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Sample records for reactor iter mcnp

  1. Report on Thermal Neutron Diffusion Length Measurement in Reactor Grade Graphite Using MCNP and COMSOL Multiphysics

    CERN Document Server

    Mirfayzi, S R

    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, based on partial and ordinary differential equation. The theoretical work includes numerical approximation methods including transcendental technique to illustrate the iteration process with the FEA method. Finally collision density of thermal neutron in graphite is measured, also specific heat relation dependability of collision density is also calculated theoretically, the thermal neutron diffusion length in graphite is evaluated at $50.85 \\pm 0.3cm$ using COMSOL Multiphysics and $50.95 \\pm 0.5cm$ using MCNP. Finally ...

  2. Report on Thermal Neutron Diffusion Length Measurement in Reactor Grade Graphite Using MCNP and COMSOL Multiphysics

    OpenAIRE

    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...

  3. Use of MCNP for characterization of reactor vessel internals waste from decommissioned nuclear reactors

    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.

  4. Pebble bed reactors simulation using MCNP: The Chinese HTR-10 reactor

    Directory of Open Access Journals (Sweden)

    SA Hosseini

    2013-09-01

    Full Text Available   Given the role of Gas-Graphite reactors as the fourth generation reactors and their recently renewed importance, in 2002 the IAEA proposed a set of Benchmarking problems. In this work, we propose a model both efficient in time and resources and exact to simulate the HTR-10 reactor using MCNP-4C code. During the present work, all of the pressing factors in PBM reactor design such as the inter-pebble leakage, fuel particle distribution and fuel pebble packing fraction effects have been taken into account to obtain an exact and easy to run model. Finally, the comparison between the results of the present work and other calculations made at INEEL proves the exactness of the proposed model.

  5. Application of MCNP for neutronic calculations at VR-1 training reactor

    Science.gov (United States)

    Huml, Ondřej; Rataj, Jan; Bílý, Tomáš

    2014-06-01

    The paper presents utilization of Monte Carlo MCNP transport code for neutronic calculations of training reactor VR-1. Results of calculations are compared with results of measurements realized during last few critical experiments with various reactor core configurations. Very good agreement between calculations and measurements is observed.

  6. The first fusion reactor: ITER

    Science.gov (United States)

    Campbell, D. J.

    2016-11-01

    Established by the signature of the ITER Agreement in November 2006 and currently under construction at St Paul-lez-Durance in southern France, the ITER project [1,2] involves the European Union (including Switzerland), China, India, Japan, the Russian Federation, South Korea and the United States. ITER (`the way' in Latin) is a critical step in the development of fusion energy. Its role is to provide an integrated demonstration of the physics and technology required for a fusion power plant based on magnetic confinement.

  7. A simulation of a pebble bed reactor core by the MCNP-4C computer code

    Directory of Open Access Journals (Sweden)

    Bakhshayesh Moshkbar Khalil

    2009-01-01

    Full Text Available Lack of energy is a major crisis of our century; the irregular increase of fossil fuel costs has forced us to search for novel, cheaper, and safer sources of energy. Pebble bed reactors - an advanced new generation of reactors with specific advantages in safety and cost - might turn out to be the desired candidate for the role. The calculation of the critical height of a pebble bed reactor at room temperature, while using the MCNP-4C computer code, is the main goal of this paper. In order to reduce the MCNP computing time compared to the previously proposed schemes, we have devised a new simulation scheme. Different arrangements of kernels in fuel pebble simulations were investigated and the best arrangement to decrease the MCNP execution time (while keeping the accuracy of the results, chosen. The neutron flux distribution and control rods worth, as well as their shadowing effects, have also been considered in this paper. All calculations done for the HTR-10 reactor core are in good agreement with experimental results.

  8. Assessment of CANDU reactor physics effects using a simplified whole-core MCNP model

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S

    2002-07-01

    A whole-core Monte Carlo n-particle (MCNP) model of a simplified CANDU reactor was developed and used to study core configurations and reactor physics phenomena of interest in CANDU safety analysis. The resulting reactivity data were compared with values derived from corresponding WIMS-AECL/RFSP, two-neutron-energy-group diffusion theory core simulations, thereby extending the range of CANDU-related code-to-code benchmark comparisons to include whole-core representations. These comparisons show a systematic discrepancy of about 6 mk between the respective absolute k{sub eff} values, but very good agreement to within about -0.15 {+-} 0.06 mk for the reactivity perturbation induced by G-core checkerboard coolant voiding. These findings are generally consistent with the results of much simpler uniform-lattice comparisons involving only WIMS-AECL and MCNP. In addition, MCNP fission-energy tallies were used to evaluate other core-wide properties, such as fuel bundle and total-channel power distributions, as well as intra-bundle details, such as outer-fuel-ring relative power densities and outer-ring fuel element azimuthal power variations, which cannot be determined directly from WIMS-AECL/RFSP core calculations. The average MCNP values for the ratio of outer fuel element to average fuel element power density agreed well with corresponding values derived from WIMS-AECL lattice-cell cases, showing a small systematic discrepancy of about 0.5 %, independent of fuel bum-up. For fuel bundles containing the highest-power fuel elements, the maximum peak-to-average outer-element azimuthal power variation was about 2.5% for cases where a statistically significant trend was observed, while much larger peak-to-average outer-element azimuthal power variations of up to around 42% were observed in low-power fuel bundles at the core/radial-neutron-reflector interface. (author)

  9. Studies on the liquid fluoride thorium reactor: Comparative neutronics analysis of MCNP6 code with SRAC95 reactor analysis code based on FUJI-U3-(0)

    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.

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

    Science.gov (United States)

    Kazeminezhad, F.; Anghaie, S.

    2008-01-01

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

  11. Thorium-based mixed oxide fuel in a pressurized water reactor: A feasibility analysis with MCNP

    Science.gov (United States)

    Tucker, Lucas Powelson

    This dissertation investigates techniques for spent fuel monitoring, and assesses the feasibility of using a thorium-based mixed oxide fuel in a conventional pressurized water reactor for plutonium disposition. Both non-paralyzing and paralyzing dead-time calculations were performed for the Portable Spectroscopic Fast Neutron Probe (N-Probe), which can be used for spent fuel interrogation. Also, a Canberra 3He neutron detector's dead-time was estimated using a combination of subcritical assembly measurements and MCNP simulations. Next, a multitude of fission products were identified as candidates for burnup and spent fuel analysis of irradiated mixed oxide fuel. The best isotopes for these applications were identified by investigating half-life, photon energy, fission yield, branching ratios, production modes, thermal neutron absorption cross section and fuel matrix diffusivity. 132I and 97Nb were identified as good candidates for MOX fuel on-line burnup analysis. In the second, and most important, part of this work, the feasibility of utilizing ThMOX fuel in a pressurized water reactor (PWR) was first examined under steady-state, beginning of life conditions. Using a three-dimensional MCNP model of a Westinghouse-type 17x17 PWR, several fuel compositions and configurations of a one-third ThMOX core were compared to a 100% UO2 core. A blanket-type arrangement of 5.5 wt% PuO2 was determined to be the best candidate for further analysis. Next, the safety of the ThMOX configuration was evaluated through three cycles of burnup at several using the following metrics: axial and radial nuclear hot channel factors, moderator and fuel temperature coefficients, delayed neutron fraction, and shutdown margin. Additionally, the performance of the ThMOX configuration was assessed by tracking cycle length, plutonium destroyed, and fission product poison concentration.

  12. Simulation of reactor noise analysis measurement for light-water critical assembly TCA using MCNP-DSP

    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)

  13. Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

    Energy Technology Data Exchange (ETDEWEB)

    Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N. [Institution Project center ITER, Moscow (Russian Federation)

    2014-08-21

    In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ–ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed.

  14. Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

    Science.gov (United States)

    Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N.

    2014-08-01

    In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ-ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed.

  15. Determination of {beta}{sub eff} using MCNP-4C2 and application to the CROCUS and PROTEUS reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vollaire, J. [European Organization for Nuclear Research CERN, CH-1211 Geneve 23 (Switzerland); Plaschy, M.; Jatuff, F. [Paul Scherrer Institut PSI, CH-5232 Villigen PSI (Switzerland); Chawla, R. [Paul Scherrer Institut PSI, CH-5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne EPFL, CH-1015 Lausanne (Switzerland)

    2006-07-01

    A new Monte Carlo method for the determination of {beta}{sub eff} has been recently developed and tested using appropriate models of the experimental reactors CROCUS and PROTEUS. The current paper describes the applied methodology and highlights the resulting improvements compared to the simplest MCNP approach, i.e. the 'prompt method' technique. In addition, the flexibility advantages of the developed method are presented. Specifically, the possibility to obtain the effective delayed neutron fraction {beta}{sub eff} per delayed neutron group, per fissioning nuclide and per reactor region is illustrated. Finally, the MCNP predictions of {beta}{sub eff} are compared to the results of deterministic calculations. (authors)

  16. DESAIN TERAS PLTN JENIS PEBBLE BED MODULAR REACTOR (PBMR) MENGGUNAKAN PAKET PROGRAM MCNP-5 PADA KONDISI BEGINNING OF LIFE

    OpenAIRE

    Ralind Re Marla; Yohannes Sardjono; Supardi Supardi

    2015-01-01

    Telah dilakukan desain teras Pembangkit Listrik Tenaga Nuklir (PLTN) untuk jenis Pebble Bed Modular Reactor (PBMR) dengan daya 70 MWe untuk keperluan proses smelter pada keadaan beginning of life (BOL). Analisis ini bertujuan untuk mengetahui persen pengkayaan, distribusi suhu dan nilai keselamatan dengan koefisien reaktivitas teras yang negatif pada reaktor jenis PBMR apabila daya reaktor 70 MWe. Analisis menggunakan program Monte Carlo N-Particle-5 (MCNP5) dan dari hasil analisis ini dihara...

  17. Designing an epithermal neutron beam for boron neutron capture therapy for a DIDO type reactor using MCNP

    Science.gov (United States)

    Ross, D.; Constantine, G.; Weaver, D. R.; Beynon, T. D.

    1993-10-01

    This paper describes work undertaken to design an epithermal neutron beam for a DIDO type reactor for use in boron neutron capture therapy, a form of cancer treatment. It involved extensive use of MCNP, a Monte Carlo computer code. Initially, calculations were made with MCNP to simulate earlier experiments with an epithermal beam on the DIDO reactor. This comparison made it possible both to validate the Monte Carlo modelling of the reactor and to gain an insight into the important features of the simulation. Following this, MCNP was used to design a filtered epithermal neutron beam facility for DIDO's largest beam tube, a 13.7 cm radius horizontal tube which extends radially away from the core. First a selection was made of the optimum filter components for the beam. Then the research concentrated on combining these filter elements to construct a practical epithermal beam design. The results suggest that the optimum method of generating the epithermal neutron source is to employ a filter combination consisting principally of liquid argon with the addition of cadmium, aluminium, titanium and possibly tin. The calculations also show that the resultant neutron beam would have a flux greater than 1.0 × 10 9 n cm -2 s -1 and have sufficiently low fast-neutron and gamma-ray contamination.

  18. Neutronic analysis for core conversion (HEU–LEU) of the low power research reactor using the MCNP4C code

    OpenAIRE

    Aldawahra Saadou; Khattab Kassem; Saba Gorge

    2015-01-01

    Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR) have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad) and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad) cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The propos...

  19. MCNP Calculations for the Shielding Design of a Beam Tube to Be Installed at the Portuguese Research Reactor

    Science.gov (United States)

    Gonçalves, I. F.; Ramalho, A. G.; Gonçalves, I. C.; Salgado, J.

    The work presented concerns the calculation of the external biological shielding for a neutron beam tube that will be installed at the Portuguese Research Reactor, RPI. This tube will have enough versatility to be used in fields so different as the analysis of the composition of samples or research work in Boron Neutron Capture Therapy, BNCT. The calculation was made by using the MCNP code. This code is a well validated and widely used code, and has therefore become an important tool in the design and optimisation work of experiences related to neutrons and gamma radiation.

  20. Comparative analysis of results between CASMO, MCNP and Serpent for a suite of Benchmark problems on BWR reactors; Analisis comparativo de resultados entre CASMO, MCNP y SERPENT para una suite de problemas Benchmark en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, J. V.; Vargas E, S.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Reyes F, M. del C.; Del Valle G, E., E-mail: vicente.xolocostli@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, UP - Adolfo Lopez Mateos, Edif. 9, 07738 Mexico D. F. (Mexico)

    2014-10-15

    In this paper a comparison is made in analyzing the suite of Benchmark problems for reactors type BWR between CASMO-4, MCNP6 and Serpent code. The Benchmark problem consists of two different geometries: a fuel cell of a pin and assembly type BWR. To facilitate the study of reactors physics in the fuel pin their nuclear characteristics are provided to detail, such as burnt dependence, the reactivity of selected nuclide, etc. With respect to the fuel assembly, the presented results are regarding to infinite multiplication factor for burning different steps and different vacuum conditions. Making the analysis of this set of Benchmark problems provides comprehensive test problems for the next fuels generation of BWR reactors with high extended burned. It is important to note that when making this comparison the purpose is to validate the methodologies used in modeling for different operating conditions, if the case is of other BWR assembly. The results will be within a range with some uncertainty, considering that does not depend on code that is used. Escuela Superior de Fisica y Matematicas of Instituto Politecnico Nacional (IPN (Mexico) has accumulated some experience in using Serpent, due to the potential of this code over other commercial codes such as CASMO and MCNP. The obtained results for the infinite multiplication factor are encouraging and motivate the studies to continue with the generation of the X S of a core to a next step a respective nuclear data library is constructed and this can be used by codes developed as part of the development project of the Mexican Analysis Platform of Nuclear Reactors AZTLAN. (Author)

  1. Comparison of KENO-VI and MCNP5 Criticality Analyses for a Lunar Regolith Clustered-Reactor System

    Science.gov (United States)

    Bess, John Darrell

    2008-01-01

    The Lunar Regolith Clustered-Reactor System design has been presented as an alternative method for providing surface power to a lunar facility using a fast-fission, heatpipe-cooled nuclear reactor. The reactor system is divided into subcritical units that can be safely launched into orbit without risk of inadvertent criticality in the event of a launch accident. The reactor subunits are emplaced into the lunar surface to form a clustered-reactor system, utilizing the regolith as both radiation shielding and neutron-reflector material. Coordinated placement of multiple subunits can provision a critical reactor system proportional to localized lunar surface power demand. Reactor units assembled using proven and tested materials in radiation environments such as UO2 fuel, stainless-steel cladding and support, and compatible liquid-metal heatpipes promote safety and reliability, with ease of manufacture and testing. Reactor power levels of approximately 100 kWth per subunit significantly reduces the negative effects of elevated temperature and radiation environments associated with single nuclear power reactors operated at higher power levels. The analysis of subunit criticality in various accident scenarios differs by up to 4% (~$6 in reactivity) between results generated using conventional criticality analysis codes, MCNP5 and KENO-VI. A demonstrated trend exists between results of the two criticality codes as accident conditions approach a multiplication factor of one. Code comparison of a tri-cluster system on the lunar surface provides comparable results with calculated system reactivity within 0.5%. Iron concentration is confirmed as the dominant element in the lunar regolith influencing system reactivity.

  2. Design of boron carbide-shielded irradiation channel of the outer irradiation channel of the Ghana Research Reactor-1 using MCNP.

    Science.gov (United States)

    Abrefah, R G; Sogbadji, R B M; Ampomah-Amoako, E; Birikorang, S A; Odoi, H C; Nyarko, B J B

    2011-01-01

    The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbide-shielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one boron carbide covered outer irradiation channel; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model, which has a boron carbide-shielded channel is to equip the Ghana Research Reactor-1 with the means of performing efficient epithermal neutron activation analysis. After the simulation, a comparison of the results from the original MCNP model for the Ghana Research Reactor-1 and the new redesigned model of the boron carbide shielded channel was made. The final effective criticality of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new boron carbide designed model was recorded as 1.00282. Also, a final prompt neutron lifetime of 1.5245 × 10(-4)s was recorded for the new boron carbide designed model while a value of 1.5571 × 10(-7)s was recorded for the original MCNP design of the GHARR-1.

  3. Estimation of doses received by operators in the 1958 RB reactor accident using the MCNP5 computer code simulation

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2012-01-01

    Full Text Available A numerical simulation of the radiological consequences of the RB reactor reactivity excursion accident, which occurred on October 15, 1958, and an estimation of the total doses received by the operators were run by the MCNP5 computer code. The simulation was carried out under the same assumptions as those used in the 1960 IAEA-organized experimental simulation of the accident: total fission energy of 80 MJ released in the accident and the frozen positions of the operators. The time interval of exposure to high doses received by the operators has been estimated. Data on the RB1/1958 reactor core relevant to the accident are given. A short summary of the accident scenario has been updated. A 3-D model of the reactor room and the RB reactor tank, with all the details of the core, created. For dose determination, 3-D simplified, homogenised, sexless and faceless phantoms, placed inside the reactor room, have been developed. The code was run for a number of neutron histories which have given a dose rate uncertainty of less than 2%. For the determination of radiation spectra escaping the reactor core and radiation interaction in the tissue of the phantoms, the MCNP5 code was run (in the KCODE option and “mode n p e”, with a 55-group neutron spectra, 35-group gamma ray spectra and a 10-group electron spectra. The doses were determined by using the conversion of flux density (obtained by the F4 tally in the phantoms to doses using factors taken from ICRP-74 and from the deposited energy of neutrons and gamma rays (obtained by the F6 tally in the phantoms’ tissue. A rough estimation of the time moment when the odour of ozone was sensed by the operators is estimated for the first time and given in Appendix A.1. Calculated total absorbed and equivalent doses are compared to the previously reported ones and an attempt to understand and explain the reasons for the obtained differences has been made. A Root Cause Analysis of the accident was done and

  4. Development and validation of a model TRIGA Mark III reactor with code MCNP5; Desarrollo y validacion de un modelo del reactor Triga Mark III con el codigo MCNP5

    Energy Technology Data Exchange (ETDEWEB)

    Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Aguilar H, F., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    The main purpose of this paper is to obtain a model of the reactor core TRIGA Mark III that accurately represents the real operating conditions to 1 M Wth, using the Monte Carlo code MCNP5. To provide a more detailed analysis, different models of the reactor core were realized by simulating the control rods extracted and inserted in conditions in cold (293 K) also including an analysis for shutdown margin, so that satisfied the Operation Technical Specifications. The position they must have the control rods to reach a power equal to 1 M Wth, were obtained from practice entitled Operation in Manual Mode performed at Instituto Nacional de Investigaciones Nucleares (ININ). Later, the behavior of the K{sub eff} was analyzed considering different temperatures in the fuel elements, achieving calculate subsequently the values that best represent the actual reactor operation. Finally, the calculations in the developed model for to obtain the distribution of average flow of thermal, epithermal and fast neutrons in the six new experimental facilities are presented. (Author)

  5. Investigation of reactivity variations of the Isfahan MNSR reactor due to variations in the thickness of the core top beryllium layer using WIMSD and MCNP codes

    Directory of Open Access Journals (Sweden)

    A Shirani

    2010-12-01

    Full Text Available In this work, the Isfahan Miniature Neutron Source Reactor (MNSR is first simulated using the WIMSD code, and its fuel burn-up after 7 years of operation ( when the reactor was revived by adding a 1.5 mm thick beryllium shim plate to the top of its core and also after 14 years of operation (total operation time of the reactor is calculated. The reactor is then simulated using the MCNP code, and its reactivity variation due to adding a 1.5 mm thick beryllium shim plate to the top of the reactor core, after 7 years of operation, is calculated. The results show good agreement with the available data collected at the revival time. Exess reactivity of the reactor at present time (after 14 years of operation and after 7 years of the the reactor revival time is also determined both experimentally and by calculation, which show good agreement, and indicate that at the present time there is no need to add any further beryllium shim plate to the top of the reactor core. Furthermore, by adding more beryllium layers with various thicknesses to the top of the reactor core, in the input program of the MCNP program, reactivity value of these layers is calculated. From these results, one can predict the necessary beryllium thickness needed to reach a desired reactivity in the MNSR reactor.

  6. Investigation of Isfahan miniature neutron source reactor (MNSR) for boron neutron capture therapy by MCNP simulation

    OpenAIRE

    S. Z. Kalantari; H Tavakoli; Nami, M.

    2015-01-01

    One of the important neutron sources for Boron Neutron Capture Therapy (BNCT) is a nuclear reactor. It needs a high flux of epithermal neutrons. The optimum conditions of the neutron spectra for BNCT are provided by the International Atomic Energy Agency (IAEA). In this paper, Miniature Neutron Source Reactor (MNSR) as a neutron source for BNCT was investigated. For this purpose, we designed a Beam Shaping Assembly (BSA) for the reactor and the neutron transport from the core of the reactor t...

  7. Neutronic analysis for core conversion (HEU–LEU of the low power research reactor using the MCNP4C code

    Directory of Open Access Journals (Sweden)

    Aldawahra Saadou

    2015-06-01

    Full Text Available Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The proposed LEU core contained the same number of fuel pins as the HEU core. All other structure materials and dimensions of HEU and LEU cores were the same except the increase in the radius of control rod material from 0.195 to 0.205 cm and keeping the outer diameter of the control rod unchanged in the LEU core. The effective multiplication factor (keff, excess reactivity (ρex, control rod worth (CRW, shutdown margin (SDM, safety reactivity factor (SRF, delayed neutron fraction (βeff and the neutron fluxes in the irradiation tubes for the existing and the potential LEU fuel were investigated. The results showed that the safety parameters and the neutron fluxes in the irradiation tubes of the LEU fuels were in good agreements with the HEU results. Therefore, the LEU fuel was validated to be a suitable choice for fuel conversion of the MNSR in the future.

  8. DESAIN TERAS PLTN JENIS PEBBLE BED MODULAR REACTOR (PBMR MENGGUNAKAN PAKET PROGRAM MCNP-5 PADA KONDISI BEGINNING OF LIFE

    Directory of Open Access Journals (Sweden)

    Ralind Re Marla

    2015-03-01

    Full Text Available Telah dilakukan desain teras Pembangkit Listrik Tenaga Nuklir (PLTN untuk jenis Pebble Bed Modular Reactor (PBMR dengan daya 70 MWe untuk keperluan proses smelter pada keadaan beginning of life (BOL. Analisis ini bertujuan untuk mengetahui persen pengkayaan, distribusi suhu dan nilai keselamatan dengan koefisien reaktivitas teras yang negatif pada reaktor jenis PBMR apabila daya reaktor 70 MWe. Analisis menggunakan program Monte Carlo N-Particle-5 (MCNP5 dan dari hasil analisis ini diharapkan dapat memenuhi syarat dalam mendukung program percepatan pembangunan kelistrikan batubara 10.000 MWe khususnya untuk proses smelter, yang tersebar merata di wilayah Indonesia. Hasil penelitian menunjukkan bahwa, faktor perlipatan efektif (k-eff Reaktor jenis PBMR daya 70 MWe mengalami kondisi kritis pada pengkayaan 5,626 % dengan nilai faktor perlipatan efektif 1,00031±0,00087 dan nilai koefisien reaktivitas suhu pada -10,0006 pcm/K. Dari hasil analisis daat disimpulkan bahwa reaktor jenis PBMR daya 70 MWe adalah aman.   ABSTRACT The core design of Nuclear Power Plant for Pebble Bed Modular Reactor (PBMR type with 70 MWe capacity power in Beginning of Life (BOL has been performed. The aim of this analysis, to know percent enrichment, temperature distribution and safety value by negative temperature coefficient at type PBMR if reactor power become lower equal to 70 MWe. This analysis was expected become one part of overview project development the power plant with 10.000 MWe of total capacity, spread evenly in territory of Indonesia especially to support of smelter industries. The results showed that, effective multiplication factor (keff with power 70 MWe critical condition at enrichment 5,626 %is 1,00031±0,00087, based on enrichment result, a value of the temperature coefficient reactivity is - 10,0006 pcm/K. Based on the results of these studies, it can beconcluded that the PBMR 70 MWe design is theoritically safe.

  9. Investigation of Isfahan miniature neutron source reactor (MNSR for boron neutron capture therapy by MCNP simulation

    Directory of Open Access Journals (Sweden)

    S.Z Kalantari

    2015-01-01

    Full Text Available One of the important neutron sources for Boron Neutron Capture Therapy (BNCT is a nuclear reactor. It needs a high flux of epithermal neutrons. The optimum conditions of the neutron spectra for BNCT are provided by the International Atomic Energy Agency (IAEA. In this paper, Miniature Neutron Source Reactor (MNSR as a neutron source for BNCT was investigated. For this purpose, we designed a Beam Shaping Assembly (BSA for the reactor and the neutron transport from the core of the reactor to the output windows of BSA was simulated by MCNPX code. To optimize the BSA performance, two sets of parameters should be evaluated, in-air and in-phantom parameters. For evaluating in-phantom parameters, a Snyder head phantom was used and biological dose rate and dose-depth curve were calculated in brain normal and tumor tissues. Our calculations showed that the neutron flux of the MNSR reactor can be used for BNCT, and the designed BSA in optimum conditions had a good therapeutic characteristic for BNCT.

  10. Optimization of Neutron Spectrum in Northwest Beam Tube of Tehran Research Reactor for BNCT, by MCNP Code

    Energy Technology Data Exchange (ETDEWEB)

    Zamani, M. [National Radiation Protection Department - NRPD, Atomic Energy Organization of Iran - AEOI, Tehran (Iran, Islamic Republic of); End of North Kargar st, Atomic Energy Organization of Iran, P.O. Box: 14155-1339, Tehran (Iran, Islamic Republic of); Kasesaz, Y.; Khalafi, H.; Shayesteh, M. [Radiation Application School, Nuclear Science and Technology Research Institute, AEOI, Tehran (Iran, Islamic Republic of)

    2015-07-01

    In order to gain the neutron spectrum with proper components specification for BNCT, it is necessary to design a Beam Shape Assembling (BSA), include of moderator, collimator, reflector, gamma filter and thermal neutrons filter, in front of the initial radiation beam from the source. According to the result of MCNP4C simulation, the Northwest beam tube has the most optimized neuron flux between three north beam tubes of Tehran Research Reactor (TRR). So, it has been chosen for this purpose. Simulation of the BSA has been done in four above mentioned phases. In each stage, ten best configurations of materials with different length and width were selected as the candidates for the next stage. The last BSA configuration includes of: 78 centimeters of air as an empty space, 40 centimeters of Iron plus 52 centimeters of heavy-water as moderator, 30 centimeters of water or 90 centimeters of Aluminum-Oxide as a reflector, 1 millimeters of lithium (Li) as thermal neutrons filter and finally 3 millimeters of Bismuth (Bi) as a filter of gamma radiation. The result of Calculations shows that if we use this BSA configuration for TRR Northwest beam tube, then the best neutron flux and spectrum will be achieved for BNCT. (authors)

  11. Calculation with MCNP of capture photon flux in VVER-1000 experimental reactor.

    Science.gov (United States)

    Töre, Candan; Ortego, Pedro

    2005-01-01

    The aim of this study is to obtain by Monte Carlo method the high energy photon flux due to neutron capture in the internals and vessel layers of the experimental reactor LR-0 located in REZ, Czech Republic, and loaded with VVER-1000 fuel. The calclated neutron, photon and photon to neutron flux ratio are compared with experimental measurements performed with a multi-parameter stilbene detector. The results show clear underestimation of photon flux in downcomer and some overestimation at vessel surface and 1/4 thickness but a good fitting for deeper points in vessel.

  12. The REBUS-MCNP linkage.

    Energy Technology Data Exchange (ETDEWEB)

    Stevens, J. G.; Nuclear Engineering Division

    2009-04-24

    The Reduced Enrichment Research and Test Reactor (RERTR) Program uses the REBUS-PC computer code to provide reactor physics and core design information such as neutron flux distributions in space, energy, and time, and to track isotopic changes in fuel and neutron absorbers with burnup. REBUS-PC models the complete fuel cycle including shuffling capability. REBUS-PC evolved using the neutronic capabilities of multi-group diffusion theory code DIF3D 9.0, but was extended to apply the continuous energy Monte Carlo code MCNP for one-group fluxes and cross-sections. The linkage between REBUS-PC and MCNP has recently been modernized and extended, as described in this manual. REBUS-PC now calls MCNP via a system call so that the user can apply any valid MCNP executable. The interface between REBUS-PC and MCNP requires minimal changes to an existing MCNP model, and little additional input. The REBUS-MCNP interface can also be used in conjunction with DIF3D neutronics to update an MCNP model with fuel compositions predicted using a DIF3D based depletion.

  13. Evaluation of the thermal neutron flux in the core of IPEN/MB-01 reactor using the code Monte Carlo (MCNP)

    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)

  14. ITER (International Thermonuclear Experimental Reactor) current drive and heating physics

    Energy Technology Data Exchange (ETDEWEB)

    Nevins, W.M.; Lindquist, W. (Lawrence Livermore National Lab., CA (USA)); Fujisawa, N.; Kimura, H. (Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan)); Hopman, H.; Rebuffi, L.; Wegrowe, J.G. (Max-Planck-Institut fuer Plasmaphysik, Garching (Germany, F.R.). NET Design Team); Parail, V.; Vdovin, V. (Gosudarstvennyj Komitet po Ispol' zovaniyu Atomnoj Ehnergii SSSR, Moscow (USSR). Inst. Atomnoj Ehn

    1990-01-01

    The ITER Current Drive and Heating (CD H) systems are required for: Ionization and current initiation; Non-inductive current ramp-up assist; Heating of the plasma; Steady-state operation with full non-inductive current drive; Current profile control; and Burn control by modulation of the auxiliary power. Steady-state current drive is the most demanding requirement, so this has driven the choice of the ITER current drive and heating systems.

  15. Modeling the effect in of criticality from changes in key parameters for small High Temperature Nuclear Reactor (U-BatteryTM) using MCNP4C

    Science.gov (United States)

    Pauzi, A. M.

    2013-06-01

    The neutron transport code, Monte Carlo N-Particle (MCNP) which was wellkown as the gold standard in predicting nuclear reaction was used to model the small nuclear reactor core called "U-batteryTM", which was develop by the University of Manchester and Delft Institute of Technology. The paper introduces on the concept of modeling the small reactor core, a high temperature reactor (HTR) type with small coated TRISO fuel particle in graphite matrix using the MCNPv4C software. The criticality of the core were calculated using the software and analysed by changing key parameters such coolant type, fuel type and enrichment levels, cladding materials, and control rod type. The criticality results from the simulation were validated using the SCALE 5.1 software by [1] M Ding and J L Kloosterman, 2010. The data produced from these analyses would be used as part of the process of proposing initial core layout and a provisional list of materials for newly design reactor core. In the future, the criticality study would be continued with different core configurations and geometries.

