupled stability in SCFR is almost zero because of a small coolant density coefficient of the fast reactor. The evaluated decay ratio of the thermal-nuclear coupled stability is 3,41 {approx} 10{sup -V} at 100% power in SCFR and 0,028 at 100% power in SCLWR. The sensitivity is investigated. It is found that the thermal-hydraulic stability is sensitive to the mass flow rate strongly and the thermal-nuclear coupled stability to the coolant density coefficient. The bottom power peak distribution makes the thermal-nuclear stability worse and the thermal-nuclear stability better. (author)
Rate-Only analysis with reactor-off data in the Double Chooz experiment
Novella, P
2013-01-01
Among ongoing reactor-based experiments, Double Chooz is unique in obtaining data when the reactor cores are brought down for maintenance. These reactor-off data allow for a clean measurement of the backgrounds of the experiment, thus being of uppermost importance for the theta13 oscillation analysis. While the oscillation results published by the collaboration in 2011 and 2012 rely on background models derived from reactor-on data, in this talk we present an independent study based on the handle provided by 7.53 days of reactor-off data. A global fit to both theta13 and the total background is performed by analyzing the observed neutrino rate as a function of the non-oscillated expected rate for different reactor power conditions. The result presented in this talk is fully consistent with the one already published by Double Chooz. As they both yield almost the same precision, this work stands as a prove of the reliability of the background estimates and the oscillation analysis of the experiment.
An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components
Energy Technology Data Exchange (ETDEWEB)
Holcomb, David Eugene [ORNL; Cetiner, Sacit M [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL
2009-11-01
This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.
Power-feedwater temperature operating domain for Sbwr applying Monte Carlo simulation
Energy Technology Data Exchange (ETDEWEB)
Aguilar M, L. A.; Quezada G, S.; Espinosa M, E. G.; Vazquez R, A.; Varela H, J. R.; Cazares R, R. I.; Espinosa P, G., E-mail: sequega@gmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)
2014-10-15
In this work the analyses of the feedwater temperature effects on reactor power in a simplified boiling water reactor (Sbwr) applying a methodology based on Monte Carlo simulation is presented. The Monte Carlo methodology was applied systematically to establish operating domain, due that the Sbwr are not yet in operation, the analysis of the nuclear and thermal-hydraulic processes must rely on numerical modeling, with the purpose of developing or confirming the design basis and qualifying the existing or new computer codes to enable reliable analyses. The results show that the reactor power is inversely proportional to the temperature of the feedwater, reactor power changes at 8% when the feed water temperature changes in 8%. (Author)
Empirical Markov Chain Monte Carlo Bayesian analysis of fMRI data.
de Pasquale, F; Del Gratta, C; Romani, G L
2008-08-01
In this work an Empirical Markov Chain Monte Carlo Bayesian approach to analyse fMRI data is proposed. The Bayesian framework is appealing since complex models can be adopted in the analysis both for the image and noise model. Here, the noise autocorrelation is taken into account by adopting an AutoRegressive model of order one and a versatile non-linear model is assumed for the task-related activation. Model parameters include the noise variance and autocorrelation, activation amplitudes and the hemodynamic response function parameters. These are estimated at each voxel from samples of the Posterior Distribution. Prior information is included by means of a 4D spatio-temporal model for the interaction between neighbouring voxels in space and time. The results show that this model can provide smooth estimates from low SNR data while important spatial structures in the data can be preserved. A simulation study is presented in which the accuracy and bias of the estimates are addressed. Furthermore, some results on convergence diagnostic of the adopted algorithm are presented. To validate the proposed approach a comparison of the results with those from a standard GLM analysis, spatial filtering techniques and a Variational Bayes approach is provided. This comparison shows that our approach outperforms the classical analysis and is consistent with other Bayesian techniques. This is investigated further by means of the Bayes Factors and the analysis of the residuals. The proposed approach applied to Blocked Design and Event Related datasets produced reliable maps of activation.
Improving Markov Chain Monte Carlo algorithms in LISA Pathfinder Data Analysis
Karnesis, N.; Nofrarias, M.; Sopuerta, C. F.; Lobo, A.
2012-06-01
The LISA Pathfinder mission (LPF) aims to test key technologies for the future LISA mission. The LISA Technology Package (LTP) on-board LPF will consist of an exhaustive suite of experiments and its outcome will be crucial for the future detection of gravitational waves. In order to achieve maximum sensitivity, we need to have an understanding of every instrument on-board and parametrize the properties of the underlying noise models. The Data Analysis team has developed algorithms for parameter estimation of the system. A very promising one implemented for LISA Pathfinder data analysis is the Markov Chain Monte Carlo. A series of experiments are going to take place during flight operations and each experiment is going to provide us with essential information for the next in the sequence. Therefore, it is a priority to optimize and improve our tools available for data analysis during the mission. Using a Bayesian framework analysis allows us to apply prior knowledge for each experiment, which means that we can efficiently use our prior estimates for the parameters, making the method more accurate and significantly faster. This, together with other algorithm improvements, will lead us to our main goal, which is no other than creating a robust and reliable tool for parameter estimation during the LPF mission.
Vyawahare, Vishwesh A.; Nataraj, P. S. V.
2013-07-01
In this paper, we report the development and analysis of some novel versions and approximations of the fractional-order (FO) point reactor kinetics model for a nuclear reactor with slab geometry. A systematic development of the FO Inhour equation, Inverse FO point reactor kinetics model, and fractional-order versions of the constant delayed neutron rate approximation model and prompt jump approximation model is presented for the first time (for both one delayed group and six delayed groups). These models evolve from the FO point reactor kinetics model, which has been derived from the FO Neutron Telegraph Equation for the neutron transport considering the subdiffusive neutron transport. Various observations and the analysis results are reported and the corresponding justifications are addressed using the subdiffusive framework for the neutron transport. The FO Inhour equation is found out to be a pseudo-polynomial with its degree depending on the order of the fractional derivative in the FO model. The inverse FO point reactor kinetics model is derived and used to find the reactivity variation required to achieve exponential and sinusoidal power variation in the core. The situation of sudden insertion of negative reactivity is analyzed using the FO constant delayed neutron rate approximation. Use of FO model for representing the prompt jump in reactor power is advocated on the basis of subdiffusion. Comparison with the respective integer-order models is carried out for the practical data. Also, it has been shown analytically that integer-order models are a special case of FO models when the order of time-derivative is one. Development of these FO models plays a crucial role in reactor theory and operation as it is the first step towards achieving the FO control-oriented model for a nuclear reactor. The results presented here form an important step in the efforts to establish a step-by-step and systematic theory for the FO modeling of a nuclear reactor.
Development of guidelines for inelastic analysis in design of fast reactor components
Energy Technology Data Exchange (ETDEWEB)
Nakamura, Kyotada [Japan Atomic Energy Agency (JAEA), 4002, Narita, O-arai, Ibaraki 311-1393 (Japan); Kasahara, Naoto [Japan Atomic Energy Agency (JAEA), 4002, Narita, O-arai, Ibaraki 311-1393 (Japan)], E-mail: kasahara.naoto@jaea.go.jp; Morishita, Masaki [Japan Atomic Energy Agency (JAEA), 4002, Narita, O-arai, Ibaraki 311-1393 (Japan); Shibamoto, Hiroshi; Inoue, Kazuhiko [The Japan Atomic Power Company, Dispatched to JAEA, 4002, Narita, O-arai, Ibaraki 311-1393 (Japan); Nakayama, Yasunari [Kawasaki Heavy Industries, Ltd., 1-1, Kawasaki, Akashi, Hyogo 673-8666 (Japan)
2008-02-15
The interim guidelines for the application of inelastic analysis to design of fast reactor components were developed. These guidelines are referred from 'Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)'. The basic policies of the guidelines are more rational predictions compared with elastic analysis approach and a guarantee of conservative results for design conditions. The guidelines recommend two kinds of constitutive equations to estimate strains conservatively. They also provide the methods for modeling load histories and estimating fatigue and creep damage based on the results of inelastic analysis. The guidelines were applied to typical design examples and their results were summarized as exemplars to support users.
Leclaire, N.; Cochet, B.; Le Dauphin, F. X.; Haeck, W.; Jacquet, O.
2014-06-01
The present paper aims at providing experimental validation for the use of the MORET 5 code for advanced concepts of reactor involving thorium and heavy water. It therefore constitutes an opportunity to test and improve the thermal-scattering data of heavy water and also to test the recent implementation of probability tables in the MORET 5 code.
Yaghini, N.; Iedema, P.D.
2015-01-01
Modeling of the mol. wt. distribution (MWD) under circumstances of low-d. polyethylene (ldPE) has been carried out for a tubular reactor under realistic non-isothermal conditions and for a series of CSTR's (Yaghini and Iedema, in press). The model allows for the existence of multiradicals and the
Modern design and safety analysis of the University of Florida Training Reactor
Energy Technology Data Exchange (ETDEWEB)
Jordan, K.A., E-mail: kjordan@ufl.edu [University of Florida, 106 UFTR Bldg., PO Box 116400, Gainesville, FL 32611-6400 (United States); Springfels, D., E-mail: dspringfels@ufl.edu [University of Florida, 106 UFTR Bldg., PO Box 116400, Gainesville, FL 32611-6400 (United States); Schubring, D., E-mail: dlschubring@ufl.edu [University of Florida, 202 Nuclear Science Building, PO Box 118300, Gainesville, FL 32611-8300 (United States)
2015-05-15
Highlights: • A new safety analysis of the University of Florida Training Reactor is presented. • This analysis uses modern codes and replaces the NRC approved analysis from 1982. • Reduction in engineering margin confirms that the UFTR is a negligible risk reactor. • Safety systems are not required to ensure that safety limits are not breached. • Negligible risk reactors are ideal for testing digital I&C equipment. - Abstract: A comprehensive series of neutronics and thermal hydraulics analyses were conducted to demonstrate the University of Florida Training Reactor (UFTR), an ARGONAUT type research reactor, as a negligible risk reactor that does not require safety-related systems or components to prevent breach of a safety limit. These analyses show that there is no credible UFTR accident that would result in major fuel damage or risk to public health and safety. The analysis was based on two limiting scenarios, whose extremity bound all other accidents of consequence: (1) the large step insertion of positive reactivity and (2) the release of fission products due to mechanical damage to a spent fuel plate. The maximum step insertion of positive reactivity was modeled using PARET/ANL software and shows a maximum peak fuel temperature of 283.2 °C, which is significantly below the failure limit of 530 °C. The exposure to the staff and general public was calculated for the worst-case fission product release scenario using the ORIGEN-S and COMPLY codes and was shown to be 6.5% of the annual limit. Impacts on reactor operations and an Instrumentation & Control System (I&C) upgrade are discussed.
Core-scale solute transport model selection using Monte Carlo analysis
Malama, Bwalya; Kuhlman, Kristopher L.; James, Scott C.
2013-06-01
Model applicability to core-scale solute transport is evaluated using breakthrough data from column experiments conducted with conservative tracers tritium (3H) and sodium-22 (22Na ), and the retarding solute uranium-232 (232U). The three models considered are single-porosity, double-porosity with single-rate mobile-immobile mass-exchange, and the multirate model, which is a deterministic model that admits the statistics of a random mobile-immobile mass-exchange rate coefficient. The experiments were conducted on intact Culebra Dolomite core samples. Previously, data were analyzed using single-porosity and double-porosity models although the Culebra Dolomite is known to possess multiple types and scales of porosity, and to exhibit multirate mobile-immobile-domain mass transfer characteristics at field scale. The data are reanalyzed here and null-space Monte Carlo analysis is used to facilitate objective model selection. Prediction (or residual) bias is adopted as a measure of the model structural error. The analysis clearly shows single-porosity and double-porosity models are structurally deficient, yielding late-time residual bias that grows with time. On the other hand, the multirate model yields unbiased predictions consistent with the late-time -5/2 slope diagnostic of multirate mass transfer. The analysis indicates the multirate model is better suited to describing core-scale solute breakthrough in the Culebra Dolomite than the other two models.
Risk analysis of gravity dam instability using credibility theory Monte Carlo simulation model.
Xin, Cao; Chongshi, Gu
2016-01-01
Risk analysis of gravity dam stability involves complicated uncertainty in many design parameters and measured data. Stability failure risk ratio described jointly by probability and possibility has deficiency in characterization of influence of fuzzy factors and representation of the likelihood of risk occurrence in practical engineering. In this article, credibility theory is applied into stability failure risk analysis of gravity dam. Stability of gravity dam is viewed as a hybrid event considering both fuzziness and randomness of failure criterion, design parameters and measured data. Credibility distribution function is conducted as a novel way to represent uncertainty of influence factors of gravity dam stability. And combining with Monte Carlo simulation, corresponding calculation method and procedure are proposed. Based on a dam section, a detailed application of the modeling approach on risk calculation of both dam foundation and double sliding surfaces is provided. The results show that, the present method is feasible to be applied on analysis of stability failure risk for gravity dams. The risk assessment obtained can reflect influence of both sorts of uncertainty, and is suitable as an index value.
Institute of Scientific and Technical Information of China (English)
ZHANG Jun; GUO Fan
2015-01-01
Tooth modification technique is widely used in gear industry to improve the meshing performance of gearings. However, few of the present studies on tooth modification considers the influence of inevitable random errors on gear modification effects. In order to investigate the uncertainties of tooth modification amount variations on system’s dynamic behaviors of a helical planetary gears, an analytical dynamic model including tooth modification parameters is proposed to carry out a deterministic analysis on the dynamics of a helical planetary gear. The dynamic meshing forces as well as the dynamic transmission errors of the sun-planet 1 gear pair with and without tooth modifications are computed and compared to show the effectiveness of tooth modifications on gear dynamics enhancement. By using response surface method, a fitted regression model for the dynamic transmission error(DTE) fluctuations is established to quantify the relationship between modification amounts and DTE fluctuations. By shifting the inevitable random errors arousing from manufacturing and installing process to tooth modification amount variations, a statistical tooth modification model is developed and a methodology combining Monte Carlo simulation and response surface method is presented for uncertainty analysis of tooth modifications. The uncertainly analysis reveals that the system’s dynamic behaviors do not obey the normal distribution rule even though the design variables are normally distributed. In addition, a deterministic modification amount will not definitely achieve an optimal result for both static and dynamic transmission error fluctuation reduction simultaneously.
Core Flow Distribution from Coupled Supercritical Water Reactor Analysis
Directory of Open Access Journals (Sweden)
Po Hu
2014-01-01
Full Text Available This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady state conditions. The possibility of moderator flow reversal is found in some hot moderator channels. Different moderator flow orifice strategies, both uniform across the core and nonuniform based on the power distribution, are explored with the goal of preventing the reversal.
Steady thermal hydraulic analysis for a molten salt reactor
Institute of Scientific and Technical Information of China (English)
ZHANG Dalin; QIU Suizheng; LIU Changliang; SU Guanghui
2008-01-01
The Molten Salt Reactor (MSR) can meet the demand of transmutation and breeding. In this study, theoretical calculation of steady thermal hydraulic characteristics of a graphite-moderated channel type MSR is conducted. The DRAGON code is adopted to calculate the axial and radial power factor firstly. The flow and heat transfer model in the fuel salt and graphite are developed on basis of the fundamental mass, momentum and energy equations. The results show the detailed flow distribution in the core, and the temperature profiles of the fuel salt, inner and outer wall in the nine typical elements along the axial flow direction are also obtained.
Energy Technology Data Exchange (ETDEWEB)
Salinger, A.G.; Shadid, J.N.; Hutchinson, S.A. [and others
1998-01-01
A numerical analysis of the deposition of gallium from trimethylgallium (TMG) and arsine in a horizontal CVD reactor with tilted susceptor and a three inch diameter rotating substrate is performed. The three-dimensional model includes complete coupling between fluid mechanics, heat transfer, and species transport, and is solved using an unstructured finite element discretization on a massively parallel computer. The effects of three operating parameters (the disk rotation rate, inlet TMG fraction, and inlet velocity) and two design parameters (the tilt angle of the reactor base and the reactor width) on the growth rate and uniformity are presented. The nonlinear dependence of the growth rate uniformity on the key operating parameters is discussed in detail. Efficient and robust algorithms for massively parallel reacting flow simulations, as incorporated into our analysis code MPSalsa, make detailed analysis of this complicated system feasible.
Application of neutron activation analysis system in Xi'an pulsed reactor
Zhang Wen Shou; Yu Qi
2002-01-01
Neutron Activation Analysis System in Xi'an Pulsed Reactor is consist of rabbit fast radiation system and experiment measurement system. The functions of neutron activation analysis are introduced. Based on the radiation system. A set of automatic data handling and experiment simulating system are built. The reliability of data handling and experiment simulating system had been verified by experiment
Seismic analysis of the APR1400 nuclear reactor system using a verified beam element model
Energy Technology Data Exchange (ETDEWEB)
Park, Jong-beom [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Park, No-Cheol, E-mail: pnch@yonsei.ac.kr [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Lee, Sang-Jeong; Park, Young-Pil [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Choi, Youngin [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 34142 (Korea, Republic of)
2017-03-15
Highlights: • A simplified beam element model is constructed based on the real dynamic characteristics of the APR1400. • Time history analysis is performed to calculate the seismic responses of the structures. • Large deformations can be observed at the in-phase mode of reactor vessel and core support barrel. - Abstract: Structural integrity is the first priority in the design of nuclear reactor internal structures. In particular, nuclear reactor internals should be designed to endure external forces, such as those due to earthquakes. Many researchers have performed finite element analyses to meet these design requirements. Generally, a seismic analysis model should reflect the dynamic characteristics of the target system. However, seismic analysis based on the finite element method requires long computation times as well as huge storage space. In this research, a beam element model was developed and confirmed based on the real dynamic characteristics of an advanced pressurized water nuclear reactor 1400 (APR1400) system. That verification process enhances the accuracy of the finite element analysis using the beam elements, remarkably. Also, the beam element model reduces seismic analysis costs. Therefore, the beam element model was used to perform the seismic analysis. Then, the safety of the APR1400 was assessed based on a seismic analysis of the time history responses of its structures. Thus, efficient, accurate seismic analysis was demonstrated using the proposed beam element model.
Analysis of Nigeria research reactor-1 thermal power calibration methods
Energy Technology Data Exchange (ETDEWEB)
Agbo, Sunday Arome; Ahmed, Yusuf Aminu; Ewa, Ita Okon; Jibrin, Yahaya [Ahmadu Bello University, Zaria (Nigeria)
2016-06-15
This paper analyzes the accuracy of the methods used in calibrating the thermal power of Nigeria Research Reactor-1 (NIRR-1), a low-power miniature neutron source reactor located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria, Nigeria. The calibration was performed at three different power levels: low power (3.6 kW), half power (15 kW), and full power (30 kW). Two methods were used in the calibration, namely, slope and heat balance methods. The thermal power obtained by the heat balance method at low power, half power, and full power was 3.7 ± 0.2 kW, 15.2 ± 1.2 kW, and 30.7 ± 2.5 kW, respectively. The thermal power obtained by the slope method at half power and full power was 15.8 ± 0.7 kW and 30.2 ± 1.5 kW, respectively. It was observed that the slope method is more accurate with deviations of 4% and 5% for calibrations at half and full power, respectively, although the linear fit (slope method) on average temperature-rising rates during the thermal power calibration procedure at low power (3.6 kW) is not fitting. As such, the slope method of power calibration is not suitable at lower power for NIRR-1.
Safety Analysis of Pb-208 Cooled 800 MWt Modified CANDLE Reactors
Su'ud, Zaki; Widiawati, Nina; Sekimoto, H.; Artoto, A.