  16. Implementation of On-the-Fly Doppler Broadening in MCNP5 for Multiphysics Simulation of Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    William Martin

    2012-11-16

    A new method to obtain Doppler broadened cross sections has been implemented into MCNP, removing the need to generate cross sections for isotopes at problem temperatures. Previous work had established the scientific feasibility of obtaining Doppler-broadened cross sections "on-the-fly" (OTF) during the random walk of the neutron. Thus, when a neutron of energy E enters a material region that is at some temperature T, the cross sections for that material at the exact temperature T are immediately obtained by interpolation using a high order functional expansion for the temperature dependence of the Doppler-broadened cross section for that isotope at the neutron energy E. A standalone Fortran code has been developed that generates the OTF library for any isotope that can be processed by NJOY. The OTF cross sections agree with the NJOY-based cross sections for all neutron energies and all temperatures in the range specified by the user, e.g., 250K - 3200K. The OTF methodology has been successfully implemented into the MCNP Monte Carlo code and has been tested on several test problems by comparing MCNP with conventional ACE cross sections versus MCNP with OTF cross sections. The test problems include the Doppler defect reactivity benchmark suite and two full-core VHTR configurations, including one with multiphysics coupling using RELAP5-3D/ATHENA for the thermal-hydraulic analysis. The comparison has been excellent, verifying that the OTF libraries can be used in place of the conventional ACE libraries generated at problem temperatures. In addition, it has been found that using OTF cross sections greatly reduces the complexity of the input for MCNP, especially for full-core temperature feedback calculations with many temperature regions. This results in an order of magnitude decrease in the number of input lines for full-core configurations, thus simplifying input preparation and reducing the potential for input errors. Finally, for full-core problems with multiphysics

  17. Neutron flux distribution inside the cylindrical core of minor excess of reactivity in the IPEN/MB-01 reactor and comparison with citation code and MCNP- 5 code

    Energy Technology Data Exchange (ETDEWEB)

    Aredes, Vitor Ottoni; Bitelli, Ulysses d' Utra; Mura, Luiz Ernesto C.; Santos, Diogo Feliciano dos; Lima, Ana Cecilia de Souza, E-mail: ubitelli@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    This study aims to determine the distribution of thermal neutron flux in the IPEN/MB-01 nuclear reactor core assembled with cylindrical core configuration of minor excess of reactivity with 568 fuel rods (28 fuel rods in diameter). The thermal neutron flux at the positions of irradiation derive from the method of reaction rate using gold foils. The experiment consists in inserting gold activations foils with and without cadmium coverage (cadmium boxes with 0.0502 cm thickness) in several positions throughout the active core. After irradiation, activity induced by nuclear reaction rates over gold foils is assessed by gamma ray spectrometry using a high-purity germanium (HPGe) detector. Experimental results are compared to those derived from calculations performed using a three dimensional CITATION diffusion code and MCNP-5 code and a proper nuclear data library. While calculated neutron flux data shows good agreement with experimental values in regions with little disturbance in the neutron flux, also showing that in the region of the reflectors of neutrons and near the control rods, the diffusion theory is not very precise. The average value of thermal neutron flux obtained experimentally compared to the calculated value by CITATION code and MCNP-5 code respectively show a difference of 1.18% and 0.84% at a nuclear power level of 74.65 ± 3.28 % watts. The average measured value of thermal neutron flux is 4.10 10{sup 8} ± 5.25% n/cm{sup 2}s. (author)

  18. Calculations of the thermal and fast neutron fluxes in the Syrian miniature neutron source reactor using the MCNP-4C code.

    Science.gov (United States)

    Khattab, K; Sulieman, I

    2009-04-01

    The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method ((197)Au (n, gamma) (198)Au and (59)Co (n, gamma) (60)Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.

  19. Beyond ITER: neutral beams for a demonstration fusion reactor (DEMO) (invited).

    Science.gov (United States)

    McAdams, R

    2014-02-01

    In the development of magnetically confined fusion as an economically sustainable power source, International Tokamak Experimental Reactor (ITER) is currently under construction. Beyond ITER is the demonstration fusion reactor (DEMO) programme in which the physics and engineering aspects of a future fusion power plant will be demonstrated. DEMO will produce net electrical power. The DEMO programme will be outlined and the role of neutral beams for heating and current drive will be described. In particular, the importance of the efficiency of neutral beam systems in terms of injected neutral beam power compared to wallplug power will be discussed. Options for improving this efficiency including advanced neutralisers and energy recovery are discussed.

  20. Beyond ITER: Neutral beams for a demonstration fusion reactor (DEMO) (invited)

    Energy Technology Data Exchange (ETDEWEB)

    McAdams, R., E-mail: roy.mcadams@ccfe.ac.uk [EURATOM/CCFE Association, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom)

    2014-02-15

    In the development of magnetically confined fusion as an economically sustainable power source, International Tokamak Experimental Reactor (ITER) is currently under construction. Beyond ITER is the demonstration fusion reactor (DEMO) programme in which the physics and engineering aspects of a future fusion power plant will be demonstrated. DEMO will produce net electrical power. The DEMO programme will be outlined and the role of neutral beams for heating and current drive will be described. In particular, the importance of the efficiency of neutral beam systems in terms of injected neutral beam power compared to wallplug power will be discussed. Options for improving this efficiency including advanced neutralisers and energy recovery are discussed.

  1. Comparison and validation of the results of the AZNHEX v.1.0 code with the MCNP code simulating the core of a fast reactor cooled with sodium; Comparacion y validacion de los resultados del codigo AZNHEX v.1.0 con el codigo MCNP simulando el nucleo de un reactor rapido refrigerado con sodio

    Energy Technology Data Exchange (ETDEWEB)

    Galicia A, J.; Francois L, J. L.; Bastida O, G. E. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Esquivel E, J., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    The development of the AZTLAN platform for the analysis and design of nuclear reactors is led by Instituto Nacional de Investigaciones Nucleares (ININ) and divided into four working groups, which have well-defined activities to achieve significant progress in this project individually and jointly. Within these working groups is the users group, whose main task is to use the codes that make up the AZTLAN platform to provide feedback to the developers, and in this way to make the final versions of the codes are efficient and at the same time reliable and easy to understand. In this paper we present the results provided by the AZNHEX v.1.0 code when simulating the core of a fast reactor cooled with sodium at steady state. The validation of these results is a fundamental part of the platform development and responsibility of the users group, so in this research the results obtained with AZNHEX are compared and analyzed with those provided by the Monte Carlo code MCNP-5, software worldwide used and recognized. A description of the methodology used with MCNP-5 is also presented for the calculation of the interest variables and the difference that is obtained with respect to the calculated with AZNHEX. (Author)

  2. ITER perspective on fusion reactor diagnostics—A spectroscopic view

    Science.gov (United States)

    De Bock, M. F. M.; Barnsley, R.; Bassan, M.; Bertalot, L.; Brichard, B.; Bukreev, I. M.; Drevon, J. M.; Le Guern, F.; Hutton, R.; Ivantsivskiy, M.; Lee, H. G.; Leipold, F.; Maquet, P.; Marot, L.; Martin, V.; Mertens, P.; Mokeev, A.; Moser, L.; Mukhin, E. E.; Pak, S.; Razdobarin, A. G.; Reichle, R.; Seon, C. R.; Seyvet, F.; Simrock, S.; Udintsev, V.; Vayakis, G.; Vorpahl, C.

    2016-08-01

    The ITER tokamak requires diagnostics that on the one hand have a high sensitivity, high spatial and temporal resolution and a high dynamic range, while on the other hand are robust enough to survive in a harsh environment. In recent years significant progress has been made in addressing critical challenges to the development of spectroscopic (but also other) diagnostics. This contribution presents an overview of recent achievements in 4 topical areas: • First mirror protection and cleaning • Nuclear confinement • Radiation mitigation strategy for optical and electronic components • Calibration strategies

  3. ITER (International Thermonuclear Experimental Reactor) shield and blanket work package report

    Energy Technology Data Exchange (ETDEWEB)

    1988-06-01

    This report summarizes nuclear-related work in support of the US effort for the International Thermonuclear Experimental Reactor (ITER) Study. The purpose of this work was to prepare for the first international ITER workshop devoted to defining a basic ITER concept that will serve as a basis for an indepth conceptual design activity over the next 2-1/2 years. Primary tasks carried out during the past year included: design improvements of the inboard shield developed for the TIBER concept, scoping studies of a variety of tritium breeding blanket options, development of necessary design guidelines and evaluation criteria for the blanket options, further safety considerations related to nuclear components and issues regarding structural materials for an ITER device. 44 refs., 31 figs., 29 tabs.

  4. The New MCNP6 Depletion Capability

    Energy Technology Data Exchange (ETDEWEB)

    Fensin, Michael Lorne [Los Alamos National Laboratory; James, Michael R. [Los Alamos National Laboratory; Hendricks, John S. [Los Alamos National Laboratory; Goorley, John T. [Los Alamos National Laboratory

    2012-06-19

    The first MCNP based inline Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. Both the MCNP5 and MCNPX codes have historically provided a successful combinatorial geometry based, continuous energy, Monte Carlo radiation transport solution for advanced reactor modeling and simulation. However, due to separate development pathways, useful simulation capabilities were dispersed between both codes and not unified in a single technology. MCNP6, the next evolution in the MCNP suite of codes, now combines the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. We describe here the new capabilities of the MCNP6 depletion code dating from the official RSICC release MCNPX 2.6.0, reported previously, to the now current state of MCNP6. NEA/OECD benchmark results are also reported. The MCNP6 depletion capability enhancements beyond MCNPX 2.6.0 reported here include: (1) new performance enhancing parallel architecture that implements both shared and distributed memory constructs; (2) enhanced memory management that maximizes calculation fidelity; and (3) improved burnup physics for better nuclide prediction. MCNP6 depletion enables complete, relatively easy-to-use depletion calculations in a single Monte Carlo code. The enhancements described here help provide a powerful capability as well as dictate a path forward for future development to improve the usefulness of the technology.

  5. Requirements for US regulatory approval of the International Thermonuclear Experimental Reactor (ITER)

    Energy Technology Data Exchange (ETDEWEB)

    Petti, D.A.; Haire, J.C.

    1993-12-01

    The International Thermonuclear Experimental Reactor (ITER) is the first fusion machine that will have sufficient decay heat and activation product inventory to pose potential nuclear safety concerns. As a result, nuclear safety and environmental issues will be much more important in the approval process for the design, siting, construction, and operation of ITER in the United States than previous fusion devices, such as the Tokamak Fusion Test Reactor. The purpose of this report is (a) to provide an overview of the regulatory approval process for a Department of Energy (DOE) nuclear facility; (b) to present the dose limits used by DOE to protect workers, the public, and the environment from the risks of exposure to radiation and hazardous materials; (c) to discuss some key nuclear safety-related issues that must be addressed early in the Engineering Design Activities (EDA) to obtain regulatory approval; and (d) to provide general guidelines to the ITER Joint Central Team (JCT) concerning the development of a regulatory framework for the ITER project.

  6. Validation of the inspections with ultrasound of the welds of the reactor of ITER vacuum vessel; Validacion de las inspecciones con ultrasonidos de las soldaduras de la Vasija de Vacio del reactor del ITER

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, A.; Fernandez, F.; Perez, C.; Sillero, J. A.

    2013-07-01

    The ITER fusion reactor vacuum vessel has thousands of welding austenitic with shapes and different manufacturing processes. The RCC-MR code, which is that applied to the manufacture of the fusion reactor, requires a volumetric test all of them. This test should be mainly by x-rays and welds where it was not possible to use this method, ultrasonic.09-06.

  7. Fusion Reactor and Fusion Reactor Materials:Concept Design of the ITER Test Blanket Modules

    Institute of Scientific and Technical Information of China (English)

    HUANGJinhua; LIZaixing; ZHUYukun; HUGang

    2003-01-01

    Performances required: prospect to be adopted in DEMO. Shielding for V.V. and TFC in ITER. Design principles: the peak temperature and stress should not exceed technical limits. The structure of test blanket modules (TBM) should be simple for easy fabrication, and TBM should be robust for reliability.

  8. Burn control of an ITER-like fusion reactor using fuzzy logic

    Science.gov (United States)

    Garcia-Amador, A. Sair; Martinell, Julio J.

    2016-10-01

    The fuel burn in a fusion reactor has to be kept at a nearly constant rate in order to have a steady power exhaust. Here, we develop a control system based on a fuzzy logic controller in order that adjusts external parameters to keep the plasma temperature and density at the design values of a reactor of the characteristics of ITER. The control parameters chosen are the D-T refueling rate, the auxiliary heating power and a neutral helium beam. We use a fuzzy controller of the Mamdani type that uses a number of membership functions appropriate to produce a response to parameter deviations that minimizes the response time. The inference rules are determined in a way to provide stabilization to all perturbations of the temperature, density and alpha particle fraction. The dynamical response of the reactor is simulated with a 0D model that uses confinement times provided by the ITER scaling. We show that the system is feedback stabilized for a large range of parameters around the nominal values. The recovery time after a departure from the steady values is of the order of one second. We compare the results with another control system based on neural networks that was developed previously. Funded by projects PAPIIT IN109115 and Conacyt 152905.

  9. Iter

    Science.gov (United States)

    Iotti, Robert

    2015-04-01

    ITER is an international experimental facility being built by seven Parties to demonstrate the long term potential of fusion energy. The ITER Joint Implementation Agreement (JIA) defines the structure and governance model of such cooperation. There are a number of necessary conditions for such international projects to be successful: a complete design, strong systems engineering working with an agreed set of requirements, an experienced organization with systems and plans in place to manage the project, a cost estimate backed by industry, and someone in charge. Unfortunately for ITER many of these conditions were not present. The paper discusses the priorities in the JIA which led to setting up the project with a Central Integrating Organization (IO) in Cadarache, France as the ITER HQ, and seven Domestic Agencies (DAs) located in the countries of the Parties, responsible for delivering 90%+ of the project hardware as Contributions-in-Kind and also financial contributions to the IO, as ``Contributions-in-Cash.'' Theoretically the Director General (DG) is responsible for everything. In practice the DG does not have the power to control the work of the DAs, and there is not an effective management structure enabling the IO and the DAs to arbitrate disputes, so the project is not really managed, but is a loose collaboration of competing interests. Any DA can effectively block a decision reached by the DG. Inefficiencies in completing design while setting up a competent organization from scratch contributed to the delays and cost increases during the initial few years. So did the fact that the original estimate was not developed from industry input. Unforeseen inflation and market demand on certain commodities/materials further exacerbated the cost increases. Since then, improvements are debatable. Does this mean that the governance model of ITER is a wrong model for international scientific cooperation? I do not believe so. Had the necessary conditions for success

  10. Influence of reactor irradiation on the mechanical behavior of ITER TF coil candidate insulation systems

    Energy Technology Data Exchange (ETDEWEB)

    Bittner-Rohrhofer, K. E-mail: kbittner@ati.ac.at; Humer, K.; Fillunger, H.; Maix, R.K.; Wang, Z.D.; Weber, H.W

    2003-09-01

    Extensive material tests have to be performed in order to obtain information on the radiation induced change in the mechanical behavior of insulating materials for the ITER Toroidal Field (TF) coil. The investigated insulation systems are R-glass fiber reinforced tapes, vacuum impregnated with a DGEBA epoxy resin and interleafed with Kapton H-foils. According to the actual operating conditions of ITER-FEAT, the systems were irradiated in the TRIGA reactor (Vienna, Austria) to neutron fluences of 5x10{sup 21} and 1x10{sup 22} m{sup -2} (E>0.1 MeV). Static tensile, short-beam-shear (SBS) as well as double-lap-shear (DLS) tests were carried out at 77 K prior to and after irradiation. Furthermore, results on swelling and weight loss as well as on the material properties under tension-tension fatigue loading conditions are presented.

  11. Note: Readout of a micromechanical magnetometer for the ITER fusion reactor.

    Science.gov (United States)

    Rimminen, H; Kyynäräinen, J

    2013-05-01

    We present readout instrumentation for a MEMS magnetometer, placed 30 m away from the MEMS element. This is particularly useful when sensing is performed in high-radiation environment, where the semiconductors in the readout cannot survive. High bandwidth transimpedance amplifiers are used to cancel the cable capacitances of several nanofarads. A frequency doubling readout scheme is used for crosstalk elimination. Signal-to-noise ratio in the range of 60 dB was achieved and with sub-percent nonlinearity. The presented instrument is intended for the steady-state magnetic field measurements in the ITER fusion reactor.

  12. Toolkit for high performance Monte Carlo radiation transport and activation calculations for shielding applications in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Serikov, A.; Fischer, U.; Grosse, D.; Leichtle, D.; Majerle, M., E-mail: arkady.serikov@kit.edu [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany)

    2011-07-01

    The Monte Carlo (MC) method is the most suitable computational technique of radiation transport for shielding applications in fusion neutronics. This paper is intended for sharing the results of long term experience of the fusion neutronics group at Karlsruhe Institute of Technology (KIT) in radiation shielding calculations with the MCNP5 code for the ITER fusion reactor with emphasizing on the use of several ITER project-driven computer programs developed at KIT. Two of them, McCad and R2S, seem to be the most useful in radiation shielding analyses. The McCad computer graphical tool allows to perform automatic conversion of the MCNP models from the underlying CAD (CATIA) data files, while the R2S activation interface couples the MCNP radiation transport with the FISPACT activation allowing to estimate nuclear responses such as dose rate and nuclear heating after the ITER reactor shutdown. The cell-based R2S scheme was applied in shutdown photon dose analysis for the designing of the In-Vessel Viewing System (IVVS) and the Glow Discharge Cleaning (GDC) unit in ITER. Newly developed at KIT mesh-based R2S feature was successfully tested on the shutdown dose rate calculations for the upper port in the Neutral Beam (NB) cell of ITER. The merits of McCad graphical program were broadly acknowledged by the neutronic analysts and its continuous improvement at KIT has introduced its stable and more convenient run with its Graphical User Interface. Detailed 3D ITER neutronic modeling with the MCNP Monte Carlo method requires a lot of computation resources, inevitably leading to parallel calculations on clusters. Performance assessments of the MCNP5 parallel runs on the JUROPA/HPC-FF supercomputer cluster permitted to find the optimal number of processors for ITER-type runs. (author)

  13. Modelling of HTR (High Temperature Reactor Pebble-Bed 10 MW to Determine Criticality as A Variations of Enrichment and Radius of the Fuel (Kernel With the Monte Carlo Code MCNP4C

    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 

  14. R&D on high-power dc reactor prototype for ITER poloidal field converter

    Energy Technology Data Exchange (ETDEWEB)

    Li, Chuan [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Song, Zhiquan; Fu, Peng [Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China); Zhang, Ming, E-mail: zhangming@hust.edu.cn [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Yu, Kexun [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Qin, Xiuqi [School of Electrical Engineering and Automation, Hefei University of Technology, Hefei 230009 (China)

    2015-10-15

    Highlights: • A new prototype design structure of dry-type air-core water-cooling reactor with epoxy resin casting technique is presented. • Theoretical analysis, finite-element simulation and prototype test verification are applied on the design. • The results of temperature rise and transient fault current test of prototypes are introduced and analyzed. • The success of tests demonstrates that the proposed structure is of high reliability and availability. - Abstract: This paper mainly introduces the research and development (R&D) of the high-power dc reactor prototype, whose functions are to limit the circulating current and ripple current in the ITER poloidal field (PF) converter. It needs to operate at rated large direct current 27.5 kA and withstand peak fault current up to 175 kA. Therefore, in order to meet the special requirements of the dynamic and thermal stability, a new prototype design structure of dry-type air-core water-cooling reactor with epoxy resin casting technique is presented, which is based on the theoretical analysis, finite-element simulation calculation and small prototype test verification. Now the full prototype has been fabricated by China industry, and the dynamic and thermal stability tests of the prototype have also been accomplished successfully. The test results are in compliance with the design and it shows the availability and feasibility of the proposed design, which may be a reference for relevant applications.

  15. MCNP Progress & Performance Improvements

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bull, Jeffrey S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-04-14

    Twenty-eight slides give information about the work of the US DOE/NNSA Nuclear Criticality Safety Program on MCNP6 under the following headings: MCNP6.1.1 Release, with ENDF/B-VII.1; Verification/Validation; User Support & Training; Performance Improvements; and Work in Progress. Whisper methodology will be incorporated into the code, and run speed should be increased.

  16. TRIPOLI-4{sup ®} Monte Carlo code ITER A-lite neutronic model validation

    Energy Technology Data Exchange (ETDEWEB)

    Jaboulay, Jean-Charles, E-mail: jean-charles.jaboulay@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Cayla, Pierre-Yves; Fausser, Clement [MILLENNIUM, 16 Av du Québec Silic 628, F-91945 Villebon sur Yvette (France); Damian, Frederic; Lee, Yi-Kang; Puma, Antonella Li; Trama, Jean-Christophe [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France)

    2014-10-15

    3D Monte Carlo transport codes are extensively used in neutronic analysis, especially in radiation protection and shielding analyses for fission and fusion reactors. TRIPOLI-4{sup ®} is a Monte Carlo code developed by CEA. The aim of this paper is to show its capability to model a large-scale fusion reactor with complex neutron source and geometry. A benchmark between MCNP5 and TRIPOLI-4{sup ®}, on the ITER A-lite model was carried out; neutron flux, nuclear heating in the blankets and tritium production rate in the European TBMs were evaluated and compared. The methodology to build the TRIPOLI-4{sup ®} A-lite model is based on MCAM and the MCNP A-lite model. Simplified TBMs, from KIT, were integrated in the equatorial-port. A good agreement between MCNP and TRIPOLI-4{sup ®} is shown; discrepancies are mainly included in the statistical error.

  17. Manufacturing and testing in reactor relevant conditions of brazed plasma facing components of the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Bisio, M. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Branca, V. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Marco, M. Di [FN s.p.a., ss 35 bis dei Giovi km 15, I-15062 Bosco Marengo (Albania) (Italy); Federici, A. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Grattarola, M. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy)]. E-mail: grattarola@ansaldo.it; Gualco, G. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Guarnone, P. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Luconi, U. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Merola, M. [EFDA, Boltzmanstr. 2, D-85748 Garching (Germany); Ozzano, C. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Pasquale, G. [FN s.p.a., ss 35 bis dei Giovi km 15, I-15062 Bosco Marengo (AL) (Italy); Poggi, P. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Rizzo, S. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Varone, F. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy)

    2005-11-15

    A fabrication route based on brazing technology has been developed for the realization of the high heat flux components for the ITER vertical target and Dome-Liner. The divertor vertical target is armoured with carbon fiber reinforced carbon and tungsten in the lower straight part and in the upper curved part, respectively. The armour material is joined to heat sinks made of precipitation hardened copper-chromium-zirconium alloy. The plasma facing units of the dome component are based on a tungsten flat tile design with hypervapotron cooling. An innovative brazing technique based on the addition of carbon fibers to the active brazing alloy, developed by Ansaldo Ricerche for applications in the field of the energy production, has been used for the carbon fiber composite to copper joint to reduce residual stresses. The tungsten-copper joint has been realized by direct casting. A proper brazing thermal cycle has been studied to guarantee the required mechanical properties of the precipitation hardened alloy after brazing. The fabrication route of plasma facing components for the ITER vertical target and dome based on the brazing technology has been proved by means of thermal fatigue tests performed on mock-ups in reactor relevant conditions.

  18. Bulk-bronzied graphites for plasma-facing components in ITER (International Thermonuclear Experimental Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Hirooka, Y.; Conn, R.W.; Doerner, R.; Khandagle, M. (California Univ., Los Angeles, CA (USA). Inst. of Plasma and Fusion Research); Causey, R.; Wilson, K. (Sandia National Labs., Livermore, CA (USA)); Croessmann, D.; Whitley, J. (Sandia National Labs., Albuquerque, NM (USA)); Holland, D.; Smolik, G. (Idaho National Engineering Lab., Idaho Falls, ID (USA)); Matsuda, T.; Sogabe, T. (Toyo Tanso Co. Ltd., O

    1990-06-01

    Newly developed bulk-boronized graphites and boronized C-C composites with a total boron concentration ranging from 1 wt % to 30 wt % have been evaluated as plasma-facing component materials for the International Thermonuclear Experimental Reactor (ITER). Bulk-boronized graphites have been bombarded with high-flux deuterium plasmas at temperatures between 200 and 1600{degree}C. Plasma interaction induced erosion of bulk-boronized graphites is observed to be a factor of 2--3 smaller than that of pyrolytic graphite, in regimes of physical sputtering, chemical sputtering and radiation enhanced sublimation. Postbombardment thermal desorption spectroscopy indicates that bulk-boronized graphites enhance recombinative desorption of deuterium, which leads to a suppression of the formation of deuterocarbon due to chemical sputtering. The tritium inventory in graphite has been found to decrease by an order of magnitude due to 10 wt % bulk-boronization at temperatures above 1000{degree}C. The critical heat flux to induce cracking for bulk-boronized graphites has been found to be essentially the same as that for non-boronized graphites. Also, 10 wt % bulk-boronization of graphite hinders air oxidation nearly completely at 800{degree}C and reduces the steam oxidation rate by a factor of 2--3 at around 1100 and 1350{degree}C. 38 refs., 5 figs.

  19. MCNP Apply in Calculating the Changing of Reactor Critical Coefficient K~ under Different Temperature%MCNP在计算温度对反应堆临界系数Keff值影响中的应用

    Institute of Scientific and Technical Information of China (English)

    曾磊; 杨波; 刘义保; 赵梦云

    2012-01-01

    堆芯内温度变化时,中子能谱、微观截面等都将相应地发生变化。所以与反应性有关的许多参数,如热中子利用系数、逃脱共振俘获概率等都是温度的函数。因而,当反应堆中各种成分的温度变化时,将引起反应堆临界系数的变化。本文利用蒙特卡罗方法的MCNP程序,从压水堆的压力容器,燃料组件和控制棒驱动组件等元件出发建模,对温度改变之于反应堆剩余反应性影响做了相应研究。模拟结果表明:随着温度的升高,反应堆剩余反应性逐渐减。并且得到结果与实际值较符合。%Reactor core temperature changing, the neutron spectrum, micro-cross-section will change accordingly. So many parameters of the reaction, such as the thermal neutron utilization factor, the probability of escaping resonance capture, are all a function of temperature. Thus, when the various components of the reactor temperature change, it will cause the reactor critical factor changing. In this paper, by using MCNP program, based on modeling on the PWR pressure container, fuel module and drive component elements, the corresponding research on the effects of reactor excess reactivity under different temperature were analyzed. The simulation results show that, with the increasing of temperature, reactor excess reactivity decrease gradually. And the results have a good fit with the practical values

  20. ITER: The International Thermonuclear Experimental Reactor and the nuclear weapons proliferation implications of thermonuclear-fusion energy systems

    OpenAIRE

    Gsponer, Andre; Hurni, Jean-Pierre

    2004-01-01

    This report contains two parts: (1) A list of "points" highlighting the strategic-political and military-technical reasons and implications of the very probable siting of ITER (the International Thermonuclear Experimental Reactor) in Japan, which should be confirmed sometimes in early 2004. (2) A technical analysis of the nuclear weapons proliferation implications of inertial- and magnetic-confinement fusion systems substantiating the technical points highlighted in the first part, and showin...

  1. Unfolding the measured neutron spectra in the irradiation chamber of the UZrH reactor using iterative method

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    In the procedure of neutron fluence measurement in the whole energy range (10-4 eV~18 MeV), in the irradiation chamber of a UZrH reactor, the neutron energy spectra are unfolded using the method of minimizing directed divergence and SAND-Ⅱ, which are used broadly at home and abroad. These methods belong to the iterative methods.In this article, the procedure of the spectra unfolding using the two methods is described in detail. The neutron spectrum distribution unfolded by the two methods agree well with each other. In the end, the major differences of the two iterative methods are compared with each other, and the main factors affecting the accuracy of the spectra unfolding with the iterative method are discussed.

  2. Conceptual core design of small traveling wave reactors based on coupled code of MCNP-ORIGEN2%基于MCNP-ORIGEN2耦合程序的小型行波堆堆芯概念设计

    Institute of Scientific and Technical Information of China (English)

    侯景景; 王世庆; 蔡云; 汪占河; 向茜; 刘海峰

    2015-01-01

    研究设计了基于中国实验快堆(China Experimental Fast Reactor,CEFR)的小型“行波”概念堆.采用中子输运程序MCNP和点燃耗程序ORIGEN2的耦合程序进行堆芯设计,重点研究了不同点火组件的富集度和不同布料方案对小型堆的物理参数的影响,设计堆芯寿期为30 a,并给出相应的倒料方案.不同点火组件富集度对比结果表明,小堆需要选取合适的富集度,富集度太低无法维持临界,而太高会影响堆芯增殖效应;而低泄漏和棋盘式布料两种方式对比结果表明,后者的增殖组件增殖效应明显高于前者.最终确定倒料周期为8a,倒料三次,堆芯实现较长寿期,且整个寿期内反应性变化小,各组件燃耗深度相对均匀,组件平均卸料燃耗深度约为238 MWD/kgHM.

  3. Investigative studies on the effects of cadmium rabbits on high enriched uranium-fueled and low enriched uranium-fueled cores of Ghana Research Reactor-1 using MCNP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Boffie, J., E-mail: jboffie@yahoo.com [Department of Nuclear Engineering and Material Science, School of Nuclear and Allied Sciences (SNAS), University of Ghana, P.O. Box AE 1, Atomic Energy, Accra (Ghana); National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG 80, Legon, Accra (Ghana); Akaho, E.H.K. [Department of Nuclear Engineering and Material Science, School of Nuclear and Allied Sciences (SNAS), University of Ghana, P.O. Box AE 1, Atomic Energy, Accra (Ghana); Nyarko, B.J.B.; Odoi, H.C.; Tuffour-Achampong, K.; Abrefah, R.G. [Department of Nuclear Engineering and Material Science, School of Nuclear and Allied Sciences (SNAS), University of Ghana, P.O. Box AE 1, Atomic Energy, Accra (Ghana); National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG 80, Legon, Accra (Ghana)

    2013-12-15

    Highlights: • The operating parameters for both the HEU core and proposed LEU core were similar. • The length of the Cd in the capsules must be increased for its use in the LEU core. • Cd rabbits can emergently be used to shut down MNSRs. - Abstract: Miniature Neutron Source Reactors (MNSRs) are noted to be among highly safe research reactors. However, because of its use of one control rod for reactivity control and shutdown purposes, alternative methods of shutting it down are important. The Ghana MNSR uses four cadmium rabbits of approximate dimensions 6.5 cm × 5.0 cm × 0.1 cm and mass of 9.48 g each to emergently shut down the reactor. The Monte Carlo N-Particle code; version 5, (MCNP5) was used to design the high enriched uranium (HEU) and low enriched uranium (LEU) cores of the MNSR with four cadmium rabbits inserted in four inner irradiation sites of each core. The operating parameters and shutdown parameters for both cores with the central control rod (CCR) either fully withdrawn or fully inserted had similar results with the HEU core having slightly better results in terms of safety. However, the results show that the four inserted cadmium rabbits make the HEU core subcritical whiles in the LEU core, it still remains critical (k{sub eff} = 1.00005 ± 0.00007). The length of the cadmium material in each cadmium rabbit must therefore be increased by at least 0.5 cm in order to attain subcriticality (k{sub eff} = 0.99989 ± 0.00006) and shutdown margin of 0.11 mk when inserted in the LEU core.

  4. Nuclear Analyses of Indian LLCB Test Blanket System in ITER

    Science.gov (United States)

    Swami, H. L.; Shaw, A. K.; Danani, C.; Chaudhuri, Paritosh

    2017-04-01

    Heading towards the Nuclear Fusion Reactor Program, India is developing Lead Lithium Ceramic Breeder (LLCB) tritium breeding blanket for its future fusion Reactor. A mock-up of the LLCB blanket is proposed to be tested in ITER equatorial port no.2, to ensure the overall performance of blanket in reactor relevant nuclear fusion environment. Nuclear analyses play an important role in LLCB Test Blanket System design & development. It is required for tritium breeding estimation, thermal-hydraulic design, coolants process design, radioactive waste management, equipment maintenance & replacement strategies and nuclear safety. The nuclear behaviour of LLCB test blanket module in ITER is predicated in terms of nuclear responses such as tritium production, nuclear heating, neutron fluxes and radiation damages. Radiation shielding capability of LLCB TBS inside and outside bio-shield was also assessed to fulfill ITER shielding requirements. In order to supports the rad-waste and safety assessment, nuclear activation analyses were carried out and radioactivity data were generated for LLCB TBS components. Nuclear analyses of LLCB TBS are performed using ITER recommended nuclear analyses codes (i.e. MCNP, EASY), nuclear cross section data libraries (i.e. FENDL 2.1, EAF) and neutronic model (ITER C-lite v.l). The paper describes a comprehensive nuclear performance of LLCB TBS in ITER.

  5. ITER: The International Thermonuclear Experimental Reactor and the nuclear weapons proliferation implications of thermonuclear-fusion energy

    CERN Document Server

    Gsponer, A; Gsponer, Andre; Hurni, Jean-Pierre

    2004-01-01

    This paper contains two parts: (I) A list of "points" highlighting the strategic-political and military-technical reasons and implications of the very probable siting of ITER (the International Thermonuclear Experimental Reactor) in Japan, which should be confirmed sometimes in early 2004. (II) A technical analysis of the nuclear weapons proliferation implications of inertial- and magnetic-confinement fusion systems substantiating the technical points highlighted in the first part, and showing that while full access to the physics of thermonuclear weapons is the main implication of ICF, full access to large-scale tritium technology is the main proliferation impact of MCF. The conclusion of the paper is that siting ITER in a country such as Japan, which already has a large separated-plutonium stockpile, and an ambitious laser-driven ICF program (comparable in size and quality to those of the United States or France) will considerably increase its latent (or virtual) nuclear weapons proliferation status, and fo...

  6. Validation suite for MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Mosteller, R. D. (Russell D.)

    2002-01-01

    Two validation suites, one for criticality and another for radiation shielding, have been defined and tested for the MCNP Monte Carlo code. All of the cases in the validation suites are based on experiments so that calculated and measured results can be compared in a meaningful way. The cases in the validation suites are described, and results from those cases are discussed. For several years, the distribution package for the MCNP Monte Carlo code1 has included an installation test suite to verify that MCNP has been installed correctly. However, the cases in that suite have been constructed primarily to test options within the code and to execute quickly. Consequently, they do not produce well-converged answers, and many of them are physically unrealistic. To remedy these deficiencies, sets of validation suites are being defined and tested for specific types of applications. All of the cases in the validation suites are based on benchmark experiments. Consequently, the results from the measurements are reliable and quantifiable, and calculated results can be compared with them in a meaningful way. Currently, validation suites exist for criticality and radiation-shielding applications.

  7. MCNP: Multigroup/adjoint capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.; Redmond, E.L. II; Palmtag, S.P.; Hendricks, J.S.

    1994-04-01

    This report discusses various aspects related to the use and validity of the general purpose Monte Carlo code MCNP for multigroup/adjoint calculations. The increased desire to perform comparisons between Monte Carlo and deterministic codes, along with the ever-present desire to increase the efficiency of large MCNP calculations has produced a greater user demand for the multigroup/adjoint capabilities. To more fully utilize these capabilities, we review the applications of the Monte Carlo multigroup/adjoint method, describe how to generate multigroup cross sections for MCNP with the auxiliary CRSRD code, describe how to use the multigroup/adjoint capability in MCNP, and provide examples and results indicating the effectiveness and validity of the MCNP multigroup/adjoint treatment. This information should assist users in taking advantage of the MCNP multigroup/adjoint capabilities.

  8. Current generation by helicons and lower hybrid waves in modern tokamaks and reactors ITER and DEMO. Scenarios, modeling and antennae

    Energy Technology Data Exchange (ETDEWEB)

    Vdovin, V. L., E-mail: vdov@nfi.kiae.ru [National Research Centre ' Kurchatov Institute,' (Russian Federation)

    2013-02-15

    The innovative concept and 3D full-wave code modeling the off-axis current drive by radio-frequency (RF) waves in large-scale tokamaks, ITER and DEMO, for steady-state operation with high efficiency is proposed. The scheme uses the helicon radiation (fast magnetosonic waves at high (20-40) ion cyclotron frequency harmonics) at frequencies of 500-700 MHz propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by helicons, in conjunction with the bootstrap current, ensure the maintenance of a given value of the total current in the stability margin q(0) {>=} 2 and q(a) {>=} 4, and will help to have regimes with a negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure {beta}{sub N} > 3 (the so-called advanced scenarios) of interest for the commercial reactor. Modeling with full-wave three-dimensional codes PSTELION and STELEC showed flexible control of the current profile in the reactor plasmas of ITER and DEMO, using multiple frequencies, the positions of the antennae and toroidal wave slow down. Also presented are the results of simulations of current generation by helicons in the DIII-D, T-15MD, and JT-60AS tokamaks. Commercially available continuous-wave klystrons of the MW/tube range are promising for commercial stationary fusion reactors. The compact antennae of the waveguide type are proposed, and an example of a possible RF system for today's tokamaks is given. The advantages of the scheme (partially tested at lower frequencies in tokamaks) are a significant decline in the role of parametric instabilities in the plasma periphery, the use of electrically strong resonator-waveguide type antennae, and substantially greater antenna-plasma coupling.

  9. Current generation by helicons and lower hybrid waves in modern tokamaks and reactors ITER and DEMO. Scenarios, modeling and antennae

    Science.gov (United States)

    Vdovin, V. L.

    2013-02-01

    The innovative concept and 3D full-wave code modeling the off-axis current drive by radio-frequency (RF) waves in large-scale tokamaks, ITER and DEMO, for steady-state operation with high efficiency is proposed. The scheme uses the helicon radiation (fast magnetosonic waves at high (20-40) ion cyclotron frequency harmonics) at frequencies of 500-700 MHz propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by helicons, in conjunction with the bootstrap current, ensure the maintenance of a given value of the total current in the stability margin q(0) ≥ 2 and q( a) ≥ 4, and will help to have regimes with a negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure β N > 3 (the so-called advanced scenarios) of interest for the commercial reactor. Modeling with full-wave three-dimensional codes PSTELION and STELEC showed flexible control of the current profile in the reactor plasmas of ITER and DEMO, using multiple frequencies, the positions of the antennae and toroidal wave slow down. Also presented are the results of simulations of current generation by helicons in the DIII-D, T-15MD, and JT-60AS tokamaks. Commercially available continuous-wave klystrons of the MW/tube range are promising for commercial stationary fusion reactors. The compact antennae of the waveguide type are proposed, and an example of a possible RF system for today's tokamaks is given. The advantages of the scheme (partially tested at lower frequencies in tokamaks) are a significant decline in the role of parametric instabilities in the plasma periphery, the use of electrically strong resonator-waveguide type antennae, and substantially greater antenna-plasma coupling.

  10. Two conceptual designs of helical fusion reactor FFHR-d1A based on ITER technologies and challenging ideas

    Science.gov (United States)

    Sagara, A.; Miyazawa, J.; Tamura, H.; Tanaka, T.; Goto, T.; Yanagi, N.; Sakamoto, R.; Masuzaki, S.; Ohtani, H.; The FFHR Design Group

    2017-08-01

    The Fusion Engineering Research Project (FERP) at the National Institute for Fusion Science (NIFS) is conducting conceptual design activities for the LHD-type helical fusion reactor FFHR-d1A. This paper newly defines two design options, ‘basic’ and ‘challenging.’ Conservative technologies, including those that will be demonstrated in ITER, are chosen in the basic option in which two helical coils are made of continuously wound cable-in-conduit superconductors of Nb3Sn strands, the divertor is composed of water-cooled tungsten monoblocks, and the blanket is composed of water-cooled ceramic breeders. In contrast, new ideas that would possibly be beneficial for making the reactor design more attractive are boldly included in the challenging option in which the helical coils are wound by connecting high-temperature REBCO superconductors using mechanical joints, the divertor is composed of a shower of molten tin jets, and the blanket is composed of molten salt FLiNaBe including Ti powers to increase hydrogen solubility. The main targets of the challenging option are early construction and easy maintenance of a large and three-dimensionally complicated helical structure, high thermal efficiency, and, in particular, realistic feasibility of the helical reactor.

  11. Design and development of in-vessel viewing periscope for ITER (International Thermonuclear Experimental Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Obara, Kenjiro; Kakudate, Satoshi; Ito, Akira; Shibanuma, Kiyoshi; Tada, Eisuke [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka , Ibaraki (Japan)

    1999-02-01

    An in-vessel viewing system is essential not only to detect and locate damage of components exposed to plasma, but also to monitor and assist in-vessel maintenance operation. In ITER, the in-vessel viewing system must be capable of operating at high temperature (200degC), under intense gamma radiation (30 kGy/h) and high vacuum or 1 bar inert gas. A periscope-type in-vessel viewing system has been chosen as a reference of the ITER in-vessel viewing system due to its wide viewing capability and durability for sever environments. According to the ITER research and development program, a full-scale radiation hard periscope with a length of 15 m has been successfully developed by the Japan Home Team. The performance tests have been shown sufficient capability at high temperature up to 250degC and radiation resistance over 100 MGy. This report describes the design and R and D results of the ITER in-vessel viewing periscope based on the development of 15-m-length radiation hard periscope. (author)

  12. Design and development of in-vessel viewing periscope for ITER (International Thermonuclear Experimental Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Obara, Kenjiro; Kakudate, Satoshi; Ito, Akira; Shibanuma, Kiyoshi; Tada, Eisuke [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka , Ibaraki (Japan)

    1999-02-01

    An in-vessel viewing system is essential not only to detect and locate damage of components exposed to plasma, but also to monitor and assist in-vessel maintenance operation. In ITER, the in-vessel viewing system must be capable of operating at high temperature (200degC), under intense gamma radiation (30 kGy/h) and high vacuum or 1 bar inert gas. A periscope-type in-vessel viewing system has been chosen as a reference of the ITER in-vessel viewing system due to its wide viewing capability and durability for sever environments. According to the ITER research and development program, a full-scale radiation hard periscope with a length of 15 m has been successfully developed by the Japan Home Team. The performance tests have been shown sufficient capability at high temperature up to 250degC and radiation resistance over 100 MGy. This report describes the design and R and D results of the ITER in-vessel viewing periscope based on the development of 15-m-length radiation hard periscope. (author)

  13. MCNP{trademark} Monte Carlo: A precis of MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Adams, K.J.

    1996-06-01

    MCNP{trademark} is a general purpose three-dimensional time-dependent neutron, photon, and electron transport code. It is highly portable and user-oriented, and backed by stringent software quality assurance practices and extensive experimental benchmarks. The cross section database is based upon the best evaluations available. MCNP incorporates state-of-the-art analog and adaptive Monte Carlo techniques. The code is documented in a 600 page manual which is augmented by numerous Los Alamos technical reports which detail various aspects of the code. MCNP represents over a megahour of development and refinement over the past 50 years and an ongoing commitment to excellence.

  14. MCNP-REN a Monte Carlo tool for neutron detector design

    CERN Document Server

    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...

  15. Whole-Core Monte Carlo Calculation Based on Fission and Surface Source Iteration Method Applied to a Fast Reactor Core Configuration

    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.

  16. MCNP-DSP users manual

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.E.

    1997-01-01

    The Monte Carlo code MCNP-DSP was developed from the Los Alamos MCNP4a code to calculate the time and frequency response statistics obtained from the {sup 252}Cf-source-driven frequency analysis measurements. This code can be used to validate calculational methods and cross section data sets from subcritical experiments. This code provides a more general model for interpretation and planning of experiments for nuclear criticality safety, nuclear safeguards, and nuclear weapons identification and replaces the use of point kinetics models for interpreting the measurements. The use of MCNP-DSP extends the usefulness of this measurement method to systems with much lower neutron multiplication factors.

  17. Breeding blanket design for ITER and prototype (DEMO) fusion reactors and breeding materials issues

    Energy Technology Data Exchange (ETDEWEB)

    Takatsu, H.; Enoeda, M. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1998-03-01

    Current status of the designs of the ITER breeding blanket and DEMO blankets is introduced placing emphasis on the breeding materials selection and related issues. The former design is based on the up-to-date design activities, as of October 1997, being performed jointly by Joint Central Team (JCT) and Home Teams (HT`s), while the latter is based on the DEMO blanket test module designs being proposed by each Party at the TBWG (Test Blanket Working Group) meetings. (J.P.N.)

  18. Hydrogen permeability technique in situ reactor irradiation for ITER structural materials

    Energy Technology Data Exchange (ETDEWEB)

    Tazhibaeva, I.L.; Shestakov, V.P.; Chikhray, E.V.; Romanenko, O.G.; Klepikov, A.K. [National Univ. of Kazakhastan, Almaty (Kazakhstan); Cherepnin, Y.S.; Kenzhin, E.A.; Basov, A.A.; Kolodeshnikov, A.A. [National Nuclear Center of Kazakhstan, Krasnoarmeyskaya (Kazakhstan)

    1995-10-01

    This work develops the technique of hydrogen diffusion parameters measurements in metal materials during the process of reactor irradiation. The possibility of irradiation stimulated hydrogen diffusion for 08Cr18Ni10Ti was obtained. Constants of diffusion, permeation and solution while the process of irradiation were measured as a result of the work. Formally calculated activation energies of diffusion and permeation were obtained to decrease while the solution heat was increasing. 5 refs., 4 figs., 1 tab.

  19. Transient Behaviour of Superconducting Magnet Systems of Fusion Reactor ITER during Safety Discharge

    Directory of Open Access Journals (Sweden)

    A. M. Miri

    2008-01-01

    Full Text Available To investigate the transient behaviour of the toroidal and poloidal field coils magnet systems of the International Thermonuclear Experimental Reactor during safety discharge, network models with lumped elements are established. Frequency-dependant values of the network elements, that is, inductances and resistances are calculated with the finite element method. That way, overvoltages can be determined. According to these overvoltages, the insulation coordination of coils has to be selected.

  20. A new MCNP{trademark} test set

    Energy Technology Data Exchange (ETDEWEB)

    Brockhoff, R.C.; Hendricks, J.S.

    1994-09-01

    The MCNP test set is used to test the MCNP code after installation on various computer platforms. For MCNP4 and MCNP4A this test set included 25 test problems designed to test as many features of the MCNP code as possible. A new and better test set has been devised to increase coverage of the code from 85% to 97% with 28 problems. The new test set is as fast as and shorter than the MCNP4A test set. The authors describe the methodology for devising the new test set, the features that were not covered in the MCNP4A test set, and the changes in the MCNP4A test set that have been made for MCNP4B and its developmental versions. Finally, new bugs uncovered by the new test set and a compilation of all known MCNP4A bugs are presented.

  1. Reactors

    CERN Document Server

    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.

  2. Radiation tests of ITER diagnostic system materials in the BOR-60 reactor

    Science.gov (United States)

    Revyakin, Yu. L.; Kosenkov, V. M.; Bender, S. E.; Belyakov, V. A.

    1996-10-01

    An in-pile experiment was conducted up to a fluence of 9.9 × 10 25 m 2 ( E ≫ 0.1 MeV) which investigated the following electrophysical characteristics of a cable with mineral insulation and nickel conductor: insulation resistance, radiation-induced current and EMF. Irradiation was also performed in the BOR-60 reactor up to a fluence 10 23 m -2 on six crystal types for the monochromator: mica, LiF, multilayered mirrors Fe/C, W/Si, Cr/C and Mo/Si. Change of the reflectivity, width and shape of diffraction reflections were investigated.

  3. MCNP and GADRAS Comparisons

    Energy Technology Data Exchange (ETDEWEB)

    Klasky, Marc Louis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Myers, Steven Charles [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); James, Michael R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mayo, Douglas R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-04-19

    To facilitate the timely execution of System Threat Reviews (STRs) for DNDO, and also to develop a methodology for performing STRs, LANL performed comparisons of several radiation transport codes (MCNP, GADRAS, and Gamma-Designer) that have been previously utilized to compute radiation signatures. While each of these codes has strengths, it is of paramount interest to determine the limitations of each of the respective codes and also to identify the most time efficient means by which to produce computational results, given the large number of parametric cases that are anticipated in performing STR's. These comparisons serve to identify regions of applicability for each code and provide estimates of uncertainty that may be anticipated. Furthermore, while performing these comparisons, examination of the sensitivity of the results to modeling assumptions was also examined. These investigations serve to enable the creation of the LANL methodology for performing STRs. Given the wide variety of radiation test sources, scenarios, and detectors, LANL calculated comparisons of the following parameters: decay data, multiplicity, device (n,γ) leakages, and radiation transport through representative scenes and shielding. This investigation was performed to understand potential limitations utilizing specific codes for different aspects of the STR challenges.

  4. Fission Matrix Capability for MCNP Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Carney, Sean E. [Los Alamos National Laboratory; Brown, Forrest B. [Los Alamos National Laboratory; Kiedrowski, Brian C. [Los Alamos National Laboratory; Martin, William R. [Los Alamos National Laboratory

    2012-09-05

    In a Monte Carlo criticality calculation, before the tallying of quantities can begin, a converged fission source (the fundamental eigenvector of the fission kernel) is required. Tallies of interest may include powers, absorption rates, leakage rates, or the multiplication factor (the fundamental eigenvalue of the fission kernel, k{sub eff}). Just as in the power iteration method of linear algebra, if the dominance ratio (the ratio of the first and zeroth eigenvalues) is high, many iterations of neutron history simulations are required to isolate the fundamental mode of the problem. Optically large systems have large dominance ratios, and systems containing poor neutron communication between regions are also slow to converge. The fission matrix method, implemented into MCNP[1], addresses these problems. When Monte Carlo random walk from a source is executed, the fission kernel is stochastically applied to the source. Random numbers are used for: distances to collision, reaction types, scattering physics, fission reactions, etc. This method is used because the fission kernel is a complex, 7-dimensional operator that is not explicitly known. Deterministic methods use approximations/discretization in energy, space, and direction to the kernel. Consequently, they are faster. Monte Carlo directly simulates the physics, which necessitates the use of random sampling. Because of this statistical noise, common convergence acceleration methods used in deterministic methods do not work. In the fission matrix method, we are using the random walk information not only to build the next-iteration fission source, but also a spatially-averaged fission kernel. Just like in deterministic methods, this involves approximation and discretization. The approximation is the tallying of the spatially-discretized fission kernel with an incorrect fission source. We address this by making the spatial mesh fine enough that this error is negligible. As a consequence of discretization we get a

  5. MCNP(TM) Version 5.

    Energy Technology Data Exchange (ETDEWEB)

    Cox, L. J. (Lawrence J.); Barrett, R. F. (Richard F.); Booth, Thomas Edward; Briesmeister, Judith F.; Brown, F. B. (Forrest B.); Bull, J. S. (Jeffrey S.); Giesler, G. C. (Gregg Carl); Goorley, J. T. (John T.); Mosteller, R. D. (Russell D.); Forster, R. A. (R. Arthur); Post, S. E. (Susan E.); Prael, R. E. (Richard E.); Selcow, Elizabeth Carol,; Sood, A. (Avneet)

    2002-01-01

    The Monte Carlo transport workhorse, MCNP, is undergoing a massive renovation at Los Alamos National Laboratory (LANL) in support of the Eolus Project of the Advanced Simulation and Computing (ASCI) Program. MCNP Version 5 (V5) (expected to be released to RSICC in Spring, 2002) will consist of a major restructuring from FORTRAN-77 (with extensions) to ANSI-standard FORTRAN-90 with support for all of the features available in the present release (MCNP-4C2/4C3). To most users, the look-and-feel of MCNP will not change much except for the improvements (improved graphics, easier installation, better online documentation). For example, even with the major format change, full support for incremental patching will still be provided. In addition to the language and style updates, MCNP V5 will have various new user features. These include improved photon physics, neutral particle radiography, enhancements and additions to variance reduction methods, new source options, and improved parallelism support (PVM, MPI, OpenMP).

  6. MCNP-DSP USERS MANUAL

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.E.

    2001-01-19

    The Monte Carlo code MCNP-DSP was developed from the Los Alamos MCNP4a code to calculate the time and frequency response statistics obtained from subcritical measurements. The code can be used to simulate a variety of subcritical measurements including source-driven noise analysis, Rossi-{alpha}, pulsed source, passive frequency analysis, multiplicity, and Feynman variance measurements. This code can be used to validate Monte Carlo methods and cross section data sets with subcritical measurements and replaces the use of point kinetics models for interpreting subcritical measurements.

  7. Dr Robert Aymar, Director of the International Thermonuclear Experimental Reactor (ITER), was nominated to succeed Professor Luciano Maiani as CERN's Director General, to take office on 1 January 2004.

    CERN Multimedia

    2002-01-01

    Dr Robert Aymar, Director of the International Thermonuclear Experimental Reactor (ITER), was nominated to succeed Professor Luciano Maiani as CERN's Director General, to take office on 1 January 2004.

  8. General purpose photoneutron production in MCNP4A

    Energy Technology Data Exchange (ETDEWEB)

    Gallmeier, F.X.

    1995-08-01

    A photoneutron production option was implemented in the MCNP4A code, mainly to supply a tool for reactor shielding calculations in beryllium and heavy water environments of complicated three-dimensional geometries. Photoneutron production cross sections for deuterium and beryllium were created. Subroutines were developed to calculate the probability of photoneutron production at photon collision sites and the energy and flight direction of the created photoneutrons. These subroutines were implemented into MCNP4A. Some small program changes were necessary for processing the input to read the photoneutron production cross sections and to install a photoneutron switch. Some arrays were installed or extended to sample photoneutron creation and loss information, and output routines were changed to give the appropriate summary tables. To verify and validate the photoneutron production data and the MCNP4A implementations, the yields of photoneutron sources were calculated and compared with experiments. In the case of deuterium-based photoneutron sources, the calculations agreed well with the experiments; the beryuium-based photoneutron source calculations were up to 30% higher compared with the measurements. More accurate beryllium photoneutron cross sections would be desirable. To apply the developed method to a real shielding problem, the fast neutron fluxes in the heavy-water-filled reflector vessel of the Advanced Neutron Source reactor were investigated and compared with published DORT calculations. Considering the complete independence between the calculations, the merely 10 to 20% lower fluxes obtained with MCNP4A, compared against the DORT results, were more than satisfactory, as the discrepancy is based primarily on differences in the calculated thermal neutron fluxes.

  9. 基于MCNP和ORIGEN2耦合程序的IHNI-1型堆裂变产物中毒及燃耗分析%The fission product poisoning and burnup calculation for IHNI-1 reactor based on coupled code of MCNP-ORIGEN2

    Institute of Scientific and Technical Information of China (English)

    张信一; 赵柱民; 江新标; 郭和伟; 陈立新; 周永茂

    2012-01-01

    To calculate the fission product poisoning and bumup of the reactor accurately, the paper sets up the coupled calculation methods based on MCNP code and ORIGEN2 code and program data translation, cross section revision and date interface codes. Making use of elaborate reactor model to calculate the fission product poisoning and bumup for in-hospital neutron irradiator mark 1 reactor.%为了准确地计算反应堆的裂变产物中毒和燃耗问题,开发了一套蒙特卡罗方法程序系统.利用通用的燃耗计算方法,基于MCNP和ORIGEN2,编写了相关的数据转换、截面修正、数据接口程序,实现了MCNP和ORIGEN2程序的耦合.采用堆芯精细结构划分,对医院中子照射器Ⅰ型堆裂变产物中毒和燃耗进行了计算分析.

  10. Evaluation of the thermal neutron flux in samples of Al–Au alloy irradiated in the carrousel channels of the TRIGA MARK I IPR-R1 reactor using MCNP code

    Energy Technology Data Exchange (ETDEWEB)

    Salomé, J.A.D.; Guerra, B.T. [Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, Av. Antônio Carlos, 6627 – PCA1 – Anexo Engenharia – Pampulha, CEP 31270-901, Belo Horizonte, MG (Brazil); Pereira, C., E-mail: claubia@nuclear.ufmg.br [Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, Av. Antônio Carlos, 6627 – PCA1 – Anexo Engenharia – Pampulha, CEP 31270-901, Belo Horizonte, MG (Brazil); Menezes, M.Â.B.C. de [Centro de Desenvolvimento da Tecnologia Nuclear, Comissão Nacional de Energia Nuclear, Campus da UFMG, Av. Antônio Carlos, 6627 31270-901, P.O. Box 941, Belo Horizonte, MG (Brazil); Silva, C.A.M. da [Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, Av. Antônio Carlos, 6627 – PCA1 – Anexo Engenharia – Pampulha, CEP 31270-901, Belo Horizonte, MG (Brazil); Dalle, H.M. [Centro de Desenvolvimento da Tecnologia Nuclear, Comissão Nacional de Energia Nuclear, Campus da UFMG, Av. Antônio Carlos, 6627 31270-901, P.O. Box 941, Belo Horizonte, MG (Brazil)

    2014-07-01

    Highlights: • The TRIGA IPR-R1 was modelled using MCNP. • The thermal neutron flux through the samples in eleven irradiation channels was obtained. • The simulated results were compared to experimental values. • The relative error, the relative trend, the z-score test and uncertainty were analysed. - Abstract: The TRIGA IPR-R1 was modelled using MCNP. The model consists of a cylinder filled with water, fuel elements, radial reflectors, central tube, control rods and neutron source. Around the core is placed the Rotary Specimen Rack (RSR) with adequate groove to insert the samples to irradiation. The values of the thermal neutron flux through the samples in eleven irradiation channels were simulated and compared to the experimental results to validate the model. After that, the values of the thermal neutron flux, in the same channels, were simulated on two horizontal planes at different heights and compared to validate the model. These channels were characterized as representative channels of the neutron flux distribution in the RSR. To evaluate the results, the relative errors, the relative trend, the z-score test and the relevance to a confidence interval of 95% were analysed. Good agreement has been obtained for the most channels when compared with the experimental results.

  11. Fusion Neutron Flux Monitor for ITER

    Institute of Scientific and Technical Information of China (English)

    YANG Jinwei; YANG Qingwei; XIAO Gongshan; ZHANG Wei; SONG Xianying; LI Xu

    2008-01-01

    Neutron flux monitor (NFM) as an important diagnostic sub-system in ITER (international thermonuclear experimental reactor) provides a global neutron source intensity, fusion power and neutron flux in real time. Three types of neutron flux monitor assemblies with different sensitivities and shielding materials have been designed. Through MCNP (Mante-Carlo neutral particle transport code) calculations, this extended system of NFM can detect the neutron flux in a range of 104 n/(cm2·s) to 1014 n/(cm2·s). It is capable of providing accurate neutron yield measurements for all operational modes encountered in the ITER experiments including the in-situ calibration. Combining both the counting mode and Campbelling (MSV; Mean Square Voltage) mode in the signal processing units, the requirement of the dynamic range (107) for these NFMs and time resolution (1 ms) can be met. Based on a uncertainty analysis, the estimated absolute measurement accuracies of the total fusion neutron yield can reach the required 10% level in both the early stage of the DD-phase and the full power DT operation mode. In the advanced DD-phase, the absolute measurement accuracy would be better than 20%.

  12. MCNP6 Fission Multiplicity with FMULT Card

    Energy Technology Data Exchange (ETDEWEB)

    Wilcox, Trevor [Los Alamos National Laboratory; Fensin, Michael Lorne [Los Alamos National Laboratory; Hendricks, John S. [Los Alamos National Laboratory; James, Michael R. [Los Alamos National Laboratory; McKinney, Gregg W. [Los Alamos National Laboratory

    2012-06-18

    With the merger of MCNPX and MCNP5 into MCNP6, MCNP6 now provides all the capabilities of both codes allowing the user to access all the fission multiplicity data sets. Detailed in this paper is: (1) the new FMULT card capabilities for accessing these different data sets; (2) benchmark calculations, as compared to experiment, detailing the results of selecting these separate data sets for thermal neutron induced fission on U-235.

  13. Current generation by helicons and LH waves in modern tokamaks and reactors FNSF-AT, ITER and DEMO. Scenarios, modeling and antennae

    Energy Technology Data Exchange (ETDEWEB)

    Vdovin, V. [NRC Kurchatov Institute Tokamak Physics Institute, Moscow (Russian Federation)

    2014-02-12

    The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20–40) IC frequency harmonics) at frequencies of 500–1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure β{sub N} > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D – Kurchatov Institute experiment on helicons CD [1].

  14. Current generation by helicons and LH waves in modern tokamaks and reactors FNSF-AT, ITER and DEMO. Scenarios, modeling and antennae

    Science.gov (United States)

    Vdovin, V.

    2014-02-01

    The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20-40) IC frequency harmonics) at frequencies of 500-1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure βN > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D - Kurchatov Institute experiment on helicons CD [1].

  15. Characteristic evaluation of high compression seismic isolator for International Thermonuclear Experimental Reactor (ITER). Verification test of sub-scaled rubber bearings. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, Hiroyuki [Hitachi Ltd., Tokyo (Japan); Nakahira, Masataka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yabana, Shuichi; Matsuda, Akihiro; Ohtori, Yasuki [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

    2001-11-01

    The International Thermonuclear Experimental Reactor (ITER) is designed to withstand the seismic load of 2 m/s{sup 2} at the ground level as a standard seismic condition. In case of severe seismic load over 2 m/s{sup 2}, an application of the seismic isolation to the tokamak building is studied so as to reduce the seismic load below 2 m/s{sup 2}. The seismic isolation with high compressive pressure of 7.35MPa to 14.7MPa is considered as a candidate, because the tokamak weight is large to the building size and the number of seismic isolator (rubber bearing) is limited in the available space of the building. Although many studies were executed in the past in order to apply the seismic isolation to the nuclear plant, the test data can not be applied to the ITER due to low compressive pressure of about 2.45MPa to 4.90MPa. Based on the above, it is therefore necessary to evaluate the various kinds of dynamic and mechanical characteristics of the rubber bearings under the high compressive pressure and to obtain the database for the design of the seismic isolation system of the ITER. The report describes the summary of the test results of the sub-scaled rubber bearings executed under the high compression condition in 1997 to 1999. (author)

  16. The measurement of neutron and neutron induced photon spectra in fusion reactor related assemblies

    CERN Document Server

    Unholzer, S; Klein, H; Seidel, K

    2002-01-01

    The spectral neutron and photon fluence (or flux) measured outside and inside of assemblies related to fusion reactor constructions are basic quantities of fusion neutronics. The comparison of measured spectra with the results of MCNP neutron and photon transport calculations allows a crucial test of evaluated nuclear data as generally used in fusion applications to be carried out. The experiments concern mixed neutron/photon fields with about the same intensity of the two components. An NE-213 scintillation spectrometer, well described by response matrices for both neutrons and photons, is used as proton-recoil and Compton spectrometer. The experiments described here in more detail address the background problematic of two applications, an iron benchmark experiment with an ns-pulsed neutron source and a deep penetration mock-up experiment for the investigation of the ITER in-board shield system. The measured spectral neutron and photon fluences are compared with spectra calculated with the MCNP code on the b...

  17. MCNP{trademark} Software Quality Assurance plan

    Energy Technology Data Exchange (ETDEWEB)

    Abhold, H.M.; Hendricks, J.S.

    1996-04-01

    MCNP is a computer code that models the interaction of radiation with matter. MCNP is developed and maintained by the Transport Methods Group (XTM) of the Los Alamos National Laboratory (LANL). This plan describes the Software Quality Assurance (SQA) program applied to the code. The SQA program is consistent with the requirements of IEEE-730.1 and the guiding principles of ISO 900.