2017-01-01
Safely analysis of 800MWt Pb-208 cooled fast reactors with natural Uranium as fuel cycle input employing axial-radial combined Modiified CANDLE burnup scheme has been performed. The analysis of unprotected loss of flow(ULOF) and unprotected rod run-out transient overpower (UTOP) are discussed. Some simulations for 800 MWt Pb-208 cooled fast reactors has been performed and the results show that the reactor can anticipate complete pumping failure inherently by reducing power through reactivity feedback and remove the rest of heat through natural circulations. Compared to the Pb-nat cooled long life Fast Reactors, Pb-208 cooled reactors have smaller Doppler but higher coolant density reactivity coefficient. In the UTOP accident case the analysis has been performed against external reactivity up to 0.003dk/k. And for ULOHS case it is assumed that the secondary cooling system has broken. During all accident the cladding temperature is the most critical. Especially for the case of UTOP accident. In addition the steam generator design has also consider excess power which may reach 50% extra during severe UTOP case..
Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event
Energy Technology Data Exchange (ETDEWEB)
Bucknor, Matthew D.; Grabaskas, David; Brunett, Acacia J.; Grelle, Austin
2016-01-01
Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended due to deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Centering on an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive reactor cavity cooling system following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. While this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability for the reactor cavity cooling system (and the reactor system in general) to the postulated transient event.
Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event
Directory of Open Access Journals (Sweden)
Matthew Bucknor
2017-03-01
Full Text Available Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general for the postulated transient event.
Advanced reactor passive system reliability demonstration analysis for an external event
Energy Technology Data Exchange (ETDEWEB)
Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.; Grelle, Austin [Argonne National Laboratory, Argonne (United States)
2017-03-15
Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.
Institute of Scientific and Technical Information of China (English)
甘佺; 吴斌; 宋婧; 程梦云; 胡丽琴
2016-01-01
Background: The Monte Carlo (MC) method is widely used in fission reactor design, because of its strongergeometry adaptability and the precise calculation result.The high-fidelity full core simulation demands for detailed fission reactor models, which is hard to build by manual and the conventional Computer Aided Design (CAD)-based modeling method.Purpose:In order tosupport the rapid design offission reactor core with MC method, and generate the detailed MC calculation models,a parameter-based and visual MC modeling method was developed in this study.Method: The method can create the detailed CAD models and convert them into MC models in a high-efficiency way. Meanwhile, the huge amounts of models in fission reactor core can be managed by different segments for supporting smooth interactions.Result: Furthermore, the method was validated by the test of creating the Accelerator Driven Sub-critical System (ADS) reactor models, and the results were agree very well by comparing the reference models.Conclusion:Depending on the test, the detailed fission models were created more conveniently than conventional method andthe numerical calculation results proved the accuracy of the new method.%蒙特卡罗程序已经广泛应用在裂变反应堆设计和验证过程中，快速获得高效的计算模型可以有效缩短反应堆的设计周期。本研究提出并实现了一种裂变堆芯快速蒙特卡罗建模的方法，该方法基于参数可视化和层次化两种建模思想快速构建出精细裂变堆芯计算机辅助设计(Computer Aided Design, CAD)模型且将其快速转换成蒙特卡罗计算模型，同时采用一种新的堆芯分段管理方法实现了大规模裂变堆模型流畅交互。基于此方法快速构建了加速器驱动次临界反应堆(Accelerator Driven Sub-critical System, ADS)的精细堆芯模型，通过与蒙特卡罗程序计算的参考结果进行对比，证明了此建模方法的高效性和可靠性。
Ma, Jianzhong; Amos, Christopher I; Warwick Daw, E
2007-09-01
Although extended pedigrees are often sampled through probands with extreme levels of a quantitative trait, Markov chain Monte Carlo (MCMC) methods for segregation and linkage analysis have not been able to perform ascertainment corrections. Further, the extent to which ascertainment of pedigrees leads to biases in the estimation of segregation and linkage parameters has not been previously studied for MCMC procedures. In this paper, we studied these issues with a Bayesian MCMC approach for joint segregation and linkage analysis, as implemented in the package Loki. We first simulated pedigrees ascertained through individuals with extreme values of a quantitative trait in spirit of the sequential sampling theory of Cannings and Thompson [Cannings and Thompson [1977] Clin. Genet. 12:208-212]. Using our simulated data, we detected no bias in estimates of the trait locus location. However, in addition to allele frequencies, when the ascertainment threshold was higher than or close to the true value of the highest genotypic mean, bias was also found in the estimation of this parameter. When there were multiple trait loci, this bias destroyed the additivity of the effects of the trait loci, and caused biases in the estimation all genotypic means when a purely additive model was used for analyzing the data. To account for pedigree ascertainment with sequential sampling, we developed a Bayesian ascertainment approach and implemented Metropolis-Hastings updates in the MCMC samplers used in Loki. Ascertainment correction greatly reduced biases in parameter estimates. Our method is designed for multiple, but a fixed number of trait loci.
A comparison of Bayesian and Monte Carlo sensitivity analysis for unmeasured confounding.
McCandless, Lawrence C; Gustafson, Paul
2017-04-06
Bias from unmeasured confounding is a persistent concern in observational studies, and sensitivity analysis has been proposed as a solution. In the recent years, probabilistic sensitivity analysis using either Monte Carlo sensitivity analysis (MCSA) or Bayesian sensitivity analysis (BSA) has emerged as a practical analytic strategy when there are multiple bias parameters inputs. BSA uses Bayes theorem to formally combine evidence from the prior distribution and the data. In contrast, MCSA samples bias parameters directly from the prior distribution. Intuitively, one would think that BSA and MCSA ought to give similar results. Both methods use similar models and the same (prior) probability distributions for the bias parameters. In this paper, we illustrate the surprising finding that BSA and MCSA can give very different results. Specifically, we demonstrate that MCSA can give inaccurate uncertainty assessments (e.g. 95% intervals) that do not reflect the data's influence on uncertainty about unmeasured confounding. Using a data example from epidemiology and simulation studies, we show that certain combinations of data and prior distributions can result in dramatic prior-to-posterior changes in uncertainty about the bias parameters. This occurs because the application of Bayes theorem in a non-identifiable model can sometimes rule out certain patterns of unmeasured confounding that are not compatible with the data. Consequently, the MCSA approach may give 95% intervals that are either too wide or too narrow and that do not have 95% frequentist coverage probability. Based on our findings, we recommend that analysts use BSA for probabilistic sensitivity analysis. Copyright © 2017 John Wiley & Sons, Ltd.
Institute of Scientific and Technical Information of China (English)
上官丹骅; 李刚; 邓力; 张宝印; 李瑞; 付元光
2015-01-01
在反应堆pin-by-pin精细建模及蒙特卡罗模拟计算研究中，由于不同栅元的功率密度差异较大，导致蒙特卡罗方法临界计算的样本在不同栅元之间的分配不均衡，由此引起栅元内的各种计数的统计误差差异较大。为使大部分栅元内计数的统计误差降至一个合理的水平，单纯增加总样本已不是一个高效的解决方法。通过在特定临界计算迭代算法的基础上改进并实现均匀裂变源算法的思想，对大亚湾压水堆pin-by-pin模型取得了具有较高效率的数值结果。本工作为具有自主知识产权的蒙特卡罗粒子输运模拟软件JMCT最终达到反应堆pin-by-pin模型(包括一系列国际基准模型)的模拟性能要求提供了一个有效的工具。%Because of a very non-uniform power distribution in core region, a very non-uniform distribution of relative un-certainties exists for tallies in Monte Carlo criticality calculations of pin-by-pin reactor model. To make a large part of cells obtain small enough relative uncertainties with reasonable time costs, increasing the total sample scale is not a good choice. By realizing a modified uniform-fission-site algorithm on the basis of source iteration algorithm of parallel Monte Carlo transport code JMCT, we obtain higher eﬃciencies for tallies in the calculations of pin-by-pin model of the Dayawang reactor plant. This work supplies a useful tool for matching the goal of simulating the benchmark pin-by-pin reactor models with a pre-described standard(the so called Kord-Smith challenge).
Energy Technology Data Exchange (ETDEWEB)
Jo, YuGwon; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)
2015-05-15
In this paper, the FSS iteration method is applied to the fast reactor where the neutron mean-free-path is around 10 times longer than that in the thermal reactor. The FSS iteration method with domain-based parallelism is tested on a two-dimensional continuous-energy fast reactor test problem. The multiplication factor and the pinwise fission-rate distributions of the FSS iteration method show good agreements with those of the conventional power method. A local domain is chosen as a cluster of 19 assemblies, taking into account the longer neutron mean-free-path. In the future, another type of local domain can be defined to take into account reflector assemblies and shield assemblies with appropriate boundary conditions. In the test problem, the multiplication factor and the pinwise fission-rate distributions of the FSS iteration method show good agreements with those of the conventional power method. Although the domain decomposition is easily achieved by the FSS iteration method, load-imbalance of local problems causes idle times in the processors. Applying the source splitting scheme and assigning different numbers of processors to local problems will reduce this problem.
Investigation of the Fundamental Constants Stability Based on the Reactor Oklo Burn-Up Analysis
Onegin, M. S.; Yudkevich, M. S.; Gomin, E. A.
2012-12-01
The burn-up of few samples of the natural Oklo reactor zones 3, 5 was calculated using the modern Monte Carlo code. We reconstructed the neutron spectrum in the core by means of the isotope ratios: 147Sm/148Sm and 176Lu/175Lu. These ratios unambiguously determine the water content and core temperature. The isotope ratio of the 149Sm in the sample calculated using this spectrum was compared with experimental one. The disagreement between these two values allows one to limit a possible shift of the low lying resonance of 149Sm. Then, these limits were converted to the limits for the change of the fine structure constant α. We have found out, that for the rate of α change, the inequality ěrt˙ {α }/α ěrt<= 5× 10-18 is fulfilled, which is one order higher than our previous limit.
Investigation of the fundamental constants stability based on the reactor Oklo burn-up analysis
Onegin, M S
2010-01-01
The burn-up for SC56-1472 sample of the natural Oklo reactor zone 3 was calculated using the modern Monte Carlo codes. We reconstructed the neutron spectrum in the core by means of the isotope ratios: $^{147}$Sm/$^{148}$Sm and $^{176}$Lu/$^{175}$Lu. These ratios unambiguously determine the spectrum index and core temperature. The effective neutron absorption cross section of $^{149}$Sm calculated using this spectrum was compared with experimental one. The disagreement between these two values allows to limit a possible shift of the low laying resonance of $^{149}$Sm even more . Then, these limits were converted to the limits for the change of the fine structure constant $\\alpha$. We found that for the rate of $\\alpha$ change the inequality $|\\delta \\dot{\\alpha}/\\alpha| \\le 5\\cdot 10^{-18}$ is fulfilled, which is of the next higher order than our previous limit.
Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle
Directory of Open Access Journals (Sweden)
Fic Adam
2015-03-01
Full Text Available Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle, which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle. The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.
Kudrolli, Haris A.
2001-04-01
A three dimensional (3D) reconstruction procedure for Positron Emission Tomography (PET) based on inverse Monte Carlo analysis is presented. PET is a medical imaging modality which employs a positron emitting radio-tracer to give functional images of an organ's metabolic activity. This makes PET an invaluable tool in the detection of cancer and for in-vivo biochemical measurements. There are a number of analytical and iterative algorithms for image reconstruction of PET data. Analytical algorithms are computationally fast, but the assumptions intrinsic in the line integral model limit their accuracy. Iterative algorithms can apply accurate models for reconstruction and give improvements in image quality, but at an increased computational cost. These algorithms require the explicit calculation of the system response matrix, which may not be easy to calculate. This matrix gives the probability that a photon emitted from a certain source element will be detected in a particular detector line of response. The ``Three Dimensional Stochastic Sampling'' (SS3D) procedure implements iterative algorithms in a manner that does not require the explicit calculation of the system response matrix. It uses Monte Carlo techniques to simulate the process of photon emission from a source distribution and interaction with the detector. This technique has the advantage of being able to model complex detector systems and also take into account the physics of gamma ray interaction within the source and detector systems, which leads to an accurate image estimate. A series of simulation studies was conducted to validate the method using the Maximum Likelihood - Expectation Maximization (ML-EM) algorithm. The accuracy of the reconstructed images was improved by using an algorithm that required a priori knowledge of the source distribution. Means to reduce the computational time for reconstruction were explored by using parallel processors and algorithms that had faster convergence rates
Derivation of landslide-triggering thresholds by Monte Carlo simulation and ROC analysis
Peres, David Johnny; Cancelliere, Antonino
2015-04-01
Rainfall thresholds of landslide-triggering are useful in early warning systems to be implemented in prone areas. Direct statistical analysis of historical records of rainfall and landslide data presents different shortcomings typically due to incompleteness of landslide historical archives, imprecise knowledge of the triggering instants, unavailability of a rain gauge located near the landslides, etc. In this work, a Monte Carlo approach to derive and evaluate landslide triggering thresholds is presented. Such an approach contributes to overcome some of the above mentioned shortcomings of direct empirical analysis of observed data. The proposed Monte Carlo framework consists in the combination of a rainfall stochastic model with hydrological and slope-stability model. Specifically, 1000-years long hourly synthetic rainfall and related slope stability factor of safety data are generated by coupling the Neyman-Scott rectangular pulses model with the TRIGRS unsaturated model (Baum et al., 2008) and a linear-reservoir water table recession model. Triggering and non-triggering rainfall events are then distinguished and analyzed to derive stochastic-input physically based thresholds that optimize the trade-off between correct and wrong predictions. For this purpose, receiver operating characteristic (ROC) indices are used. An application of the method to the highly landslide-prone area of the Peloritani mountains in north-eastern Sicily (Italy) is carried out. A threshold for the area is derived and successfully validated by comparison with thresholds proposed by other researchers. Moreover, the uncertainty in threshold derivation due to variability of rainfall intensity within events and to antecedent rainfall is investigated. Results indicate that variability of intensity during rainfall events influences significantly rainfall intensity and duration associated with landslide triggering. A representation of rainfall as constant-intensity hyetographs globally leads to
LOFT reactor vessel 290/sup 0/ downcomer stalk instrument penetration flange stress analysis
Energy Technology Data Exchange (ETDEWEB)
Finicle, D.P.
1978-06-06
The LOFT Reactor Vessel 290/sup 0/ Downcomer Stalk Instrument Penetration Flange Stress Analysis has been completed using normal operational and blowdown loading. A linear elastic analysis was completed using simplified hand analysis techniques. The analysis was in accordance with the 1977 ASME Boiler and Pressure Vessel Code, Section III, for a Class 1 component. Loading included internal pressure, bolt preload, and thermal gradients due to normal operating and blowdown.
Radiochemical analysis of concrete samples for decommission of nuclear reactors
Energy Technology Data Exchange (ETDEWEB)
Zapata-Garcia, Daniel; Wershofen, Herbert [Physikalisch-Technische Bundesanstalt (PTB), Bundesallee 100 38116, Braunschweig (Germany); Larijani, Cyrus; Sobrino-Petrirena, Maitane; Garcia-Miranda, Maria; Jerome, Simon M. [National Physical Laboratory (NPL), Hampton Road, Teddington, Middlesex, TW11 0LW (United Kingdom)
2014-07-01
Decommissioning of the oldest nuclear power reactors are some of the most challenging technological legacy issues many countries will face in forthcoming years, as many power reactors reach the end of their design lives. Decommissioning of nuclear reactors generates large amounts of waste that need to be classified according to their radioactive content. Approximately 10 % of the contaminated material ends up in different repositories (depending on their level of contamination) while the rest is decontaminated, measured and released into the environment or sent for recycling. Such classification needs to be done accurately in order to ensure that both the personnel involved in decommissioning and the population at large are not needlessly exposed to radiation or radioactive material and to minimise the environmental impact of such work. However, too conservative classification strategies should not be applied, in order to make proper use of radioactive waste repositories since space is limited and the full process must be cost-effective. Implicit in decommissioning and classification of waste is the need to analyse large amounts of material which usually combine a complex matrix with a non-homogeneous distribution of the radionuclides. Because the costs involved are large, it is possible to make great savings by the adoption of best available practices, such as the use of validated methods for on-site measurements and simultaneous determination of more than one radionuclide whenever possible. The work we present deals with the development and the validation of a procedure for the simultaneous determination of {sup 241}Am, plutonium isotopes, uranium isotopes and {sup 90}Sr in concrete samples. Samples are firstly ground and fused with LiBO{sub 2} and Li{sub 2}B{sub 4}O{sub 7}. After dissolution of the fused sample, silicate and alkaline elements are removed followed by radiochemical separation of the target radionuclides using extraction chromatography. Measurement
Lim, J. T.; Gold, H. J.; Wilkerson, G. G.; Raper, C. D. Jr; Raper CD, J. r. (Principal Investigator)
1989-01-01
We describe the application of a strategy for conducting a sensitivity analysis for a complex dynamic model. The procedure involves preliminary screening of parameter sensitivities by numerical estimation of linear sensitivity coefficients, followed by generation of a response surface based on Monte Carlo simulation. Application is to a physiological model of the vegetative growth of soybean plants. The analysis provides insights as to the relative importance of certain physiological processes in controlling plant growth. Advantages and disadvantages of the strategy are discussed.
Iiyama, Taku; Hagi, Kousuke; Urushibara, Takafumi; Ozeki, Sumio
2009-01-01
The intermolecular structure of C(2)H(5)OH molecules confined in slit-shaped graphitic micropore of activated carbon fiber was investigated by in situ X-ray diffraction (XRD) measurement and reverse Monte Carlo (RMC) analysis. The pseudo-3-dimensional intermolecular structure Of C(2)H(5)OH adsorbed in the micropores was determined by applying the RMC analysis to XRD data, assuming a simple slit-shaped space composed of double graphene sheets. The results were consistent with conventional Mont...
2013-01-01
Neutron diffusion length in reactor grade graphite is measured both experimentally and theoretically. The experimental work includes Monte Carlo (MC) coding using 'MCNP' and Finite Element Analysis (FEA) coding suing 'COMSOL Multiphysics' and Matlab. The MCNP code is adopted to simulate the thermal neutron diffusion length in a reactor moderator of 2m x 2m with slightly enriched uranium ($^{235}U$), accompanied with a model designed for thermal hydraulic analysis using point kinetic equations...
Micro-structural study and Rietveld analysis of fast reactor fuels: U-Mo fuels
Chakraborty, S.; Choudhuri, G.; Banerjee, J.; Agarwal, Renu; Khan, K. B.; Kumar, Arun
2015-12-01
U-Mo alloys are the candidate fuels for both research reactors and fast breeder reactors. In-reactor performance of the fuel depends on the microstructural stability and thermal properties of the fuel. To improve the fuel performance, alloying elements viz. Zr, Mo, Nb, Ti and fissium are added in the fuel. The first reactor fuels are normally prepared by injection casting. The objective of this work is to compare microstructure, phase-fields and hardness of as-cast four different U-Mo alloy (2, 5, 10 and 33 at.% Mo) fuels with the equilibrium microstructure of the alloys. Scanning electron microscope with energy dispersive spectrometer and optical microscope have been used to characterize the morphology of the as-cast and annealed alloys. The monoclinic α'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. A comparison of metallographic and Rietveld analysis of as-cast (dendritic microstructure) and annealed U-33 at.% Mo alloy, corresponding to intermetallic compound, has been reported here for the first time. This study will provide in depth understanding of microstructural and phase evolution of U-Mo alloys as fast reactor fuel.
Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems
Energy Technology Data Exchange (ETDEWEB)
Hwang, Dae Hyun; Seo, K. W
2006-01-15
A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC.
Energy Technology Data Exchange (ETDEWEB)
Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Prescott, Steven [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cogliati, Joshua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinoshita, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States)
2014-08-01
The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impact of these factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the application of a RISMC detailed demonstration case study for an emergent issue using the RAVEN and RELAP-7 tools. This case study looks at the impact of a couple of challenges to a hypothetical pressurized water reactor, including: (1) a power uprate, (2) a potential loss of off-site power followed by the possible loss of all diesel generators (i.e., a station black-out event), (3) and earthquake induces station-blackout, and (4) a potential earthquake induced tsunami flood. The analysis is performed by using a set of codes: a thermal-hydraulic code (RELAP-7), a flooding simulation tool (NEUTRINO) and a stochastic analysis tool (RAVEN) – these are currently under development at the Idaho National Laboratory.