  18. CONDOR-CITVAP-MCNP calculation line description

    Energy Technology Data Exchange (ETDEWEB)

    Villarino, Eduardo Anibal [INVAP S.E., San Carlos de Bariloche (Argentina)

    2002-07-01

    A general description of the CONDOR-CITVAP-MCNP calculation line is given. This calculation line starts at cross section library and allows burnup dependent detailed calculation using MCNP. This calculation line is divided in two main methodologies: CONDOR-CITVAP that allows the 3-Dimensional core burnup calculation and MCNP that performs detailed transport calculations, both methodologies are coupled using the NDDUMP code. A short description of the used codes are given: CONDOR code performs the cell calculation, generating burnup dependent macroscopic cross section and burnup dependent numerical densities per material. CITVAP codes perform the burnup dependent core calculation, including the fuel management and calculates the burnup distribution per material. NDDUMP code generates materials burnup dependent numerical densities to be used by MCNP code. This paper presents a detailed description of the CONDOR-CITVAP-MCNP calculation line and a numerical comparison of the proposed methodology. (author)

  19. Material Assessment for ITER's Collective Thomson Scattering first mirror

    Energy Technology Data Exchange (ETDEWEB)

    Santos, R.; Policarpo, H.; Goncalves, B.; Varela, P. [Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Universidade de Lisboa, 1049-001 Lisboa (Portugal); Nonboel, E.; Klinkby, E.; Lauritzen, B. [Center for Nuclear Technologies, Technical University of Denmark (Denmark); Romanets, Y.; Luis, R.; Vaz, P. [Centro de Ciencias e Tecnologias Nucleares, Instituto Superior Tecnico, Universidade de Lisboa (Portugal)

    2015-07-01

    The International Thermonuclear Energy Reactor (ITER) Collective Thomson Scattering (CTS) system is a diagnostic instrument that measures plasma density and velocity through Thomson scattering of microwave radiation. Some of the key components of the CTS are quasi-optical mirrors that are used to produce astigmatic beam patterns, which have impact on the strength and spatial resolution of the diagnostic signal. The mirrors are exposed to neutron radiation, which may alter the quality of the signal received. In this work, three different materials (molybdenum (Mo), stainless steel 316 (SS-316) and tungsten (W)) are considered for the first mirror of the CTS. The objective is to access which of the material studied are best suited for this mirror, considering different neutron radiation loads simulated scenarios defined by ITER, based on the resultant stresses and temperature distributions. For it, the neutron irradiation, and subsequent heat-load on the mirrors are simulated using the Monte Carlo N-Particle (MCNP) code. Based on the MCNP heat-load results, transient thermal-structural Finite Element Analysis (FEA) of the mirror over a 400 s discharge, with and without cooling on the rear side, are conducted using in commercial FEA software ANSYS. Results show that of the tested materials Mo and W are the most suitable material for this application. Even though, this study does not yet consider the variation of the material properties with temperature, it presents a quick initial satisfactory assessment that may be considered in subsequent and more complex analysis. (authors)

  20. MCNP5 study on kinetics parameters of coupled fast-thermal system HERBE

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2011-01-01

    Full Text Available New validation of the well-known Monte Carlo code MCNP5 against measured criticality and kinetics data for the coupled fast-thermal HERBE System at the Reactor B critical assembly is shown in this paper. Results of earlier calculations of these criticality and kinetics parameters, done by combination of transport and diffusion codes using two-dimension geometry model are compared to results of new calculations carried out by the MCNP5 code in three-dimension geometry. Satisfactory agreements in comparison of new results with experimental data, in spite complex heterogeneous composition of the HERBE core, are achieved confirming that MCNP5 code could apply successfully to study on HERBE kinetics parameters after uncertainties in impurities in material compositions and positions of fuel elements in fast zone were removed.

  1. Monte Carlo importance sampling for the MCNP{trademark} general source

    Energy Technology Data Exchange (ETDEWEB)

    Lichtenstein, H.

    1996-01-09

    Research was performed to develop an importance sampling procedure for a radiation source. The procedure was developed for the MCNP radiation transport code, but the approach itself is general and can be adapted to other Monte Carlo codes. The procedure, as adapted to MCNP, relies entirely on existing MCNP capabilities. It has been tested for very complex descriptions of a general source, in the context of the design of spent-reactor-fuel storage casks. Dramatic improvements in calculation efficiency have been observed in some test cases. In addition, the procedure has been found to provide an acceleration to acceptable convergence, as well as the benefit of quickly identifying user specified variance-reduction in the transport that effects unstable convergence.

  2. 3D neutronic calculations: CAD-MCNP methodology applied to vessel activation in KOYO-F

    Science.gov (United States)

    Herreras, Y.; Lafuente, A.; Sordo, F.; Cabellos, O.; Perlado, J. M.

    2008-05-01

    This paper presents a methodology for 3D neutronic calculations suitable for complex and extensive geometries. The geometry of the system design is first fully modelled with a CAD program, and subsequently processed through a MCNP-CAD interface in order to generate an MCNP geometry file. Neutronic irradiation results are finally achieved running the MCNPX program, where the geometry input card used is directly the MCNP-CAD interface output. This methodology enables accurate neutronic calculations for complex geometries characterised by high detail levels. This procedure will be applied to the Fast Ignition Fusion Reactor KOYO-F to determine first neutron fluxes calculations along the blanket as well as the material activation in the reduced martensitic 9Cr-1Mo steel vessel.

  3. 3D neutronic calculations: CAD-MCNP methodology applied to vessel activation in KOYO-F

    Energy Technology Data Exchange (ETDEWEB)

    Herreras, Y; Cabellos, O; Perlado, J M [Instituto de Fusion Nuclear (DENIM)/ETSII/Universidad Politecnica, Madrid (Spain); Lafuente, A; Sordo, F [Universidad Politecnica de Madrid (UPM), Madrid (Spain)], E-mail: yuri@denim.upm.es

    2008-05-15

    This paper presents a methodology for 3D neutronic calculations suitable for complex and extensive geometries. The geometry of the system design is first fully modelled with a CAD program, and subsequently processed through a MCNP-CAD interface in order to generate an MCNP geometry file. Neutronic irradiation results are finally achieved running the MCNPX program, where the geometry input card used is directly the MCNP-CAD interface output. This methodology enables accurate neutronic calculations for complex geometries characterised by high detail levels. This procedure will be applied to the Fast Ignition Fusion Reactor KOYO-F to determine first neutron fluxes calculations along the blanket as well as the material activation in the reduced martensitic 9Cr-1Mo steel vessel.

  4. ITER test programme

    Science.gov (United States)

    Abdou, M.; Baker, C.; Casini, G.

    1991-07-01

    The International Thermonuclear Experimental Reactor (ITER) was designed to operate in two phases. The first phase, which lasts for 6 years, is devoted to machine checkout and physics testing. The second phase lasts for 8 years and is devoted primarily to technology testing. This report describes the technology test program development for ITER, the ancillary equipment outside the torus necessary to support the test modules, the international collaboration aspects of conducting the test program on ITER, the requirements on the machine major parameters and the R and D program required to develop the test modules for testing in ITER.

  5. Status of ITER task T213 collaborative irradiation screening experiment on Cu/SS joints in the Russian Federation SM-2-reactor

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, D.J. [Pacific Northwest National Lab., Richland, WA (United States); Fabritsiev, S.A. [D.V. Efremov Inst., St. Petersburg (Russian Federation); Pokrovsky, A.S. [SRIAR, Dimitrovgrad (Russian Federation); Zinkle, S.J. [Oak Ridge National Lab., TN (United States)] [and others

    1996-04-01

    Specimen fabrication is underway for an irradiation screening experiment planned to start in January 1996 in the SM-2 reactor in Dimitrovgrad, Russia. The purpose of the experiment is to evaluate the effects of neutron irradiation at ITER-relevant temperatures on the bond integrity performance of Cu/SS and Be/Cu joints, as well as to further investigate the base metal properties of irradiated copper alloys. Specimens from each of the four ITER parties (U.S., EU, japan, and RF) will be irradiated to a dose of {approx}0.2 dpa at two different temperatures, 150 and 300{degrees}C. The specimens will consist of Cu/SS and Be/Cu joints in several different geometries, as well as a large number of specimens from the base materials. Fracture toughness data on base metal and Cu/SS bonded specimens will be obtained from specimens supplied by the U.S. Due to lack of material, the Be/Cu specimens supplied by the U.S will only be irradiated as TEM disks.

  6. MatMCNP: A Code for Producing Material Cards for MCNP

    Energy Technology Data Exchange (ETDEWEB)

    DePriest, Kendall Russell [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Saavedra, Karen C. [American Structurepoint, Inc., Indianapolis, IN (United States)

    2014-09-01

    A code for generating MCNP material cards (MatMCNP) has been written and verified for naturally occurring, stable isotopes. The program allows for material specification as either atomic or weight percent (fractions). MatMCNP also permits the specification of enriched lithium, boron, and/or uranium. In addition to producing the material cards for MCNP, the code calculates the atomic (or number) density in atoms/barn-cm as well as the multiplier that should be used to convert neutron and gamma fluences into dose in the material specified.

  7. New gamma and neutron measurements and MCNP simulations

    Energy Technology Data Exchange (ETDEWEB)

    Crovisier, Ph.; Camus, L.; Marty, P. [CEA Centre de Valduc, Is sur Tille (France). Service de Protection contre les Rayonnements; Groetz, J.E [Univ. de Franche Comte, Besancon (France). Laboratoire de Microanalyses Nucleaires

    2003-05-01

    To take into account the criticality risk, the Radiation Protection Service of the CEA Valduc center has developed a new method allowing quickly fixed fissile material mass determination in complex configurations where the other classical techniques, such as gamma spectrometry, cannot be easily used (contaminated areas, large thickness shield protection). Then, the Radiation Protection Service in collaboration with the Nuclear Microanalyses Laboratory carried out ambient dose equivalent rate measurements coupled with a MCNP simulations in order to estimate 'holdup' nuclear materials. The methodology used is described below: Choice of measurement devices (gamma or neutron) according to the detection limits. Use of calibrated dose rate meters and new neutron spectrometer ROSPEC (measurement references and uncertainties). Ambient dose equivalent rate measurements should be performed at different locations in the vicinity of the system studied. Complete geometry system, shields and sources locations (if it's possible) should be modeled accurately in MCNP simulations. Ambient dose equivalent rate calculations at each measurement locations and for each source described are performed by using the MCNP code. All these measurements and calculations allow to set up a linear equations system with activities sources (mass) as unknowns. Due to the measurement uncertainties, this system cannot be exactly solved but by an iterative approach. The fissile material characteristics (i.e isotopic abundance, chemical form, nuclides) located in the system are very important to enable us the nuclear material mass estimations. Previously, these features can be determined by smears radiological analyses or by knowing the elaborated nuclear materials in the concerned plant. For the first time, this new method was successfully used to study a vessel containing metal plutonium located on the walls. The second estimation concerned the 'holdup' fissile material in a

  8. MCNP APPLICATIONS FOR THE 21ST CENTURY

    Energy Technology Data Exchange (ETDEWEB)

    G. MCKINNEY; T. BOOTH; ET AL

    2000-10-01

    The Los Alamos National Laboratory (LANL) Monte Carlo N-Particle radiation transport code, MCNP, has become an international standard for a wide spectrum of neutron, photon, and electron radiation transport applications. The latest version of the code, MCNP 4C, was released to the Radiation Safety Information Computational Center (RSICC) in February 2000. This paper describes the code development philosophy, new features and capabilities, applicability to various problems, and future directions.

  9. MCNP application for the 21 century

    Energy Technology Data Exchange (ETDEWEB)

    McKinney, M.C. [and others

    2000-08-01

    The Los Alamos National Laboratory (LANL) Monte Carlo N-Particle radiation transport code, MCNP, has become an international standard for a wide spectrum of neutron, photon, and electron radiation transport applications. The latest version of the code, MCNP 4C, was released to the Radiation Safety Information Computational Center (RSICC) in February 2000. This paper describes the code development philosophy, new features and capabilities, applicability to various problems, and future directions.

  10. ITER at Cadarache; ITER a Cadarache

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-06-15

    This public information document presents the ITER project (International Thermonuclear Experimental Reactor), the definition of the fusion, the international cooperation and the advantages of the project. It presents also the site of Cadarache, an appropriate scientifical and economical environment. The last part of the documentation recalls the historical aspect of the project and the today mobilization of all partners. (A.L.B.)

  11. MCNP5 and GEANT4 comparisons for preliminary Fast Neutron Pencil Beam design at the University of Utah TRIGA system

    Science.gov (United States)

    Adjei, Christian Amevi

    The main objective of this thesis is twofold. The starting objective was to develop a model for meaningful benchmarking of different versions of GEANT4 against an experimental set-up and MCNP5 pertaining to photon transport and interactions. The following objective was to develop a preliminary design of a Fast Neutron Pencil Beam (FNPB) Facility to be applicable for the University of Utah research reactor (UUTR) using MCNP5 and GEANT4. The three various GEANT4 code versions, GEANT4.9.4, GEANT4.9.3, and GEANT4.9.2, were compared to MCNP5 and the experimental measurements of gamma attenuation in air. The average gamma dose rate was measured in the laboratory experiment at various distances from a shielded cesium source using a Ludlum model 19 portable NaI detector. As it was expected, the gamma dose rate decreased with distance. All three GEANT4 code versions agreed well with both the experimental data and the MCNP5 simulation. Additionally, a simple GEANT4 and MCNP5 model was developed to compare the code agreements for neutron interactions in various materials. Preliminary FNPB design was developed using MCNP5; a semi-accurate model was developed using GEANT4 (because GEANT4 does not support the reactor physics modeling, the reactor was represented as a surface neutron source, thus a semi-accurate model). Based on the MCNP5 model, the fast neutron flux in a sample holder of the FNPB is obtained to be 6.52×107 n/cm2s, which is one order of magnitude lower than gigantic fast neutron pencil beam facilities existing elsewhere. The MCNP5 model-based neutron spectrum indicates that the maximum expected fast neutron flux is at a neutron energy of ~1 MeV. In addition, the MCNP5 model provided information on gamma flux to be expected in this preliminary FNPB design; specifically, in the sample holder, the gamma flux is to be expected to be around 108 γ/cm 2s, delivering a gamma dose of 4.54×103 rem/hr. This value is one to two orders of magnitudes below the gamma

  12. Iterating skeletons

    DEFF Research Database (Denmark)

    Dieterle, Mischa; Horstmeyer, Thomas; Berthold, Jost;

    2012-01-01

    block inside a bigger structure. In this work, we present a general framework for skeleton iteration and discuss requirements and variations of iteration control and iteration body. Skeleton iteration is expressed by synchronising a parallel iteration body skeleton with a (likewise parallel) state...

  13. Establishment and Verification of MCNP Neutron Transport Model About Tianwan Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    ZHOU; Qi

    2012-01-01

    <正>In order to calculating the neutron flux in the surveillance box and reactor pressure vessel of the Tianwan NPP, we need to build up the neutron transport model by using the Monte Carlo code MCNP. The core of the NPP is very complicated for modeling so we put forward some assumptions that can simplify the neutron transport model. A lot of calculation works have been done to prove that the assumptions are right and suitable.

  14. CTEx Beowulf cluster for MCNP performance

    Energy Technology Data Exchange (ETDEWEB)

    Gonzaga, Roberto N.; Amorim, Aneuri S. de; Balthar, Mario Cesar V. [Centro Tecnologico do Exercito (CTEx), Divisao de Defesa Quimica, Biologica e Nuclear, Rio de Janeiro, RJ (Brazil)

    2011-07-01

    This work is an introduction to the CTEx Nuclear Defense Department's Beowulf Cluster. Building a Beowulf Cluster is a complex learning process that greatly depends upon your hardware and software requirements. The feasibility and efficiency of performing MCNP5 calculations with a small, heterogeneous computing cluster built in Red Hat's Fedora Linux operating system personal computers (PC) are explored. The performance increases that may be expected with such clusters are estimated for cases that typify general radiation transport calculations. Our results show that the speed increase from additional slave PCs is nearly linear up to 10 processors. The pre compiled parallel binary version of MCNP uses the Message-Passing Interface (MPI) protocol. The use of this pre compiled parallel version of MCNP5 with the MPI protocol on a small, heterogeneous computing cluster built from Red Hat's Fedora Linux operating system PCs is the subject of this work. (author)

  15. Adjoint-Based Uncertainty Quantification with MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Seifried, Jeffrey E. [Univ. of California, Berkeley, CA (United States)

    2011-09-01

    This work serves to quantify the instantaneous uncertainties in neutron transport simulations born from nuclear data and statistical counting uncertainties. Perturbation and adjoint theories are used to derive implicit sensitivity expressions. These expressions are transformed into forms that are convenient for construction with MCNP6, creating the ability to perform adjoint-based uncertainty quantification with MCNP6. These new tools are exercised on the depleted-uranium hybrid LIFE blanket, quantifying its sensitivities and uncertainties to important figures of merit. Overall, these uncertainty estimates are small (< 2%). Having quantified the sensitivities and uncertainties, physical understanding of the system is gained and some confidence in the simulation is acquired.

  16. New methods for neutron response calculations with MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Hendricks, J.S. [Los Alamos National Lab., NM (United States). Applied Theoretical and Computational Physics Div.

    1997-05-01

    MCNP4B was released for international distribution in February, 1997. The author summarized the new MCNP4B features since the release of MCNP4A over three years earlier and compare some results. Then he describes new methods being developed for future code releases. The focus is methods and applications of ex-core neutron response calculations.

  17. VVER-440 Ex-Core Neutron Transport Calculations by MCNP-5 Code and Comparison with Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Borodkin, Pavel; Khrennikov, Nikolay [Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS) Malaya Krasnoselskaya ul., 2/8, bld. 5, 107140 Moscow (Russian Federation)

    2008-07-01

    Ex-core neutron transport calculations are needed to evaluate radiation loading parameters (neutron fluence, fluence rate and spectra) on the in-vessel equipment, reactor pressure vessel (RPV) and support constructions of VVER type reactors. Due to these parameters are used for reactor equipment life-time assessment, neutron transport calculations should be carried out by precise and reliable calculation methods. In case of RPVs, especially, of first generation VVER-440s, the neutron fluence plays a key role in the prediction of RPV lifetime. Main part of VVER ex-core neutron transport calculations are performed by deterministic and Monte-Carlo methods. This paper deals with precise calculations of the Russian first generation VVER-440 by MCNP-5 code. The purpose of this work was an application of this code for expert calculations, verification of results by comparison with deterministic calculations and validation by neutron activation measured data. Deterministic discrete ordinates DORT code, widely used for RPV neutron dosimetry and many times tested by experiments, was used for comparison analyses. Ex-vessel neutron activation measurements at the VVER-440 NPP have provided space (in azimuth and height directions) and neutron energy (different activation reactions) distributions data for experimental (E) validation of calculated results. Calculational intercomparison (DORT vs. MCNP-5) and comparison with measured values (MCNP-5 and DORT vs. E) have shown agreement within 10-15% for different space points and reaction rates. The paper submits a discussion of results and makes conclusions about practice use of MCNP-5 code for ex-core neutron transport calculations in expert analysis. (authors)

  18. Reactor

    Science.gov (United States)

    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.

  19. Iterating skeletons

    DEFF Research Database (Denmark)

    Dieterle, Mischa; Horstmeyer, Thomas; Berthold, Jost;

    2012-01-01

    Skeleton-based programming is an area of increasing relevance with upcoming highly parallel hardware, since it substantially facilitates parallel programming and separates concerns. When parallel algorithms expressed by skeletons involve iterations – applying the same algorithm repeatedly...... block inside a bigger structure. In this work, we present a general framework for skeleton iteration and discuss requirements and variations of iteration control and iteration body. Skeleton iteration is expressed by synchronising a parallel iteration body skeleton with a (likewise parallel) state......-based iteration control, where both skeletons offer supportive type safety by dedicated types geared towards stream communication for the iteration. The skeleton iteration framework is implemented in the parallel Haskell dialect Eden. We use example applications to assess performance and overhead....

  20. Remote maintenance development for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Tada, Eisuke [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Shibanuma, Kiyoshi

    1998-04-01

    This paper describes the overall ITER remote maintenance design concept developed mainly for in-vessel components such as diverters and blankets, and outlines the ITER R and D program to develop remote handling equipment and radiation hard components. Reactor structures inside the ITER cryostat must be maintained remotely due to DT operation, making remote handling technology basic to reactor design. The overall maintenance scenario and design concepts have been developed, and maintenance design feasibility, including fabrication and testing of full-scale in-vessel remote maintenance handling equipment and tool, is being verified. (author)

  1. Development of Monteburns: A Code That Links MCNP and ORIGEN2 in an Automated Fashion for Burnup Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Holly R. Trellue

    1998-12-01

    Monteburns is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code 0RIGEN2. Monteburns produces many criticality and burnup computational parameters based on material feed/removal specifications, power(s), and time intervals. This code processes input from the user indicating the system geometry, initial material compositions, feed/removal, and other code-specific parameters. Results from MCNP, 0RIGEN2, and other calculations are then output successively as the code runs. The principle function of monteburns is to first transfer one-group cross sections and fluxes from MCNP to 0RIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from 0RIGEN2 back to MCNP in a repeated, cyclic fashion. The main requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with 0RIGEN2 and other calculations are performed by monteburns. This report presents the results obtained from the benchmarking of monteburns to measured and previously obtained data from traditional Light Water Reactor systems. The majority of the differences seen between the two were less than five percent. These were primarily a result of variances in cross sections between MCNP, cross section libraries used by other codes, and observed values. With this understanding, this code can now be used with confidence for burnup calculations in three-dimensional systems. It was designed for use in the Accelerator Transmutation of Waste project at Los Alamos National Laboratory but is also being applied to the analysis of isotopic production/destruction of transuranic actinides in a reactor system. The code has now been shown to sufficiently support these calculations.

  2. Evaluation of Geometric Progression (GP Buildup Factors using MCNP Codes (MCNP6.1 and MCNP5-1.60

    Directory of Open Access Journals (Sweden)

    Kim Kyung-O

    2016-01-01

    Full Text Available The gamma-ray buildup factors of three-dimensional point kernel code (QAD-CGGP are re-evaluated by using MCNP codes (MCNP6.1 and MCNPX5-1.60 and ENDF/B-VI.8 photoatomic data, which cover an energy range of 0.015–15 MeV and an iron thickness of 0.5–40 Mean Free Path (MFP. These new data are fitted to the Geometric Progression (GP fitting function and are then compared with ANS standard data equipped with QAD-CGGP. In addition, a simple benchmark calculation was performed to compare the QAD-CGGP results applied with new and existing buildup factors based on the MCNP codes. In the case of the buildup factors of low-energy gamma-rays, new data are evaluated to be about 5% higher than the existing data. In other cases, these new data present a similar trend based on the specific penetration depth, while existing data continuously increase beyond that depth. In a simple benchmark, the calculations using the existing data were slightly underestimated compared to the reference data at a deep penetration depth. On the other hand, the calculations with new data were stabilized with an increasing penetration depth, despite a slight overestimation at a shallow penetration depth.

  3. Use experiences of MCNP in nuclear energy study. 2. Review of variance reduction techniques

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Kiyoshi; Yamamoto, Toshihiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [eds.

    1998-03-01

    `MCNP Use Experience` Working Group was established in 1996 under the Special Committee on Nuclear Code Evaluation. This year`s main activity of the working group has been focused on the review of variance reduction techniques of Monte Carlo calculations. This working group dealt with the variance reduction techniques of (1) neutron and gamma ray transport calculation of fusion reactor system, (2) concept design of nuclear transmutation system using accelerator, (3) JMTR core calculation, (4) calculation of prompt neutron decay constant, (5) neutron and gamma ray transport calculation for exposure evaluation, (6) neutron and gamma ray transport calculation of shielding system, etc. Furthermore, this working group started an activity to compile `Guideline of Monte Carlo Calculation` which will be a standard in the future. The appendices of this report include this `Guideline`, the use experience of MCNP 4B and examples of Monte Carlo calculations of high energy charged particles. The 11 papers are indexed individually. (J.P.N.)

  4. ANALISIS LAJU DOSIS NEUTRON REAKTOR PLTN PWR 1000 MWe MENGGUNAKAN PROGRAM MCNP

    Directory of Open Access Journals (Sweden)

    Amir Hamzah

    2015-03-01

    Full Text Available Dalam rangka menyongsong PLTN pertama di Indonesia, dilakukan kajian dan analisis berbagai aspek teknologi reaktor tersebut. Tujuan dari penelitian ini adalah menentukan laju dosis neutron di luar perisai biologik reaktor PLTN PWR 1000 MWe yang merupakan bagian dari kegiatan besar di atas. Data hasil analisis laju dosis radiasi pada posisi tertentu sangat dibutuhkan untuk menunjukkan tingkat paparan radiasi di posisi tersebut. Analisis laju dosis neutron ditentukan berdasarkan hasil analisis fluks dan spektrum neutron. Analisis fluks dan spektrum neutron di teras reaktor daya PWR 1000 Mwe dilakukan menggunakan program MCNP. Model perhitungan yang dilakukan meliputi 9 zona material yaitu, teras, air, selimut, air, tong, air, bejana tekan, beton dan lapisan udara luar. Penentuan distribusi fluks dan spektrum neutron dilakukan ke arah radial hingga di luar perisai beton dengan akurasi antara 10% hingga 30% dalam tiap kelompok energi yang jumlahnya 1 dan 50 kelompok. Hasil analisis laju dosis neutron di permukaan perisai biologik reaktor PLTN PWR 1000 MWe pada kondisi reaktor beroperasi daya penuh sudah di bawah nilai batas keselamatan. Maka dapat disimpulkan bahwa dari segi paparan radiasi neutron, penggunaan perisai radiasi beton setebal dua meter sudah memenuhi persyaratan keselamatan. Kata kunci: PLTN PWR, fluks neutron, perisai, laju dosis neutron, MCNP.   In order to meet the first nuclear power plant in Indonesia, it has been conducted a study and analysis of various aspects of reactor technology. The purpose of this study was to determine the neutron dose rates at the outside of biological shield of NPP PWR 1000 MWe reactor that is a part of the activities described above. The analysis data of radiation dose rate at a specific position is needed to show the level of radiation exposure in those positions. Analysis neutron dose rate is determined based on the results of the analysis of neutron flux. Analysis of flux and neutron spectrum in

  5. Criticality Calculations with MCNP6 - Practical Lectures

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3); Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3)

    2016-11-29

    These slides are used to teach MCNP (Monte Carlo N-Particle) usage to nuclear criticality safety analysts. The following are the lecture topics: course information, introduction, MCNP basics, criticality calculations, advanced geometry, tallies, adjoint-weighted tallies and sensitivities, physics and nuclear data, parameter studies, NCS validation I, NCS validation II, NCS validation III, case study 1 - solution tanks, case study 2 - fuel vault, case study 3 - B&W core, case study 4 - simple TRIGA, case study 5 - fissile mat. vault, criticality accident alarm systems. After completion of this course, you should be able to: Develop an input model for MCNP; Describe how cross section data impact Monte Carlo and deterministic codes; Describe the importance of validation of computer codes and how it is accomplished; Describe the methodology supporting Monte Carlo codes and deterministic codes; Describe pitfalls of Monte Carlo calculations; Discuss the strengths and weaknesses of Monte Carlo and Discrete Ordinants codes; The diffusion theory model is not strictly valid for treating fissile systems in which neutron absorption, voids, and/or material boundaries are present. In the context of these limitations, identify a fissile system for which a diffusion theory solution would be adequate.

  6. CGMF & FREYA Verification in MCNP6

    Energy Technology Data Exchange (ETDEWEB)

    Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-01-05

    At the present time, the new and updated fission event generators included in MCNP6.2 have been verified to be functioning properly through a variety of detailed tests. This work describes the detailed verification steps taken to ensure these complicated fission event generators, FREYA and CGMF, are integrated into MCNP6 properly. Ultimately, with the knowledge that MCNP6 is making use of these models appropriately, we can now begin to validate the models against benchmarked experiments. Some benchmarks, including criticality and subcritical experiments interested in multiplication and bulk counting rates, are easy to model and understand but are likely insensitive to the detailed nature of these models. It will take some new measurements with coincidence detection capabilities to be able to stress the physics within each of these fission event generator models. Once the models are validated and it is understood where the models can truly be predictive, then we can study what SNM observables can be characterized for nonproliferation applications.

  7. MCNP4B{sup {trademark}} verification and validation

    Energy Technology Data Exchange (ETDEWEB)

    Hendricks, J.S.; Court, J.D.

    1996-08-01

    Several new features and bug fixes have been incorporated into the new release of MCNP. As required by the MCNP Software Quality Assurance Plan, these changes to the code and the test set are documented here for user reference. This document summarizes the new MCNP4B features and corrections, separated into major and minor groupings. Also included are a code cleanup section and a section delineating problems identified in LA-12839 which have not been corrected. Finally, we document the MCNP4B test set modifications and explain how test set coverage has been improved.

  8. Neutronic analysis for in situ calibration of ITER in-vessel neutron flux monitor with microfission chamber

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, Masao, E-mail: ishikawa.masao@jaea.go.jp [Fusion Research and Development Directorate, Japan Atomic Energy Agency, Ibaraki 311-0193 (Japan); Kondoh, Takashi; Kusama, Yoshinori [Fusion Research and Development Directorate, Japan Atomic Energy Agency, Ibaraki 311-0193 (Japan); Bertalot, Luciano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: ► Neutronic analysis is performed for in situ calibration of the microfission chamber (MFC). ► The source transfer system deigned in this study does not affect MFC detection efficiency. ► The rotation method is appropriate for full calibration because the calibration time is shorter. ► But, point-by-point method should be performed to check the accuracy of the MCNP model. ► Combination of two methods are important to perform in situ calibration efficiently. -- Abstract: Neutronic analysis is performed for in situ calibration of the microfission chamber (MFC), which is the in-vessel neutron-flux monitor at the International Thermonuclear Experimental Reactor (ITER). We present the design of the transfer system for a neutron generator, which consists of two toroidal rings and a neutron-generator holder, and estimate the effect of the system on MFC detection efficiency through neutronic analysis with the Monte Carlo N-particle (MCNP) code. The result indicates that the designed transfer system does not affect MFC detection efficiency. In situ calibrations by the point-by-point method and by the rotation method are also simulated and compared by neutronic analysis. The results indicate that the rotation method is appropriate for full calibration because the calibration time is shorter (all neutron-flux monitors can be calibrated simultaneously). However, the rotation method makes it difficult to compare the results with neutronic analysis, so the point-by-point method should be performed prior to full calibration to check the accuracy of the MCNP model.

  9. The MCNP6 Analytic Criticality Benchmark Suite

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Codes Group

    2016-06-16

    Analytical benchmarks provide an invaluable tool for verifying computer codes used to simulate neutron transport. Several collections of analytical benchmark problems [1-4] are used routinely in the verification of production Monte Carlo codes such as MCNP® [5,6]. Verification of a computer code is a necessary prerequisite to the more complex validation process. The verification process confirms that a code performs its intended functions correctly. The validation process involves determining the absolute accuracy of code results vs. nature. In typical validations, results are computed for a set of benchmark experiments using a particular methodology (code, cross-section data with uncertainties, and modeling) and compared to the measured results from the set of benchmark experiments. The validation process determines bias, bias uncertainty, and possibly additional margins. Verification is generally performed by the code developers, while validation is generally performed by code users for a particular application space. The VERIFICATION_KEFF suite of criticality problems [1,2] was originally a set of 75 criticality problems found in the literature for which exact analytical solutions are available. Even though the spatial and energy detail is necessarily limited in analytical benchmarks, typically to a few regions or energy groups, the exact solutions obtained can be used to verify that the basic algorithms, mathematics, and methods used in complex production codes perform correctly. The present work has focused on revisiting this benchmark suite. A thorough review of the problems resulted in discarding some of them as not suitable for MCNP benchmarking. For the remaining problems, many of them were reformulated to permit execution in either multigroup mode or in the normal continuous-energy mode for MCNP. Execution of the benchmarks in continuous-energy mode provides a significant advance to MCNP verification methods.

  10. MCNP simulations of material exposure experiments (u)

    Energy Technology Data Exchange (ETDEWEB)

    Temple, Brian A [Los Alamos National Laboratory

    2010-12-08

    Simulations of proposed material exposure experiments were performed using MCNP6. The experiments will expose ampules containing different materials of interest with radiation to observe the chemical breakdown of the materials. Simulations were performed to map out dose in materials as a function of distance from the source, dose variation between materials, dose variation due to ampule orientation, and dose variation due to different source energy. This write up is an overview of the simulations and will provide guidance on how to use the data in the spreadsheet.

  11. Neutronic analysis for bolometers in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Suarez, A., E-mail: alejandro.suarez@iter.org [CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Reichle, R.; Loughlin, M.; Polunovskiy, E.; Walsh, M. [ITER Organization, Route de Vinon sur Verdon, 13115, St. Paul lez Durance (France)

    2013-10-15

    Highlights: ► Radiation damage calculations for the bolometers in ITER. ► Redesign of the bolometric diagnostic in EPP01. ► New bolometer radiation damage values in EPP01 in the safe zone. -- Abstract: Neutronic considerations in ITER have such importance that they drive the design of many diagnostics and components of the machine, and bolometers are not an exception. Bolometer cameras will be installed on the vacuum vessel, viewing the plasma through the gaps between blanket modules, divertor, equatorial and upper port plugs. The ITER reference bolometer sensors are of a resistive type. For this study it is assumed that they are composed of a thin silicon nitride carrier film and platinum resistors disposed in a Wheatstone bridge configuration. Their assumed radiation hardness is 0.1 dpa. Neutronic calculations were performed with the Monte Carlo program MCNP5, the FENDL 2.1 nuclear data library and the latest B-lite ITER neutronic model with the appropriate modifications using the CAD to MCNP converter MCAM. A complete characterization of the neutron fluxes in all the bolometer locations and the calculation of neutron damage were performed. Values above the failure threshold damage were obtained for some of the bolometers, leading to a complete redesign of some parts of the bolometric system in order to extend its lifetime.

  12. Installation of the ITER committee industry. Participants guide; Installation du Comite industrie ITER. Dossier des participants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-07-01

    ITER is an international project to design and build an experimental fusion reactor based on the tokamak concept. This guide presents the ITER project and objectives and the associated organizations in France, the recommendations and actions for ITER, the industrial mobilization, the industrial committee and its members, technological sheets for the enterprises and the statistical document of the SESSI. (A.L.B.)

  13. Using MCNP in the design of neutron sources and neutron beams

    Energy Technology Data Exchange (ETDEWEB)

    Hergenreder, Daniel F.; Lecot, Carlos A.; Lovotti, Osvaldo P. [INVAP S.A., San Carlos de Bariloche (Argentina). Nuclear Projects Department. Nuclear Engineering Division

    2002-07-01

    The calculation methodology used to design cold, thermal and hot neutron sources and their associated neutron beam transport systems is presented. The design goal is to evaluate the performance of the neutron sources, their beam tubes and neutron guides at specific experimental locations in the reactor hall as well as in the neutron hall. The Monte Carlo method is a unique and powerful tool to transport neutrons. Its use in a bootstrap scheme appears to be an appropriate solution for this type of system. The proper use of MCNP as the main tool leads to a fast and reliable method to perform calculations in a relatively short time with low statistical errors. (author)

  14. MCNP6 Cosmic & Terrestrial Background Particle Fluxes -- Release 4

    Energy Technology Data Exchange (ETDEWEB)

    McMath, Garrett E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Nuclear Engineering and Nonproliferation Div.; McKinney, Gregg W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Nuclear Engineering and Nonproliferation Div.; Wilcox, Trevor [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Nuclear Engineering and Nonproliferation Div.

    2015-01-23

    Essentially a set of slides, the presentation begins with the MCNP6 cosmic-source option, then continues with the MCNP6 transport model (atmospheric, terrestrial) and elevation scaling. It concludes with a few slides on results, conclusions, and suggestions for future work.

  15. Semi-Analytical Benchmarks for MCNP6

    Energy Technology Data Exchange (ETDEWEB)

    Grechanuk, Pavel Aleksandrovi [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-11-07

    Code verification is an extremely important process that involves proving or disproving the validity of code algorithms by comparing them against analytical results of the underlying physics or mathematical theory on which the code is based. Monte Carlo codes such as MCNP6 must undergo verification and testing upon every release to ensure that the codes are properly simulating nature. Specifically, MCNP6 has multiple sets of problems with known analytic solutions that are used for code verification. Monte Carlo codes primarily specify either current boundary sources or a volumetric fixed source, either of which can be very complicated functions of space, energy, direction and time. Thus, most of the challenges with modeling analytic benchmark problems in Monte Carlo codes come from identifying the correct source definition to properly simulate the correct boundary conditions. The problems included in this suite all deal with mono-energetic neutron transport without energy loss, in a homogeneous material. The variables that differ between the problems are source type (isotropic/beam), medium dimensionality (infinite/semi-infinite), etc.

  16. Benchmarking MCNP and TRIPOLI with PGNAA measurements

    Science.gov (United States)

    Carasco, C.; Perot, B.; Sikora, A.; Mauerhofer, E.; Havenith, A.; Payan, E.; Kettler, J.; Kring, T.; Ma, J. L.

    2014-06-01

    The French Alternative Energies and Atomic Energy Commission (CEA Cadarache), the Forschungszentrum Jülich GmbH (FZJ), and the RWTH Aachen University (RWTH) are involved in a cooperation aiming at characterizing toxic and reactive elements in radioactive waste packages by means of Prompt Gamma Neutron Activation Analysis (PGNAA). The design of an optimized measurement system and the assessment of its performances for realistic scenarios can be conveniently studied by numerical Monte Carlo simulation, provided the model and nuclear data offer a sufficient precision. Previous studies performed with MCNP have shown that when the nuclear data libraries lack of precision, relevant results can still be obtained by performing calculations in multiple steps (by first determining the radiative capture rate, and transporting the induced gamma toward the detector) and by injecting valid gamma-ray production data in-between [1]. In such cases, it is interesting to compare the results obtained with different codes. In the present paper, we propose to compare the MCNP and TRIPOLI codes with measurements obtained in MEDINA (Multi Element Detection based on Instrumental Neutron Activation), which is the new FZJ PGNAA facility [2]. The aim of the measurement campaign is to assess capture gamma rays of toxic elements that can be found in 200 L waste drums which are expected for geological repository.

  17. An assessment of the MCNP4C weight window

    Energy Technology Data Exchange (ETDEWEB)

    Christopher N. Culbertson; John S. Hendricks

    1999-12-01

    A new, enhanced weight window generator suite has been developed for MCNP version 4C. The new generator correctly estimates importances in either a user-specified, geometry-independent, orthogonal grid or in MCNP geometric cells. The geometry-independent option alleviates the need to subdivide the MCNP cell geometry for variance reduction purposes. In addition, the new suite corrects several pathologies in the existing MCNP weight window generator. The new generator is applied in a set of five variance reduction problems. The improved generator is compared with the weight window generator applied in MCNP4B. The benefits of the new methodology are highlighted, along with a description of its limitations. The authors also provide recommendations for utilization of the weight window generator.

  18. Development of Burnup Calculation Code for Pebble-bed High Temperature Reactor at Equilibrium State%球床高温堆平衡态燃耗计算程序的开发

    Institute of Scientific and Technical Information of China (English)

    朱贵凤; 邹杨; 李明海; 严睿; 彭红花; 徐洪杰

    2015-01-01

    The burnup calculation code PBRE coupling MCNP5 and ORIGEN2 was developed for pebble‐bed high temperature reactor at equilibrium state ,and it can be used to analyze the neutronic performance of equilibrium core .The iteration method was optimized in order to save Monte Carlo calculation time ,and the convergence can be reached in 10 iterative steps .The average discharged burnup for HTR‐10 is consistent with literature ,and it indicates that the PBRE is suitable to analyze the burnup for pebble‐bed reactor at equilibrium state .%基于MCNP5和ORIGEN2耦合方法,开发了平衡态下球床高温堆的燃耗计算程序PBRE ,用于堆的性能价值分析。为节省蒙特卡罗计算时间,对迭代收敛的方法进行优化,使之可在10个迭代步内收敛。使用PBRE对清华大学H T R‐10进行建模计算,得到的平均卸料燃耗深度与文献报道值一致,表明PBRE程序适用于球床堆平衡态的燃耗分析。

  19. High gamma-rays irradiation tests of critical components for ITER (International Thermonuclear Experimental Reactor) in-vessel remote handling system

    Energy Technology Data Exchange (ETDEWEB)

    Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan)] [and others

    1999-02-01

    In ITER, the in-vessel remote handling is inevitably required to assemble and maintain the activated in-vessel components due to deuterium and tritium operation. Since the in-vessel remote handling system has to be operated under the intense of gamma ray irradiation, the components of the remote handling system are required to have radiation hardness so as to allow maintenance operation for a sufficient length of time under the ITER in-vessel environments. For this, the Japan, European and Russian Home Teams have extensively conducted gamma ray irradiation tests and quality improvements including optimization of material composition through ITER R and D program in order to develop radiation hard components which satisfy the doses from 10 MGy to 100 MGy at a dose rate of 1 x 10{sup 6} R/h (ITER R and D Task: T252). This report describes the latest status of radiation hard component development which has been conducted by the Japan Home Team in the ITER R and D program. The number of remote handling components tested is about seventy and these are categorized into robotics (Subtask 1), viewing system (Subtask 2) and common components (Subtask 3). The irradiation tests, including commercial base products for screening, modified products and newly developed products to improve the radiation hardness, were carried out using the gamma ray irradiation cells in Takasaki Establishment, JAERI. As a result, the development of the radiation hard components which can be tolerable for high temperature and gamma radiation has been well progressed, and many components, such as AC servo motor with ceramics insulated wire, optical periscope and CCD camera, have been newly developed. (author)

  20. Implementation and qualification of MCNP 5 through the intercomparison with the benchmark for the calculation of critical systems Godiva and Jezebel; Implementacao e qualificacao do MCNP5 atraves da intercomparacao com o benchmark para o calculo dos sistemas criticos Godiva e Jezebel

    Energy Technology Data Exchange (ETDEWEB)

    Lara, Rafael G.; Maiorino, Jose R., E-mail: rafael.lara@aluno.ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais Aplicadas

    2013-07-01

    This work aimed at the implementation and qualification of MCNP code in a supercomputer of the Universidade Federal do ABC, so that may be available a next-generation simulation tool for precise calculations of nuclear reactors and systems subject to radiation. The implementation of this tool will have multidisciplinary applications, covering various areas of engineering (nuclear, aerospace, biomedical), radiation physics and others.

  1. CAD-Based Shielding Analysis for ITER Port Diagnostics

    Science.gov (United States)

    Serikov, Arkady; Fischer, Ulrich; Anthoine, David; Bertalot, Luciano; De Bock, Maartin; O'Connor, Richard; Juarez, Rafael; Krasilnikov, Vitaly

    2017-09-01

    Radiation shielding analysis conducted in support of design development of the contemporary diagnostic systems integrated inside the ITER ports is relied on the use of CAD models. This paper presents the CAD-based MCNP Monte Carlo radiation transport and activation analyses for the Diagnostic Upper and Equatorial Port Plugs (UPP #3 and EPP #8, #17). The creation process of the complicated 3D MCNP models of the diagnostics systems was substantially accelerated by application of the CAD-to-MCNP converter programs MCAM and McCad. High performance computing resources of the Helios supercomputer allowed to speed-up the MCNP parallel transport calculations with the MPI/OpenMP interface. The found shielding solutions could be universal, reducing ports R&D costs. The shield block behind the Tritium and Deposit Monitor (TDM) optical box was added to study its influence on Shut-Down Dose Rate (SDDR) in Port Interspace (PI) of EPP#17. Influence of neutron streaming along the Lost Alpha Monitor (LAM) on the neutron energy spectra calculated in the Tangential Neutron Spectrometer (TNS) of EPP#8. For the UPP#3 with Charge eXchange Recombination Spectroscopy (CXRS-core), an excessive neutron streaming along the CXRS shutter, which should be prevented in further design iteration.

  2. Burn Control of Magnetically Confined Fusion Plasma 2. Burn Control in Tokamak Fusion Reactors 2.3 Burn Control in ITER

    Science.gov (United States)

    Fujisawa, Noboru

    The issue of burn control in FDR-ITER, the design of which was completed in 1998, is introduced, Controllability was studied based on the ID transport code, PRETOR, during the burn phase with self-ignition, as well as during start-up and shut-down. The results of the present study have helped us to identify the importance of controlling the fuel supply, impurity injection, and heating power to maintain fusion power and power to the divertor.

  3. Effects of Spatial Variations in Packing Fraction on Reactor Physics Parameters in Pebble-Bed Reactors

    Energy Technology Data Exchange (ETDEWEB)

    William K. Terry; A. M. Ougouag; Farzad Rahnema; Michael Scott McKinley

    2003-04-01

    The well-known spatial variation of packing fraction near the outer boundary of a pebble-bed reactor core is cited. The ramifications of this variation are explored with the MCNP computer code. It is found that the variation has negligible effects on the global reactor physics parameters extracted from the MCNP calculations for use in analysis by diffusion-theory codes, but for local reaction rates the effects of the variation are naturally important. Included is some preliminary work in using first-order perturbation theory for estimating the effect of the spatial variation of packing fraction on the core eigenvalue and the fision density distribution.

  4. MOCUP: MCNP-ORIGEN2 coupled utility program

    Energy Technology Data Exchange (ETDEWEB)

    Moore, R.L.; Schnitzler, B.G.; Wemple, C.A. [and others

    1995-09-30

    MOCUP is a system of external processors that allow for a limited treatment of the temporal composition of the user-selected MCNP cells in a time-dependent flux environment. The ORIGEN2 code computes the time-dependent compositions of these individually selected MCNP cells. All data communication between the two codes is accomplished through the MCNP and ORIGEN2 input/output files, the MOCUP Processor Output files, and two user supplied tables. MOCUP is either command line or interactively driven. The interactive interface is based on the portable XII window environment and the Motif tool kit. MOCUP was constructed so that no modifications to either MCNP or ORIGEN2 were necessary. Section 4 of the writeup contains the input instructions needed to set up the MOCUP run. MOCUP is extremely useful for analysts who perform isotope production, material transformation, and depletion and isotope analyses on complex, non-lattice geometries, and uniform and non-uniform lattices.

  5. Criticality calculations with MCNP{trademark}: A primer

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, C.D. II; Busch, R.D.; Briesmeister, J.F.; Forster, R.A. [New Mexico Univ., Albuquerque, NM (United States)

    1994-06-06

    With the closure of many experimental facilities, the nuclear criticality safety analyst increasingly is required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, in many cases, the analyst has little experience with the specific codes available at his/her facility. This primer will help you, the analyst, understand and use the MCNP Monte Carlo code for nuclear criticality safety analyses. It assumes that you have a college education in a technical field. There is no assumption of familiarity with Monte Carlo codes in general or with MCNP in particular. Appendix A gives an introduction to Monte Carlo techniques. The primer is designed to teach by example, with each example illustrating two or three features of MCNP that are useful in criticality analyses. Beginning with a Quickstart chapter, the primer gives an overview of the basic requirements for MCNP input and allows you to run a simple criticality problem with MCNP. This chapter is not designed to explain either the input or the MCNP options in detail; but rather it introduces basic concepts that are further explained in following chapters. Each chapter begins with a list of basic objectives that identify the goal of the chapter, and a list of the individual MCNP features that are covered in detail in the unique chapter example problems. It is expected that on completion of the primer you will be comfortable using MCNP in criticality calculations and will be capable of handling 80 to 90 percent of the situations that normally arise in a facility. The primer provides a set of basic input files that you can selectively modify to fit the particular problem at hand.

  6. Energy and spatial dependence of MCNP simulations for ZED-2 critical experiments

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)], E-mail: kozierk@aecl.ca

    2008-07-01

    MCNP simulations of ZED-2 critical experiments provide a good test of the reliability of the nuclear data involved in the simulation of reactor physics phenomena of importance to CANDU reactors, particularly the coolant void reactivity. Recent work has therefore focused on the impact of the new ENDF/B-VII.0 nuclear data library. One feature of this library is the provision of thermal scattering law data for UO{sub 2}. Initial MCNP results using preliminary ACE-format data files for UO{sub 2} thermal scattering suggested that a consistent reduction was obtained in the coolant void reactivity simulation bias, especially for ZED-2 critical experiments involving slightly enriched uranium (0.95 wt% {sup 235}U) and H{sub 2}O/air coolant. However, subsequent work using UO{sub 2} thermal scattering data files that correctly include the coherent elastic scattering component indicated that the net reactivity impact is quite small. The present work extends this investigation to examine in detail the energy dependence of the impact of the UO{sub 2} thermal scattering data and, more generally, the energy and spatial dependence of the coolant void reactivity simulation bias for some of these experiments. In addition, results are presented using MCNPX with an improved treatment for thermal scattering. It is found that the net reactivity impact results from the cancellation of larger positive and negative effects at different energies and in different fuel regions, and which generally highlight the reactor physics changes that occur when the coolant is removed. (author)

  7. Verification of MCNP6.2 for Nuclear Criticality Safety Applications

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-05-10

    Several suites of verification/validation benchmark problems were run in early 2017 to verify that the new production release of MCNP6.2 performs correctly for nuclear criticality safety applications (NCS). MCNP6.2 results for several NCS validation suites were compared to the results from MCNP6.1 [1] and MCNP6.1.1 [2]. MCNP6.1 is the production version of MCNP® released in 2013, and MCNP6.1.1 is the update released in 2014. MCNP6.2 includes all of the standard features for NCS calculations that have been available for the past 15 years, along with new features for sensitivity-uncertainty based methods for NCS validation [3]. Results from the benchmark suites were compared with results from previous verification testing [4-8]. Criticality safety analysts should consider testing MCNP6.2 on their particular problems and validation suites. No further development of MCNP5 is planned. MCNP6.1 is now 4 years old, and MCNP6.1.1 is now 3 years old. In general, released versions of MCNP are supported only for about 5 years, due to resource limitations. All future MCNP improvements, bug fixes, user support, and new capabilities are targeted only to MCNP6.2 and beyond.

  8. ITER helium ash accumulation

    Energy Technology Data Exchange (ETDEWEB)

    Hogan, J.T.; Hillis, D.L.; Galambos, J.; Uckan, N.A. (Oak Ridge National Lab., TN (USA)); Dippel, K.H.; Finken, K.H. (Forschungszentrum Juelich GmbH (Germany, F.R.). Inst. fuer Plasmaphysik); Hulse, R.A.; Budny, R.V. (Princeton Univ., NJ (USA). Plasma Physics Lab.)

    1990-01-01

    Many studies have shown the importance of the ratio {upsilon}{sub He}/{upsilon}{sub E} in determining the level of He ash accumulation in future reactor systems. Results of the first tokamak He removal experiments have been analysed, and a first estimate of the ratio {upsilon}{sub He}/{upsilon}{sub E} to be expected for future reactor systems has been made. The experiments were carried out for neutral beam heated plasmas in the TEXTOR tokamak, at KFA/Julich. Helium was injected both as a short puff and continuously, and subsequently extracted with the Advanced Limiter Test-II pump limiter. The rate at which the He density decays has been determined with absolutely calibrated charge exchange spectroscopy, and compared with theoretical models, using the Multiple Impurity Species Transport (MIST) code. An analysis of energy confinement has been made with PPPL TRANSP code, to distinguish beam from thermal confinement, especially for low density cases. The ALT-II pump limiter system is found to exhaust the He with maximum exhaust efficiency (8 pumps) of {approximately}8%. We find 1<{upsilon}{sub He}/{upsilon}{sub E}<3.3 for the database of cases analysed to date. Analysis with the ITER TETRA systems code shows that these values would be adequate to achieve the required He concentration with the present ITER divertor He extraction system.

  9. Criticality calculations with MCNP{sup TM}: A primer

    Energy Technology Data Exchange (ETDEWEB)

    Mendius, P.W. [ed.; Harmon, C.D. II; Busch, R.D.; Briesmeister, J.F.; Forster, R.A.

    1994-08-01

    The purpose of this Primer is to assist the nuclear criticality safety analyst to perform computer calculations using the Monte Carlo code MCNP. Because of the closure of many experimental facilities, reliance on computer simulation is increasing. Often the analyst has little experience with specific codes available at his/her facility. This Primer helps the analyst understand and use the MCNP Monte Carlo code for nuclear criticality analyses. It assumes no knowledge of or particular experience with Monte Carlo codes in general or with MCNP in particular. The document begins with a Quickstart chapter that introduces the basic concepts of using MCNP. The following chapters expand on those ideas, presenting a range of problems from simple cylinders to 3-dimensional lattices for calculating keff confidence intervals. Input files and results for all problems are included. The Primer can be used alone, but its best use is in conjunction with the MCNP4A manual. After completing the Primer, a criticality analyst should be capable of performing and understanding a majority of the calculations that will arise in the field of nuclear criticality safety.

  10. Availability of MCNP & MATLAB for reconstructing the water-vapor two-phase flow pattern in neutron radiography

    Institute of Scientific and Technical Information of China (English)

    FENG Qixi; FENG Quanke; TAKESHI Kawai

    2008-01-01

    The China Advanced Research Reactor (CARR) is scheduled to be operated in the autumn of 2008. In this paper, we report preparations for installing the neutron radiography instrument (NRI) and for utilizing it efficiently. The 2-D relative neutron intensity profiles for the water-vapor two-phase flow inside the tube were obtained using the MCNP code without influence of γ-ray and electronic-noise. The MCNP simulation of the 2-D neutron intensity profile for the water-vapor two-phase flow was demonstrated. The simulated 2-D neutron intensity profiles could be used as the benchmark data base by calibrating part of the data measured by the CARR-NRI. The 3-D objective images allow us to understand the flow pattern more clearly and it is reconstructed using the MATLAB through the threshold transformation techniques. And thus it is concluded that the MCNP code and the MATLAB are very useful for constructing the benchmark data base for the investigation of the water-vapor two-phase flow using the CARR-NRI.

  11. MCNP Super Lattice Method for VHTR ORIGEN2.2 Nuclear Library Improvement Based on ENDF/B-VII

    Energy Technology Data Exchange (ETDEWEB)

    G. S. Chang; J. R. Parry

    2010-10-01

    The advanced Very High Temperature gas-cooled Reactor (VHTR) achieves simplification of safety through reliance on innovative features and passive systems. One of the VHTRs innovative features is the reliance on ceramic-coated fuel particles to retain the fission products under extreme accident conditions. The effect of the random fuel kernel distribution in the fuel prismatic block creates a double-heterogeneous lattice, which needs to be addressed through the use of the newly developed prismatic super Kernel-by-Kernel Fuel (KbKF) lattice model method. Based on the new ENDF/B-VII nuclear cross section evaluated data, the developed KbKF super lattice model was then used with MCNP to calculate the material isotopes neutron reaction rates, such as, (n,?); (n,n’); (n,2n’); (n,f); (n,p); (n,?). Then, the MCNP-calculated results are rearranged to generate a set of new libraries “VHTRXS.lib,” for the ORIGEN2.2 isotopes depletion and build-up analysis code. The libraries contain one group cross section data for the structural light elements, actinides, and fission products that can be applied in the VHTR related fuel burnup and material transmutation analysis codes. The efficiency and ease of use of the MCNP method to generate and update the ORIGEN2.2 one-group spectrum weighed cross section library for VHTR was demonstrated.

  12. Conceptual design description for the tritium recovery system for the US ITER (International Thermonuclear Experimental Reactor) Li sub 2 O/Be water cooled blanket

    Energy Technology Data Exchange (ETDEWEB)

    Finn, P.A.; Sze, D.K. (Argonne National Lab., IL (USA). Fusion Power Program); Clemmer, R.G. (Pacific Northwest Lab., Richland, WA (USA))

    1990-11-01

    The tritium recovery system for the US ITER Li{sub 2}O/Be water cooled blanket processes two separate helium purge streams to recover tritium from the Li{sub 2}O zones and the Be zones of the blanket, to process the waste products, and to recirculate the helium back to the blanket. The components are selected to minimize the tritium inventory of the recovery system, and to minimize waste products. The system is robust to either an increase in the tritium release rate or to an in-leak of water in the purge system. Three major components were used to process these streams, first, 5A molecular sieves at {minus}196{degree}C separate hydrogen from the helium, second, a solid oxide electrolysis unit is used to reduce all molecular water, and third, a palladium/silver diffuser is used to ensure that only hydrogen (H{sub 2}, HT) species reach the cryogenic distillation unit. Other units are present to recover tritium from waste products but the three major components are the basis of the blanket tritium recovery system. 32 refs.

  13. Continuous energy cross section library for MCNP/MCNPX based on JENDL high energy file 2007; FXJH7

    OpenAIRE

    佐々 敏信; 菅原 隆徳; 小迫 和明; 深堀 智生

    2008-01-01

    The latest JENDL High Energy File (JENDL/HE) was released in 2007 to respond the requirements of reaction data in high energy range up to several GeV to design accelerator facilities such as accelerator-driven systems and research complex like J-PARC. To apply the JENDL/HE-2007 file to the design study, the cross section library of FXJH7 series was constructed from the JENDL/HE file for the calculation using MCNP and MCNPX codes which are widely used in the field of nuclear reactors, fusion r...

  14. Lecture Notes on Criticality Safety Validation Using MCNP & Whisper

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-11

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

  15. Recent Developments in the MCNP-POLIMI Postprocessing Code

    Energy Technology Data Exchange (ETDEWEB)

    Pozzi, S.A.

    2004-12-17

    The design and analysis of measurements performed with organic scintillators rely on the use of Monte Carlo codes to simulate the interaction of neutrons and photons, originating from fission and other reactions, with the materials present in the system and the radiation detectors. MCNP-PoliMi is a modification of the MCNP-4c code that models the physics of secondary particle emission from fission and other processes realistically. This characteristic allows for the simulation of the higher moments of the distribution of the number of neutrons and photons in a multiplying system. The present report describes the recent additions to the MCNP-PoliMi post-processing code. These include the simulation of detector dead time, multiplicity, and third order statistics.

  16. Features of MCNP6 Relevant to Medical Radiation Physics

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, H. Grady III [Los Alamos National Laboratory; Goorley, John T. [Los Alamos National Laboratory

    2012-08-29

    MCNP (Monte Carlo N-Particle) is a general-purpose Monte Carlo code for simulating the transport of neutrons, photons, electrons, positrons, and more recently other fundamental particles and heavy ions. Over many years MCNP has found a wide range of applications in many different fields, including medical radiation physics. In this presentation we will describe and illustrate a number of significant recently-developed features in the current version of the code, MCNP6, having particular utility for medical physics. Among these are major extensions of the ability to simulate large, complex geometries, improvement in memory requirements and speed for large lattices, introduction of mesh-based isotopic reaction tallies, advances in radiography simulation, expanded variance-reduction capabilities, especially for pulse-height tallies, and a large number of enhancements in photon/electron transport.

  17. 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.

  18. Geometry creation for MCNP by Sabrina and XSM

    Energy Technology Data Exchange (ETDEWEB)

    Van Riper, K.A.

    1994-02-01

    The Monte Carlo N-Particle transport code MCNP is based on a surface description of 3-dimensional geometry. Cells are defined in terms of boolean operations on signed quadratic surfaces. MCNP geometry is entered as a card image file containing coefficients of the surface equations and a list of surfaces and operators describing cells. Several programs are available to assist in creation of the geometry specification, among them Sabrina and the new ``Smart Editor`` code XSM. We briefly describe geometry creation in Sabrina and then discuss XSM in detail. XSM is under development; our discussion is based on the state of XSM as of January 1, 1994.

  19. An Electron/Photon/Relaxation Data Library for MCNP6

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, III, H. Grady [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-08-07

    The capabilities of the MCNP6 Monte Carlo code in simulation of electron transport, photon transport, and atomic relaxation have recently been significantly expanded. The enhancements include not only the extension of existing data and methods to lower energies, but also the introduction of new categories of data and methods. Support of these new capabilities has required major additions to and redesign of the associated data tables. In this paper we present the first complete documentation of the contents and format of the new electron-photon-relaxation data library now available with the initial production release of MCNP6.

  20. Construction Safety Forecast for ITER

    Energy Technology Data Exchange (ETDEWEB)

    cadwallader, lee charles

    2006-11-01

    The International Thermonuclear Experimental Reactor (ITER) project is poised to begin its construction activity. This paper gives an estimate of construction safety as if the experiment was being built in the United States. This estimate of construction injuries and potential fatalities serves as a useful forecast of what can be expected for construction of such a major facility in any country. These data should be considered by the ITER International Team as it plans for safety during the construction phase. Based on average U.S. construction rates, ITER may expect a lost workday case rate of < 4.0 and a fatality count of 0.5 to 0.9 persons per year.

  1. Overview of recent nuclear analyses for the Upper ECH launcher in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Serikov, A., E-mail: serikov@inr.fzk.d [Association FZK-EURATOM, Forschungszentrum Karlsruhe, P.O. Box 3640, D-76021 Karlsruhe (Germany); Fischer, U.; Grosse, D.; Heidinger, R.; Kleefeldt, K.; Spaeh, P.; Strauss, D.; Vaccaro, A. [Association FZK-EURATOM, Forschungszentrum Karlsruhe, P.O. Box 3640, D-76021 Karlsruhe (Germany)

    2010-12-15

    An overview is given for the analyses of the nuclear physics characteristics in support of the development of the Electron Cyclotron Heating (ECH) launcher installed in the ITER upper port. The launcher's Quasi-Optical (QO) system of millimeter-wave guides represents a pathway for the neutron streaming which results in radiation loads on the launcher internals, mainly on the front steering (FS) mirrors and neighboring ITER components, in particular the vacuum vessel (VV) and the superconducting magnets surrounding the launcher. The radiation transport calculations were performed with the Monte Carlo code MCNP5 employing the recent MCNP geometry model of ITER called Alite and the FENDL-2.1 nuclear data. A dedicated CAD model of the QO ECH launcher, generated with CATIA V5, was converted at FZK into the MCNP5 geometry representation using the McCad conversion tool. The neutron and photon fluxes and critical nuclear responses, such as nuclear heating, neutron damage, and the helium production rate were calculated in the paper and compared with the ITER nuclear design limits. On the basis of the results obtained from the nuclear analyses, it is concluded that the recent design of the QO ECH launcher satisfies the ITER radiation requirements, and thus, from this point of view, can be operated safely.

  2. Performance of the MTR core with MOX fuel using the MCNP4C2 code.

    Science.gov (United States)

    Shaaban, Ismail; Albarhoum, Mohamad

    2016-08-01

    The MCNP4C2 code was used to simulate the MTR-22 MW research reactor and perform the neutronic analysis for a new fuel namely: a MOX (U3O8&PuO2) fuel dispersed in an Al matrix for One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) in its core. Its new characteristics were compared to its original characteristics based on the U3O8-Al fuel. Experimental data for the neutronic parameters including criticality relative to the MTR-22 MW reactor for the original U3O8-Al fuel at nominal power were used to validate the calculated values and were found acceptable. The achieved results seem to confirm that the use of MOX fuel in the MTR-22 MW will not degrade the safe operational conditions of the reactor. In addition, the use of MOX fuel in the MTR-22 MW core leads to reduce the uranium fuel enrichment with (235)U and the amount of loaded (235)U in the core by about 34.84% and 15.21% for the ONT and TNTs cases, respectively.

  3. Processing of W-Cu functionally graded materials (FGM) through the powder metallurgy route: application as plasma facing components for ITER-like thermonuclear fusion reactor; Elaboration de materiaux W-Cu a gradient de proprietes fonctionnelles (FGM) par metallurgie des poudres: application en tant que composants face au plasma de machines de fusion thermonucleaire de type Iter

    Energy Technology Data Exchange (ETDEWEB)

    Raharijaona, J.J.