2013-10-25
... COMMISSION Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors AGENCY... Systems for Light-Water-Cooled Nuclear Power Reactors,'' in which the NRC made editorial corrections and... analysis for liquid and gaseous radwaste system components for light water nuclear power...
The effect of load imbalances on the performance of Monte Carlo algorithms in LWR analysis
Energy Technology Data Exchange (ETDEWEB)
Siegel, A.R., E-mail: siegela@mcs.anl.gov [Argonne National Laboratory, Nuclear Engineering Division (United States); Argonne National Laboratory, Mathematics and Computer Science Division (United States); Smith, K., E-mail: kord@mit.edu [Massachusetts Institute of Technology, Department of Nuclear Science and Engineering (United States); Romano, P.K., E-mail: romano7@mit.edu [Massachusetts Institute of Technology, Department of Nuclear Science and Engineering (United States); Forget, B., E-mail: bforget@mit.edu [Massachusetts Institute of Technology, Department of Nuclear Science and Engineering (United States); Felker, K., E-mail: felker@mcs.anl.gov [Argonne National Laboratory, Mathematics and Computer Science Division (United States)
2013-02-15
A model is developed to predict the impact of particle load imbalances on the performance of domain-decomposed Monte Carlo neutron transport algorithms. Expressions for upper bound performance “penalties” are derived in terms of simple machine characteristics, material characterizations and initial particle distributions. The hope is that these relations can be used to evaluate tradeoffs among different memory decomposition strategies in next generation Monte Carlo codes, and perhaps as a metric for triggering particle redistribution in production codes.
Development of Monte Carlo code for coincidence prompt gamma-ray neutron activation analysis
Han, Xiaogang
Prompt Gamma-Ray Neutron Activation Analysis (PGNAA) offers a non-destructive, relatively rapid on-line method for determination of elemental composition of bulk and other samples. However, PGNAA has an inherently large background. These backgrounds are primarily due to the presence of the neutron excitation source. It also includes neutron activation of the detector and the prompt gamma rays from the structure materials of PGNAA devices. These large backgrounds limit the sensitivity and accuracy of PGNAA. Since most of the prompt gamma rays from the same element are emitted in coincidence, a possible approach for further improvement is to change the traditional PGNAA measurement technique and introduce the gamma-gamma coincidence technique. It is well known that the coincidence techniques can eliminate most of the interference backgrounds and improve the signal-to-noise ratio. A new Monte Carlo code, CEARCPG has been developed at CEAR to simulate gamma-gamma coincidence spectra in PGNAA experiment. Compared to the other existing Monte Carlo code CEARPGA I and CEARPGA II, a new algorithm of sampling the prompt gamma rays produced from neutron capture reaction and neutron inelastic scattering reaction, is developed in this work. All the prompt gamma rays are taken into account by using this new algorithm. Before this work, the commonly used method is to interpolate the prompt gamma rays from the pre-calculated gamma-ray table. This technique works fine for the single spectrum. However it limits the capability to simulate the coincidence spectrum. The new algorithm samples the prompt gamma rays from the nucleus excitation scheme. The primary nuclear data library used to sample the prompt gamma rays comes from ENSDF library. Three cases are simulated and the simulated results are benchmarked with experiments. The first case is the prototype for ETI PGNAA application. This case is designed to check the capability of CEARCPG for single spectrum simulation. The second
The use of Monte Carlo analysis for exposure assessment of an estuarine food web
Energy Technology Data Exchange (ETDEWEB)
Iannuzzi, T.J.; Shear, N.M.; Harrington, N.W.; Henning, M.H. [McLaren/Hart Environmental Engineering Corp., Portland, ME (United States). ChemRisk Div.
1995-12-31
Despite apparent agreement within the scientific community that probabilistic methods of analysis offer substantially more informative exposure predictions than those offered by the traditional point estimate approach, few risk assessments conducted or approved by state and federal regulatory agencies have used probabilistic methods. Among the likely deterrents to application of probabilistic methods to ecological risk assessment is the absence of ``standard`` data distributions that are considered applicable to most conditions for a given ecological receptor. Indeed, point estimates of ecological exposure factor values for a limited number of wildlife receptors have only recently been published. The Monte Carlo method of probabilistic modeling has received increasing support as a promising technique for characterizing uncertainty and variation in estimates of exposure to environmental contaminants. An evaluation of literature on the behavior, physiology, and ecology of estuarine organisms was conducted in order to identify those variables that most strongly influence uptake of xenobiotic chemicals from sediments, water and food sources. The ranges, central tendencies, and distributions of several key parameter values for polychaetes (Nereis sp.), mummichog (Fundulus heteroclitus), blue crab (Callinectes sapidus), and striped bass (Morone saxatilis) in east coast estuaries were identified. Understanding the variation in such factors, which include feeding rate, growth rate, feeding range, excretion rate, respiration rate, body weight, lipid content, food assimilation efficiency, and chemical assimilation efficiency, is critical to the understanding the mechanisms that control the uptake of xenobiotic chemicals in aquatic organisms, and to the ability to estimate bioaccumulation from chemical exposures in the aquatic environment.
Documented Safety Analysis Addendum for the Neutron Radiography Reactor Facility Core Conversion
Energy Technology Data Exchange (ETDEWEB)
Boyd D. Christensen
2009-05-01
The Neutron Radiography Reactor Facility (NRAD) is a Training, Research, Isotope Production, General Atomics (TRIGA) reactor which was installed in the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) in the mid 1970s. The facility provides researchers the capability to examine both irradiated and non-irradiated materials in support of reactor fuel and components programs through non-destructive neutron radiography examination. The facility has been used in the past as one facet of a suite of reactor fuels and component examination facilities available to researchers at the INL and throughout the DOE complex. The facility has also served various commercial research activities in addition to the DOE research and development support. The reactor was initially constructed using Fuel Lifetime Improvement Program (FLIP)- type highly enriched uranium (HEU) fuel obtained from the dismantled Puerto Rico Nuclear Center (PRNC) reactor. In accordance with international non-proliferation agreements, the NRAD core will be converted to a low enriched uranium (LEU) fuel and will continue to utilize the PRNC control rods, control rod drives, startup source, and instrument console as was previously used with the HEU core. The existing NRAD Safety Analysis Report (SAR) was created and maintained in the preferred format of the day, combining sections of both DOE-STD-3009 and Nuclear Regulatory Commission Regulatory Guide 1.70. An addendum was developed to cover the refueling and reactor operation with the LEU core. This addendum follows the existing SAR format combining required formats from both the DOE and NRC. This paper discusses the project to successfully write a compliant and approved addendum to the existing safety basis documents.
Hoffmann, Max J.; Engelmann, Felix; Matera, Sebastian
2017-01-01
Lattice kinetic Monte Carlo simulations have become a vital tool for predictive quality atomistic understanding of complex surface chemical reaction kinetics over a wide range of reaction conditions. In order to expand their practical value in terms of giving guidelines for the atomic level design of catalytic systems, it is very desirable to readily evaluate a sensitivity analysis for a given model. The result of such a sensitivity analysis quantitatively expresses the dependency of the turnover frequency, being the main output variable, on the rate constants entering the model. In the past, the application of sensitivity analysis, such as degree of rate control, has been hampered by its exuberant computational effort required to accurately sample numerical derivatives of a property that is obtained from a stochastic simulation method. In this study, we present an efficient and robust three-stage approach that is capable of reliably evaluating the sensitivity measures for stiff microkinetic models as we demonstrate using the CO oxidation on RuO2(110) as a prototypical reaction. In the first step, we utilize the Fisher information matrix for filtering out elementary processes which only yield negligible sensitivity. Then we employ an estimator based on the linear response theory for calculating the sensitivity measure for non-critical conditions which covers the majority of cases. Finally, we adapt a method for sampling coupled finite differences for evaluating the sensitivity measure for lattice based models. This allows for an efficient evaluation even in critical regions near a second order phase transition that are hitherto difficult to control. The combined approach leads to significant computational savings over straightforward numerical derivatives and should aid in accelerating the nano-scale design of heterogeneous catalysts.
Farobie, Obie; Matsumura, Yukihiko
2015-11-01
In this study, energy analysis was conducted for the production of biodiesel in a spiral reactor using supercritical tert-butyl methyl ether (MTBE). This study aims to determine the net energy ratio (NER) and energy efficiency for the production of biodiesel using supercritical MTBE and to verify the effectiveness of the spiral reactor in terms of heat recovery efficiency. The analysis results revealed that the NER for this process was 0.92. Meanwhile, the energy efficiency was 0.98, indicating that the production of biodiesel in a spiral reactor using supercritical MTBE is an energy-efficient process. By comparing the energy supply required for biodiesel production between spiral and conventional reactors, the spiral reactor was more efficient than the conventional reactor.
Comparative study of Thermal Hydraulic Analysis Codes for Pressurized Water Reactors
Energy Technology Data Exchange (ETDEWEB)
Kim, Yang Hoon; Jang, Mi Suk; Han, Kee Soo [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)
2015-05-15
Various codes are used for the thermal hydraulic analysis of nuclear reactors. The use of some codes among these is limited by user and some codes are not even open to general person. Thus, the use of alternative code is considered for some analysis. In this study, simple thermal hydraulic behaviors are analyzed using three codes to show that alternative codes are possible for the analysis of nuclear reactors. We established three models of the simple u-tube manometer using three different codes. RELAP5 (Reactor Excursion and Leak Analysis Program), SPACE (Safety and Performance Analysis CodE for nuclear power Plants), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are selected for this analysis. RELAP5 is widely used codes for the analysis of system behavior of PWRs. SPACE has been developed based on RELAP5 for the analysis of system behavior of PWRs and licensing of the code is in progress. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. The internal behavior of u-tube manometer was analyzed by RELAP5, SPACE and GOTHIC codes. The general transient behavior was similar among 3 codes. However, the stabilized status of the transient period analyzed by REPAP5 was different from the other codes. It would be resulted from the different physical models used in the other codes, which is specialized for the multi-phase thermal hydraulic behavior analysis.
Validation of SCALE and the TRITON Depletion Sequence for Gas-Cooled Reactor Analysis
Energy Technology Data Exchange (ETDEWEB)
DeHart, Mark D [ORNL; Pritchard, Megan L [ORNL
2008-01-01
The very-high-temperature reactor (VHTR) is an advanced reactor concept that uses graphite-moderated fuel and helium gas as a coolant. At present there are two primary VHTR reactor designs under consideration for development: in the pebble-bed reactor, a core is loaded with 'pebbles' consisting of 6 cm diameter spheres, while in a high-temperature gas-cooled reactor, fuel rods are placed within prismatic graphite blocks. In both systems, fuel elements (spheres or rods) are comprised of tristructural-isotropic (TRISO) fuel particles. The TRISO particles are either dispersed in the matrix of a graphite pebble for the pebble-bed design or molded into compacts/rods that are then inserted into the hexagonal graphite blocks for the prismatic concept. Two levels of heterogeneity exist in such fuel designs: (1) microspheres of TRISO particles dispersed in a graphite matrix of a cylindrical or spherical shape, and (2) neutron interactions at the rod-to-rod or sphere-to-sphere level. Such double heterogeneity (DH) provides a challenge to multigroup cross-section processing methods, which must treat each level of heterogeneity separately. A new capability to model doubly heterogeneous systems was added to the SCALE system in the release of Version 5.1. It was included in the control sequences CSAS and CSAS6, which use the Monte Carlo codes KENO V.a and KENO-VI, respectively, for three-dimensional neutron transport analyses and in the TRITON sequence, which uses the two-dimensional lattice physics code NEWT along with both versions of KENO for transport and depletion analyses. However, the SCALE 5.1 version of TRITON did not support the use of the DH approach for depletion. This deficiency has been addressed, and DH depletion will be available as an option in the upcoming release of SCALE 6. At present Oak Ridge National Laboratory (ORNL) staff are developing a set of calculations that may be used to validate SCALE for DH calculations. This paper discusses the
Advances in thermal hydraulic and neutronic simulation for reactor analysis and safety
Energy Technology Data Exchange (ETDEWEB)
Tentner, A.M.; Blomquist, R.N.; Canfield, T.R.; Ewing, T.F.; Garner, P.L.; Gelbard, E.M.; Gross, K.C.; Minkoff, M.; Valentin, R.A.
1993-03-01
This paper describes several large-scale computational models developed at Argonne National Laboratory for the simulation and analysis of thermal-hydraulic and neutronic events in nuclear reactors and nuclear power plants. The impact of advanced parallel computing technologies on these computational models is emphasized.
CFD analysis and flow model reduction for surfactant production in helix reactor
Nikačević, N.M.; Thielen, L.; Twerda, A.; Hof, P.M.J. van den
2014-01-01
Flow pattern analysis in a spiral Helix reactor is conducted, for the application in the commercial surfactant production. Step change response curves (SCR) were obtained from numerical tracer experiments by three-dimensional computational fluid dynamics (CFD) simulations. Non-reactive flow is simul
Energy Technology Data Exchange (ETDEWEB)
Sievers, D.; Kuhn, E.; Tucker, M.; Stickel, J.; Wolfrum, E.
2013-06-01
Measurement and analysis of residence time distribution (RTD) data is the focus of this study where data collection methods were developed specifically for the pretreatment reactor environment. Augmented physical sampling and automated online detection methods were developed and applied. Both the measurement techniques themselves and the produced RTD data are presented and discussed.
Comparative analysis of using natural and radiogenic lead as heat-transfer agent in fast reactors
Laas, R. A.; Gizbrekht, R. V.; Komarov, P. A.; Nesterov, V. N.
2016-06-01
Fast reactors with lead coolant have several advantages over analogues. Performance can be further improved by replacement of natural composition lead with radiogenic one. Thus, two main issues need to be addressed: induced radioactivity in coolant and efficient neutron multiplication factor in the core will be changed and need to be estimated. To address these issues analysis of the scheme of the nuclear transformations in the lead heat-transfer agent in the process of radiation was carried out. Induced radioactivity of radiogenic and natural lead has been studied. It is shown that replacement of lead affects multiplication factor in a certain way. Application of radiogenic lead can significantly affect reactor operation.
Spent fuel acceptance scenarios devoted to shutdown reactors: A preliminary analysis
Energy Technology Data Exchange (ETDEWEB)
Wood, T.W.; Plummer, A.M.; Dippold, D.G.; Short, S.M. (Pacific Northwest Lab., Richland, WA (USA); Battelle Memorial Inst., Columbus, OH (USA). Office of Transportation Systems and Planning; Pacific Northwest Lab., Richland, WA (USA))
1989-10-01
Spent fuel acceptance schedules and the allocation of federal acceptance capacity among commercial nuclear power reactors have important operational and cost consequences for reactor operators. Alternative allocation schemes were investigated to some extent in DOE's MRS Systems Study. The current study supplements these analyses for a class of acceptance schemes in which the acceptance capacity of the federal radioactive waste management system is allocated principally to shutdown commercial power reactors, and extends the scope of analysis to include considerations of at-reactor cask loading rates. The operational consequences of these schemes for power reactors, as measured in terms of quantity of spent fuel storage requirement above storage pool capacities and number of years of pool operations after last discharge, are estimated, as are the associated utility costs. This study does not attempt to examine the inter-utility equity considerations involved in departures from the current oldest-fuel-first (OFF) allocation rule as specified in the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste.'' In the sense that the alternative allocations are more economically efficient than OFF, however, they approximate the allocations that could result from free exchange of acceptance rights among utilities. Such a process would result in the preservation of inter-utility equity. 13 refs., 9 figs., 9 tabs.
Validation of SCALE for High Temperature Gas-Cooled Reactors Analysis
Energy Technology Data Exchange (ETDEWEB)
Ilas, Germina [ORNL; Ilas, Dan [ORNL; Kelly, Ryan P [ORNL; Sunny, Eva E [ORNL
2012-08-01
This report documents verification and validation studies carried out to assess the performance of the SCALE code system methods and nuclear data for modeling and analysis of High Temperature Gas-Cooled Reactor (HTGR) configurations. Validation data were available from the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhE Handbook), prepared by the International Reactor Physics Experiment Evaluation Project, for two different HTGR designs: prismatic and pebble bed. SCALE models have been developed for HTTR, a prismatic fuel design reactor operated in Japan and HTR-10, a pebble bed reactor operated in China. The models were based on benchmark specifications included in the 2009, 2010, and 2011 releases of the IRPhE Handbook. SCALE models for the HTR-PROTEUS pebble bed configuration at the PROTEUS critical facility in Switzerland have also been developed, based on benchmark specifications included in a 2009 IRPhE draft benchmark. The development of the SCALE models has involved a series of investigations to identify particular issues associated with modeling the physics of HTGRs and to understand and quantify the effect of particular modeling assumptions on calculation-to-experiment comparisons.
Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12
Energy Technology Data Exchange (ETDEWEB)
Aji, Indarta Kuncoro, E-mail: indartaaji@s.itb.ac.id [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia); Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Permana, Sidik [Nuclear Physics & Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia)
2015-09-30
Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF{sub 2}-ThF{sub 4}-{sup 233}UF{sub 4} respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.
Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12
Aji, Indarta Kuncoro; Waris, Abdul; Permana, Sidik
2015-09-01
Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF2-ThF4-233UF4 respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.
Fuel burnup analysis for IRIS reactor using MCNPX and WIMS-D5 codes
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.
Millimeter-Wave Thermal Analysis Development and Application to GEN IV Reactor Materials
Energy Technology Data Exchange (ETDEWEB)
Wosko, Paul; Sundram, S. K.
2012-10-16
New millimeter-wave thermal analysis instrumentation has been developed and studied for characterization of materials required for diverse fuel and structural needs in high temperature reactor environments such as the Next Generation Nuclear Plant (NGNP). A two-receiver 137 GHz system with orthogonal polarizations for anisotropic resolution of material properties has been implemented at MIT. The system was tested with graphite and silicon carbide specimens at temperatures up to 1300 ºC inside an electric furnace. The analytic and hardware basis for active millimeter-wave radiometry of reactor materials at high temperature has been established. Real-time, non contact measurement sensitivity to anisotropic surface emissivity and submillimeter surface displacement was demonstrated. The 137 GHz emissivity of reactor grade graphite (NBG17) from SGL Group was found to be low, ~ 5 %, in the 500 – 1200 °C range and increases by a factor of 2 to 4 with small linear grooves simulating fracturing. The low graphite emissivity would make millimeter-wave active radiometry a sensitive diagnostic of graphite changes due to environmentally induced stress fracturing, swelling, or corrosion. The silicon carbide tested from Ortek, Inc. was found to have a much higher emissivity at 137 GHz of ~90% Thin coatings of silicon carbide on reactor grade graphite supplied by SGL Group were found to be mostly transparent to millimeter-waves, increasing the 137 GHz emissivity of the coated reactor grade graphite to about ~14% at 1250 ºC.
Directory of Open Access Journals (Sweden)
Marziye Ebrahimkhani
2016-02-01
Full Text Available Calculation of the core neutronic parameters is one of the key components in all nuclear reactors. In this research, the energy spectrum and spatial distribution of the neutron flux in a uranium target have been calculated. In addition, sensitivity of the core neutronic parameters in accelerator-driven subcritical advanced liquid metal reactors, such as electron beam energy (Ee and source multiplication coefficient (ks, has been investigated. A Monte Carlo code (MCNPX_2.6 has been used to calculate neutronic parameters such as effective multiplication coefficient (keff, net neutron multiplication (M, neutron yield (Yn/e, energy constant gain (G0, energy gain (G, importance of neutron source (φ∗, axial and radial distributions of neutron flux, and power peaking factor (Pmax/Pave in two axial and radial directions of the reactor core for four fuel loading patterns. According to the results, safety margin and accelerator current (Ie have been decreased in the highest case of ks, but G and φ∗ have increased by 88.9% and 21.6%, respectively. In addition, for LP1 loading pattern, with increasing Ee from 100 MeV up to 1 GeV, Yn/e and G improved by 91.09% and 10.21%, and Ie and Pacc decreased by 91.05% and 10.57%, respectively. The results indicate that placement of the Np–Pu assemblies on the periphery allows for a consistent keff because the Np–Pu assemblies experience less burn-up.
Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis
Energy Technology Data Exchange (ETDEWEB)
Farmer, Mitchell T. [Argonne National Lab. (ANL), Argonne, IL (United States); Bunt, R. [Southern Nuclear, Atlanta, GA (United States); Corradini, M. [Univ. of Wisconsin, Madison, WI (United States); Ellison, Paul B. [GE Power and Water, Duluth, GA (United States); Francis, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Gabor, John D. [Erin Engineering, Walnut Creek, CA (United States); Gauntt, R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Henry, C. [Fauske and Associates, Burr Ridge, IL (United States); Linthicum, R. [Exelon Corp., Chicago, IL (United States); Luangdilok, W. [Fauske and Associates, Burr Ridge, IL (United States); Lutz, R. [PWR Owners Group (PWROG); Paik, C. [Fauske and Associates, Burr Ridge, IL (United States); Plys, M. [Fauske and Associates, Burr Ridge, IL (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rempe, J. [Rempe and Associates LLC, Idaho Falls, ID (United States); Robb, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Wachowiak, R. [Electric Power Research Inst. (EPRI), Knovville, TN (United States)
2015-01-31
The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).
TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis
Energy Technology Data Exchange (ETDEWEB)
Liles, D.R.; Mahaffy, J.H.
1984-02-01
The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided.
Energy Technology Data Exchange (ETDEWEB)
Kim, Hyun Jo; Lee, Sung Ho; Lee, Byung Doo; Jung, Juang [KAERI, Daejeon (Korea, Republic of)
2016-05-15
KiJang research reactor (KJRR) will be constructed to produce the radioisotope such as Mo-99 etc., provide the neutron transmutation doping (NTD) service of silicon, and develop the core technologies of research reactor. In this paper, the features of the process and nuclear material flow are reviewed and the material balance area (MBA) and key measurement point (KMP) are established based on the nuclear material flow. Also, this paper reviews the approach on safeguards and nuclear material accountancy at the facility level for Safeguards-by-Design at research reactor handling item count and bulk material. In this paper, MBA and KMPs are established through the analysis on facility features and major process at KJRR handling item count and bulk material. Also, this paper reviews the IAEA safeguards implementation and nuclear material accountancy at KJRR. It is necessary to discuss the safeguards approach on the fresh FM target assemblies and remaining uranium in the intermediate level liquid wastes.
A Monte Carlo approach to Beryllium-7 solar neutrino analysis with KamLAND
Grant, Christopher Peter
Terrestrial measurements of neutrinos produced by the Sun have been of great interest for over half a century because of their ability to test the accuracy of solar models. The first solar neutrinos detected with KamLAND provided a measurement of the 8B solar neutrino interaction rate above an analysis threshold of 5.5 MeV. This work describes efforts to extend KamLAND's detection sensitivity to solar neutrinos below 1 MeV, more specifically, those produced with an energy of 0.862 MeV from the 7Be electron-capture decay. Many of the difficulties in measuring solar neutrinos below 1 MeV arise from backgrounds caused abundantly by both naturally occurring, and man-made, radioactive nuclides. The primary nuclides of concern were 210Bi, 85Kr, and 39Ar. Since May of 2007, the KamLAND experiment has undergone two separate purification campaigns. During both campaigns a total of 5.4 ktons (about 6440 m3) of scintillator was circulated through a purification system, which utilized fractional distillation and nitrogen purging. After the purification campaign, reduction factors of 1.5 x 103 for 210Bi and 6.5 x 10 4 for 85Kr were observed. The reduction of the backgrounds provided a unique opportunity to observe the 7Be solar neutrino rate in KamLAND. An observation required detailed knowledge of the detector response at low energies, and to accomplish this, a full detector Monte Carlo simulation, called KLG4sim, was utilized. The optical model of the simulation was tuned to match the detector response observed in data after purification, and the software was optimized for the simulation of internal backgrounds used in the 7Be solar neutrino analysis. The results of this tuning and estimates from simulations of the internal backgrounds and external backgrounds caused by radioactivity on the detector components are presented. The first KamLAND analysis based on Monte Carlo simulations in the energy region below 2 MeV is shown here. The comparison of the chi2 between the null
Hazard analysis for 300 Area N Reactor Fuel Fabrication and Storage Facilty
Energy Technology Data Exchange (ETDEWEB)
Johnson, D.J.; Brehm, J.R.
1994-01-25
This hazard analysis (HA) has been prepared for the 300 Area N Reactor Fuel Fabrication and Storage Facility (Facility), in compliance with the requirements of Westinghouse Hanford Company (Westinghouse Hanford) controlled manual WHC-CM-4-46, Nonreactor Facility Safety Analysis Manual, and to the direction of WHC-IP-0690, Safety Analysis and Regulation Desk Instructions, (WHC 1992). An HA identifies potentially hazardous conditions in a facility and the associated potential accident scenarios. Unlike the Facility hazard classification documented in WHC-SD-NR-HC-004, Hazard Classification for 300 Area N Reactor Fuel Fabrication and Storage Facility, (Huang 1993), which is based on unmitigated consequences, credit is taken in an HA for administrative controls or engineered safety features planned or in place. The HA is the foundation for the accident analysis. The significant event scenarios identified by this HA will be further evaluated in a subsequent accident analysis.
Uncertainty analysis in the simulation of an HPGe detector using the Monte Carlo Code MCNP5
Energy Technology Data Exchange (ETDEWEB)
Gallardo, Sergio; Pozuelo, Fausto; Querol, Andrea; Verdu, Gumersindo; Rodenas, Jose, E-mail: sergalbe@upv.es [Universitat Politecnica de Valencia, Valencia, (Spain). Instituto de Seguridad Industrial, Radiofisica y Medioambiental (ISIRYM); Ortiz, J. [Universitat Politecnica de Valencia, Valencia, (Spain). Servicio de Radiaciones. Lab. de Radiactividad Ambiental; Pereira, Claubia [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear
2013-07-01
A gamma spectrometer including an HPGe detector is commonly used for environmental radioactivity measurements. Many works have been focused on the simulation of the HPGe detector using Monte Carlo codes such as MCNP5. However, the simulation of this kind of detectors presents important difficulties due to the lack of information from manufacturers and due to loss of intrinsic properties in aging detectors. Some parameters such as the active volume or the Ge dead layer thickness are many times unknown and are estimated during simulations. In this work, a detailed model of an HPGe detector and a petri dish containing a certified gamma source has been done. The certified gamma source contains nuclides to cover the energy range between 50 and 1800 keV. As a result of the simulation, the Pulse Height Distribution (PHD) is obtained and the efficiency curve can be calculated from net peak areas and taking into account the certified activity of the source. In order to avoid errors due to the net area calculation, the simulated PHD is treated using the GammaVision software. On the other hand, it is proposed to use the Noether-Wilks formula to do an uncertainty analysis of model with the main goal of determining the efficiency curve of this detector and its associated uncertainty. The uncertainty analysis has been focused on dead layer thickness at different positions of the crystal. Results confirm the important role of the dead layer thickness in the low energy range of the efficiency curve. In the high energy range (from 300 to 1800 keV) the main contribution to the absolute uncertainty is due to variations in the active volume. (author)
Energy Technology Data Exchange (ETDEWEB)
Abanades, A., E-mail: abanades@etsii.upm.es [Grupo de Modelizacion de Sistemas Termoenergeticos, ETSII, Universidad Politecnica de Madrid, c/Ramiro de Maeztu, 7, 28040 Madrid (Spain); Alvarez-Velarde, F.; Gonzalez-Romero, E.M. [Centro de Investigaciones Medioambientales y Tecnologicas (CIEMAT), Avda. Complutense, 40, Ed. 17, 28040 Madrid (Spain); Ismailov, K. [Tokyo Institute of Technology, 2-12-1, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Lafuente, A. [Grupo de Modelizacion de Sistemas Termoenergeticos, ETSII, Universidad Politecnica de Madrid, c/Ramiro de Maeztu, 7, 28040 Madrid (Spain); Nishihara, K. [Transmutation Section, J-PARC Center, JAEA, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Saito, M. [Tokyo Institute of Technology, 2-12-1, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Stanculescu, A. [International Atomic Energy Agency (IAEA), Vienna (Austria); Sugawara, T. [Transmutation Section, J-PARC Center, JAEA, Tokai-mura, Ibaraki-ken 319-1195 (Japan)
2013-01-15
Highlights: Black-Right-Pointing-Pointer TARC experiment benchmark capture rates results. Black-Right-Pointing-Pointer Utilization of updated databases, included ADSLib. Black-Right-Pointing-Pointer Self-shielding effect in reactor design for transmutation. Black-Right-Pointing-Pointer Effect of Lead nuclear data. - Abstract: The design of Accelerator Driven Systems (ADS) requires the development of simulation tools that are able to describe in a realistic way their nuclear performance and transmutation rate capability. In this publication, we present an evaluation of state of the art Monte Carlo design tools to assess their performance concerning transmutation of long-lived fission products. This work, performed under the umbrella of the International Atomic Energy Agency, analyses two important aspects for transmutation systems: moderation on Lead and neutron captures of {sup 99}Tc, {sup 127}I and {sup 129}I. The analysis of the results shows how shielding effects due to the resonances at epithermal energies of these nuclides affects strongly their transmutation rate. The results suggest that some research effort should be undertaken to improve the quality of Iodine nuclear data at epithermal and fast neutron energy to obtain a reliable transmutation estimation.
Energy Technology Data Exchange (ETDEWEB)
Hogerton, John
1964-01-01
This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.
Adjoint-based uncertainty quantification and sensitivity analysis for reactor depletion calculations
Stripling, Hayes Franklin
Depletion calculations for nuclear reactors model the dynamic coupling between the material composition and neutron flux and help predict reactor performance and safety characteristics. In order to be trusted as reliable predictive tools and inputs to licensing and operational decisions, the simulations must include an accurate and holistic quantification of errors and uncertainties in its outputs. Uncertainty quantification is a formidable challenge in large, realistic reactor models because of the large number of unknowns and myriad sources of uncertainty and error. We present a framework for performing efficient uncertainty quantification in depletion problems using an adjoint approach, with emphasis on high-fidelity calculations using advanced massively parallel computing architectures. This approach calls for a solution to two systems of equations: (a) the forward, engineering system that models the reactor, and (b) the adjoint system, which is mathematically related to but different from the forward system. We use the solutions of these systems to produce sensitivity and error estimates at a cost that does not grow rapidly with the number of uncertain inputs. We present the framework in a general fashion and apply it to both the source-driven and k-eigenvalue forms of the depletion equations. We describe the implementation and verification of solvers for the forward and ad- joint equations in the PDT code, and we test the algorithms on realistic reactor analysis problems. We demonstrate a new approach for reducing the memory and I/O demands on the host machine, which can be overwhelming for typical adjoint algorithms. Our conclusion is that adjoint depletion calculations using full transport solutions are not only computationally tractable, they are the most attractive option for performing uncertainty quantification on high-fidelity reactor analysis problems.
Shimada, M.; Yamada, Y.; Itoh, M.; Yatagai, T.
2001-09-01
Measurement of melanin and blood concentration in human skin is needed in the medical and the cosmetic fields because human skin colour is mainly determined by the colours of melanin and blood. It is difficult to measure these concentrations in human skin because skin has a multi-layered structure and scatters light strongly throughout the visible spectrum. The Monte Carlo simulation currently used for the analysis of skin colour requires long calculation times and knowledge of the specific optical properties of each skin layer. A regression analysis based on the modified Beer-Lambert law is presented as a method of measuring melanin and blood concentration in human skin in a shorter period of time and with fewer calculations. The accuracy of this method is assessed using Monte Carlo simulations.
Calculation of Reactor Kinetic Parameters with Monte Carlo Method%反应堆动态参数的蒙特卡罗计算研究
Institute of Scientific and Technical Information of China (English)
王冠博; 刘汉刚; 王侃; 刘永康; 曾和荣; 杨鑫
2012-01-01
Basic conceptions of kinetic parameters, including effective delayed neutron fraction (βeff), effective neutron generation time (Aeff) and a eigenvalue, and Monte Carlo calculation methods for these values are systematically introduced in this paper. Βeff is obtained with a "Prompt Method". Perturbation method is chosen to obtain Aeff. And then a eigenvalue is obtained by two ways, (I) prompt neutron density attenuation, in other words "direct simulation of time evolvement", (ii) indirect method using the result of kp and neutron generation time. Linear fitting is used to get the critical ac eigenvalues which match well with experimental ones. And uncertainties of kinetic parameters with different methods using Monte Carlo method are also analyzed.%介绍缓发中子有效份额(βeff)、有效中子代时间(∧eff)和α本征值的概念及其蒙特卡罗程序计算方法.采用Prompt Method方法计算得到βeff;微扰法得到∧eff；采用瞬发中子密度衰减直接拟合法和间接求解法得到α本征值；将各种反应性状态下的α拟合得到临界α本征值,并与实验测量的α,值进行比对,结果符合很好；并对动态参数蒙特卡罗程序计算的各种方法进行不确定度分析.
Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor
Energy Technology Data Exchange (ETDEWEB)
Zarifi, Ehsan; Sepanloo, Kamran [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor and Nuclear Safety School; Jahanfarnia, Golamreza [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch
2017-05-15
This study is aimed to perform the coupled thermal-hydraulic/neutronic analysis of nanofluids as the coolant in the hot fuel assembly of VVER-1000 reactor core. Water-based nanofluid containing various volume fractions of Al{sub 2}O{sub 3} nanoparticle is analyzed. WIMS and CITATION codes are used for neutronic simulation of the reactor core, calculating neutron flux and thermal power distribution. In the thermal-hydraulic modeling, the porous media approach is used to analyze the thermal behavior of the reactor core and the subchannel analysis is used to calculate the hottest fuel assembly thermal-hydraulic parameters. The derived conservation equations for coolant and conduction heat transfer equation for fuel and clad are discretized by Finite volume method and solved numerically using visual FORTRAN program. Finally the analysis results for nanofluids and pure water are compared together. The achieved results show that at low concentration (0.1 percent volume fraction) alumina is the optimum nanoparticles for normal reactor operation.
Simulation of a Neutron Noise Analysis Method for the Detection of Reactor Internals Vibration
Energy Technology Data Exchange (ETDEWEB)
Christian, Robby [Korea Advanced Institue of Science and Technology, Daejeon (Korea, Republic of); Song, Seon Ho [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)
2013-10-15
The results were compared against expected hypothesis. This simulation technique developed in C++ programming environment can successfully illustrate the principle of the neutron noise analysis as a vibration monitoring method. The addition of a white noise signal spectrum into the neutron flux data may result in a better coherence analysis. Examination of the phase data on adjacent and opposite flux pairs may be used to determine the vibration mode 3 session. Safety aspect is always highly demanded in any nuclear power plants operation. To achieve a high level of safety, it is desirable to perform preventive measures instead of corrective ones. One of these measures is the monitoring of reactor internals vibration characteristics. Any changes in the vibration signatures indicates an anomaly in the reactor internals. One proven method for this purpose is by analyzing the neutron flux sensed by ex-core detectors around the reactor core. Standards and guides have been written on the proper conduct of this method. The American Society of Mechanical Engineers (ASME) published two similar guides in the ASME OM-S/G-2007 document. Part 5 focuses on specifically monitoring the core support barrel axial preload. Part 23 elaborates on monitoring of reactor internals vibrations. U. S. Nuclear Regulatory Commission (NRC) issued a Regulatory Guide 1.20 on Comprehensive Vibration Assessment Program (CVAP). To understand the principle of neutron noise analysis on vibration monitoring, a simple neutron-transport model was simulated.
Reactor Pressure Vessel Fracture Analysis Capabilities in Grizzly
Energy Technology Data Exchange (ETDEWEB)
Spencer, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Backman, Marie [Univ. of Tennessee, Knoxville, TN (United States); Chakraborty, Pritam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hoffman, William [Univ. of Idaho, Moscow, ID (United States)
2015-03-01
Efforts have been underway to develop fracture mechanics capabilities in the Grizzly code to enable it to be used to perform deterministic fracture assessments of degraded reactor pressure vessels (RPVs). Development in prior years has resulted a capability to calculate -integrals. For this application, these are used to calculate stress intensity factors for cracks to be used in deterministic linear elastic fracture mechanics (LEFM) assessments of fracture in degraded RPVs. The -integral can only be used to evaluate stress intensity factors for axis-aligned flaws because it can only be used to obtain the stress intensity factor for pure Mode I loading. Off-axis flaws will be subjected to mixed-mode loading. For this reason, work has continued to expand the set of fracture mechanics capabilities to permit it to evaluate off-axis flaws. This report documents the following work to enhance Grizzly’s engineering fracture mechanics capabilities for RPVs: • Interaction Integral and -stress: To obtain mixed-mode stress intensity factors, a capability to evaluate interaction integrals for 2D or 3D flaws has been developed. A -stress evaluation capability has been developed to evaluate the constraint at crack tips in 2D or 3D. Initial verification testing of these capabilities is documented here. • Benchmarking for axis-aligned flaws: Grizzly’s capabilities to evaluate stress intensity factors for axis-aligned flaws have been benchmarked against calculations for the same conditions in FAVOR. • Off-axis flaw demonstration: The newly-developed interaction integral capabilities are demon- strated in an application to calculate the mixed-mode stress intensity factors for off-axis flaws. • Other code enhancements: Other enhancements to the thermomechanics capabilities that relate to the solution of the engineering RPV fracture problem are documented here.
Monte Carlo-based multiphysics coupling analysis of x-ray pulsar telescope
Li, Liansheng; Deng, Loulou; Mei, Zhiwu; Zuo, Fuchang; Zhou, Hao
2015-10-01
X-ray pulsar telescope (XPT) is a complex optical payload, which involves optical, mechanical, electrical and thermal disciplines. The multiphysics coupling analysis (MCA) plays an important role in improving the in-orbit performance. However, the conventional MCA methods encounter two serious problems in dealing with the XTP. One is that both the energy and reflectivity information of X-ray can't be taken into consideration, which always misunderstands the essence of XPT. Another is that the coupling data can't be transferred automatically among different disciplines, leading to computational inefficiency and high design cost. Therefore, a new MCA method for XPT is proposed based on the Monte Carlo method and total reflective theory. The main idea, procedures and operational steps of the proposed method are addressed in detail. Firstly, it takes both the energy and reflectivity information of X-ray into consideration simultaneously. And formulate the thermal-structural coupling equation and multiphysics coupling analysis model based on the finite element method. Then, the thermalstructural coupling analysis under different working conditions has been implemented. Secondly, the mirror deformations are obtained using construction geometry function. Meanwhile, the polynomial function is adopted to fit the deformed mirror and meanwhile evaluate the fitting error. Thirdly, the focusing performance analysis of XPT can be evaluated by the RMS. Finally, a Wolter-I XPT is taken as an example to verify the proposed MCA method. The simulation results show that the thermal-structural coupling deformation is bigger than others, the vary law of deformation effect on the focusing performance has been obtained. The focusing performances of thermal-structural, thermal, structural deformations have degraded 30.01%, 14.35% and 7.85% respectively. The RMS of dispersion spot are 2.9143mm, 2.2038mm and 2.1311mm. As a result, the validity of the proposed method is verified through
Monte Carlo analysis of a control technique for a tunable white lighting system
DEFF Research Database (Denmark)
Chakrabarti, Maumita; Thorseth, Anders; Jepsen, Jørgen
2017-01-01
A simulated colour control mechanism for a multi-coloured LED lighting system is presented. The system achieves adjustable and stable white light output and allows for system-to-system reproducibility after application of the control mechanism. The control unit works using a pre-calibrated lookup...... table for an experimentally realized system, with a calibrated tristimulus colour sensor. A Monte Carlo simulation is used to examine the system performance concerning the variation of luminous flux and chromaticity of the light output. The inputs to the Monte Carlo simulation, are variations of the LED...... peak wavelength, the LED rated luminous flux bin, the influence of the operating conditions, ambient temperature, driving current, and the spectral response of the colour sensor. The system performance is investigated by evaluating the outputs from the Monte Carlo simulation. The outputs show...