    2009-11-15

    The aim of this study was to study and optimize the sintering of W-Cu graded composition materials, for first wall of ITER-like thermonuclear reactor application. The graded composition in the material generates graded functional properties (Functionally Graded Materials - FGM). Rough thermomechanical calculations have shown the interest of W-Cu FGM to improve the lifetime of Plasma Facing Components (PFC). To process W-Cu FGM, powder metallurgy route was analyzed and optimized from W-CuO powder mixtures. The influence of oxide reduction on the sintering of powder mixtures was highlighted. An optimal heating treatment under He/H{sub 2} atmosphere was determined. The sintering mechanisms were deduced from the analysis of the effect of the Cu-content. Sintering of W-Cu materials with a graded composition and grain size has revealed two liquid migration steps: i) capillary migration, after the Cu-melting and, ii) expulsion of liquid, at the end of sintering, from the dense part to the porous part, due to the continuation of W-skeleton sintering. These two steps were confirmed by a model based on capillary pressure calculation. In addition, thermal conductivity measurements were conducted on sintered parts and showed values which gradually increase with the Cu-content. Hardness tests on a polished cross-section in the bulk are consistent with the composition profiles obtained and the differential grain size. (author)

  4. MCNP6. Simulating Correlated Data in Fission Events

    Energy Technology Data Exchange (ETDEWEB)

    Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Sood, Avneet [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-12-03

    This report is a series of slides discussing the MCNP6 code and its status in simulating fission. Applications of interest include global security and nuclear nonproliferation, detection of special nuclear material (SNM), passive and active interrogation techniques, and coincident neutron and photon leakage.

  5. Duplicating MC-15 Output with Python and MCNP

    Energy Technology Data Exchange (ETDEWEB)

    McSpaden, Alexander Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-08-23

    Two Python scripts have been written that process the output files of MCNP6 into a format that mimics the list-mode output of Los Alamos National Laboratory’s MC-15 and NPOD neutron detection systems. This report details the methods implemented in these scripts and instructions on their use.

  6. ETR/ITER systems code

    Energy Technology Data Exchange (ETDEWEB)

    Barr, W.L.; Bathke, C.G.; Brooks, J.N.; Bulmer, R.H.; Busigin, A.; DuBois, P.F.; Fenstermacher, M.E.; Fink, J.; Finn, P.A.; Galambos, J.D.; Gohar, Y.; Gorker, G.E.; Haines, J.R.; Hassanein, A.M.; Hicks, D.R.; Ho, S.K.; Kalsi, S.S.; Kalyanam, K.M.; Kerns, J.A.; Lee, J.D.; Miller, J.R.; Miller, R.L.; Myall, J.O.; Peng, Y-K.M.; Perkins, L.J.; Spampinato, P.T.; Strickler, D.J.; Thomson, S.L.; Wagner, C.E.; Willms, R.S.; Reid, R.L. (ed.)

    1988-04-01

    A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs.

  7. Requirements for ITER diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Young, K.M.

    1991-01-01

    The development and design of plasma diagnostics for the International Thermonuclear Experimental Reactor (ITER) present a formidable challenge for experimental plasma physicists. The large plasma size, the high central density and temperature and the very high thermal wall loadings provide new challenges for present measurement techniques and lead to a search for new methods. But the physics and control requirements for the long burn phase of the discharge, combined with very limited access to the plasma, constrained by the requirement for radiation shielding of the coils and sharing of access ports with heating and current drive power, remote manipulation, fueling and turn blanket modules, make for very difficult design choices. An initial attempt at these choices has been made by an international team of diagnostic physicists, gathering together in a series of three workshops during the ITER Conceptual Design Activity. This paper is based on that report and provides a summary of its most important points. To provide a background against which to place the diagnostic requirements and design concepts, the ITER device, its most important plasma properties and the proposed experimental program will be described. The specifications for the measurement of the plasma parameters and the proposed diagnostics for these measurements will then be addressed, followed by some examples of the design concepts that have been proposed. As a result of these design studies, it was clear that there were many uncertainties associated with these concepts, particularly because of the nuclear radiation environment, so that a Research and Development Program for diagnostic hardware was established. It will also be briefly summarized.

  8. ITER中子通量监测器的优化计算%Optical Calculations of Neutron Flux Monitor for ITER

    Institute of Scientific and Technical Information of China (English)

    李初; 王强; 兰礼; 刘虓瀚; 曾军; 刘艺琴; 罗小兵

    2012-01-01

    中子通量监测器(NFM)可实现ITER实时的中子通量测定,转化得到聚变功率,功率密度,等离子体温度等.获得NFM探测效率对能量的相对平坦响应对准确诊断十分必要.论文针对特定的NFM裂变室结构,运用MCNP—4C对裂变室包裹层慢化剂/屏蔽材料种类及厚度进行了优化计算.这些工作对探测器裂变室结构的优化设计实验标定及定型具有重要意义.%The Neutron Flux Monitor(NFM) can provide the real - time flux of ITER( International Thermonuclear Experimental Reactor ) , and get the fusion power and temperature of the plasma after transformation. A relative flat energy response curve of neutron detection efficiency is essential for accurate diagnosis of NFM in ITER. The paper makes an optimal computation on thickness of different moderator/ shielding material with the MCNP - 4C as to the specific structure of NFM fission chamber. It is significant for the optimal design and the experimental calibration of the NFM

  9. Possible Improvements to MCNP6 and its CEM/LAQGSM Event-Generators

    Energy Technology Data Exchange (ETDEWEB)

    Mashnik, Stepan Georgievich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-08-04

    This report is intended to the MCNP6 developers and sponsors of MCNP6. It presents a set of suggested possible future improvements to MCNP6 and to its CEM03.03 and LAQGSM03.03 event-generators. A few suggested modifications of MCNP6 are quite simple, aimed at avoiding possible problems with running MCNP6 on various computers, i.e., these changes are not expected to change or improve any results, but should make the use of MCNP6 easier; such changes are expected to require limited man-power resources. On the other hand, several other suggested improvements require a serious further development of nuclear reaction models, are expected to improve significantly the predictive power of MCNP6 for a number of nuclear reactions; but, such developments require several years of work by real experts on nuclear reactions.

  10. Comparison of MCNP4B and WIMS-AECL calculations of coolant-void-reactivity effects for uniform lattices of CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S

    1999-05-01

    This paper compares the results of coolant-void reactivity (CVR) reactor-physics calculations performed using the Monte Carlo N-particle transport code, MCNP version 4B, with those obtained using Atomic Energy of Canada Limited's (AECL's) latest version of the Winfrith improved multigroup scheme (WIMS) code, WIMS-AECL version 2-5c. Cross sections derived from the evaluated nuclear data file version B-VI (ENDF/B-VI) are used for both the WIMS-AECL and MCNP4B calculations. The comparison is made for uniform lattices at room temperature containing either fresh natural uranium or mixed oxide (MOX) 37-element CANDU fuel. The MOX fuel composition corresponds roughly to that of irradiated CANDU fuel at a burnup of about 4500 MWd/tU. The level of agreement between the CVR predictions of WIMS-AECL and MCNP4B is studied as a function of lattice buckling (a measure of the curvature of the neutron-flux distribution) over the range from 0.0 to 4.1 m{sup -2} . For the cases studied, it is found that the absolute k values calculated by WIMS-AECL are higher than those of MCNP4B by several mk (1 mk is a change of 0.001 in k), amounts that depend on the fuel type being modelled and the particular cross-section data used. However, the agreement between WIMS-AECL and MCNP4B is much better for the CVR (i.e., the {delta}k on coolant voiding), and is relatively insensitive to the fuel type. (author)

  11. Comparison of MCNP4B and WIMS-AECL calculations of coolant-void-reactivity effects for uniform lattices of CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    1999-07-01

    This paper compares the results of coolant-void reactivity (CVR) reactor-physics calculations performed using the Monte Carlo N-particle transport code, MCNP version 4B, with those obtained using Atomic Energy of Canada Limited's (AECL's) latest version of the Winfrith improved multigroup scheme (WIMS) code, WIMS-AECL version 2-5c. Cross sections derived from the evaluated nuclear data file version B-VT (ENDF/B-VI) are used for both the WIMS-AECL and MCNP4B calculations. The comparison is made for uniform lattices at room temperature containing either fresh natural uranium or mixed oxide (MOX) 37-element CANDU fuel. The MOX fuel composition corresponds roughly to that of irradiated CANDU fuel at a burnup of about 4500 MWd/tU. The level of agreement between the CVR predictions of WIMS-AECL and MCNP4B is studied as a function of lattice buckling (a measure of the curvature of the neutron-flux distribution) over the range from 0.0 to 4.1 m{sup -2}. For the cases studied, it is found that the absolute keff values calculated by WIMS-AECL are higher than those of MCNP4B by several mk (1 mk is a change of 0.001 in keff), amounts that depend on the fuel type being modelled and the particular cross-section data used. However, the agreement between WIMS-AECL and MCNP4B is much better for the CVR (i.e., the {delta}keff on coolant voiding), and is relatively insensitive to the fuel type. (author)

  12. Modelling and shielding analysis of the neutral beam injector ports in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Pereslavtsev, P., E-mail: pavel.pereslavtsev@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Fischer, U. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Loughlin, M. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Lu, Lei [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Polunovskiy, E. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Vielhaber, S. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2015-10-15

    Highlights: • The engineering CAD models of the NBI ports were simplified on the CATIA platform. • CAD to MCNP model convesion was done making use of McCAD converting tool. • The new NBI port model was integrated into 80° A-lite ITER torus sector model. • The nuclear responces important for the safety issues were assessed. - Abstract: A new MCNP geometry model of the ITER Neutral Beam Injection (NBI) ports was developed starting from the latest engineering CAD models provided by ITER. The model includes 3 heating (HNBI) ports and one diagnostic port (DNBI), and extends up to the bio-shield. The engineering CAD models were simplified on the CATIA platform according to the neutronic requirements and then converted into MCNP geometry making use of the McCad conversion tool. Finally, the new NBI port model was integrated into an available 80° A-lite ITER torus sector model. The nuclear analysis performed on the basis of this model provides the following nuclear responses: the neutron flux distribution in all NBI ports, the nuclear heating distribution in all NBI ducts; the nuclear heating and radiation loads to the TFC magnets; the radiation damage and gas production in the VV; and the distribution of the shutdown dose rate inside the cryostat.

  13. Nuclear analysis of ITER Test Blanket Module Port Plug

    Energy Technology Data Exchange (ETDEWEB)

    Villari, Rosaria, E-mail: rosaria.villari@enea.it [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Kim, Byoung Yoon; Barabash, Vladimir; Giancarli, Luciano; Levesy, Bruno; Loughlin, Michael; Merola, Mario [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 Saint Paul-lez-Durance Cedex (France); Moro, Fabio [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Pascal, Romain [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 Saint Paul-lez-Durance Cedex (France); Petrizzi, Luigino [European Commission, DG Research & Innovation G5, CDMA 00/030, B-1049 Brussels (Belgium); Polunovsky, Eduard; Van Der Laan, Jaap G. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 Saint Paul-lez-Durance Cedex (France)

    2015-10-15

    Highlights: • 3D nuclear analysis of the ITER TBM Port Plug (PP). • Calculations of neutron fluxes, nuclear heating, damage and He-production in TBM PP components. • Shutdown dose rate assessment with Advanced D1S method considering different configurations. • Potential design improvements to reduce the shutdown dose rate in the port interspace. - Abstract: Nuclear analyses have been performed for the ITER Test Blanket Module Port Plug (TBM PP) using the MCNP-5 Monte Carlo Code. A detailed 3D model of the TBM Port Plug with dummy TBM has been integrated into the ITER MCNP model (B-lite v.3). Neutron fluxes, nuclear heating, helium production and neutron damage have been calculated in all the TBM PP components. Global shutdown dose rate calculations have also been performed with Advanced D1S method for different configurations of the TBM PP system. This paper presents the results of these analyses and discusses potential design improvements aiming to further reduce the shutdown dose rate in the port interspace.

  14. Characterization of control rod worths and fuel rod power peaking factors in the university of Utah TRIGA Mark I reactor

    Directory of Open Access Journals (Sweden)

    Alroumi Fawaz

    2016-01-01

    Full Text Available Control rod reactivity (worths for the three control rods and fuel rod power peaking factors in the University of Utah research reactor (100 kW TRIGA Mark I are characterized using the AGENT code system and the results described in this paper. These values are compared to the MCNP6 and existing experimental measurements. In addition, the eigenvalue, neutron spatial flux distributions and reaction rates are analyzed and discussed. The AGENT code system is widely benchmarked for various reactor types and complexities in their geometric arrangements of the assemblies and reactor core material distributions. Thus, it is used as a base methodology to evaluate neutronics variables of the research reactor at the University of Utah. With its much shorter computation time than MCNP6, AGENT provides agreement with the MCNP6 within a 0.5 % difference for the eigenvalue and a maximum difference of 10% in the power peaking factor values. Differential and integral control rod worths obtained by AGENT show well agreement with MCNP6 and the theoretical model. However, regulating the control rod worth is somewhat overestimated by both MCNP6 and AGENT models when compared to the experimental/theoretical values. In comparison to MCNP6, the total control rod worths and shutdown margin obtained with AGENT show better agreement to the experimental values.

  15. Optimised Iteration in Coupled Monte Carlo - Thermal-Hydraulics Calculations

    Science.gov (United States)

    Hoogenboom, J. Eduard; Dufek, Jan

    2014-06-01

    This paper describes an optimised iteration scheme for the number of neutron histories and the relaxation factor in successive iterations of coupled Monte Carlo and thermal-hydraulic reactor calculations based on the stochastic iteration method. The scheme results in an increasing number of neutron histories for the Monte Carlo calculation in successive iteration steps and a decreasing relaxation factor for the spatial power distribution to be used as input to the thermal-hydraulics calculation. The theoretical basis is discussed in detail and practical consequences of the scheme are shown, among which a nearly linear increase per iteration of the number of cycles in the Monte Carlo calculation. The scheme is demonstrated for a full PWR type fuel assembly. Results are shown for the axial power distribution during several iteration steps. A few alternative iteration method are also tested and it is concluded that the presented iteration method is near optimal.

  16. A Verification of MCNP6 FMESH Tally Capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Swift, Alicia L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McKigney, Edward A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Schirato, Richard C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Robinson, Alex Philip [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Temple, Brian Allen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-02-10

    This work serves to verify the MCNP6 FMESH capability through comparison to two types of data. FMESH tallies, binned in time, were generated on an ideal detector face for neutrons undergoing a single scatter in a graphite target. For verification, FMESH results were compared to analytic calculations of the nonrelativistic TOF for elastic and inelastic single neutron scatters (TOF for the purposes of this paper is the time for a neutron to travel from its scatter location in the graphite target to the detector face). FMESH tally results were also compared to F4 tally results, an MNCP tally that calculates fluence in the same way as the FMESH tally. The FMESH tally results agree well with the analytic results and the F4 tally; hence, it is believed that, for simple geometries, MCNP6 FMESH tallies represent the physics of neutron scattering very well.

  17. Hybrid parallel code acceleration methods in full-core reactor physics calculations

    Energy Technology Data Exchange (ETDEWEB)

    Courau, T.; Plagne, L.; Ponicot, A. [EDF R and D, 1, Avenue du General de Gaulle, 92141 Clamart Cedex (France); Sjoden, G. [Nuclear and Radiological Engineering, Georgia Inst. of Technology, Atlanta, GA 30332 (United States)

    2012-07-01

    When dealing with nuclear reactor calculation schemes, the need for three dimensional (3D) transport-based reference solutions is essential for both validation and optimization purposes. Considering a benchmark problem, this work investigates the potential of discrete ordinates (Sn) transport methods applied to 3D pressurized water reactor (PWR) full-core calculations. First, the benchmark problem is described. It involves a pin-by-pin description of a 3D PWR first core, and uses a 8-group cross-section library prepared with the DRAGON cell code. Then, a convergence analysis is performed using the PENTRAN parallel Sn Cartesian code. It discusses the spatial refinement and the associated angular quadrature required to properly describe the problem physics. It also shows that initializing the Sn solution with the EDF SPN solver COCAGNE reduces the number of iterations required to converge by nearly a factor of 6. Using a best estimate model, PENTRAN results are then compared to multigroup Monte Carlo results obtained with the MCNP5 code. Good consistency is observed between the two methods (Sn and Monte Carlo), with discrepancies that are less than 25 pcm for the k{sub eff}, and less than 2.1% and 1.6% for the flux at the pin-cell level and for the pin-power distribution, respectively. (authors)

  18. New Challenge to the Existing Fuelling Technique for ITER

    Institute of Scientific and Technical Information of China (English)

    DENGBaiquan; PENGLilin

    2003-01-01

    A new challenge to the existing pellet fuelling technique for the International Thermonuclear Experimental Reactor (ITER) has been discussed by computing ablation rate using Kuteev lentil model with considering the kinetic effects. Numerical integral results show the existing pellet injection technique will be difficult to meet the deep fuelling requirement for the reactor relevant plasma of ITER. As high as a pressure of 254 MPa should be applied to the pellet accelerator of 2-meters long single stage pneumatic gun,

  19. Energetic Light Fragment Production Capability in MCNP6

    CERN Document Server

    Kerby, Leslie M; Gudima, Konstantin K; Sierk, Arnold J; Bull, Jeffrey S; James, Michael R

    2016-01-01

    The goal of this research is to enable MCNP6 to produce high-energy light fragments. These energetic light fragments may be emitted by our models through three processes: Fermi breakup, preequilibrium, and coalescence. We explore the emission of light fragments through each of these mechanisms and demonstrate an improved agreement with experimental data achieved by extending precompound models to include emission of fragments heavier than $^4$He.

  20. Neutronics Analysis of the ITER In-Vessel Viewing System

    CERN Document Server

    Turner, Andrew; Puiu, Adrian

    2013-01-01

    The In-Vessel Viewing System (IVVS) in ITER consists of six identical units which are deployed between pulses or during shutdown, to perform visual examination and metrology of plasma facing components. The system is housed in dedicated ports at B1 level, with deployment at the level between the divertor cassettes and the lowermost outboard blanket modules. Boron carbide shielding blocks are envisaged to protect the sensitive components of the IVVS from damage during operations, and personnel from radiation fields. In order to progress the design of the IVVS beyond the pre-conceptual stage, analyses were conducted using MCNP to determine the acceptability of a series of different shielding configurations.

  1. ITER faces further five-year delay

    Science.gov (United States)

    Clery, Daniel

    2016-06-01

    The €14bn ITER fusion reactor currently under construction in Cadarache, France, will require an additional cash injection of €4.6bn if it is to start up in 2025 - a target date that is already five years later than currently scheduled.

  2. Nuclear Reactors

    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.

  3. Embeddings of Iteration Trees

    OpenAIRE

    Mitchell, William

    1992-01-01

    This paper, dating from May 1991, contains preliminary (and unpublishable) notes on investigations about iteration trees. They will be of interest only to the specialist. In the first two sections I define notions of support and embeddings for tree iterations, proving for example that every tree iteration is a direct limit of finite tree iterations. This is a generalization to models with extenders of basic ideas of iterated ultrapowers using only ultrapowers. In the final section (which is m...

  4. Comparison of CAP88 and MCNP for Overhead Gamma-emitting Plumes

    Energy Technology Data Exchange (ETDEWEB)

    Mcnaughton, Michael [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Gillis, Jessica Mcdonnel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McClory, Aysha Reede [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Whicker, Jeffrey Jay [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fuehne, David Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-01-08

    The purpose of this paper is to use the Monte Carlo N-Particle Code (MCNP) to investigate the dose from gamma-emitting radionuclides such as Carbon-11 when a plume passes overhead. MCNP results are compared with results from the EPA program, CAP88. In some cases, typically near the source during stable conditions, the CAP88 results are less than the MCNP results. However, in the case of a receptor 800 m from a source at the Los Alamos Neutron Science Center (LANSCE), the CAP88 result is greater than the MCNP result.

  5. Reactor Dosimetry State of the Art 2008

    Science.gov (United States)

    Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

    2009-08-01

    . Williams, A. P. Ribaric and T. Schnauber. Agile high-fidelity MCNP model development techniques for rapid mechanical design iteration / J. A. Kulesza.Extension of Raptor-M3G to r-8-z geometry for use in reactor dosimetry applications / M. A. Hunter, G. Longoni and S. L. Anderson. In vessel exposure distributions evaluated with MCNP5 for Atucha II / J. M. Longhino, H. Blaumann and G. Zamonsky. Atucha I nuclear power plant azimutal ex-vessel flux profile evaluation / J. M. Longhino ... [et al.]. UFTR thermal column characterization and redesign for maximized thermal flux / C. Polit and A. Haghighat. Activation counter using liquid light-guide for dosimetry of neutron burst / M. Hayashi ... [et al.]. Control rod reactivity curves for the annular core research reactor / K. R. DePriest ... [et al.]. Specification of irradiation conditions in VVER-440 surveillance positions / V. Kochkin ... [et al.]. Simulations of Mg-Ar ionisation and TE-TE ionisation chambers with MCNPX in a straightforward gamma and beta irradiation field / S. Nievaart ... [et al.]. The change of austenitic stainless steel elements content in the inner parts of VVER-440 reactor during operation / V. Smutný, J. Hep and P. Novosad. Fast neutron environmental spectrometry using disk activation / G. Lövestam ... [et al.]. Optimization of the neutron activation detector location scheme for VVER-lOOO ex-vessel dosimetry / V. N. Bukanov ... [et al.]. Irradiation conditions for surveillance specimens located into plane containers installed in the WWER-lOOO reactor of unit 2 of the South-Ukrainian NPP / O. V. Grytsenko. V. N. Bukanov and S. M. Pugach. Conformity between LRO mock-ups and VVERS NPP RPV neutron flux attenuation / S. Belousov. Kr. Ilieva and D. Kirilova. FLUOLE: a new relevant experiment for PWR pressure vessel surveillance / D. Beretz ... [et al.]. Transport of neutrons and photons through the iron and water layers / M. J. Kost'ál ... [et al.]. Condition evaluation of spent nuclear fuel assemblies

  6. On the safety of ITER accelerators.

    Science.gov (United States)

    Li, Ge

    2013-01-01

    Three 1 MV/40A accelerators in heating neutral beams (HNB) are on track to be implemented in the International Thermonuclear Experimental Reactor (ITER). ITER may produce 500 MWt of power by 2026 and may serve as a green energy roadmap for the world. They will generate -1 MV 1 h long-pulse ion beams to be neutralised for plasma heating. Due to frequently occurring vacuum sparking in the accelerators, the snubbers are used to limit the fault arc current to improve ITER safety. However, recent analyses of its reference design have raised concerns. General nonlinear transformer theory is developed for the snubber to unify the former snubbers' different design models with a clear mechanism. Satisfactory agreement between theory and tests indicates that scaling up to a 1 MV voltage may be possible. These results confirm the nonlinear process behind transformer theory and map out a reliable snubber design for a safer ITER.

  7. Study on Modeling Technology in Digital Reactor System

    Institute of Scientific and Technical Information of China (English)

    刘晓平; 罗月童; 童莉莉

    2004-01-01

    Modeling is the kernel part of a digital reactor system. As an extensible platform for reactor conceptual design, it is very important to study modeling technology and develop some kind of tools to speed up preparation of all classical computing models. This paper introduces the background of the project and basic conception of digital reactor. MCAM is taken as an example for modeling and its related technologies used are given. It is an interface program for MCNP geometry model developed by FDS team (ASIPP & HUT), and designed to run on windows system. MCAM aims at utilizing CAD technology to facilitate creation of MCNP geometry model. There have been two ways for MCAM to utilize CAD technology:(1) Making use of user interface technology in aid of generation of MCNP geometry model;(2) Making use of existing 3D CAD model to accelerate creation of MCNP geometry model. This paper gives an overview of MCAM's major function. At last, several examples are given to demonstrate MCAM's various capabilities.

  8. Household energy consumption: the future is in our hands. ITER, an experimental fusion reactor. Do CO{sub 2}-free energies exist? Liquefied natural gas, king of the gas market; Consommation d'energie domestique: prenons l'avenir entre nos mains. ITER, un reacteur experimental de fusion. Existe-t-il des energies sans CO{sub 2}? Le gaz naturel liquefie, force motrice du marche du gaz

    Energy Technology Data Exchange (ETDEWEB)

    Anon

    2008-07-01

    This issue of Alternatives newsletter features 4 main articles dealing with: 1 - Household energy consumption - the future is in our hands: With energy resources growing scarcer and more expensive, everyone has a duty to conserve energy. Because combating global warming also means adopting simple habits and using the right equipment - with help from our governments to lead us to change. A practical look at what we can do. 2 - ITER, an experimental fusion reactor: The entire international community is trying to reproduce here on Earth the fusion of hydrogen atoms occurring naturally in the Sun, lured by the promise of a virtually inexhaustible source of energy. More on ITER from the project's Director General. 3 - Do CO{sub 2}-free energies exist?: As nations struggle to reduce greenhouse gas emissions, the question is moot. Environmental engineer Jean-Marc Jancovici gives us his point of view. 4 - Liquefied natural gas, king of the gas market: LNG's many advantages are enticing industry to develop supply routes and infrastructure to meet strong demand. But the race for LNG is not without its limits.

  9. Digital Spectrometric System for Characterization of Mixed Neutron - Gamma Field in the Experimental Reactor LR-0

    Science.gov (United States)

    Mravec, Filip; Matej, Zdenek; Cvachovec, Frantisek; Kostal, Michal; Veskrna, Martin; Prenosil, Vaclav

    2016-02-01

    LR-0 reactor is an experimental reactor in NRI Rez, Czech Republic. So far an analog apparatus was used for measurements of the space-energy distribution of the neutron gamma mixed field inside the reactor vessel. Recently we measured in LR-0 with fully digital apparatus using Agilent digitizer and compared our results with older established results from analog apparatus and also with MCNP calculations.

  10. Assessment of evaluated (n,d) energy-angle elastic scattering distributions using MCNP simulations of critical measurements and simplified calculation benchmarks

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario (Canada)

    2008-07-01

    Different evaluated (n,d) energy-angle elastic scattering distributions produce k-effective differences in MCNP5 simulations of critical experiments involving heavy water (D{sub 2}O) of sufficient magnitude to suggest a need for new (n,d) scattering measurements and/or distributions derived from modern theoretical nuclear models, especially at neutron energies below a few MeV. The present work focuses on the small reactivity change of < 1 mk that is observed in the MCNP5 D{sub 2}O coolant-void-reactivity calculation bias for simulations of two pairs of critical experiments performed in the ZED-2 reactor at the Chalk River Laboratories when different nuclear data libraries are used for deuterium. The deuterium data libraries tested include Endf/B-VII.0, Endf/B-VI.4, JENDL-3.3 and a new evaluation, labelled Bonn-B, which is based on recent theoretical nuclear-model calculations. Comparison calculations were also performed for a simplified, two-region, spherical model having an inner, 250-cm radius, homogeneous sphere of UO{sub 2}, without and with deuterium, and an outer 20-cm-thick deuterium reflector. A notable observation from this work is the reduction of about 0.4 mk in the MCNP5 ZED-2 CVR calculation bias that is obtained when the O-in-UO{sub 2} thermal scattering data comes from Endf-B-VII.0. (author)

  11. Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP; Calculos de criticidad de un contenedor de combustible generico para ensambles combustibles PWR, mediante el codigo MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Vargas E, S.; Esquivel E, J.; Ramirez S, J. R., E-mail: samuel.vargas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

  12. MCNP5 CALCULATIONS REPLICATING ARH-600 NITRATE DATA

    Energy Technology Data Exchange (ETDEWEB)

    FINFROCK SH

    2011-10-25

    This report serves to extend the previous document: 'MCNP Calculations Replicating ARH-600 Data' by replicating the nitrate curves found in ARH-600. This report includes the MCNP models used, the calculated critical dimension for each analyzed parameter set, and the resulting data libraries for use with the CritView code. As with the ARH-600 data, this report is not meant to replace the analysis of the fissile systems by qualified criticality personnel. The M CNP data is presented without accounting for the statistical uncertainty (although this is typically less than 0.001) or bias and, as such, the application of a reasonable safety margin is required. The data that follows pertains to the uranyl nitrate and plutonium nitrate spheres, infinite cylinders, and infinite slabs of varying isotopic composition, reflector thickness, and molarity. Each of the cases was modeled in MCNP (version 5.1.40), using the ENDF/B-VI cross section set. Given a molarity, isotopic composition, and reflector thickness, the fissile concentration and diameter (or thicknesses in the case of the slab geometries) were varied. The diameter for which k-effective equals 1.00 for a given concentration could then be calculated and graphed. These graphs are included in this report. The pages that follow describe the regions modeled, formulas for calculating the various parameters, a list of cross-sections used in the calculations, a description of the automation routine and data, and finally the data output. The data of most interest are the critical dimensions of the various systems analyzed. This is presented graphically, and in table format, in Appendix B. Appendix C provides a text listing of the same data in a format that is compatible with the CritView code. Appendices D and E provide listing of example Template files and MCNP input files (these are discussed further in Section 4). Appendix F is a complete listing of all of the output data (i.e., all of the analyzed dimensions and

  13. Validation of the Monte Carlo code MCNP-DSP

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.E.; Mihalczo, J.T. [Oak Ridge National Lab., TN (United States)

    1996-09-12

    Several calculations were performed to validate MCNP-DSP, which is a Monte Carlo code that calculates all the time and frequency analysis parameters associated with the {sup 252}Cf-source-driven time and frequency analysis method. The frequency analysis parameters are obtained in two ways: directly by Fourier transforming the detector responses and indirectly by taking the Fourier transform of the autocorrelation and cross-correlation functions. The direct and indirect Fourier processing methods were shown to produce the same frequency spectra and convergence, thus verifying the way to obtain the frequency analysis parameters from the time sequences of detector pulses. (Author).

  14. Uncertainty analysis in MCNP5 calculations for brachytherapy treatment

    Energy Technology Data Exchange (ETDEWEB)

    Gerardy, I., E-mail: gerardy@isib.be [Institut Superieur Industriel de Bruxelles, 150, Rue Royale, B-1000 Brussels (Belgium); Rodenas, J.; Gallardo, S. [Departamento de Ingenieria Quimica y Nuclear, Universidad Politecnica de Valencia (Spain)

    2011-08-15

    The Monte Carlo (MC) method can be applied to simulate brachytherapy treatment planning. The MCNP5 code gives, together with results, a statistical uncertainty associated with them. However, the latter is not the only existing uncertainty related to the simulation and other uncertainties must be taken into account. A complete analysis of all sources of uncertainty having some influence on results of the simulation of brachytherapy treatment is presented in this paper. This analysis has been based on the recommendations of the American Association for Physicist in Medicine (AAPM) and of the International Standard Organisation (ISO).

  15. Pressure vessel calculations for VVER-440 reactors.

    Science.gov (United States)

    Hordósy, G; Hegyi, Gy; Keresztúri, A; Maráczy, Cs; Temesvári, E; Vértes, P; Zsolnay, E

    2005-01-01

    For the determination of the fast neutron load of the reactor pressure vessel a mixed calculational procedure was developed. The procedure was applied to the Unit II of Paks NPP, Hungary. The neutron source on the outer surfaces of the reactor was determined by a core design code, and the neutron transport calculations outside the core were performed by the Monte Carlo code MCNP. The reaction rate in the activation detectors at surveillance positions and at the cavity were calculated and compared with measurements. In most cases, fairly good agreement was found.

  16. Preliminary consideration of CFETR ITER-like case diagnostic system.

    Science.gov (United States)

    Li, G S; Yang, Y; Wang, Y M; Ming, T F; Han, X; Liu, S C; Wang, E H; Liu, Y K; Yang, W J; Li, G Q; Hu, Q S; Gao, X

    2016-11-01

    Chinese Fusion Engineering Test Reactor (CFETR) is a new superconducting tokamak device being designed in China, which aims at bridging the gap between ITER and DEMO, where DEMO is a tokamak demonstration fusion reactor. Two diagnostic cases, ITER-like case and towards DEMO case, have been considered for CFETR early and later operating phases, respectively. In this paper, some preliminary consideration of ITER-like case will be presented. Based on ITER diagnostic system, three versions of increased complexity and coverage of the ITER-like case diagnostic system have been developed with different goals and functions. Version A aims only machine protection and basic control. Both of version B and version C are mainly for machine protection, basic and advanced control, but version C has an increased level of redundancy necessary for improved measurements capability. The performance of these versions and needed R&D work are outlined.

  17. Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation

    Energy Technology Data Exchange (ETDEWEB)

    Pecchia, M.; D' Auria, F. [San Piero A Grado Nuclear Research Group GRNSPG, Univ. of Pisa, via Diotisalvi, 2, 56122 - Pisa (Italy); Mazzantini, O. [Nucleo-electrica Argentina Societad Anonima NA-SA, Buenos Aires (Argentina)

    2012-07-01

    Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3D{sup C}/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI for performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)

  18. MCNP: a general Monte Carlo code for neutron and photon transport

    Energy Technology Data Exchange (ETDEWEB)

    Forster, R.A.; Godfrey, T.N.K.

    1985-01-01

    MCNP is a very general Monte Carlo neutron photon transport code system with approximately 250 person years of Group X-6 code development invested. It is extremely portable, user-oriented, and a true production code as it is used about 60 Cray hours per month by about 150 Los Alamos users. It has as its data base the best cross-section evaluations available. MCNP contains state-of-the-art traditional and adaptive Monte Carlo techniques to be applied to the solution of an ever-increasing number of problems. Excellent user-oriented documentation is available for all facets of the MCNP code system. Many useful and important variants of MCNP exist for special applications. The Radiation Shielding Information Center (RSIC) in Oak Ridge, Tennessee is the contact point for worldwide MCNP code and documentation distribution. A much improved MCNP Version 3A will be available in the fall of 1985, along with new and improved documentation. Future directions in MCNP development will change the meaning of MCNP to Monte Carlo N Particle where N particle varieties will be transported.

  19. MCNP6 and DRiFT modeling efforts for the NEUANCE/DANCE detector array

    Energy Technology Data Exchange (ETDEWEB)

    Pinilla, Maria Isabel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-01-30

    This report seeks to study and benchmark code predictions against experimental data; determine parameters to match MCNP-simulated detector response functions to experimental stilbene measurements; add stilbene processing capabilities to DRiFT; and improve NEUANCE detector array modeling and analysis using new MCNP6 and DRiFT features.

  20. ITER power electrical networks; Sistemas electricos de alimentacion a los consumidores del ITER

    Energy Technology Data Exchange (ETDEWEB)

    Sejas Portela, S.

    2011-07-01

    The ITER project (International Thermonuclear Experimental Reactor) is an international effort to research and development to design, build and operate an experimental facility to demonstrate the scientific and technological possibility of obtaining useful energy from the physical phenomenon known as nuclear fusion.

  1. MCNP modelling of a combined neutron/gamma counter

    CERN Document Server

    Bourva, L C A; Ottmar, H; Weaver, D R

    1999-01-01

    A series of Monte Carlo neutron calculations for a combined gamma/passive neutron coincidence counter has been performed. This type of device, part of a suite of non-destructive assay instruments utilised for the enforcement of the Euratom nuclear safeguards within the European Union, is to be used for high accuracy measurements of the plutonium content of small samples of nuclear materials. The multi-purpose Monte Carlo N-particle (MCNP) code version 4B has been used to model in detail the neutron coincidence detector and to investigate the leakage self-multiplication of PuO sub 2 and mixed U-Pu oxide (MOX) reference samples used to calibrate the instrument. The MCNP calculations have been used together with a neutron coincidence counting interpretative model to determine characteristic parameters of the detector. A comparative study to both experimental and previous numerical results has been performed. Sensitivity curves of the variation of the detector's efficiency, epsilon, to, alpha, the ratio of (alpha...

  2. MCNP6 fragmentation of light nuclei at intermediate energies

    CERN Document Server

    Mashnik, Stepan G

    2014-01-01

    Fragmentation reactions induced on light target nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below are studied with the latest Los Alamos Monte Carlo transport code MCNP6 and with its cascade-exciton model (CEM) and Los Alamos version of the quark-gluon string model (LAQGSM) event generators, version 03.03, used as stand-alone codes. Such reactions are involved in different applications, like cosmic-ray-induced single event upsets (SEU's), radiation protection, and cancer therapy with proton and ion beams, among others; therefore, it is important that MCNP6 simulates them as well as possible. CEM and LAQGSM assume that intermediate-energy fragmentation reactions on light nuclei occur generally in two stages. The first stage is the intranuclear cascade (INC), followed by the second, Fermi breakup disintegration of light excited residual nuclei produced after INC. Both CEM and LAQGSM account also for coalescence of light fragments (complex particles) up to He4 from energetic nucleons ...

  3. Estimation and interpretation of k{sub eff} confidence intervals in MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Urbatsch, T.J. [Michigan Univ., Ann Arbor, MI (United States). Dept. of Nuclear Engineering; Forster, R.A.; Prael, R.E.; Beckman, R.J. [Los Alamos National Lab., NM (United States)

    1995-07-01

    MCNP has three different, but correlated, estimators for Calculating k{sub eff} in nuclear criticality calculations: collision, absorption, and track length estimators. The combination of these three estimators, the three-combined k{sub eff} estimator, is shown to be the best k{sub eff} estimator available in MCNP for estimating k{sub eff} confidence intervals. Theoretically, the Gauss-Markov Theorem provides a solid foundation for MCNP`s three-combined estimator. Analytically, a statistical study, where the estimates are drawn using a known covariance matrix, shows that the three-combined estimator is superior to the individual estimator with the smallest variance. The importance of MCNP`s batch statistics is demonstrated by an investigation of the effects of individual estimator variance bias on the combination of estimators, both heuristically with the analytical study and emprically with MCNP.

  4. Testing the Delayed Gamma Capability in MCNP6

    Energy Technology Data Exchange (ETDEWEB)

    Weldon, Robert A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fensin, Michael L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McKinney, Gregg W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-10-28

    . We examine five different decay chains (two-stage decay to stable) and show the predictability of the MCNP6 delayed gamma feature. Results do show that while the default delayed gamma calculations available in the MCNP6 1.0 release can give accurate results for some isotopes (e.g., 137Ba), the percent differences between the closed form analytic solutions and the MCNP6 calculations were often >40% (28Mg, 28Al, 42K, 47Ca, 47Sc, 60Co). With the MCNP6 1.1 Beta release, the tenth entry on the DBCN card allows improved calculation within <5% as compared to the closed form analytic solutions for immediate parent emissions and transient equilibrium systems. While the tenth entry on the DBCN card for MCNP6 1.1 gives much better results for transient equilibrium systems and parent emissions in general, it does little to improve daughter emissions of secular equilibrium systems. Finally, hypotheses were presented as to why daughter emissions of secular equilibrium systems might be mispredicted in some cases and not in others.

  5. Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Fast Salt Reactor (MFSR)

    Science.gov (United States)

    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.

  6. RADIATION DOSIMETRY AT THE BNL HIGH FLUX BEAM REACTOR AND MEDICAL RESEARCH REACTOR.

    Energy Technology Data Exchange (ETDEWEB)

    HOLDEN,N.E.

    1999-09-10

    RADIATION DOSIMETRY MEASUREMENTS HAVE BEEN PERFORMED OVER A PERIOD OF MANY YEARS AT THE HIGH FLUX BEAM REACTOR (HFBR) AND THE MEDICAL RESEARCH REACTOR (BMRR) AT BROOKHAVEN NATIONAL LABORATORY TO PROVIDE INFORMATION ON THE ENERGY DISTRIBUTION OF THE NEUTRON FLUX, NEUTRON DOSE RATES, GAMMA-RAY FLUXES AND GAMMA-RAY DOSE RATES. THE MCNP PARTICLE TRANSPORT CODE PROVIDED MONTE CARLO RESULTS TO COMPARE WITH VARIOUS DOSIMETRY MEASUREMENTS PERFORMED AT THE EXPERIMENTAL PORTS, AT THE TREATMENT ROOMS AND IN THE THIMBLES AT BOTH HFBR AND BMRR.

  7. Approximate Modified Policy Iteration

    CERN Document Server

    Scherrer, Bruno; Ghavamzadeh, Mohammad; Geist, Matthieu

    2012-01-01

    Modified policy iteration (MPI) is a dynamic programming (DP) algorithm that contains the two celebrated policy and value iteration methods. Despite its generality, MPI has not been thoroughly studied, especially its approximation form which is used when the state and/or action spaces are large or infinite. In this paper, we propose three approximate MPI (AMPI) algorithms that are extensions of the well-known approximate DP algorithms: fitted-value iteration, fitted-Q iteration, and classification-based policy iteration. We provide an error propagation analysis for AMPI that unifies those for approximate policy and value iteration. We also provide a finite-sample analysis for the classification-based implementation of AMPI (CBMPI), which is more general (and somehow contains) than the analysis of the other presented AMPI algorithms. An interesting observation is that the MPI's parameter allows us to control the balance of errors (in value function approximation and in estimating the greedy policy) in the fina...

  8. Applied iterative methods

    CERN Document Server

    Hageman, Louis A

    2004-01-01

    This graduate-level text examines the practical use of iterative methods in solving large, sparse systems of linear algebraic equations and in resolving multidimensional boundary-value problems. Assuming minimal mathematical background, it profiles the relative merits of several general iterative procedures. Topics include polynomial acceleration of basic iterative methods, Chebyshev and conjugate gradient acceleration procedures applicable to partitioning the linear system into a "red/black" block form, adaptive computational algorithms for the successive overrelaxation (SOR) method, and comp

  9. Helias reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Beidler, C.D. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Grieger, G. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Harmeyer, E. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Kisslinger, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Karulin, N. [Nuclear Fusion Institute, Moscow (Russian Federation); Maurer, W. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany); Nuehrenberg, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Rau, F. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Sapper, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Wobig, H. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

    1995-10-01

    The present status of Helias reactor studies is characterised by the identification and investigation of specific issues which result from the particular properties of this type of stellarator. On the technical side these are issues related to the coil system, while physics studies have concentrated on confinement, alpha-particle behaviour and ignition conditions. The usual assumptions have been made in those fields which are common to all toroidal fusion reactors: blanket and shield, refuelling and exhaust, safety and economic aspects. For blanket and shield sufficient space has been provided, a detailed concept will be developed in future. To date more emphasis has been placed on scoping and parameter studies as opposed to fixing a specific set of parameters and providing a detailed point study. One result of the Helias reactor studies is that physical dimensions are on the same order as those of tokamak reactors. However, it should be noticed that this comparison is difficult in view of the large spectrum of tokamak reactors ranging from a small reactor like Aries, to a large device such as SEAFP. The notion that the large aspect ratio of 10 or more in Helias configurations also leads to large reactors is misleading, since the large major radius of 22 m is compensated by the average plasma radius of 1.8 m and the average coil radius of 5 m. The plasma volume of 1400 m{sup 3} is about the same as the ITER reactor and the magnetic energy of the coil system is about the same or even slightly smaller than envisaged in ITER. (orig.)

  10. MCNP(TM) Release 6.1.1 beta: Creating and Testing the Code Distribution

    Energy Technology Data Exchange (ETDEWEB)

    Cox, Lawrence J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Casswell, Laura [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-06-12

    This report documents the preparations for and testing of the production release of MCNP6™1.1 beta through RSICC at ORNL. It addresses tests on supported operating systems (Linux, MacOSX, Windows) with the supported compilers (Intel, Portland Group and gfortran). Verification and Validation test results are documented elsewhere. This report does not address in detail the overall packaging of the distribution. Specifically, it does not address the nuclear and atomic data collection, the other included software packages (MCNP5, MCNPX and MCNP6) and the collection of reference documents.

  11. Compact fusion reactors

    CERN Document Server

    CERN. Geneva

    2015-01-01

    Fusion research is currently to a large extent focused on tokamak (ITER) and inertial confinement (NIF) research. In addition to these large international or national efforts there are private companies performing fusion research using much smaller devices than ITER or NIF. The attempt to achieve fusion energy production through relatively small and compact devices compared to tokamaks decreases the costs and building time of the reactors and this has allowed some private companies to enter the field, like EMC2, General Fusion, Helion Energy, Lawrenceville Plasma Physics and Lockheed Martin. Some of these companies are trying to demonstrate net energy production within the next few years. If they are successful their next step is to attempt to commercialize their technology. In this presentation an overview of compact fusion reactor concepts is given.

  12. Study of heat and synchrotron radiation transport in fusion tokamak plasmas. Application to the modelling of steady state and fast burn termination scenarios for the international experimental fusion reactor ITER

    Energy Technology Data Exchange (ETDEWEB)

    Villar Colome, J. [Association Euratom-CEA, Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee]|[Universitat Polytechnica de Catalunya (Spain)

    1997-12-01

    The aim of this thesis is to give a global scope of the problem of energy transport within a thermonuclear plasma in the context of its power balance and the implications when modelling ITER operating scenarios. This is made in two phases. First, by furnishing new elements to the existing models of heat and synchrotron radiation transport in a thermonuclear plasma. Second, by applying the improved models to plasma engineering studies of ITER operating scenarios. The scenarios modelled are the steady state operating point and the transient that appears to have the biggest technological implications: the fast burn termination. The conduction-convection losses are modelled through the energy confinement time. This parameter is empirically obtained from the existing experimental data, since the underlying mechanisms are not well understood. In chapter 2 an expression for the energy confinement time is semi-analytically deduced from the Rebut-Lallia-Watkins local transport model. The current estimates of the synchrotron radiation losses are made with expressions of the dimensionless transparency factor deduced from a 0-dimensional cylindrical model proposed by Trubnikov in 1979. In chapter 3 realistic hypothesis for the cases of cylindrical and toroidal geometry are included in the model to deduce compact explicit expressions for the fast numerical computation of the synchrotron radiation losses. Numerical applications are provided for the cylindrical case. The results are checked against the existing models. In chapter 4, the nominal operating point of ITER and its thermal stability is studied by means of a 0-dimensional burn model of the thermonuclear plasma in ignition. This model is deduced by the elements furnished by the plasma particle and power balance. Possible heat overloading on the plasma facing components may provoke severe structural damage, implying potential safety problems related to tritium inventory and metal activation. In chapter 5, the assessment

  13. Iteration, Not Induction

    Science.gov (United States)

    Dobbs, David E.

    2009-01-01

    The main purpose of this note is to present and justify proof via iteration as an intuitive, creative and empowering method that is often available and preferable as an alternative to proofs via either mathematical induction or the well-ordering principle. The method of iteration depends only on the fact that any strictly decreasing sequence of…

  14. The reactor core TRIGA Mark-III with fuels type 30/20; El nucleo del reactor TRIGA Mark-III con combustible tipo 30/20

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F., E-mail: fortunato.aguilar@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    This work describes the calculation series carried out with the program MCNP5 in order to define the configuration of the reactor core with fuels 30/20 (fuels with 30% of uranium content in the Or-Zr-H mixture and a nominal enrichment of 20%). To select the configuration of the reactor core more appropriate to the necessities and future uses of the reactor, the following criterions were taken into account: a) the excess in the reactor reactivity, b) the switch out margin and c) to have new irradiation facilities inside the reactor core. Taking into account these criterions is proceeded to know the characteristics of the components that form the reactor core (dimensions, geometry, materials, densities and positions), was elaborated a base model of the reactor core, for the MCNP5 code, with a configuration composed by 85 fuel elements, 4 control bars and the corresponding structural elements. The high reactivity excess obtained with this model, gave the rule to realize other models of the reactor core in which the reactivity excess and the switch out margin were approximate to the values established in the technical specifications of the reactor operation. Several models were realized until finding the satisfactory model; this is composite for 74 fuels, 4 control bars and 6 additional experimental positions inside the reactor core. (Author)

  15. Developing an interface between MCNP and McStas for simulation of neutron moderators

    DEFF Research Database (Denmark)

    Klinkby, Esben Bryndt; Lauritzen, Bent; Nonbøl, Erik;

    2012-01-01

    Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using MCNP/X whereas simulations of neutron transport and instrument performance are carried out by neutron ray tracing codes such as McStas. The coupling between the two simulations suites...... typically consists of providing analytical fits from MCNP/X neutron spectra to McStas. This method is generally successful, but as will be discussed in the this paper, there are limitations and a more direct coupling between MCNP/X andMcStas could allow for more accurate simulations of e.g. complex...... moderator geometries, interference between beamlines as well as shielding requirements along the neutron guides. In this paper different possible interfaces between McStas and MCNP/X are discussed and first preliminary performance results are shown....

  16. Current status of MCNP6 as a simulation tool useful for space and accelerator applications

    CERN Document Server

    Mashnik, S G; Hughes, H G; Prael, R E; Sierk, A J

    2012-01-01

    For the past several years, a major effort has been undertaken at Los Alamos National Laboratory (LANL) to develop the transport code MCNP6, the latest LANL Monte-Carlo transport code representing a merger and improvement of MCNP5 and MCNPX. We emphasize a description of the latest developments of MCNP6 at higher energies to improve its reliability in calculating rare-isotope production, high-energy cumulative particle production, and a gamut of reactions important for space-radiation shielding, cosmic-ray propagation, and accelerator applications. We present several examples of validation and verification of MCNP6 compared to a wide variety of intermediate- and high-energy experimental data on reactions induced by photons, mesons, nucleons, and nuclei at energies from tens of MeV to about 1 TeV/nucleon, and compare to results from other modern simulation tools.

  17. A Patch to MCNP5 for Multiplication Inference: Description and User Guide

    Energy Technology Data Exchange (ETDEWEB)

    Solomon, Jr., Clell J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-05-05

    A patch to MCNP5 has been written to allow generation of multiple neutrons from a spontaneous-fission event and generate list-mode output. This report documents the implementation and usage of this patch.

  18. NEPHTIS: 2D/3D validation elements using MCNP4c and TRIPOLI4 Monte-Carlo codes

    Energy Technology Data Exchange (ETDEWEB)

    Courau, T.; Girardi, E. [EDF R and D/SINETICS, 1av du General de Gaulle, F92141 Clamart CEDEX (France); Damian, F.; Moiron-Groizard, M. [DEN/DM2S/SERMA/LCA, CEA Saclay, F91191 Gif-sur-Yvette CEDEX (France)

    2006-07-01

    High Temperature Reactors (HTRs) appear as a promising concept for the next generation of nuclear power applications. The CEA, in collaboration with AREVA-NP and EDF, is developing a core modeling tool dedicated to the prismatic block-type reactor. NEPHTIS (Neutronics Process for HTR Innovating System) is a deterministic codes system based on a standard two-steps Transport-Diffusion approach (APOLLO2/CRONOS2). Validation of such deterministic schemes usually relies on Monte-Carlo (MC) codes used as a reference. However, when dealing with large HTR cores the fission source stabilization is rather poor with MC codes. In spite of this, it is shown in this paper that MC simulations may be used as a reference for a wide range of configurations. The first part of the paper is devoted to 2D and 3D MC calculations of a HTR core with control devices. Comparisons between MCNP4c and TRIPOLI4 MC codes are performed and show very consistent results. Finally, the last part of the paper is devoted to the code to code validation of the NEPHTIS deterministic scheme. (authors)

  19. Dosimetric characterization of a brachytherapy applicator using MCNP5 modelisation and in-phantom measurements.

    Science.gov (United States)

    Gerardy, I; Ródenas, J; van Dycke, M; Gallardo, S; Ceccolini, Elisa

    2010-01-01

    A gynaecological applicator consisting of a metallic intra-uterine tube with a plastic vaginal applicator and an HDR Ir-192 source have been simulated with MCNP5 (Monte Carlo code). A solid phantom has been designed to perform measurements around the applicator with radiochromic films. The isodose curves obtained are compared with curves calculated with the F4MESH tally of MCNP5 with a good agreement. A pinpoint ionization chamber has been used to evaluate dose at some reference points.

  20. Developing an interface between MCNP and McStas for simulation of neutron moderators

    OpenAIRE

    Klinkby, Esben Bryndt; Lauritzen, Bent; Nonbøl, Erik; Willendrup, Peter Kjær

    2012-01-01

    Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using MCNP/X whereas simulations of neutron transport and instrument performance are carried out by neutron ray tracing codes such as McStas. The coupling between the two simulations suites typically consists of providing analytical fits from MCNP/X neutron spectra to McStas. This method is generally successful, but as will be discussed in the this paper, there are limitations and a more dire...

  1. A comparison between the Monte Carlo radiation transport codes MCNP and MCBEND

    Energy Technology Data Exchange (ETDEWEB)

    Sawamura, Hidenori; Nishimura, Kazuya [Computer Software Development Co., Ltd., Tokyo (Japan)

    2001-01-01

    In Japan, almost of all radiation analysts are using the MCNP code and MVP code on there studies. But these codes have not had automatic variance reduction. MCBEND code made by UKAEA have automatic variance reduction. And, MCBEND code is user friendly more than other Monte Carlo Radiation Transport Codes. Our company was first introduced MCBEND code in Japan. Therefore, we compared with MCBEND code and MCNP code about functions and production capacity. (author)

  2. MCNP6 Simulation of Light and Medium Nuclei Fragmentation at Intermediate Energies

    Energy Technology Data Exchange (ETDEWEB)

    Mashnik, Stepan Georgievich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kerby, Leslie Marie [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-05-22

    MCNP6, the latest and most advanced LANL Monte Carlo transport code, representing a merger of MCNP5 and MCNPX, is actually much more than the sum of those two computer codes; MCNP6 is available to the public via RSICC at Oak Ridge, TN, USA. In the present work, MCNP6 was validated and verified (V&V) against different experimental data on intermediate-energy fragmentation reactions, and results by several other codes, using mainly the latest modifications of the Cascade-Exciton Model (CEM) and of the Los Alamos version of the Quark-Gluon String Model (LAQGSM) event generators CEM03.03 and LAQGSM03.03. It was found that MCNP6 using CEM03.03 and LAQGSM03.03 describes well fragmentation reactions induced on light and medium target nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below, and can serve as a reliable simulation tool for different applications, like cosmic-ray-induced single event upsets (SEU’s), radiation protection, and cancer therapy with proton and ion beams, to name just a few. Future improvements of the predicting capabilities of MCNP6 for such reactions are possible, and are discussed in this work.

  3. Validation and verification of MCNP6 as a new simulation tool useful for medical applications

    Energy Technology Data Exchange (ETDEWEB)

    Mashnik, Stepan G [Los Alamos National Laboratory

    2011-01-06

    MCNP6, the latest and most advanced LANL transport code, representing a merger of MCNP5 and MCNPX has been Validated and Verified (V&V) against different experimental data and results by other codes relevant to medical applications. In the present work, we V&V MCNP6 using mainly the latest modifications of the Cascade-Exciton Model (CEM) and of the Los Alamos version of the Quark-Gluon String Model (LAQGSM) event generators CEM03.02 and LAQGSM03.03. We found that MCNP6 describes well data of interest for medical applications measured on both thin and thick targets and agrees very well with similar results obtained with other codes; MCNP6 may be a very useful tool for medical applications We plan to make MCNP6 available to the public via RSICC at Oak Ridge in the middle of 2011 but we are allowed to provide it to friendly US Beta-users outside LANL already now.

  4. PTRANSP Tests Of TGLF And Predictions For ITER

    Energy Technology Data Exchange (ETDEWEB)

    Robert V. Budny, Xingqiu Yuan, S. Jardin, G. Hammett, G. Staebler, members of the ITPA Transport and Confinement Topical Group, and JET EFDA Contributions

    2012-02-28

    One of the physics goals for ITER is to achieve high fusion power PDT at a high gain QDT. This goal is important for studying the physics of reactor-relevant burning plasmas. Simulations of plasma performance in ITER can help achieve this goal by aiding in the design of systems such as diagnostics and in planning ITER plasma regimes. Simulations can indicate areas where further research in theory and experiments is needed. To have credible simulations integrated modeling is necessary since plasma profiles and applied heating, torque, and current drive are strongly coupled.

  5. Final Report on ITER Task Agreement 81-10

    Energy Technology Data Exchange (ETDEWEB)

    Brad J. Merrill

    2009-01-01

    An International Thermonuclear Experimental Reactor (ITER) Implementing Task Agreement (ITA) on Magnet Safety was established between the ITER International Organization (IO) and the Idaho National Laboratory (INL) Fusion Safety Program (FSP) during calendar year 2004. The objectives of this ITA were to add new capabilities to the MAGARC code and to use this updated version of MAGARC to analyze unmitigated superconductor quench events for both poloidal field (PF) and toroidal field (TF) coils of the ITER design. This report documents the completion of the work scope for this ITA. Based on the results obtained for this ITA, an unmitigated quench event in an ITER larger PF coil does not appear to be as severe an accident as in an ITER TF coil.

  6. IER-163 Post-Experiment MCNP Calculations (U)

    Energy Technology Data Exchange (ETDEWEB)

    Favorite, Jeffrey A. [Los Alamos National Laboratory

    2012-06-04

    IER-163 has been modeled with high fidelity in MCNP6. The model k{sub eff} was high, as in other similar calculations. The fission ratio {sup 238}U(n,f)/{sup 235}U(n,f) was 12.6% too small compared with measurements; the ratio {sup 239}Pu(n,f)/{sup 235}U(n,f) was 11.5% too small compared with measurements; the iridium ratio {sup 193}Ir(n,n{prime})/{sup 191}Ir(n,{gamma}) was 16.4% too large; and the gold ratios {sup 197}Au(n,2n)/{sup 197}Au(n,{gamma}), {sup 197}Au(n,2n)/{sup 235}U(n,f), and {sup 197}Au(n,{gamma})/{sup 235}U(n,f) were within one standard deviation of the measured values. It is suggested that the calculated {sup 235}U fission rate is too large and the calculated {sup 238}U fission rate is too small.

  7. Extension and Test of Limits on MCNP Geometry Description%MCNP程序几何描述能力扩展及应用测试

    Institute of Scientific and Technical Information of China (English)

    刘镇洲; 李刚; 邓力; 柴晓明

    2013-01-01

    为使MCNP程序能模拟数百万规模的反应堆“pin-by-pin”问题和医学体素模型,本文对MCNP程序进行了改进,使几何块、几何面数量可扩展.改进后的程序对硼中子俘获治疗(BNCT)的人体大脑进行几何建模,栅元数量达百万量级;计算了大脑的中子、光子吸收剂量率随深度的变化,为大脑BNCT提供理论支持.此外,对百万规模的“Like n But”重复结构模型进行了串、并行测试,验证了几何规模扩展后程序计算的正确性.%To set up models for 'pin-by-pin' facilities (reactors) and human phantoms with millions of cells by using the MCNP code, the MCNP code was modified and limits on the number of cell, surface, material, etc. were extended. The Snyder head phantom for boron neutron capture therapy was modeled. Neutron and induced gamma-ray absorbed dose in human brain was calculated. Besides, a critical calculation model of repeated structures described by the 'Like n But' cards was calculated employing both series and parallel computing. As a conclusion, modifications of the MCNP code were proved to be correct.

  8. Approximate iterative algorithms

    CERN Document Server

    Almudevar, Anthony Louis

    2014-01-01

    Iterative algorithms often rely on approximate evaluation techniques, which may include statistical estimation, computer simulation or functional approximation. This volume presents methods for the study of approximate iterative algorithms, providing tools for the derivation of error bounds and convergence rates, and for the optimal design of such algorithms. Techniques of functional analysis are used to derive analytical relationships between approximation methods and convergence properties for general classes of algorithms. This work provides the necessary background in functional analysis a

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

    Science.gov (United States)

    Castanier, Eric; Paterne, Loic; Louis, Céline

    2017-09-01

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

  10. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  11. Neutron Measurement Instrumentation Development at KIT for the European ITER TBM

    Energy Technology Data Exchange (ETDEWEB)

    Klix, A.; Fischer, U.; Raj, P.; Reimann, Th.; Szalkai, D.; Tian, K. [Association KIT-EURATOM, Karlsruhe Institute of Technology, D-76344 Eggenstein-Leopoldshafen (Germany); Angelone, M. [Associazione ENEA-EURATOM sulla Fusione, ENEA C.R., I-00044 Frascati (Italy); Gehre, D. [Technical University of Dresden, D-01069 Dresden (Germany); Lyoussi, A. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France)

    2015-07-01

    Fusion power reactors will rely on the internal production of the fuel tritium from lithium in the tritium breeding blanket. Test Blanket Modules (TBM) will be installed in ITER with the aim to investigate the nuclear performance of different breeding blanket designs. Currently there is no fully qualified nuclear instrumentation available for the measurement of neutron fluxes and tritium production rates which would be able to withstand the harsh environment conditions in the TBM such as high temperature (>400 deg. C) and, depending on the operation scenario, intense radiation levels. As partner of the European Consortium on Nuclear Data and Measurement Techniques in the framework of several F4E specific grants and contracts, KIT and ENEA have jointly studied the possibility to develop and test detectors suitable to operate in ITER-TBMs. Here we present an overview of ongoing work on three types of neutron flux monitors under development for the TBMs with focus on the KIT activities. A neutron activation system (NAS) with pneumatic sample transport could provide absolute neutron flux measurements in selected positions. A test system for investigating activation materials with short half-lives was constructed at the DT neutron generator laboratory of Technical University of Dresden to investigate the neutronics aspects. Several irradiations have been performed with focus on the simultaneous measurement of the extracted activated probes. An engineering assessment of a TBM NAS in the conceptual design phase has been done which considered issues of design requirements and integration. Last but not least, a mechanical test bench is under construction at KIT which will address issues of driving the activation probes, solutions for loading the system etc. experimentally. Self-powered neutron detectors (SPND) are widely applied in fission reactor monitoring, and the commercially available SPNDs are sensitive to thermal neutrons. We are investigating novel materials for

  12. Design of In-vessel neutron monitor using micro fission chambers for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Nishitani, Takeo; Kasai, Satoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ebisawa, Katsuyuki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Walker, Chris [ITER Joint Central Team, Garching (Germany)

    2001-10-01

    A neutron monitor using micro fission chambers to be installed inside the vacuum vessel has been designed for compact ITER (ITER-FEAT). We investigated the responses of the micro fission chambers to find the suitable position of micro fission chambers by a neutron Monte Carlo calculation using MCNP version 4b code. It was found that the averaged output of the micro fission chambers behind blankets at upper outboard and lower outboard is insensitive to the changes in the plasma position and the neutron source profile. A set of {sup 235}U micro fission chamber and ''blank'' detector which is a fissile material free detector to identify noise issues such as from {gamma}-rays are installed behind blankets. Employing both pulse counting mode and Campbelling mode in the electronics, the ITER requirement of 10{sup 7} dynamic range with 1 ms temporal resolution can be accomplished. The in-situ calibration has been simulated by MCNP calculation, where a point source of 14 MeV neutrons is moving on the plasma axis. It was found that the direct calibration is possible by using a neutron generator with an intensity of 10{sup 11} n/s. The micro fission chamber system can meet the required 10% accuracy for a fusion power monitor. (author)

  13. Monte Carlo Calculation of Core Reactivity and Fluxes for the Development of the BNCT Neutron Source at the Kyiv Research Reactor

    Science.gov (United States)

    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.