Monte Carlo analysis of a control technique for a tunable white lighting system
DEFF Research Database (Denmark)
Chakrabarti, Maumita; Thorseth, Anders; Jepsen, Jørgen
2017-01-01
A simulated colour control mechanism for a multi-coloured LED lighting system is presented. The system achieves adjustable and stable white light output and allows for system-to-system reproducibility after application of the control mechanism. The control unit works using a pre-calibrated lookup...... table for an experimentally realized system, with a calibrated tristimulus colour sensor. A Monte Carlo simulation is used to examine the system performance concerning the variation of luminous flux and chromaticity of the light output. The inputs to the Monte Carlo simulation, are variations of the LED...... peak wavelength, the LED rated luminous flux bin, the influence of the operating conditions, ambient temperature, driving current, and the spectral response of the colour sensor. The system performance is investigated by evaluating the outputs from the Monte Carlo simulation. The outputs show...
Monte-Carlo based Uncertainty Analysis For CO2 Laser Microchanneling Model
Prakash, Shashi; Kumar, Nitish; Kumar, Subrata
2016-09-01
CO2 laser microchanneling has emerged as a potential technique for the fabrication of microfluidic devices on PMMA (Poly-methyl-meth-acrylate). PMMA directly vaporizes when subjected to high intensity focused CO2 laser beam. This process results in clean cut and acceptable surface finish on microchannel walls. Overall, CO2 laser microchanneling process is cost effective and easy to implement. While fabricating microchannels on PMMA using a CO2 laser, the maximum depth of the fabricated microchannel is the key feature. There are few analytical models available to predict the maximum depth of the microchannels and cut channel profile on PMMA substrate using a CO2 laser. These models depend upon the values of thermophysical properties of PMMA and laser beam parameters. There are a number of variants of transparent PMMA available in the market with different values of thermophysical properties. Therefore, for applying such analytical models, the values of these thermophysical properties are required to be known exactly. Although, the values of laser beam parameters are readily available, extensive experiments are required to be conducted to determine the value of thermophysical properties of PMMA. The unavailability of exact values of these property parameters restrict the proper control over the microchannel dimension for given power and scanning speed of the laser beam. In order to have dimensional control over the maximum depth of fabricated microchannels, it is necessary to have an idea of uncertainty associated with the predicted microchannel depth. In this research work, the uncertainty associated with the maximum depth dimension has been determined using Monte Carlo method (MCM). The propagation of uncertainty with different power and scanning speed has been predicted. The relative impact of each thermophysical property has been determined using sensitivity analysis.
PROMSAR: A backward Monte Carlo spherical RTM for the analysis of DOAS remote sensing measurements
Palazzi, E.; Petritoli, A.; Giovanelli, G.; Kostadinov, I.; Bortoli, D.; Ravegnani, F.; Sackey, S. S.
A correct interpretation of diffuse solar radiation measurements made by Differential Optical Absorption Spectroscopy (DOAS) remote sensors require the use of radiative transfer models of the atmosphere. The simplest models consider radiation scattering in the atmosphere as a single scattering process. More realistic atmospheric models are those which consider multiple scattering and their application is useful and essential for the analysis of zenith and off-axis measurements regarding the lowest layers of the atmosphere, such as the boundary layer. These are characterized by the highest values of air density and quantities of particles and aerosols acting as scattering nuclei. A new atmospheric model, PROcessing of Multi-Scattered Atmospheric Radiation (PROMSAR), which includes multiple Rayleigh and Mie scattering, has recently been developed at ISAC-CNR. It is based on a backward Monte Carlo technique which is very suitable for studying the various interactions taking place in a complex and non-homogeneous system like the terrestrial atmosphere. PROMSAR code calculates the mean path of the radiation within each layer in which the atmosphere is sub-divided taking into account the large variety of processes that solar radiation undergoes during propagation through the atmosphere. This quantity is then employed to work out the Air Mass Factor (AMF) of several trace gases, to simulate in zenith and off-axis configurations their slant column amounts and to calculate the weighting functions from which informations about the gas vertical distribution is obtained using inversion methods. Results from the model, simulations and comparisons with actual slant column measurements are presented and discussed.
Software for neutron activation analysis at reactor IBR-2, FLNP, JINR
Zlokazov, V B
2004-01-01
A Delphi program suite, developed for processing gamma-spectra of induced activity of nuclei, obtained from the neutron activation measurements at the reactor IBR-2, FLNF, JINR, is reported. This suite contains components, intended for carrying out all the operations of the analysis cycle, starling with a data acquisition program for gamma -spectrometers Gamma (written in C++ Builder) and including Delphi programs for steps of the analysis. (6 refs).
Energy Technology Data Exchange (ETDEWEB)
NONE
1994-10-01
The objective is to sample and analyze for the presence of gamma emitting isotopes and hazardous constituents within certain areas of the Test Reactor Area (TRA), prior to D and D activities. The TRA is composed of three major reactor facilities and three smaller reactors built in support of programs studying the performance of reactor materials and components under high neutron flux conditions. The Materials Testing Reactor (MTR) and Engineering Test Reactor (ETR) facilities are currently pending D/D. Work consists of pre-D and D sampling of designated TRA (primarily ETR) process areas. This report addresses only a limited subset of the samples which will eventually be required to characterize MTR and ETR and plan their D and D. Sampling which is addressed in this document is intended to support planned D and D work which is funded at the present time. Biased samples, based on process knowledge and plant configuration, are to be performed. The multiple process areas which may be potentially sampled will be initially characterized by obtaining data for upstream source areas which, based on facility configuration, would affect downstream and as yet unsampled, process areas. Sampling and analysis will be conducted to determine the level of gamma emitting isotopes and hazardous constituents present in designated areas within buildings TRA-612, 642, 643, 644, 645, 647, 648, 663; and in the soils surrounding Facility TRA-611. These data will be used to plan the D and D and help determine disposition of material by D and D personnel. Both MTR and ETR facilities will eventually be decommissioned by total dismantlement so that the area can be restored to its original condition.
Cluster Monte Carlo and numerical mean field analysis for the water liquid-liquid phase transition
Mazza, Marco G.; Stokely, Kevin; Strekalova, Elena G.; Stanley, H. Eugene; Franzese, Giancarlo
2009-04-01
Using Wolff's cluster Monte Carlo simulations and numerical minimization within a mean field approach, we study the low temperature phase diagram of water, adopting a cell model that reproduces the known properties of water in its fluid phases. Both methods allow us to study the thermodynamic behavior of water at temperatures, where other numerical approaches - both Monte Carlo and molecular dynamics - are seriously hampered by the large increase of the correlation times. The cluster algorithm also allows us to emphasize that the liquid-liquid phase transition corresponds to the percolation transition of tetrahedrally ordered water molecules.
NUMERICAL ANALYSIS OF THERMAL STRATIFICATION IN THE UPPER PLENUM OF THE MONJU FAST REACTOR
Directory of Open Access Journals (Sweden)
SEOK-KI CHOI
2013-04-01
Full Text Available A numerical analysis of thermal stratification in the upper plenum of the MONJU fast breeder reactor was performed. Calculations were performed for a 1/6 simplified model of the MONJU reactor using the commercial code, CFX-13. To better resolve the geometrically complex upper core structure of the MONJU reactor, the porous media approach was adopted for the simulation. First, a steady state solution was obtained and the transient solutions were then obtained for the turbine trip test conducted in December 1995. The time dependent inlet conditions for the mass flow rate and temperature were provided by JAEA. Good agreement with the experimental data was observed for steady state solution. The numerical solution of the transient analysis shows the formation of thermal stratification within the upper plenum of the reactor vessel during the turbine trip test. The temporal variations of temperature were predicted accurately by the present method in the initial rapid coastdown period (∼300 seconds. However, transient numerical solutions show a faster thermal mixing than that observed in the experiment after the initial coastdown period. A nearly homogenization of the temperature field in the upper plenum is predicted after about 900 seconds, which is a much shorter-term thermal stratification than the experimental data indicates. This discrepancy may be due to the shortcoming of the turbulence models available in the CFX-13 code for a natural convection flow with thermal stratification.
Energy Technology Data Exchange (ETDEWEB)
O' Kula, K.R.; Sharp, D.A. (Westinghouse Savannah River Co., Aiken, SC (United States)); Amos, C.N.; Wagner, K.C.; Bradley, D.R. (Science Applications International Corp., Albuquerque, NM (United States))
1992-01-01
A three-level Probabilistic Safety Assessment (PSA) of production reactor operation has been underway since 1985 at the US Department of Energy's Savannah River Site (SRS). The goals of this analysis are to: Analyze existing margins of safety provided by the heavy-water reactor (HWR) design challenged by postulated severe accidents; Compare measures of risk to the general public and onsite workers to guideline values, as well as to those posed by commercial reactor operation; and Develop the methodology and database necessary to prioritize improvements to engineering safety systems and components, operator training, and engineering projects that contribute significantly to improving plant safety. PSA technical staff from the Westinghouse Savannah River Company (WSRC) and Science Applications International Corporation (SAIC) have performed the assessment despite two obstacles: A variable baseline plant configuration and power level; and a lack of technically applicable code methodology to model the SRS reactor conditions. This paper discusses the detailed effort necessary to modify the requisite codes before accident analysis insights for the risk assessment were obtained.
DEFF Research Database (Denmark)
Mladenovska, Zuzana; Dabrowski, Slawomir; Ahring, Birgitte Kiær
2003-01-01
and a higher removal of VS than the reactor treating manure. Microbial population analysis done by cultivation - most probable number (MPN) test and specific methanogenic activity (SMA) measurement, revealed higher MPN and increased SMA of methanogenic populations of biomass from the reactor codigesting manure...
Safety analysis of 5 MW IEAR-1 reactor; Analise de seguranca do reator IEA-R1 a 5 MW
Energy Technology Data Exchange (ETDEWEB)
Silva, Antonio T. e; Maprelian, Eduardo; Rodrigues, Antonio C.I.; Cabral, Eduardo L.L.; Molnary, Leslie de; Mesquita, Ricardo N.; Mendonca, Arlindo G. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Dept. de Reatores. E-mail: teixeira@net.ipen.br
2000-07-01
This paper presents the methods and procedures utilized in the safety analysis of IEA-R1 research reactor. Four postulated accidents are quantitatively analyzed, being the fuel channel blockage accident considered as the Maximum Credible Accident for the reactor. The potential accident consequences and the criteria for radiological doses acceptance are evaluated and discussed. (author)
Energy Technology Data Exchange (ETDEWEB)
Dunn, Floyd E. [Argonne National Lab. (ANL), Argonne, IL (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States). Nuclear Reactor Lab.; Wilson, Erik [Argonne National Lab. (ANL), Argonne, IL (United States)
2016-12-01
The STAT code was written to automate many of the steady-state thermal hydraulic safety calculations for the MIT research reactor, both for conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel and for future fuel re-loads after the conversion. A Monte-Carlo statistical propagation approach is used to treat uncertainties in important parameters in the analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings on avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis has been performed by analyzing the margin to ONB for each axial node of each stripe of each plate of each element in the core.
Monte Carlo homogenized limit analysis model for randomly assembled blocks in-plane loaded
Milani, Gabriele; Lourenço, Paulo B.
2010-11-01
A simple rigid-plastic homogenization model for the limit analysis of masonry walls in-plane loaded and constituted by the random assemblage of blocks with variable dimensions is proposed. In the model, blocks constituting a masonry wall are supposed infinitely resistant with a Gaussian distribution of height and length, whereas joints are reduced to interfaces with frictional behavior and limited tensile and compressive strength. Block by block, a representative element of volume (REV) is considered, constituted by a central block interconnected with its neighbors by means of rigid-plastic interfaces. The model is characterized by a few material parameters, is numerically inexpensive and very stable. A sub-class of elementary deformation modes is a-priori chosen in the REV, mimicking typical failures due to joints cracking and crushing. Masonry strength domains are obtained equating the power dissipated in the heterogeneous model with the power dissipated by a fictitious homogeneous macroscopic plate. Due to the inexpensiveness of the approach proposed, Monte Carlo simulations can be repeated on the REV in order to have a stochastic estimation of in-plane masonry strength at different orientations of the bed joints with respect to external loads accounting for the geometrical statistical variability of blocks dimensions. Two cases are discussed, the former consisting on full stochastic REV assemblages (obtained considering a random variability of both blocks height an length) and the latter assuming the presence of a horizontal alignment along bed joints, i.e. allowing blocks height variability only row by row. The case of deterministic blocks height (quasi-periodic texture) can be obtained as a subclass of this latter case. Masonry homogenized failure surfaces are finally implemented in an upper bound FE limit analysis code for the analysis at collapse of entire walls in-plane loaded. Two cases of engineering practice, consisting on the prediction of the failure
The use of experimental data in an MTR-type nuclear reactor safety analysis
Day, Simon E.
Reactivity initiated accidents (RIAs) are a category of events required for research reactor safety analysis. A subset of this is unprotected RIAs in which mechanical systems or human intervention are not credited in the response of the system. Light-water cooled and moderated MTR-type ( i.e., aluminum-clad uranium plate fuel) reactors are self-limiting up to some reactivity insertion limit beyond which fuel damage occurs. This characteristic was studied in the Borax and Spert reactor tests of the 1950s and 1960s in the USA. This thesis considers the use of this experimental data in generic MTR-type reactor safety analysis. The approach presented herein is based on fundamental phenomenological understanding and uses correlations in the reactor test data with suitable account taken for differences in important system parameters. Specifically, a semi-empirical approach is used to quantify the relationship between the power, energy and temperature rise response of the system as well as parametric dependencies on void coefficient and the degree of subcooling. Secondary effects including the dependence on coolant flow are also examined. A rigorous curve fitting approach and error assessment is used to quantify the trends in the experimental data. In addition to the initial power burst stage of an unprotected transient, the longer term stability of the system is considered with a stylized treatment of characteristic power/temperature oscillations (chugging). A bridge from the HEU-based experimental data to the LEU fuel cycle is assessed and outlined based on existing simulation results presented in the literature. A cell-model based parametric study is included. The results are used to construct a practical safety analysis methodology for determining reactivity insertion safety limits for a light-water moderated and cooled MTR-type core.
Analysis of Moderator System Failure Accidents by Using New Method for Wolsong-1 CANDU 6 Reactor
Energy Technology Data Exchange (ETDEWEB)
Jin, Dongsik; Kim, Jonghyun; Cho, Cheonhwey [Atomic Creative Technology Co., Ltd., Daejeon (Korea, Republic of); Kim, Sungmin [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)
2013-05-15
To reconfirm the safety of moderator system failure accidents, the safety analysis by using the reactor physics code, RFSP-IST, coupled with the thermal hydraulics code, CATHENA is performed additionally. In the present paper, the newly developed analysis method is briefly described and the results obtained from the moderator system failure accident simulations for Wolsong-1 CANDU 6 reactor by using the new method are summarized. The safety analysis of the moderator system failure accidents for Wolsong-1 CANDU 6 reactor was carried out by using the new code system, i. e., CATHENA and RFSP-IST, instead of the non-IST old codes, namely, SMOKIN G-2 and MODSTBOIL. The analysis results by using the new method revealed as same with the results by using the old method that the fuel integrity is warranted because the localized power peak remained well below the limits and, most importantly, the reactor operation enters into the self-shutdown mode due to the substantial loss of moderator D{sub 2}O inventory from the moderator system. In the analysis results obtained by using the old method, it was predicted that the ROP trip conditions occurred for the transient cases which are also studied in the present paper. But, in the new method, it was found that the ROP trip conditions did not occur. Consequently, in the safety analysis performed additionally by using the new method, the safety of moderator system failure accidents was reassured. In the future, the new analysis method by using the IST codes instead of the non-IST old codes for the moderator system failure accidents is strongly recommended.
Lai, Bo-Lun; Sheu, Rong-Jiun; Lin, Uei-Tyng
2015-05-01
Monte Carlo simulations are generally considered the most accurate method for complex accelerator shielding analysis. Simplified models based on point-source line-of-sight approximation are often preferable in practice because they are intuitive and easy to use. A set of shielding data, including source terms and attenuation lengths for several common targets (iron, graphite, tissue, and copper) and shielding materials (concrete, iron, and lead) were generated by performing Monte Carlo simulations for 100-300 MeV protons. Possible applications and a proper use of the data set were demonstrated through a practical case study, in which shielding analysis on a typical proton treatment room was conducted. A thorough and consistent comparison between the predictions of our point-source line-of-sight model and those obtained by Monte Carlo simulations for a 360° dose distribution around the room perimeter showed that the data set can yield fairly accurate or conservative estimates for the transmitted doses, except for those near the maze exit. In addition, this study demonstrated that appropriate coupling between the generated source term and empirical formulae for radiation streaming can be used to predict a reasonable dose distribution along the maze. This case study proved the effectiveness and advantage of applying the data set to a quick shielding design and dose evaluation for proton therapy accelerators.
A Bayesian analysis of rare B decays with advanced Monte Carlo methods
Energy Technology Data Exchange (ETDEWEB)
Beaujean, Frederik
2012-11-12
Searching for new physics in rare B meson decays governed by b {yields} s transitions, we perform a model-independent global fit of the short-distance couplings C{sub 7}, C{sub 9}, and C{sub 10} of the {Delta}B=1 effective field theory. We assume the standard-model set of b {yields} s{gamma} and b {yields} sl{sup +}l{sup -} operators with real-valued C{sub i}. A total of 59 measurements by the experiments BaBar, Belle, CDF, CLEO, and LHCb of observables in B{yields}K{sup *}{gamma}, B{yields}K{sup (*)}l{sup +}l{sup -}, and B{sub s}{yields}{mu}{sup +}{mu}{sup -} decays are used in the fit. Our analysis is the first of its kind to harness the full power of the Bayesian approach to probability theory. All main sources of theory uncertainty explicitly enter the fit in the form of nuisance parameters. We make optimal use of the experimental information to simultaneously constrain theWilson coefficients as well as hadronic form factors - the dominant theory uncertainty. Generating samples from the posterior probability distribution to compute marginal distributions and predict observables by uncertainty propagation is a formidable numerical challenge for two reasons. First, the posterior has multiple well separated maxima and degeneracies. Second, the computation of the theory predictions is very time consuming. A single posterior evaluation requires O(1s), and a few million evaluations are needed. Population Monte Carlo (PMC) provides a solution to both issues; a mixture density is iteratively adapted to the posterior, and samples are drawn in a massively parallel way using importance sampling. The major shortcoming of PMC is the need for cogent knowledge of the posterior at the initial stage. In an effort towards a general black-box Monte Carlo sampling algorithm, we present a new method to extract the necessary information in a reliable and automatic manner from Markov chains with the help of hierarchical clustering. Exploiting the latest 2012 measurements, the fit
Passive Gamma Analysis of the Boiling-Water-Reactor Assemblies
Energy Technology Data Exchange (ETDEWEB)
Vo, Duc Ta [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Favalli, Andrea [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2016-03-31
Passive gamma analysis can be used to determine BU and CT of BWR assembly. The analysis is somewhat more complicated and less effective than similar method for PWR assemblies. From the measurements along the lengths of the BWR1 and BWR9 assemblies, there are hints that we may be able to use their information to help improve the model functions for better results.
Major update of Safety Analysis Report for Thai Research Reactor-1/Modification 1
Energy Technology Data Exchange (ETDEWEB)
Tippayakul, Chanatip [Thailand Institute of Nuclear Technology, Bangkok (Thailand)
2013-07-01
Thai Research Reactor-1/Modification 1 (TRR-1/M1) was converted from a Material Testing Reactor in 1975 and it had been operated by Office of Atom for Peace (OAP) since 1977 until 2007. During the period, Office of Atom for Peace had two duties for the reactor, that is, to operate and to regulate the reactor. However, in 2007, there was governmental office reformation which resulted in the separation of the reactor operating organization from the regulatory body in order to comply with international standard. The new organization is called Thailand Institute of Nuclear Technology (TINT) which has the mission to promote peaceful utilization of nuclear technology while OAP remains essentially the regulatory body. After the separation, a new ministerial regulation was enforced reflecting a new licensing scheme in which TINT has to apply for a license to operate the reactor. The safety analysis report (SAR) shall be submitted as part of the license application. The ministerial regulation stipulates the outlines of the SAR almost equivalent to IAEA standard 35-G1. Comparing to the IAEA 35-G1 standard, there were several incomplete and missing chapters in the original SAR of TRR1/M1. The major update of the SAR was therefore conducted and took approximately one year. The update work included detail safety evaluation of core configuration which used two fuel element types, the classification of systems, structures and components (SSC), the compilation of detail descriptions of all SSCs and the review and evaluation of radiation protection program, emergency plan and emergency procedure. Additionally, the code of conduct and operating limits and conditions were revised and finalized in this work. A lot of new information was added to the SAR as well, for example, the description of commissioning program, information on environmental impact assessment, decommissioning program, quality assurance program and etc. Due to the complexity of this work, extensive knowledge was
Analysis of Postulated Core Meltdown of an SRP Reactor - Final Report
Energy Technology Data Exchange (ETDEWEB)
Durant, W.S.; Brown, R.J.
1970-10-01
An analysis was made to determine the consequences of a postulated accident in which the core of a Savannah River Plant reactor melts down following the loss of coolant. The study was made to determine (1) the potential damage to the reactor building that could impair its integrity for confining activity and (2) the need for additional facilities to prevent the activity confinement system from being overheated by the decay heat in the debris. A preliminary report on this analysis was issued previously. The sequence of events during and following the loss of coolant has now been studied in more detail, and a computer program has been written and used to investigate transient heating effects. This is the final report of the analysis and presents the conclusions.
Computer program uses Monte Carlo techniques for statistical system performance analysis
Wohl, D. P.
1967-01-01
Computer program with Monte Carlo sampling techniques determines the effect of a component part of a unit upon the overall system performance. It utilizes the full statistics of the disturbances and misalignments of each component to provide unbiased results through simulated random sampling.
Progress report on neutron activation analysis at Dalat Nuclear Research Reactor
Energy Technology Data Exchange (ETDEWEB)
Tuan, Nguyen Ngoc [Nuclear Research Institute, Dalat (Viet Nam)
2003-03-01
Neutron Activation Analysis (NAA) is one of most powerful techniques for the simultaneous multi-elements analysis. This technique has been studied and applied to analyze major, minor and trace elements in Geological, Biological and Environmental samples at Dalat Nuclear Research Reactor. At the sixth Workshop, February 8-11, 1999, Yojakarta, Indonesia we had a report on Current Status of Neutron Activation Analysis using Dalat Nuclear Research Reactor. Another report on Neutron Activation Analysis at the Dalat Nuclear Research Reactor also was presented at the seventh Workshop in Taejon, Korea from November 20-24, 2000. So in this report, we would like to present the results obtained of the application of NAA at NRI for one year as follows: (1) Determination of the concentrations of noble, rare earth, uranium, thorium and other elements in Geological samples according to requirement of clients particularly the geologists, who want to find out the mineral resources. (2) The analysis of concentration of radionuclides and nutrient elements in foodstuffs to attend the program on Asian Reference Man. (3) The evaluation of the contents of trace elements in crude oil and basement rock samples to determine original source of the oil. (4) Determination of the elemental composition of airborne particle in the Ho Chi Minh City for studying air pollution. The analytical data of standard reference material, toxic elements and natural radionuclides in seawater are also presented. (author)
Suzuki, Teppei; Tani, Yuji; Ogasawara, Katsuhiko
2016-07-25
Consistent with the "attention, interest, desire, memory, action" (AIDMA) model of consumer behavior, patients collect information about available medical institutions using the Internet to select information for their particular needs. Studies of consumer behavior may be found in areas other than medical institution websites. Such research uses Web access logs for visitor search behavior. At this time, research applying the patient searching behavior model to medical institution website visitors is lacking. We have developed a hospital website search behavior model using a Bayesian approach to clarify the behavior of medical institution website visitors and determine the probability of their visits, classified by search keyword. We used the website data access log of a clinic of internal medicine and gastroenterology in the Sapporo suburbs, collecting data from January 1 through June 31, 2011. The contents of the 6 website pages included the following: home, news, content introduction for medical examinations, mammography screening, holiday person-on-duty information, and other. The search keywords we identified as best expressing website visitor needs were listed as the top 4 headings from the access log: clinic name, clinic name + regional name, clinic name + medical examination, and mammography screening. Using the search keywords as the explaining variable, we built a binomial probit model that allows inspection of the contents of each purpose variable. Using this model, we determined a beta value and generated a posterior distribution. We performed the simulation using Markov Chain Monte Carlo methods with a noninformation prior distribution for this model and determined the visit probability classified by keyword for each category. In the case of the keyword "clinic name," the visit probability to the website, repeated visit to the website, and contents page for medical examination was positive. In the case of the keyword "clinic name and regional name," the
Predictive uncertainty analysis of a saltwater intrusion model using null-space Monte Carlo
Herckenrath, Daan; Langevin, Christian D.; Doherty, John
2011-01-01
Because of the extensive computational burden and perhaps a lack of awareness of existing methods, rigorous uncertainty analyses are rarely conducted for variable-density flow and transport models. For this reason, a recently developed null-space Monte Carlo (NSMC) method for quantifying prediction uncertainty was tested for a synthetic saltwater intrusion model patterned after the Henry problem. Saltwater intrusion caused by a reduction in fresh groundwater discharge was simulated for 1000 randomly generated hydraulic conductivity distributions, representing a mildly heterogeneous aquifer. From these 1000 simulations, the hydraulic conductivity distribution giving rise to the most extreme case of saltwater intrusion was selected and was assumed to represent the "true" system. Head and salinity values from this true model were then extracted and used as observations for subsequent model calibration. Random noise was added to the observations to approximate realistic field conditions. The NSMC method was used to calculate 1000 calibration-constrained parameter fields. If the dimensionality of the solution space was set appropriately, the estimated uncertainty range from the NSMC analysis encompassed the truth. Several variants of the method were implemented to investigate their effect on the efficiency of the NSMC method. Reducing the dimensionality of the null-space for the processing of the random parameter sets did not result in any significant gains in efficiency and compromised the ability of the NSMC method to encompass the true prediction value. The addition of intrapilot point heterogeneity to the NSMC process was also tested. According to a variogram comparison, this provided the same scale of heterogeneity that was used to generate the truth. However, incorporation of intrapilot point variability did not make a noticeable difference to the uncertainty of the prediction. With this higher level of heterogeneity, however, the computational burden of
Energy Technology Data Exchange (ETDEWEB)
Lee, Jae Han; Yoo, Bong
2000-11-01
It is important to establish a highly accurate technique of evaluating the sloshing behavior of liquid sodium coolant during earthquake for structural integrity of KALIMER reactor vessel and internals. The analysis procedure of sloshing behaviors is established using finite element computer program ANSYS, and the effectiveness of the procedure is confirmed by comparison with theoretical and experimental results in the literature. The analysis results agree well with experimental ones. Based on the procedure, the sloshing characteristics of liquid sodium coolant in the KALIMER reactor vessel including reactor internal components are evaluated. The maximum response height of sodium free surface at the reactor vessel is about 55cm when subjected to horizontal safe shutdown earthquake (SSE) of 0.3g for seismically isolated reactor building.
Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 3-Surry Unit 1 Cycle 2
Energy Technology Data Exchange (ETDEWEB)
Bowman, S.M.
1995-01-01
The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using selected critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations in this report is based on the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies in the critical configurations were calculated using the SAS2H analytical sequence of the SCALE-4 system. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code module was used to extract the necessary isotopic densities from the SAS2H results and to provide the data in the format required by the SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of the cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) of each case. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all the calculations. This volume of the report documents the SCALE system analysis of two reactor critical configurations for Surry Unit 1 Cycle 2. This unit and cycle were chosen for a previous analysis using a different methodology because detailed isotopics from multidimensional reactor calculations were available from the Virginia Power Company. These data permitted a direct comparison of criticality calculations using the utility-calculated isotopics with those using the isotopics generated by the SCALE-4
CFD analysis and flow model reduction for surfactant production in helix reactor
Directory of Open Access Journals (Sweden)
Nikačević N.M.
2015-01-01
Full Text Available Flow pattern analysis in a spiral Helix reactor is conducted, for the application in the commercial surfactant production. Step change response curves (SCR were obtained from numerical tracer experiments by three-dimensional computational fluid dynamics (CFD simulations. Non-reactive flow is simulated, though viscosity is treated as variable in the direction of flow, as it increases during the reaction. The design and operating parameters (reactor diameter, number of coils and inlet velocity are varied in CFD simulations, in order to examine the effects on the flow pattern. Given that 3D simulations are not practical for fast computations needed for optimization, scale-up and control, CFD flow model is reduced to one-dimensional axial dispersion (AD model with spatially variable dispersion coefficient. Dimensionless dispersion coefficient (Pe is estimated under different conditions and results are analyzed. Finally, correlation which relates Pe number with Reynolds number and number of coils from the reactor entrance is proposed for the particular reactor application and conditions.
Design and Analysis of the Power Control System of the Fast Zero Energy Reactor FR-0
Energy Technology Data Exchange (ETDEWEB)
Schuh, N.J.H.
1966-12-15
This report describes the power control by means of the fine-control rod and the design of the control system of the fast zero energy reactor FR-0 located in Studsvik, Sweden. System requirements and some operational conditions were used as design criteria. Manual and automatic control is possible. Variable electronic end-stops for the control rod have been designed, because of the special construction of the reactor and control rod. Noise in the control system caused by the reactor, detector and electronics caused disturbances of the control system at the lower power levels. The noise power-spectrum was measured. Statistical design methods, using the measured noise power spectrum, were used to design filters, which will reduce the influence of the noise at the lower power levels. Root Loci sketches and Bode diagrams were used for stability analyses. The system was simulated on an analogue computer, taking into account even nonlinearities of the control system and noise. Typical cases of reactor operation were simulated and stability analysis performed.
Directory of Open Access Journals (Sweden)
Junli Gou
2009-01-01
Full Text Available A theoretical investigation on the thermal hydraulic characteristics of a new type of passive residual heat removal system (PRHRS, which is connected to the reactor coolant system via the secondary side of the steam generator, for an integral pressurized water reactor is presented in this paper. Three-interknited natural circulation loops are adopted by this PRHRS to remove the residual heat of the reactor core after a reactor trip. Based on the one-dimensional model and a simulation code (SCPRHRS, the transient behaviors of the PRHRS as well as the effects of the height difference between the steam generator and the heat exchanger and the heat transfer area of the heat exchanger are studied in detail. Through the calculation analysis, it is found that the calculated parameter variation trends are reasonable. The higher height difference between the steam generator and the residual heat exchanger and the larger heat transfer area of the residual heat exchanger are favorable to the passive residual heat removal system.
Analysis of the key enzymes of butyric and acetic acid fermentation in biogas reactors.
Gabris, Christina; Bengelsdorf, Frank R; Dürre, Peter
2015-09-01
This study aimed at the investigation of the mechanisms of acidogenesis, which is a key process during anaerobic digestion. To expose possible bottlenecks, specific activities of the key enzymes of acidification, such as acetate kinase (Ack, 0.23-0.99 U mg(-1) protein), butyrate kinase (Buk, < 0.03 U mg(-1) protein) and butyryl-CoA:acetate-CoA transferase (But, 3.24-7.64 U mg(-1) protein), were determined in cell free extracts of biogas reactor content from three different biogas reactors. Furthermore, the detection of Ack was successful via Western blot analysis. Quantification of corresponding functional genes encoding Buk (buk) and But (but) was not feasible, although an amplification was possible. Thus, phylogenetic trees were constructed based on respective gene fragments. Four new clades of possible butyrate-producing bacteria were postulated, as well as bacteria of the genera Roseburia or Clostridium identified. The low Buk activity was in contrast to the high specific But activity in the analysed samples. Butyrate formation via Buk activity does barely occur in the investigated biogas reactor. Specific enzyme activities (Ack, Buk and But) in samples drawn from three different biogas reactors correlated with ammonia and ammonium concentrations (NH₃ and NH₄(+)-N), and a negative dependency can be postulated. Thus, high concentrations of NH₃ and NH₄(+)-N may lead to a bottleneck in acidogenesis due to decreased specific acidogenic enzyme activities.
Thermal-hydraulics and safety analysis of sectored compact reactor for lunar surface power
Energy Technology Data Exchange (ETDEWEB)
Schriener, T. M. [Inst. for Space and Nuclear Power Studies, Univ. of New Mexico, Albuquerque, NM (United States); Chemical and Nuclear Engineering Dept., Univ. of New Mexico, Albuquerque, NM (United States); El-Genk, M. S. [Inst. for Space and Nuclear Power Studies, Univ. of New Mexico, Albuquerque, NM (United States); Chemical and Nuclear Engineering Dept., Univ. of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Dept., Univ. of New Mexico, Albuquerque, NM (United States)
2012-07-01
The liquid NaK-cooled, fast-neutron spectrum, Sectored Compact Reactor (SCoRe-N 5) concept has been developed at the Univ. of New Mexico for lunar surface power applications. It is loaded with highly enriched UN fuel pins in a triangular lattice, and nominally operates at exit and inlet coolant temperatures of 850 K and 900 K. This long-life reactor generates up to 1 MWth continuously for {>=} 20 years. To avoid a single point failure in reactor cooling, the core is divided into 6 sectors that are neutronically and thermally coupled, but hydraulically independent. This paper performs a 3-D the thermal-hydraulic analysis of SCoRe--N 5 at nominal operation temperatures and a power level of 1 MWth. In addition, the paper investigates the potential of continuing reactor operation at a lower power in the unlikely event that one sector in the core experiences a loss of coolant (LOC). Redesigning the core with a contiguous steel matrix enhances the cooling of the sector experiencing a LOC. Results show that with a core sector experiencing a LOC, SCORE-N 5 could continue operating safely at a reduced power of 166.6 kWth. (authors)
Evans, Michael N
2008-07-01
A reactor for converting cellulose into carbon monoxide for subsequent oxygen isotopic analysis via continuous flow isotope ratio mass spectrometry is described. The system employs an induction heater to produce temperatures >or=1500 degrees C within a molybdenum foil crucible positioned by boron nitride (BN) spacers within a quartz outer sleeve. For samples of a homogeneous working standard cellulose between 300 and 400 microg in size, the blank/signal ratio is <5%, and the long-term precision is 0.30 per thousand (N = 232). For samples of 30 to 100 microg in size, a gas pressure sintered silicon nitride (Si(3)N(4)) outer sleeve replaces the quartz sleeve, the BN spacers are not used, and 6.0-grade carrier He must be used to minimize the blank signal. With these modifications a blank/sample ratio of <5% and long-term precision of 0.30 per thousand (N = 144) are obtained. These results are similar to those achieved using standard high-temperature furnaces, but the reactor is simpler to pack, the system is more economical to run, and samples as small as 30 microg cellulose may be measured. For both reactors memory is significant in the subsequent sample and is believed to be due to exchange with reactor oxygen at temperatures above 1000 degrees C. Further applications might include online preparation of other materials requiring temperatures of 1500-2600 degrees C.
Analysis of the key enzymes of butyric and acetic acid fermentation in biogas reactors
Gabris, Christina; Bengelsdorf, Frank R; Dürre, Peter
2015-01-01
This study aimed at the investigation of the mechanisms of acidogenesis, which is a key process during anaerobic digestion. To expose possible bottlenecks, specific activities of the key enzymes of acidification, such as acetate kinase (Ack, 0.23–0.99 U mg−1 protein), butyrate kinase (Buk, biogas reactor content from three different biogas reactors. Furthermore, the detection of Ack was successful via Western blot analysis. Quantification of corresponding functional genes encoding Buk (buk) and But (but) was not feasible, although an amplification was possible. Thus, phylogenetic trees were constructed based on respective gene fragments. Four new clades of possible butyrate-producing bacteria were postulated, as well as bacteria of the genera Roseburia or Clostridium identified. The low Buk activity was in contrast to the high specific But activity in the analysed samples. Butyrate formation via Buk activity does barely occur in the investigated biogas reactor. Specific enzyme activities (Ack, Buk and But) in samples drawn from three different biogas reactors correlated with ammonia and ammonium concentrations (NH3 and NH4+-N), and a negative dependency can be postulated. Thus, high concentrations of NH3 and NH4+-N may lead to a bottleneck in acidogenesis due to decreased specific acidogenic enzyme activities. PMID:26086956
CFD Analysis of the Fuel Temperature in High Temperature Gas-Cooled Reactors
Energy Technology Data Exchange (ETDEWEB)
In, W. K.; Chun, T. H.; Lee, W. J.; Chang, J. H. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
2005-07-01
High temperature gas-cooled reactors (HTGR) have received a renewed interest as potential sources for future energy needs, particularly for a hydrogen production. Among the HTGRs, the pebble bed reactor (PBR) and a prismatic modular reactor (PMR) are considered as the nuclear heat source in Korea's nuclear hydrogen development and demonstration project. PBR uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. PMR uses graphite fuel blocks which contain cylindrical fuel compacts consisting of the fuel particles. The fuel blocks also contain coolant passages and locations for absorber and control material. The maximum fuel temperature in the core hot spot is one of the important design parameters for both PBR and PMR. The objective of this study is to predict the fuel temperature distributions in PBR and PMR using a computational fluid dynamics(CFD) code, CFX-5. The reference reactor designs used in this analysis are PBMR400 and GT-MHR600.
Consequence analysis of core meltdown accidents in liquid metal fast reactor
Energy Technology Data Exchange (ETDEWEB)
Suk, S.D.; Hahn, D. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
2001-07-01
Core disruptive accidents have been investigated at Korea Atomic Energy Research Institute(KAERI) as part of work to demonstrate the inherent and ultimate safety of the conceptual design of the Korea Advanced Liquid Metal Reactor(KALIMER), a 150 Mw pool-type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. In this study, a simple method was developed using a modified Bethe-Tait method to simulate the kinetics and hydraulic behavior of a homogeneous spherical core over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion. Calculations of energy release during excursions in the sodium-voided core of the KALIMER were subsequently performed using the method for various reactivity insertion rates up to 100 $/s, which has been widely considered to be the upper limit of ramp rates due to fuel compaction. Benchmark calculations were made to compare with the results of more detailed analysis for core meltdown energetics of the oxide fuelled fast reactor. A set of parametric studies was also performed to investigate the sensitivity of the results on the various thermodynamics and reactor parameters. (author)
Energy Technology Data Exchange (ETDEWEB)
Muswema, J.L., E-mail: jeremie.muswem@unikin.ac.cd [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Ekoko, G.B. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Lukanda, V.M. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Democratic Republic of the Congo' s General Atomic Energy Commission, P.O. Box AE1 (Congo, The Democratic Republic of the); Lobo, J.K.-K. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Darko, E.O. [Radiation Protection Institute, Ghana Atomic Energy Commission, P.O. Box LG 80, Legon, Accra (Ghana); Boafo, E.K. [University of Ontario Institute of Technology, 2000 Simcoe St. North, Oshawa, ONL1 H7K4 (Canada)
2015-01-15
Highlights: • Atmospheric dispersion modeling for two credible accidents of the TRIGA Mark II research reactor in Kinshasa (TRICO II) was performed. • Radiological safety analysis after the postulated initiating events (PIE) was also carried out. • The Karlsruhe KORIGEN and the HotSpot Health Physics codes were used to achieve the objectives of this study. • All the values of effective dose obtained following the accident scenarios were below the regulatory limits for reactor staff members and the public, respectively. - Abstract: The source term from the 1 MW TRIGA Mark II research reactor core of the Democratic Republic of the Congo was derived in this study. An atmospheric dispersion modeling followed by radiation dose calculation were performed based on two possible postulated accident scenarios. This derivation was made from an inventory of peak radioisotope activities released in the core by using the Karlsruhe version of isotope generation code KORIGEN. The atmospheric dispersion modeling was performed with HotSpot code, and its application yielded to radiation dose profile around the site using meteorological parameters specific to the area under study. The two accident scenarios were picked from possible accident analyses for TRIGA and TRIGA-fueled reactors, involving the case of destruction of the fuel element with highest activity release and a plane crash on the reactor building as the worst case scenario. Deterministic effects of these scenarios are used to update the Safety Analysis Report (SAR) of the reactor, and for its current version, these scenarios are not yet incorporated. Site-specific meteorological conditions were collected from two meteorological stations: one installed within the Atomic Energy Commission and another at the National Meteorological Agency (METTELSAT), which is not far from the site. Results show that in both accident scenarios, radiation doses remain within the limits, far below the recommended maximum effective
Robinson 2 reactor vessel: pressurized thermal shock analysis for a small-break LOCA
Energy Technology Data Exchange (ETDEWEB)
Marston, T.; Griesbach, T.; Chao, J.; Chexal, B.; Norris, D.; Nickell, B.; Layman, B.
1984-08-01
A best-estimate Pressurized Thermal Shock (PTS) analysis was performed for a three-inch diameter hot-leg small-break loss-of-coolant accident for the Robinson 2 plant. This plant specific analysis was performed using EPRI's linked set of codes for PTS analysis. The analysis shows that with the H.B. Robinson 2 reactor pressure vessel, a hot-leg small-break loss-of-coolant accident does not pose a significant health or safety concern to the public for at least 40 years of operation.
Development of a steady thermal-hydraulic analysis code for the China Advanced Research Reactor
Institute of Scientific and Technical Information of China (English)
TIAN Wenxi; QIU Suizheng; GUO Yun; SU Guanghui; JIA Dounan; LIU Tiancai; ZHANG Jianwei
2007-01-01
A multi-channel model steady-state thermalhydraulic analysis code was developed for the China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed mass flow distribution in the core was obtained. The result shows that structure size plays the most important role in mass flow distribution, and the influence of core power could be neglected under singlephase flow. The temperature field of the fuel element under unsymmetrical cooling condition was also obtained, which is necessary for further study such as stress analysis, etc. Of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of the mean and hot channel was carried out and it is proved that all thermal-hydraulic parameters satisfy the "Safety design regulation of CARR".
A Flooding Induced Station Blackout Analysis for a Pressurized Water Reactor Using the RISMC Toolkit
Directory of Open Access Journals (Sweden)
Diego Mandelli
2015-01-01
Full Text Available In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code called NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. In addition, the impact of power uprate is determined in terms of both core damage probability and safety margins.
Integrated physics analysis of plasma start-up scenario of helical reactor FFHR-d1
Goto, T.; Miyazawa, J.; Sakamoto, R.; Seki, R.; Suzuki, C.; Yokoyama, M.; Satake, S.; Sagara, A.; The FFHR Design Group
2015-06-01
1D physics analysis of the plasma start-up scenario of the large helical device (LHD)-type helical reactor FFHR-d1 was conducted. The time evolution of the plasma profile is calculated using a simple model based on the LHD experimental observations. A detailed assessment of the magnetohydrodynamic equilibrium and neo-classical energy loss was conducted using the integrated transport analysis code TASK3D. The robust controllability of the fusion power was confirmed by feedback control of the pellet fuelling and a simple staged variation of the external heating power with a small number of simple diagnostics (line-averaged electron density, edge electron density and fusion power). A baseline operation control scenario (plasma start-up and steady-state sustainment) of the FFHR-d1 reactor for both self-ignition and sub-ignition operation modes was demonstrated.
Bhunia, Puspendu; Ghangrekar, M M
2008-05-01
Studies have been undertaken to explore the applicability of different kinetic models for the performance appraisal of upflow anaerobic sludge blanket (UASB) reactors treating wastewater in the range of 300-4000 mg COD/l. Three kinetic models namely, Monod, Grau second-order, and Haldane model are considered for the analysis. Both linear and nonlinear regressions have been performed to examine the best-fit among the kinetic models. In this process, five error analysis methods have been used to analyze the data. Apart from optimization of kinetic coefficients with minimization of associated errors, prediction of effluent COD has also been undertaken to verify the applicability of kinetic models. In both the cases, Grau second-order model is found to be the best class of fit for wide range of data sets in UASB reactor.
Yang, P.; Ng, T. L.; Yang, W.
2015-12-01
Effective water resources management depends on the reliable estimation of the uncertainty of drought events. Confidence intervals (CIs) are commonly applied to quantify this uncertainty. A CI seeks to be at the minimal length necessary to cover the true value of the estimated variable with the desired probability. In drought analysis where two or more variables (e.g., duration and severity) are often used to describe a drought, copulas have been found suitable for representing the joint probability behavior of these variables. However, the comprehensive assessment of the parameter uncertainties of copulas of droughts has been largely ignored, and the few studies that have recognized this issue have not explicitly compared the various methods to produce the best CIs. Thus, the objective of this study to compare the CIs generated using two widely applied uncertainty estimation methods, bootstrapping and Markov Chain Monte Carlo (MCMC). To achieve this objective, (1) the marginal distributions lognormal, Gamma, and Generalized Extreme Value, and the copula functions Clayton, Frank, and Plackett are selected to construct joint probability functions of two drought related variables. (2) The resulting joint functions are then fitted to 200 sets of simulated realizations of drought events with known distribution and extreme parameters and (3) from there, using bootstrapping and MCMC, CIs of the parameters are generated and compared. The effect of an informative prior on the CIs generated by MCMC is also evaluated. CIs are produced for different sample sizes (50, 100, and 200) of the simulated drought events for fitting the joint probability functions. Preliminary results assuming lognormal marginal distributions and the Clayton copula function suggest that for cases with small or medium sample sizes (~50-100), MCMC to be superior method if an informative prior exists. Where an informative prior is unavailable, for small sample sizes (~50), both bootstrapping and MCMC
Monte Carlo analysis of an ODE Model of the Sea Urchin Endomesoderm Network
Directory of Open Access Journals (Sweden)
Klipp Edda
2009-08-01
Full Text Available Abstract Background Gene Regulatory Networks (GRNs control the differentiation, specification and function of cells at the genomic level. The levels of interactions within large GRNs are of enormous depth and complexity. Details about many GRNs are emerging, but in most cases it is unknown to what extent they control a given process, i.e. the grade of completeness is uncertain. This uncertainty stems from limited experimental data, which is the main bottleneck for creating detailed dynamical models of cellular processes. Parameter estimation for each node is often infeasible for very large GRNs. We propose a method, based on random parameter estimations through Monte-Carlo simulations to measure completeness grades of GRNs. Results We developed a heuristic to assess the completeness of large GRNs, using ODE simulations under different conditions and randomly sampled parameter sets to detect parameter-invariant effects of perturbations. To test this heuristic, we constructed the first ODE model of the whole sea urchin endomesoderm GRN, one of the best studied large GRNs. We find that nearly 48% of the parameter-invariant effects correspond with experimental data, which is 65% of the expected optimal agreement obtained from a submodel for which kinetic parameters were estimated and used for simulations. Randomized versions of the model reproduce only 23.5% of the experimental data. Conclusion The method described in this paper enables an evaluation of network topologies of GRNs without requiring any parameter values. The benefit of this method is exemplified in the first mathematical analysis of the complete Endomesoderm Network Model. The predictions we provide deliver candidate nodes in the network that are likely to be erroneous or miss unknown connections, which may need additional experiments to improve the network topology. This mathematical model can serve as a scaffold for detailed and more realistic models. We propose that our method can
Energy Technology Data Exchange (ETDEWEB)
Lange, Carsten; Hurtado, Antonio [Technische Univ. Dresden (Germany). Chair of Hydrogen and Nuclear Energy; Schuster, Roland [Kernkraftwerk Brunsbuettel GmbH und Co. oHG, Brunsbuettel (Germany); Lukas, Bernard [EnBW Kernkraft GmbH, Philippsburg (Germany). Kernkraftwerk Philippsburg; Aguirre, Carlos [Kernkraftwerk Leibstadt AG, Aargau (Switzerland); Hennig, Dieter
2012-11-15
Modern theoretical methods for analysing the stability behaviour of Boiling Water Reactors (BWRs) are relatively reliable. The analysis is performed by comprehensive validated system codes comprising 3D core models and one-dimensional thermal-hydraulic parallel channel models in the frequency (linearized models) or time domain. Nevertheless the spontaneous emergence of stable or unstable periodic orbits as solutions of the coupled nonlinear differential equations determining the stability properties of the coupled thermal-hydraulic and neutron kinetic (highly) nonlinear BWR system is a surprising phenomenon, and it is worth thinking about the mathematical background controlling such behaviour. In particular the coexistence of different types of solutions, such as the coexistence of unstable limit cycles and stable fixed points, are states of stability, not all nuclear engineers are familiar with. Hence the part I of this paper is devoted to the mathematical background of linear and nonlinear stability analysis and introduces a novel efficient approach to treat the nonlinear BWR stability behaviour with both system codes and so-called (advanced) reduced order models (ROMs). The efficiency of this approach, called the RAM-ROM method, will be demonstrated by some results of stability analyses for different power plants. (orig.)
Fluid Dynamic Analysis of Shaken and Mechanically Stirred Reactors by Means of Optical Techniques
Pieralisi, Irene
2016-01-01
This research aims at deepening the knowledge of the hydrodynamics developing within some specific mixing system configurations. In particular, a standard geometry stirred vessel, an unbaffled stirred vessel of unconventional geometry and a shaken bioreactor are studied. Optical techniques (LDA, PIV) are employed in order to obtain detailed information on the mean and turbulent characteristics of the flow developing within the different lab-scale reactors, while subsequent data analysis a...
Menke, M. M.; Judd, B. R.
1973-01-01
The development policy for thermionic reactors to provide electric propulsion and power for space exploration was analyzed to develop a logical procedure for selecting development alternatives that reflect the technical feasibility, JPL/NASA project objectives, and the economic environment of the project. The partial evolution of a decision model from the underlying philosophy of decision analysis to a deterministic pilot phase is presented, and the general manner in which this decision model can be employed to examine propulsion development alternatives is illustrated.
Analysis of dashpot performance for rotating control drums of a lithium cooled fast reactor concept
Wenzler, C. J.
1972-01-01
A dashpot was incorporated in the design of the drive train of the rotating control drum to prevent shock damage to the control drum and drive train at the termination of a scram action. A rotating vane dashpot using reactor coolant lithium as a damping fluid appears to be the best candidate of the various damping devices explored. A performance analysis, results and discussion of vane type dashpots are presented.
Energy Technology Data Exchange (ETDEWEB)
Brandt, Gerhard [University of Oxford (United Kingdom); Krauss, Frank [IPPP Durham (United Kingdom); Lacker, Heiko; Leyton, Michael; Mamach, Martin; Schulz, Holger; Weyh, Daniel [Humboldt University of Berlin (Germany)
2012-07-01
Monte Carlo (MC) event generators are widely employed in the analysis of experimental data also for LHC in order to predict the features of observables and test analyses with them. These generators rely on phenomenological models containing various parameters which are free in certain ranges. Variations of these parameters relative to their default lead to uncertainties on the predictions of the event generators and, in turn, on the results of any experimental data analysis making use of the event generator. A Generalized method for quantifying a certain class of these generator based uncertainties will be presented in this talk. We study for the SHERPA event generator the effect on the analysis results from uncertainties in the choice of the merging and factorization scale. The quantification is done within an example ATLAS analysis measuring underlying event UE properties in Z-boson production limited to low transverse momenta (p{sub T}{sup Z}<3 GeV) of the Z-boson. The analysis extracts event-shape distributions from charged particles in the event that do not belong to the Z decay for generate Monte Carlo event and data which are unfolded back to the generator level.
Accident progression event tree analysis for postulated severe accidents at N Reactor
Energy Technology Data Exchange (ETDEWEB)
Wyss, G.D.; Camp, A.L.; Miller, L.A.; Dingman, S.E.; Kunsman, D.M. (Sandia National Labs., Albuquerque, NM (USA)); Medford, G.T. (Science Applications International Corp., Albuquerque, NM (USA))
1990-06-01
A Level II/III probabilistic risk assessment (PRA) has been performed for N Reactor, a Department of Energy (DOE) production reactor located on the Hanford reservation in Washington. The accident progression analysis documented in this report determines how core damage accidents identified in the Level I PRA progress from fuel damage to confinement response and potential releases the environment. The objectives of the study are to generate accident progression data for the Level II/III PRA source term model and to identify changes that could improve plant response under accident conditions. The scope of the analysis is comprehensive, excluding only sabotage and operator errors of commission. State-of-the-art methodology is employed based largely on the methods developed by Sandia for the US Nuclear Regulatory Commission in support of the NUREG-1150 study. The accident progression model allows complex interactions and dependencies between systems to be explicitly considered. Latin Hypecube sampling was used to assess the phenomenological and systemic uncertainties associated with the primary and confinement system responses to the core damage accident. The results of the analysis show that the N Reactor confinement concept provides significant radiological protection for most of the accident progression pathways studied.
Pumps modelling of a sodium fast reactor design and analysis of hydrodynamic behavior
Directory of Open Access Journals (Sweden)
Ordóñez Ródenas José
2016-01-01
Full Text Available One of the goals of Generation IV reactors is to increase safety from those of previous generations. Different research platforms have been identified the need to improve the reliability of the simulation tools to ensure the capability of the plant to accommodate the design basis transients established in preliminary safety studies. The paper describes the modelling of primary pumps in advanced sodium cooled reactors using the TRACE code. Following the implementation of the models, the results obtained in the analysis of different design basis transients are compared with the simplifying approximations used in reference models. The paper shows the process to obtain a consistent pump model of the ESFR (European Sodium Fast Reactor design and the analysis of loss of flow transients triggered by pumps coast–down analyzing the thermal hydraulic neutronic coupled system response. A sensitivity analysis of the system pressure drops effect and the other relevant parameters that influence the natural convection after the pumps coast–down is also included.
N-Reactor (U-metal) Fuel Characteristics for Disposal Criticality Analysis
Energy Technology Data Exchange (ETDEWEB)
Taylor, Larry Lorin
2000-05-01
DOE-owned spent nuclear fuels encompass many fuel types. In an effort to facilitate criticality analysis for these various fuel types, they were categorized into nine characteristic fuel groups with emphasis on fuel matrix composition. Out of each fuel group, a representative fuel type was chosen for analysis as a bounding case within that fuel group. Generally, burnup data, fissile enrichments, and total fuel and fissile mass govern the selection of the representative or candidate fuel within that group. Additionally, the criticality analysis will also require data to support design of the canister internals, thermal, and radiation shielding. The purpose of this report is to consolidate and provide in a concise format, material and information/data needed to perform supporting analyses to qualify N-Reactor fuels for acceptance into the designated repository. The N Reactor fuels incorporate zirconium cladding and uranium metal with unique fabrication details in terms of physical size, and method of construction. The fuel construction and post-irradiation handling have created attendant issues relative to cladding failure in the underwater storage environment. These fuels were comprised of low-enriched metal (0.947 to 1.25 wt% 235U) that were originally intended to generate weapons-grade plutonium for national defense. Modifications in subsequent fuel design and changes in the mode of reactor operation in later years were focused more toward power production.
An analysis on the theory of pulse oximetry by Monte Carlo simulation
Fan, Shangchun; Cai, Rui; Xing, Weiwei; Liu, Changting; Chen, Guangfei; Wang, Junfeng
2008-10-01
The pulse oximetry is a kind of electronic instrument that measures the oxygen saturation of arterial blood and pulse rate by non-invasive techniques. It enables prompt recognition of hypoxemia. In a conventional transmittance type pulse oximeter, the absorption of light by oxygenated and reduced hemoglobin is measured at two wavelength 660nm and 940nm. But the accuracy and measuring range of the pulse oximeter can not meet the requirement of clinical application. There are limitations in the theory of pulse oximetry, which is proved by Monte Carlo method. The mean paths are calculated in the Monte Carlo simulation. The results prove that the mean paths are not the same between the different wavelengths.
A Game Theory Based on Monte Carlo Analysis for Optimizing Evacuation Routing in Complex Scenes
Directory of Open Access Journals (Sweden)
Wenhui Li
2015-01-01
Full Text Available With more complex structures and denser populations, congestion is a crucial factor in estimating evacuation clearance time. This paper presents a novel evacuation model that implements a game theory combining the greatest entropy optimization criterion with stochastic Monte Carlo methods to optimize the congestion problem and other features of emergency evacuation planning. We introduce the greatest entropy criterion for convergence to Nash equilibrium in the n-person noncooperative game. The process of managing the conflict problem is divided into two steps. In the first step, we utilize Monte Carlo methods to evaluate the risk degree of each route. In the second step, we propose an improved method based on game theory, which obtains an optimal solution to guide the evacuation of all agents from the building.
Evaluation of CASMO-3 and HELIOS for Fuel Assembly Analysis from Monte Carlo Code
Energy Technology Data Exchange (ETDEWEB)
Shim, Hyung Jin; Song, Jae Seung; Lee, Chung Chan
2007-05-15
This report presents a study comparing deterministic lattice physics calculations with Monte Carlo calculations for LWR fuel pin and assembly problems. The study has focused on comparing results from the lattice physics code CASMO-3 and HELIOS against those from the continuous-energy Monte Carlo code McCARD. The comparisons include k{sub inf}, isotopic number densities, and pin power distributions. The CASMO-3 and HELIOS calculations for the k{sub inf}'s of the LWR fuel pin problems show good agreement with McCARD within 956pcm and 658pcm, respectively. For the assembly problems with Gadolinia burnable poison rods, the largest difference between the k{sub inf}'s is 1463pcm with CASMO-3 and 1141pcm with HELIOS. RMS errors for the pin power distributions of CASMO-3 and HELIOS are within 1.3% and 1.5%, respectively.
Monte Carlo analysis of a low power domino gate under parameter fluctuation
Institute of Scientific and Technical Information of China (English)
Wang Jinhui; Wu Wuchen; Gong Na; Hou Ligang; Peng Xiaohong; Gao Daming
2009-01-01
Using the multiple-parameter Monte Carlo method, the effectiveness of the dual threshold voltage technique (DTV) in low power domino logic design is analyzed. Simulation results indicate that under significant temperature and process fluctuations, DTV is still highly effective in reducing the total leakage and active power consumption for domino gates with speed loss. Also, regarding power and delay characteristics, different structure domino gates with DTV have different robustness against temperature and process fluctuation.
Monte Carlo analysis of a low power domino gate under parameter fluctuation
Energy Technology Data Exchange (ETDEWEB)
Wang Jinhui; Wu Wuchen; Hou Ligang; Peng Xiaohong; Gao Daming [VLSI and System Laboratory, Beijing University of Technology, Beijing 100124 (China); Gong Na, E-mail: wangjinhui888@emails.bjut.edu.c [Department of Computer Science and Engineering, State University of New York at Buffalo, Buffalo 14260, NY (United States)
2009-12-15
Using the multiple-parameter Monte Carlo method, the effectiveness of the dual threshold voltage technique (DTV) in low power domino logic design is analyzed. Simulation results indicate that under significant temperature and process fluctuations, DTV is still highly effective in reducing the total leakage and active power consumption for domino gates with speed loss. Also, regarding power and delay characteristics, different structure domino gates with DTV have different robustness against temperature and process fluctuation. (semiconductor integrated circuits)
MONTE CARLO ANALYSIS FOR PREDICTION OF NOISE FROM A CONSTRUCTION SITE
Directory of Open Access Journals (Sweden)
Zaiton Haron
2009-06-01
Full Text Available The large number of operations involving noisy machinery associated with construction site activities result in considerable variation in the noise levels experienced at receiver locations. This paper suggests an approach to predict noise levels generated from a site by using a Monte Carlo approach. This approach enables the determination of details regarding the statistical uncertainties associated with noise level predictions or temporal distributions. This technique could provide the basis for a generalised prediction technique and a simple noise management tool.
Monte Carlo analysis of Gunn oscillations in narrow and wide band-gap asymmetric nanodiodes
González, T.; Iñiguez-de-la Torre, I.; Pardo, D.; Mateos, J.; Song, A. M.
2009-11-01
By means of Monte Carlo simulations we show the feasibility of asymmetric nonlinear planar nanodiodes for the development of Gunn oscillations. For channel lengths about 1 μm, oscillation frequencies around 100 GHz are predicted in InGaAs diodes, being significantly higher, around 400 GHz, in the case of GaN structures. The DC to AC conversion efficiency is found to be higher than 1% for the fundamental and second harmonic frequencies in GaN diodes.
Energy Technology Data Exchange (ETDEWEB)
Yokoyama, Kenji; Hazama, Taira; Chiba, Go; Ohki, Shigeo; Ishikawa, Makoto [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center
2002-12-01
In the core design of fast reactors (FRs), it is very important to improve the prediction accuracy of the nuclear characteristics for both reducing cost and ensuring reliability of FR plants. A nuclear reactor analysis code system for FRs has been developed by the Japan Nuclear Cycle Development Institute (JNC). This paper describes the outline of the calculation models and methods in the system consisting of several analysis codes, such as the cell calculation code CASUP, the core calculation code TRITAC and the sensitivity analysis code SAGEP. Some examples of verification results and improvement of the design accuracy are also introduced based on the measurement data from critical assemblies, e.g, the JUPITER experiment (USA/Japan), FCA (Japan), MASURCA (France), and BFS (Russia). Furthermore, application fields and future plans, such as the development of new generation nuclear constants and applications to MA{center_dot}FP transmutation, are described. (author)
A Markov chain Monte Carlo method family in incomplete data analysis
Directory of Open Access Journals (Sweden)
Vasić Vladimir V.
2003-01-01
Full Text Available A Markov chain Monte Carlo method family is a collection of techniques for pseudorandom draws out of probability distribution function. In recent years, these techniques have been the subject of intensive interest of many statisticians. Roughly speaking, the essence of a Markov chain Monte Carlo method family is generating one or more values of a random variable Z, which is usually multidimensional. Let P(Z = f(Z denote a density function of a random variable Z, which we will refer to as a target distribution. Instead of sampling directly from the distribution f, we will generate [Z(1, Z(2..., Z(t,... ], in which each value is, in a way, dependant upon the previous value and where the stationary distribution will be a target distribution. For a sufficient value of t, Z(t will be approximately random sampling of the distribution f. A Markov chain Monte Carlo method family is useful when direct sampling is difficult, but when sampling of each value is not.
2010-03-22
... New Reactors Based on Probabilistic Risk Assessment AGENCY: Nuclear Regulatory Commission (NRC... Probabilistic Risk Assessment,'' (Agencywide Documents Access and Management System (ADAMS) Accession No..., ``Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,'' issued...
Rodriguez, E.; Lopes, A.; Fdz-Polanco, M.; Stams, A.J.M.; Garcia Encina, P.A.
2012-01-01
The microbial communities (Bacteria and Archaea) established in an anaerobic fluidized bed reactor used to treat synthetic vinasse (betaine, glucose, acetate, propionate, and butyrate) were characterized by denaturing gradient gel electrophoresis (DGGE) and phylogenetic analysis. This study was focu
Uncertainty Analysis of Light Water Reactor Fuel Lattices
Directory of Open Access Journals (Sweden)
C. Arenas
2013-01-01
Full Text Available The study explored the calculation of uncertainty based on available cross-section covariance data and computational tool on fuel lattice levels, which included pin cell and the fuel assembly models. Uncertainty variations due to temperatures changes and different fuel compositions are the main focus of this analysis. Selected assemblies and unit pin cells were analyzed according to the OECD LWR UAM benchmark specifications. Criticality and uncertainty analysis were performed using TSUNAMI-2D sequence in SCALE 6.1. It was found that uncertainties increase with increasing temperature, while kinf decreases. This increase in the uncertainty is due to the increase in sensitivity of the largest contributing reaction of uncertainty, namely, the neutron capture reaction 238U(n, γ due to the Doppler broadening. In addition, three types (UOX, MOX, and UOX-Gd2O3 of fuel material compositions were analyzed. A remarkable increase in uncertainty in kinf was observed for the case of MOX fuel. The increase in uncertainty of kinf in MOX fuel was nearly twice the corresponding value in UOX fuel. The neutron-nuclide reaction of 238U, mainly inelastic scattering (n, n′, contributed the most to the uncertainties in the MOX fuel, shifting the neutron spectrum to higher energy compared to the UOX fuel.
Chham, E; El Bardouni, T; Benaalilou, K; Boukhal, H; El Bakkari, B; Boulaich, Y; El Younoussi, C; Nacir, B
2016-10-01
This study was conducted to improve the capacity of radioisotope production in the Moroccan TRIGA Mark II research reactor, which is considered as one of the most important applications of research reactors. The aim of this study is to enhance the utilization of TRIGA core in the field of neutron activation and ensure an economic use of the fuel. The main idea was to create an additional irradiation channel (IC) inside the core. For this purpose, three new core configurations are proposed, which differ according to the IC position in the core. Thermal neutron flux distribution and other neutronic safety parameters such as power peaking factors, excess reactivity, and control rods worth reactivity were calculated using the Monte Carlo N-Particle Transport (MCNP) code and neutron cross-section library based on ENDF/B-VII evaluation. The calculated thermal flux in the central thimble (CT) and in the added IC for the reconfigured core is compared with the thermal flux in the CT of the existing core, which is taken as a reference. The results show that all the obtained fluxes in CTs are very close to the reference value, while a remarkable difference is observed between the fluxes in the new ICs and reference. This difference depends on the position of IC in the reactor core. To demonstrate that the Moroccan TRIGA reactor could safely operate at 2MW, with new configurations based on new ICs, different safety-related thermal-hydraulic parameters were investigated. The PARET model was used in this study to verify whether the safety margins are met despite the new modifications of the core. The results show that it is possible to introduce new ICs safely in the reactor core, because the obtained values of the parameters are largely far from compromising the safety of the reactor.
Analysis of Multiple Spurious Operation Scenarios for Decay Heat Removal Function of CANDU Reactors
Energy Technology Data Exchange (ETDEWEB)
Lee, Youngseung; Bae, Yeon-kyoung; Kim, Myungsu [KHNP CRI, Daejeon (Korea, Republic of)
2016-10-15
The worst fire broke out in the Browns Ferry Nuclear Power Plant on March 22, 1975. A fire occurrence in a nuclear power plant has recognized a latently serious incident. Nuclear power plants should achieve and maintain the safe shutdown conditions during and after the occurrence of a fire. Functions of the safe shutdown are five such as the shutdown function, the decay heat removal function, the containment function, monitoring and control function, and the supporting function for CANDU type reactors. The purpose of this paper is to analyze that the decay heat removal function of the safe shutdown functions for CANDU type reactors is achieved under the fire induced multiple spurious operation. The scenarios of the fire induced multiple spurious operations (MSO) for the systems used for the decay heat cooling were analyzed. Additionally, Integrated Severe Accident Analysis code for CANDU plants (ISAAC) for determining success criteria of thermal hydraulic analysis was used. Decay heat cooling systems of CANDU reactors are the auxiliary feedwater system, the emergency water supply system, and the shutdown cooling system. A big fire can threat the safety of nuclear power plants, and safe shutdown conditions. The regulatory body in Korea requires the fire hazard analysis including fire induced MSOs. The safe shutdown functions for CANDU reactors are the shutdown function, the decay heat removal function, the containment function, the monitoring and control function, and the supporting service function. The number of spurious operations for the auxiliary feedwater system is more than six and that for the emergency water supply system is one. Additionally, misoperations for the shutdown cooling system are more than two. Accordingly, if total nine components could be spuriously operated, the decay heat removal function would be lost entirely.
The pressurization transient analysis for Lungmen advanced boiling water reactor using RETRAN-02
Energy Technology Data Exchange (ETDEWEB)
Tsai, C.-W., E-mail: d937121@oz.nthu.edu.t [Department of Engineering and System Science, National Tsing Hua University, No. 101, Sec. 2, Kuang Fu Road, Hsinchu 30013, Taiwan (China); Shih Chunkuan [Department of Engineering and System Science, National Tsing Hua University, No. 101, Sec. 2, Kuang Fu Road, Hsinchu 30013, Taiwan (China); Institute of Nuclear Engineering and Science, National Tsing Hua University, No. 101, Sec. 2, Kuang Fu Road, Hsinchu 30013, Taiwan (China); Wang, J.-R.; Lin, H.-T. [Institute of Nuclear Energy Research, No. 1000, Wenhua Rd., Longtan Township, Taoyuan County 32546, Taiwan (China); Cheng, S.-C. [Department of Nuclear Engineering, Taiwan Power Company, No. 242, Sec. 3, Roosevelt Rd., Taipei City 10016, Taiwan (China)
2010-10-15
A RETRAN-02 model was devised and benchmarked against the preliminary safety analysis report (PSAR) for the Lungmen nuclear power plant roughly 10 years ago. During these years, the fuel design, some of the reactor vessel designs, and control systems have since been revised. The Lungmen RETRAN-02 model has also been modified with updated information when available. This study uses the analytical results of the final safety analysis report (FSAR) to benchmark the Lungmen RETRAN-02 plant model. Five transients, load rejection (LR), turbine trip (TT), main steam line isolation valves closure (MSIVC), loss of feedwater flow (LOFF), and one turbine control valve closure (OTCVC), were utilized to validate the Lungmen RETRAN-02 model. Moreover, due to the strong coupling effect between neutron dynamics and the thermal-hydraulic response during pressurization of transients, the one-dimensional kinetic model with the cross-section data library is used to simulate the coupling effect. The analytical results show good agreement in trends between the RETRAN-02 calculation and the Lungmen FSAR data. Based on the benchmark of these design-basis transients, the modified Lungmen RETRAN-02 model has been adjusted to a level of confidence for analysis of pressure increase transients. Analytical results indicate that the Lungmen advanced boiling water reactor (ABWR) design satisfied design criteria, i.e., vessel pressure and hot shutdown capability. However, a slight difference exists in the simulation of the water level for cases with changes in water levels. The Lungmen RETRAN-02 model tends to predict the change in water level at a slower rate than that in the Lungmen FSAR. There is also a slight difference in void reactivity response toward vessel pressure change in both simulations, which causes the calculated neutron flux before reactor shutdown to differ to some degree when the reactor experiences a rapid pressure increase. Further studies will be performed in the future using
Energy Technology Data Exchange (ETDEWEB)
Love, E.F.; Pauley, K.A.; Reid, B.D.
1995-09-01
This study describes the use of the Monte Carlo Neutron-Photon (MCNP) code for determining activation levels of irradiated reactor vessel internals hardware. The purpose of the analysis is to produce data for the Department of Energy`s Greater-Than-Class C Low-Level Radioactive Waste Program. An MCNP model was developed to analyze the Yankee Rowe reactor facility. The model incorporates reactor geometry, material compositions, and operating history data acquired from Yankee Atomic Electric Company. In addition to the base activation analysis, parametric studies were performed to determine the sensitivity of activation to specific parameters. A component sampling plan was also developed to validate the model results, although the plan was not implemented. The calculations for the Yankee Rowe reactor predict that only the core baffle and the core support plates will be activated to levels above the Class C limits. The parametric calculations show, however, that the large uncertainties in the material compositions could cause errors in the estimates that could also increase the estimated activation level of the core barrel to above the Class C limits. Extrapolation of the results to other reactor facilities indicates that in addition to the baffle and support plates, core barrels may also be activated to above Class C limits; however the classification will depend on the specific operating conditions of the reactor and the specific material compositions of the metal, as well as the use of allowable concentration averaging practices in packaging and classifying the waste.
Energy Technology Data Exchange (ETDEWEB)
B. Boer; A. M. Ougouag
2010-09-01
The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge
Burnup analysis of the VVER-1000 reactor using thorium-based fuel
Energy Technology Data Exchange (ETDEWEB)
Korkmaz, Mehmet E.; Agar, Osman; Bueyueker, Eylem [Karamanoglu Mehmetbey Univ., Karaman (Turkey). Faculty of Kamil Ozdag Science
2014-12-15
This paper aims to investigate {sup 232}Th/{sup 233}U fuel cycles in a VVER-1000 reactor through calculation by computer. The 3D core geometry of VVER-1000 system was designed using the Serpent Monte Carlo 1.1.19 Code. The Serpent Code using parallel programming interface (Message Passing Interface-MPI), was run on a workstation with 12-core and 48 GB RAM. {sup 232}Th/{sup 235}U/{sup 238}U oxide mixture was considered as fuel in the core, when the mass fraction of {sup 232}Th was increased as 0.05-0.1-0.2-0.3-0.4 respectively, the mass fraction of {sup 238}U equally was decreased. In the system, the calculations were made for 3 000 MW thermal power. For the burnup analyses, the core is assumed to deplete from initial fresh core up to a burnup of 16 MWd/kgU without refuelling considerations. In the burnup calculations, a burnup interval of 360 effective full power days (EFPDs) was defined. According to burnup, the mass changes of the {sup 232}Th, {sup 233}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 241}Am and {sup 244}Cm were evaluated, and also flux and criticality of the system were calculated in dependence of the burnup rate.
Energy Technology Data Exchange (ETDEWEB)
M. G. McKellar; E. A. Harvego; A. M. Gandrik
2010-11-01
An updated reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production has been developed. The HTE plant is powered by a high-temperature gas-cooled reactor (HTGR) whose configuration and operating conditions are based on the latest design parameters planned for the Next Generation Nuclear Plant (NGNP). The current HTGR reference design specifies a reactor power of 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 322°C and 750°C, respectively. The reactor heat is used to produce heat and electric power to the HTE plant. A Rankine steam cycle with a power conversion efficiency of 44.4% was used to provide the electric power. The electrolysis unit used to produce hydrogen includes 1.1 million cells with a per-cell active area of 225 cm2. The reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes a steam-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The overall system thermal-to-hydrogen production efficiency (based on the higher heating value of the produced hydrogen) is 42.8% at a hydrogen production rate of 1.85 kg/s (66 million SCFD) and an oxygen production rate of 14.6 kg/s (33 million SCFD). An economic analysis of this plant was performed with realistic financial and cost estimating The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.03/kg of hydrogen was calculated assuming an internal rate of return of 10% and a debt to equity ratio of 80%/20% for a reactor cost of $2000/kWt and $2.41/kg of hydrogen for a reactor cost of $1400/kWt.
Nuclear data evaluation and group constant generation for reactor analysis
Energy Technology Data Exchange (ETDEWEB)
Kim, Jung Doh; Kil, Chung Sub [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1994-12-01
Data testing of ENDF/B-VI.2 was performed and ACE-format continuous point-wise cross section library from ENDF/B-VI.2 for MCNP was validated through CSEWG benchmark and power plant mockup experiments. The calculated k-effective of ORNL-1, -2, -3, -4 and -10 with ENDF/B-VI are low by about 0.5% but those of L-7, -8, -9, -10 and -11 show good agreement with experiments. Overall results for uranium core with ENDF/B-VI is low in critically than with ENDF/B-V. The calculated results with ENDF/B-VI for PNL-6 {approx} 12 of plutonium core and PNL-30 {approx} 35 of mixed oxide core show good agreement with the experiments. The results of critically calculation for fast core benchmark do not show large difference between ENDF/B-VI and -V. But the calculated results of reaction rate ratio with ENDF/B-VI are improved, compared with ENDF/B-V. The calculated power distribution for VENUS PWR mockup core and typical BWR core of GE with both of ENDF/B-VI and -V agree well with measured values. From the above results, newly generated MCNP library from ENDF/B-VI is useful for nuclear and shielding design and analysis. 5 figs, 13 tabs, 11 refs. (Author).
Seismic Analysis Issues in Design Certification Applications for New Reactors
Energy Technology Data Exchange (ETDEWEB)
Miranda, M.; Morante, R.; Xu, J.
2011-07-17
The licensing framework established by the U.S. Nuclear Regulatory Commission under Title 10 of the Code of Federal Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” provides requirements for standard design certifications (DCs) and combined license (COL) applications. The intent of this process is the early reso- lution of safety issues at the DC application stage. Subsequent COL applications may incorporate a DC by reference. Thus, the COL review will not reconsider safety issues resolved during the DC process. However, a COL application that incorporates a DC by reference must demonstrate that relevant site-specific de- sign parameters are within the bounds postulated by the DC, and any departures from the DC need to be justified. This paper provides an overview of several seismic analysis issues encountered during a review of recent DC applications under the 10 CFR Part 52 process, in which the authors have participated as part of the safety review effort.
Low-cost activation analysis at small research reactors
Westphal, G P; Lemmel, H; Niedermaier, M R; Joestl, K; Schröder, P; Böck, H H; Schachner, H; Klapfer, E
2003-01-01
A software implementation of a loss-free counting multichannel analyzer, storing immediately into the multimegabyte memory of a low-cost 486 or Pentium type PC, enables the real-time control of a rabbit system as well as the collection of up to 1000 pairs of simultaneously recorded loss-corrected and non-corrected spectra of 16 k channels each, in a true sequence without time gaps in between, at throughput rates of up to 200 kc/s. Intended for activation analysis of short-lived isomeric transitions, the system renders possible peak to background optimizations and separations of lines with different half-lives without an a priori knowledge of sample composition by summing up appropriate numbers of spectra over appropriate intervals of time. By automatically adapting the noise filtering time to individual pulse intervals, the Preloaded Digital Filter (PLDF) combines low- to medium-rate resolutions comparable to those of high-quality Gaussian amplifiers with throughput rates of up to 100 kc/s, and high-rate reso...
Analysis of three-dimensional thermo-hydraulic phenomena in the reactor core of LMFBR
Energy Technology Data Exchange (ETDEWEB)
Hu, S.; Lee, Y. B.; Jang, W. P.; Ha, K. S.; Jung, H. Y. [KAERI, Taejon (Korea, Republic of)
2004-07-01
The mismatch between power and flow under the transient condition of LMFBR (Liquid Metal cooled Fast Breeder Reactor) core results in thermal stratification in hot pool. Since the fluid of hot pool enters IHXs, the temperature distribution of hot pool can alter the overall system response, therefore three-dimensional analysis of thermo-hydraulic phenomena is necessary. In this study, the thermo-hydraulic phenomena under normal operating condition and unprotected transient condition of LMFBR is investigated using which is the three-dimensional analysis code, COMMIX-1AR/P. The basic input data is based on the design data of KALIMER-600, which is sodium-cooled fast breeder reactor developed by KAERI. COMMIX-1AR/P code has not a reactivity model and the power and core flowrate must be supplied in the input data. In this study, results of SSC-K calculation is used. The temperature and velocity distributions are calculated and compared with those of SSC-K calculation results. The UTOF(Unprotected Loss Of Flow) accident is calculated using COMMIX-1AR/P and the temperature and velocity distributions in the total reactor core are calculated and the natural circulation mode under this transient condition is investigated.
Energy Technology Data Exchange (ETDEWEB)
Carpenter, D.C.
1998-01-01
This bibliography provides a list of references on finite element and related methods analysis in reactor physics computations. These references have been published in scientific journals, conference proceedings, technical reports, thesis/dissertations and as chapters in reference books from 1971 to the present. Both English and non-English references are included. All references contained in the bibliography are sorted alphabetically by the first author`s name and a subsort by date of publication. The majority of the references relate to reactor physics analysis using the finite element method. Related topics include the boundary element method, the boundary integral method, and the global element method. All aspects of reactor physics computations relating to these methods are included: diffusion theory, deterministic radiation and neutron transport theory, kinetics, fusion research, particle tracking in finite element grids, and applications. For user convenience, many of the listed references have been categorized. The list of references is not all inclusive. In general, nodal methods were purposely excluded, although a few references do demonstrate characteristics of finite element methodology using nodal methods (usually as a non-conforming element basis). This area could be expanded. The author is aware of several other references (conferences, thesis/dissertations, etc.) that were not able to be independently tracked using available resources and thus were not included in this listing.