  14. Detailed 3-D nuclear analysis of ITER outboard blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Bohm, Tim, E-mail: tdbohm@wisc.edu [Fusion Technology Institute, University of Wisconsin-Madison, Madison, WI (United States); Davis, Andrew; Sawan, Mohamed; Marriott, Edward; Wilson, Paul [Fusion Technology Institute, University of Wisconsin-Madison, Madison, WI (United States); Ulrickson, Michael; Bullock, James [Formerly, Fusion Technology, Sandia National Laboratories, Albuquerque, NM (United States)

    2015-10-15

    Highlights: • Nuclear analysis was performed on detailed CAD models placed in a 40 degree model of ITER. • The regions examined include BM09, the upper ELM coil region (BM11–13), the neutral beam (NB) region (BM13–16), and BM18. • The results show that VV nuclear heating exceeds limits in the NB and upper ELM coil regions. • The results also show that the level of He production in parts of BM18 exceeds limits. • These calculations are being used to modify the design of the ITER blanket modules. - Abstract: In the ITER design, the blanket modules (BM) provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40 degree partially homogenized ITER global model. The regions analyzed include BM09, BM16 near the heating neutral beam injection (HNB) region, BM11–13 near the upper ELM coil region, and BM18. For the BM16 HNB region, the VV nuclear heating behind the NB region exceeds the design limit by up to 80%. For the BM11–13 region, the nuclear heating of the VV exceeds the design limit by up to 45%. For BM18, the results show that He production does not meet the limit necessary for re-welding. The results presented in this work are being used by the ITER Organization Blanket and Tokamak Integration groups to modify the BM design in the cases where limits are exceeded.

  15. Fusion neutron diagnostics on ITER tokamak

    Science.gov (United States)

    Bertalot, L.; Barnsley, R.; Direz, M. F.; Drevon, J. M.; Encheva, A.; Jakhar, S.; Kashchuk, Y.; Patel, K. M.; Arumugam, A. P.; Udintsev, V.; Walker, C.; Walsh, M.

    2012-04-01

    ITER is an experimental nuclear reactor, aiming to demonstrate the feasibility of nuclear fusion realization in order to use it as a new source of energy. ITER is a plasma device (tokamak type) which will be equipped with a set of plasma diagnostic tools to satisfy three key requirements: machine protection, plasma control and physics studies by measuring about 100 different parameters. ITER diagnostic equipment is integrated in several ports at upper, equatorial and divertor levels as well internally in many vacuum vessel locations. The Diagnostic Systems will be procured from ITER Members (Japan, Russia, India, United States, Japan, Korea and European Union) mainly with the supporting structures in the ports. The various diagnostics will be challenged by high nuclear radiation and electromagnetic fields as well by severe environmental conditions (ultra high vacuum, high thermal loads). Several neutron systems with different sensitivities are foreseen to measure ITER expected neutron emission from 1014 up to almost 1021 n/s. The measurement of total neutron emissivity is performed by means of Neutron Flux Monitors (NFM) installed in diagnostic ports and by Divertor Neutron Flux Monitors (DNFM) plus MicroFission Chambers (MFC) located inside the vacuum vessel. The neutron emission profile is measured with radial and vertical neutron cameras. Spectroscopy is accomplished with spectrometers looking particularly at 2.5 and 14 MeV neutron energy. Neutron Activation System (NAS), with irradiation ends inside the vacuum vessel, provide neutron yield data. A calibration strategy of the neutron diagnostics has been developed foreseeing in situ and cross calibration campaigns. An overview of ITER neutron diagnostic systems and of the associated challenging engineering and integration issues will be reported.

  16. Modification to the Monte Carlo N-Particle (MCNP) Visual Editor (MCNPVised) to Read in Computer Aided Design (CAD) Files

    Energy Technology Data Exchange (ETDEWEB)

    Randolph Schwarz; Leland L. Carter; Alysia Schwarz

    2005-08-23

    Monte Carlo N-Particle Transport Code (MCNP) is the code of choice for doing complex neutron/photon/electron transport calculations for the nuclear industry and research institutions. The Visual Editor for Monte Carlo N-Particle is internationally recognized as the best code for visually creating and graphically displaying input files for MCNP. The work performed in this grant was used to enhance the capabilities of the MCNP Visual Editor to allow it to read in both 2D and 3D Computer Aided Design (CAD) files, allowing the user to electronically generate a valid MCNP input geometry.

  17. Modification to the Monte Carlo N-Particle (MCNP) Visual Editor (MCNPVised) to Read in Computer Aided Design (CAD) Files

    Energy Technology Data Exchange (ETDEWEB)

    Randolph Schwarz; Leland L. Carter; Alysia Schwarz

    2005-08-23

    Monte Carlo N-Particle Transport Code (MCNP) is the code of choice for doing complex neutron/photon/electron transport calculations for the nuclear industry and research institutions. The Visual Editor for Monte Carlo N-Particle is internationally recognized as the best code for visually creating and graphically displaying input files for MCNP. The work performed in this grant was used to enhance the capabilities of the MCNP Visual Editor to allow it to read in both 2D and 3D Computer Aided Design (CAD) files, allowing the user to electronically generate a valid MCNP input geometry.

  18. Influence of the magnetic field profile on ITER conductor testing

    NARCIS (Netherlands)

    Nijhuis, A.; Ilyin, Y.; Kate, ten H.H.J.

    2006-01-01

    We performed simulations with the numerical CUDI-CICC code on a typical short ITER (International Thermonuclear Experimental Reactor) conductor test sample of dual leg configuration, as usually tested in the SULTAN test facility, and made a comparison with the new EFDA-Dipole test facility offering

  19. An iterative method for determination of a minimal eigenvalue

    DEFF Research Database (Denmark)

    Kristiansen, G.K.

    1968-01-01

    Kristiansen (1963) has discussed the convergence of a group of iterative methods (denoted the Equipoise methods) for the solution of reactor criticality problems. The main result was that even though the methods are said to work satisfactorily in all practical cases, examples of divergence can...

  20. Conceptual design of a clinical BNCT beam in an adjacent dry cell of the Jozef Stefan Institute TRIGA reactor

    NARCIS (Netherlands)

    Maucec, M

    2000-01-01

    The MCNP4B Monte Carlo transport code is used in a feasibility study of the epithermal neutron boron neutron capture therapy facility in the thermalizing column of the 250-kW TRIGA Mark II reactor at the Jozef Stefan Institute (JSI). To boost the epithermal neutron flux at the reference irradiation

  1. Conceptual design of a clinical BNCT beam in an adjacent dry cell of the Jozef Stefan Institute TRIGA reactor

    NARCIS (Netherlands)

    Maucec, M

    2000-01-01

    The MCNP4B Monte Carlo transport code is used in a feasibility study of the epithermal neutron boron neutron capture therapy facility in the thermalizing column of the 250-kW TRIGA Mark II reactor at the Jozef Stefan Institute (JSI). To boost the epithermal neutron flux at the reference irradiation

  2. Development of structural design criteria for ITER.

    Energy Technology Data Exchange (ETDEWEB)

    Majumdar, S.

    1998-06-22

    The irradiation environment experienced by the in-vessel components of fusion reactors such as HER presents structural design challenges not envisioned in the development of existing structural design criteria such as the ASME Code or RCC-MR. From the standpoint of design criteria, the most significant issues stem from the irradiation-induced changes in material properties, specifically the reduction of ductility, strain hardening capability, and fracture toughness with neutron irradiation. Recently, Draft 7 of the interim ITER structural design criteria (ISDC), which provide new rules for guarding against such problems, was released for trial use by the ITER designers. The new rules, which were derived from a simple model based on the concept of elastic follow up factor, provide primary and secondary stress limits as functions of uniform elongation and ductility. The implication of these rules on the allowable surface heat flux on typical first walls made of type 316 stainless steel and vanadium alloys are discussed.

  3. ITER Safety Analyses with ISAS

    Science.gov (United States)

    Gulden, W.; Nisan, S.; Porfiri, M.-T.; Toumi, I.; de Gramont, T. Boubée

    1997-06-01

    Detailed analyses of accident sequences for the International Thermonuclear Experimental Reactor (ITER), from an initiating event to the environmental release of activity, have involved in the past the use of different types of computer codes in a sequential manner. Since these codes were developed at different time scales in different countries, there is no common computing structure to enable automatic data transfer from one code to the other, and no possibility exists to model or to quantify the effect of coupled physical phenomena. To solve this problem, the Integrated Safety Analysis System of codes (ISAS) is being developed, which allows users to integrate existing computer codes in a coherent manner. This approach is based on the utilization of a command language (GIBIANE) acting as a “glue” to integrate the various codes as modules of a common environment. The present version of ISAS allows comprehensive (coupled) calculations of a chain of codes such as ATHENA (thermal-hydraulic analysis of transients and accidents), INTRA (analysis of in-vessel chemical reactions, pressure built-up, and distribution of reaction products inside the vacuum vessel and adjacent rooms), and NAUA (transport of radiological species within buildings and to the environment). In the near future, the integration of S AFALY (simultaneous analysis of plasma dynamics and thermal behavior of in-vessel components) is also foreseen. The paper briefly describes the essential features of ISAS development and the associated software architecture. It gives first results of a typical ITER accident sequence, a loss of coolant accident (LOCA) in the divertor cooling loop inside the vacuum vessel, amply demonstrating ISAS capabilities.

  4. The neutron standard fields at the BR1 reactor at SCK.CEN

    Energy Technology Data Exchange (ETDEWEB)

    Wagemans, J.; Malambu, E.; Borms, L. [SCKCEN, Boeretang 200, 2400 Mol (Belgium)

    2011-07-01

    The BR1 research reactor at SCK-CEN is characterized by a wide variety of irradiation possibilities, a large reactor core, and strong flexibility in its operation. A full MCNP model of BR1 has been recently developed in order to complement the results that can be obtained from activation dosimetry. After a general presentation of the reactor, this paper pays particular attention to its standard {sup 235}U(n,f) fast neutron field MARK III. This irradiation field is a useful tool for integral measurements and for detector calibrations. With the support of MCNP calculations, irradiations in MARK III can be directly referred to the pure {sup 235}U(n,f) fast neutron spectrum. (authors)

  5. Methods for modeling impact-induced reactivity changes in small reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Tallman, Tyler N.; Radel, Tracy E.; Smith, Jeffrey A.; Villa, Daniel L.; Smith, Brandon M. (U. of Wisconsin, Madison, WI); Radel, Ross F.; Lipinski, Ronald J.; Wilson, Paul Philip Hood (U. of Wisconsin, Madison, WI)

    2010-10-01

    This paper describes techniques for determining impact deformation and the subsequent reactivity change for a space reactor impacting the ground following a potential launch accident or for large fuel bundles in a shipping container following an accident. This technique could be used to determine the margin of subcriticality for such potential accidents. Specifically, the approach couples a finite element continuum mechanics model (Pronto3D or Presto) with a neutronics code (MCNP). DAGMC, developed at the University of Wisconsin-Madison, is used to enable MCNP geometric queries to be performed using Pronto3D output. This paper summarizes what has been done historically for reactor launch analysis, describes the impact criticality analysis methodology, and presents preliminary results using representative reactor designs.

  6. Robust iterative methods

    Energy Technology Data Exchange (ETDEWEB)

    Saadd, Y.

    1994-12-31

    In spite of the tremendous progress achieved in recent years in the general area of iterative solution techniques, there are still a few obstacles to the acceptance of iterative methods in a number of applications. These applications give rise to very indefinite or highly ill-conditioned non Hermitian matrices. Trying to solve these systems with the simple-minded standard preconditioned Krylov subspace methods can be a frustrating experience. With the mathematical and physical models becoming more sophisticated, the typical linear systems which we encounter today are far more difficult to solve than those of just a few years ago. This trend is likely to accentuate. This workshop will discuss (1) these applications and the types of problems that they give rise to; and (2) recent progress in solving these problems with iterative methods. The workshop will end with a hopefully stimulating panel discussion with the speakers.

  7. 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.)

  8. Quantum Iterated Function Systems

    CERN Document Server

    Lozinski, A; Slomczynski, W; Lozinski, Artur; Zyczkowski, Karol; Slomczynski, Wojciech

    2003-01-01

    Iterated functions system (IFS) is defined by specifying a set of functions in a classical phase space, which act randomly on the initial point. In an analogous way, we define quantum iterated functions system (QIFS), where functions act randomly with prescribed probabilities in the Hilbert space. In a more general setting a QIFS consists of completely positive maps acting in the space of density operators. We present exemplary classical IFSs, the invariant measure of which exhibits fractal structure, and study properties of the corresponding QIFSs and their invariant state.

  9. ITER1O kA高温超导电流引线测试装置低温系统的研究%Cryogenic System Development for Test Facility of 10 kA High-Temperature-Superconductor Current Lead in International Thermonuclear Experimental Reactor

    Institute of Scientific and Technical Information of China (English)

    倪清; 毕延芳; 丁开忠; 冯汉升; 周挺志; 沈光; 刘承连; 黄雄一; 宋云涛

    2011-01-01

    The cryogenic system,dedicated to the testing facility of the 10 Ka high-temperature-superconductor current lead (HTS-CL) in international thermonuclear experimental reactor,has been successfully developed. The cryogenic system includes a 500W/4.5K helium refrigerator, a vacuum dewar, cryogenic modules (cryogenic control valves, sub-cooler, electrical heaters, and thermal shields) ,carburetors,a set of cryogenic pipelines,and control unit. The discussions focused on the technical requirements and the design considerations: such as the design of vacuum Dewar and the sub-cooler, and the scheme of the cooling circuit. The field-test results of the newly-developed cryogenic system show that it is capable of doing a good job.%为ITER CC 10 kA高温超导电流引线服务的低温性能测试装置已研制完成,并成功运行.其低温系统主要由500W/4.5 K氦制冷机,真空杜瓦,低温组件(低温阀门,过冷槽,管道加热器,热防护层),汽化器及低温传输管线等部分组成.本文对真空杜瓦和过冷槽进行设计,并讨论该低温系统的冷却流程方案,最后通过电流引线10 kA稳态实验结果对低温系统的运行效果进行分析,结果表明该低温系统运行稳定,能满足ITER CC电流引线的测试需要.

  10. MCNP6 Simulation of Reactions of Interest to FRIB, Medical, and Space Applications

    CERN Document Server

    Mashnik, Stepan G

    2014-01-01

    The latest, production, version of the Los Alamos Monte Carlo N-Particle transport code MCNP6 has been used to simulate a variety of particle-nucleus and nucleus-nucleus reactions of academic and applied interest to the Facility for Rare Isotope Beams (FRIB), medical isotope production, space-radiation shielding, cosmic-ray propagation, and accelerator applications, including several reactions induced by radioactive isotopes, analyzing production of both stable and radioactive residual nuclei. Here, we discuss examples of validation and verification of MCNP6 compared to recent neutron spectra measured at the Heavy Ion Medical Accelerator in Chiba, Japan; to spectra of light fragments from several reactions measured recently at GANIL, France; INFN Laboratori Nazionali del Sud, Catania, Italy; COSY of the Julich Research Center, Germany; and to cross sections of products from several reactions measured lately at GSI, Darmstadt, Germany; ITEP, Moscow, Russia; LANSCE, LANL, Los Alamos, USA. As a rule, MCNP6 provi...

  11. Simulation of Photon energy Spectra Using MISC, SOURCES, MCNP and GADRAS

    Energy Technology Data Exchange (ETDEWEB)

    Tucker, Lucas P. [Los Alamos National Laboratory; Shores, Erik F. [Los Alamos National Laboratory; Myers, Steven C. [Los Alamos National Laboratory; Felsher, Paul D. [Los Alamos National Laboratory; Garner, Scott E. [Los Alamos National Laboratory; Solomon, Clell J. Jr. [Los Alamos National Laboratory

    2012-08-14

    The detector response functions included in the Gamma Detector Response and Analysis Software (GADRAS) are a valuable resource for simulating radioactive source emission spectra. Application of these response functions to the results of three-dimensional transport calculations is a useful modeling capability. Using a 26.2 kg shell of depleted uranium (DU) as a simple test problem, this work illustrates a method for manipulating current tally results from MCNP into the GAM file format necessary for a practical link to GADRAS detector response functions. MISC (MCNP Intrinsic Source Constructor) and SOURCES 4C were used to develop photon and neutron source terms for subsequent MCNP transport, and the resultant spectrum is shown to be in good agreement with that from GADRAS. A 1 kg DU sphere was also modeled with the method described here and showed similarly encouraging results.

  12. Antineutrino emission and gamma background characteristics from a thermal research reactor

    CERN Document Server

    Bui, V M; Fallot, M; Communeau, V; Cormon, S; Estienne, M; Lenoir, M; Peuvrel, N; Shiba, T; Cucoanes, A S; Elnimr, M; Martino, J; Onillon, A; Porta, A; Pronost, G; Remoto, A; Thiolliere, N; Yermia, F; Zakari-Issoufou, A -A

    2016-01-01

    The detailed understanding of the antineutrino emission from research reactors is mandatory for any high sensitivity experiments either for fundamental or applied neutrino physics, as well as a good control of the gamma and neutron backgrounds induced by the reactor operation. In this article, the antineutrino emission associated to a thermal research reactor: the OSIRIS reactor located in Saclay, France, is computed in a first part. The calculation is performed with the summation method, which sums all the contributions of the beta decay branches of the fission products, coupled for the first time with a complete core model of the OSIRIS reactor core. The MCNP Utility for Reactor Evolution code was used, allowing to take into account the contributions of all beta decayers in-core. This calculation is representative of the isotopic contributions to the antineutrino flux which can be found at research reactors with a standard 19.75\\% enrichment in $^{235}$U. In addition, the required off-equilibrium correction...

  13. Validation of MCNP NPP Activation Simulations for Decommissioning Studies by Analysis of NPP Neutron Activation Foil Measurement Campaigns

    Directory of Open Access Journals (Sweden)

    Volmert Ben

    2016-01-01

    Full Text Available In this paper, an overview of the Swiss Nuclear Power Plant (NPP activation methodology is presented and the work towards its validation by in-situ NPP foil irradiation campaigns is outlined. Nuclear Research and consultancy Group (NRG in The Netherlands has been given the task of performing the corresponding neutron metrology. For this purpose, small Aluminium boxes containing a set of circular-shaped neutron activation foils have been prepared. After being irradiated for one complete reactor cycle, the sets have been successfully retrieved, followed by gamma-spectrometric measurements of the individual foils at NRG. Along with the individual activities of the foils, the reaction rates and thermal, intermediate and fast neutron fluence rates at the foil locations have been determined. These determinations include appropriate corrections for gamma self-absorption and neutron self-shielding as well as corresponding measurement uncertainties. The comparison of the NPP Monte Carlo calculations with the results of the foil measurements is done by using an individual generic MCNP model functioning as an interface and allowing the simulation of individual foil activation by predetermined neutron spectra. To summarize, the comparison between calculation and measurement serve as a sound validation of the Swiss NPP activation methodology by demonstrating a satisfying agreement between measurement and calculation. Finally, the validation offers a chance for further improvements of the existing NPP models by ensuing calibration and/or modelling optimizations for key components and structures.

  14. Implementation of a tree algorithm in MCNP code for nuclear well logging applications.

    Science.gov (United States)

    Li, Fusheng; Han, Xiaogang

    2012-07-01

    The goal of this paper is to develop some modeling capabilities that are missing in the current MCNP code. Those missing capabilities can greatly help for some certain nuclear tools designs, such as a nuclear lithology/mineralogy spectroscopy tool. The new capabilities to be developed in this paper include the following: zone tally, neutron interaction tally, gamma rays index tally and enhanced pulse-height tally. The patched MCNP code also can be used to compute neutron slowing-down length and thermal neutron diffusion length.

  15. Estimation and interpretation of k{sub eff} confidence intervals in MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Urbatsch, T.J. [Univ. of Michigan, Ann Arbor, MI (United States). Dept. of Nuclear Engineering; Forster, R.A.; Prael, R.E.; Beckman, R.J. [Los Alamos National Lab., NM (United States)

    1995-11-01

    MCNP`s criticality methodology and some basic statistics are reviewed. Confidence intervals are discussed, as well as how to build them and their importance in the presentation of a Monte Carlo result. The combination of MCNP`s three k{sub eff} estimators is shown, theoretically and empirically, by statistical studies and examples, to be the best k{sub eff} estimator. The method of combining estimators is based on a solid theoretical foundation, namely, the Gauss-Markov Theorem in regard to the least squares method. The confidence intervals of the combined estimator are also shown to have correct coverage rates for the examples considered.

  16. Iterative List Decoding

    DEFF Research Database (Denmark)

    Justesen, Jørn; Høholdt, Tom; Hjaltason, Johan

    2005-01-01

    We analyze the relation between iterative decoding and the extended parity check matrix. By considering a modified version of bit flipping, which produces a list of decoded words, we derive several relations between decodable error patterns and the parameters of the code. By developing a tree...... of codewords at minimal distance from the received vector, we also obtain new information about the code....

  17. Iterative software kernels

    Energy Technology Data Exchange (ETDEWEB)

    Duff, I.

    1994-12-31

    This workshop focuses on kernels for iterative software packages. Specifically, the three speakers discuss various aspects of sparse BLAS kernels. Their topics are: `Current status of user lever sparse BLAS`; Current status of the sparse BLAS toolkit`; and `Adding matrix-matrix and matrix-matrix-matrix multiply to the sparse BLAS toolkit`.

  18. Reactor Physics

    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.

  19. MCNP and other nuclear codes output graphical representation using python scripts; Representacion grafica de outputs de MCNP y codigos nucleares mediante el uso de scripts en python

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.

    2016-08-01

    Due to the lack of graphical representation capability of same nuclear codes like MCNP of GOTHIC, widely used in the industry, the following article describes the development of an interface to use a graphical representation open source (Paraview) with the outputs generated by the nuclear codes. Moreover, this article aims at describing the advantage of this type of visualization programs for the modeling and decision making in the calculation. (Author)

  20. Simulation of a nuclear densimeter using the Monte Carlo MCNP-4C code; Simulacao de um densimetro nuclear utilizando o codigo Monte Carlo MCNP-4C

    Energy Technology Data Exchange (ETDEWEB)

    Penna, Rodrigo [UNI-BH, Belo Horizonte, MG (Brazil). Dept. de Ciencias Biologicas, Ambientais e da Saude (DCBAS/DCET); Silva, Clemente Jose Gusmao Carneiro da [Universidade Estadual de Santa Cruz, UESC, Ilheus, BA (Brazil); Gomes, Paulo Mauricio Costa [Universidade FUMEC, Belo Horizonte, MG (Brazil)

    2008-07-01

    Viability of building a nuclear wood densimeter based on low energy photons Compton scattering was done using Monte Carlo code (MCNP- 4C). It is simulated a collimated 60 keV beam of gamma rays emitted by {sup 241}Am source reaching wood blocks. Backscattered radiation by these blocks was calculated. Photons scattered were correlated with blocks of different wood densities. Results showed a linear relationship on wood density and scattered photons, therefore the viability of this wood densimeter. (author)

  1. Nuclear analysis of the ITER Cryopump Ports

    Energy Technology Data Exchange (ETDEWEB)

    Moro, Fabio, E-mail: fabio.moro@enea.it [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Villari, Rosaria; Flammini, Davide [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Antipenkov, Alexander; Dremel, Matthias; Levesy, Bruno; Loughlin, Michael [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul-lez-Durance (France); Juarez, Rafael; Perez, Lucia [UNED, Energetic Engineering Department, C/Juan del Rosal 12, Madrid (Spain); Petrizzi, Luigino [European Commission, DG Research & Innovation G5, CDMA 00/030, B-1049 Brussels (Belgium)

    2015-10-15

    Highlights: • Evaluation the shielding effectiveness of the TCPHs by means of 3-D neutrons and gamma maps. • Assessment of the nuclear heating induced by neutron and photons on the TCP and TCPHs. • Calculation of the dose rate at 12 days after shutdown in the maintenance area of the Lower Ports with the Advanced D1S method, in order to verify the design target (100 μSv/h). • Potential improvements of the shielding configuration aimed at the reduction of the dose level in the Port Cell have been proposed and discussed. - Abstract: The ITER machine will be equipped with 6 torus Cryopumps (TCP) that are positioned in their housings (TCPH) and integrated into the cryostat walls at B1 level in the port cells. A comprehensive nuclear analysis of the Cryopump Ports #4 and #12 has been carried out by means of the MCNP-5 Monte Carlo code in a full 3-D geometry, providing guidelines for the design of the embedded components. Radiation transport calculations have been performed in order to determine the radiation field inside the Lower Ports, up the Port Cell: 3-D neutrons and gamma maps have been provided in order to evaluate the shielding effectiveness of the TCPHs. Nuclear heating induced by neutron and photons have been estimated on the TCP and TCPH to assess the nuclear loads during plasma operations. The shutdown dose rate in the maintenance area of the Lower Ports has been assessed with the Advanced D1S method to verify the design limits.

  2. Reactor safeguards

    CERN Document Server

    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

  3. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  4. Monte Carlo radiation shielding and activation analyses for the Diagnostic Equatorial Port Plug in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Serikov, A., E-mail: arkady.serikov@kit.edu [Karlsruhe Institute of Technology KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Fischer, U.; Leichtle, D. [Karlsruhe Institute of Technology KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Pitcher, C.S. [ITER Organization, Route de Vinon sur Verdon, 13115, St. Paul lez Durance (France)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer Systematic neutronics analyses were conducted to assess the ITER Equatorial Port Plug radiation shielding performance. Black-Right-Pointing-Pointer Shielding optimization was achieved by parametric analyses of several design variants using the MCNP5, FISPACT-2007, and R2Smesh codes. Black-Right-Pointing-Pointer Dominant effect of radiation streaming along the port plug gaps was recognized. Black-Right-Pointing-Pointer Combination of the gap labyrinths and streaming stoppers or rails reduces shutdown doses by 2 orders of magnitude. Black-Right-Pointing-Pointer Using the proposed shielding, the shutdown dose in the ITER port interspace is less than the personnel access limit of 100 {mu}Sv/h. - Abstract: This paper addresses neutronics aspects of the design development of the Diagnostic Generic Equatorial Port Plug (EPP) in ITER. To secure the personnel access at the EPP back-end interspace, parametric neutronics analyses of the EPP radiation environment have been performed and practical shielding solutions have been found. Radiation transport was performed with the Monte Carlo MCNP5 code. Activation calculations were conducted with the FISPACT-2007 inventory code. The R2Smesh approach was applied to couple transport and activation calculations. Newly created EPP local MCNP5 model was devised by extracting the EPP and adjacent blanket modules from the ITER Alite-4.1 model with proper modification of the EPP geometry in accordance with recent 3D CAD CATIA model. The EPP local model reproduces the EPP neutronically important features and allows investigation of the EPP neutronics effects in isolation from all other ITER components. Thorough EPP parametric analyses revealed dominant effect of gaps around EPP and several EPP design improvements were implemented as the outcomes of the analyses. Gap labyrinths and streaming stoppers inserted into the gaps were shown are capable to reduce the shutdown dose rate which is below the 100

  5. Iterative Algorithms for Nonexpansive Mappings

    Directory of Open Access Journals (Sweden)

    Yao Yonghong

    2008-01-01

    Full Text Available Abstract We suggest and analyze two new iterative algorithms for a nonexpansive mapping in Banach spaces. We prove that the proposed iterative algorithms converge strongly to some fixed point of .

  6. Safety analysis of the US dual coolant liquid lead lithium ITER test blanket module

    Science.gov (United States)

    Merrill, Brad; Reyes, Susana; Sawan, Mohamed; Wong, Clement

    2007-07-01

    The US is proposing a prototype of a dual coolant liquid lead-lithium DEMO blanket concept for testing in the International Thermonuclear Experimental Reactor (ITER) as an ITER test blanket module (TBM). Because safety considerations are an integral part of the design process to ensure that this TBM does not adversely impact the safety of ITER, a safety assessment has been conducted for this TBM and its ancillary systems as requested by the ITER project. Four events were selected by the ITER international team (IT) to address specific reactor safety concerns, such as vaccum vessel (VV) pressurization, confinement building pressure build-up, TBM decay heat removal capability, tritium and activation products release from the TBM system and hydrogen and heat production from chemical reactions. This paper summarizes the results of this safety assessment conducted with the MELCOR computer code.

  7. Detailed 3-D nuclear analysis of ITER blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Bohm, T.D., E-mail: tdbohm@wisc.edu [University of Wisconsin-Madison, Madison, WI (United States); Sawan, M.E.; Marriott, E.P.; Wilson, P.P.H. [University of Wisconsin-Madison, Madison, WI (United States); Ulrickson, M.; Bullock, J. [Sandia National Laboratories, Albuquerque, NM (United States)

    2014-10-15

    In ITER, the blanket modules (BM) are arranged around the plasma to provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. As a part of the BM design process, nuclear analysis is required to determine the level of nuclear heating, helium production, and radiation damage in the BM. Additionally, nuclear heating in the VV is also important for assessing the BM design. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40-degree partially homogenized ITER global model. The regions analyzed include BM01, the neutral beam injection (NB) region, and the upper port region. For BM01, the results show that He production meets the limit necessary for re-welding, and the VV heating behind BM01 is acceptable. For the NBI region, the VV nuclear heating behind the NB region exceeds the design limit by a factor of two. For the upper port region, the nuclear heating of the VV exceeds the design limit by up to 20%. The results presented in this work are being used to modify the BM design in the cases where limits are exceeded.

  8. ITER: Promises unkept ? (2/2)

    CERN Document Server

    CERN. Geneva

    2008-01-01

    Fusion power as the source of energy on Earth has been the dream of mankind ever since the principles were understood. ITER, the Latin word for “the way”, is the world’s largest Fusion device presently under construction in Cadarache, France. Supported by the People’s Republic of China, the European Atomic Energy Community, India, Japan, the Republic of Korea, the Russian Federation, and the United States of America, an international organization was founded after the signature of the Joint ITER Agreement in October of 2006. The goal is to build a Fusion reactor with a power amplification of 10, a total fusion power of 500 MW or more operating at extended burn times of 400-3000 seconds, with Deuterium and Tritium as its basic fuel. Following a short introduction into fusion science principles, the history of thermo nuclear fusion will be covered. Finally more recent construction projects around the world, their latest achievements and the path to ITER will be described. Technological and scientific c...

  9. ITER: Promises unkept ? (1/2)

    CERN Document Server

    CERN. Geneva

    2008-01-01

    Fusion power as the source of energy on Earth has been the dream of mankind ever since the principles were understood. ITER, the Latin word for “the way”, is the world’s largest Fusion device presently under construction in Cadarache, France. Supported by the People’s Republic of China, the European Atomic Energy Community, India, Japan, the Republic of Korea, the Russian Federation, and the United States of America, an international organization was founded after the signature of the Joint ITER Agreement in October of 2006. The goal is to build a Fusion reactor with a power amplification of 10, a total fusion power of 500 MW or more operating at extended burn times of 400-3000 seconds, with Deuterium and Tritium as its basic fuel. Following a short introduction into fusion science principles, the history of thermo nuclear fusion will be covered. Finally more recent construction projects around the world, their latest achievements and the path to ITER will be described. Technological and scientific c...

  10. Iterative supervirtual refraction interferometry

    KAUST Repository

    Al-Hagan, Ola

    2014-05-02

    In refraction tomography, the low signal-to-noise ratio (S/N) can be a major obstacle in picking the first-break arrivals at the far-offset receivers. To increase the S/N, we evaluated iterative supervirtual refraction interferometry (ISVI), which is an extension of the supervirtual refraction interferometry method. In this method, supervirtual traces are computed and then iteratively reused to generate supervirtual traces with a higher S/N. Our empirical results with both synthetic and field data revealed that ISVI can significantly boost up the S/N of far-offset traces. The drawback is that using refraction events from more than one refractor can introduce unacceptable artifacts into the final traveltime versus offset curve. This problem can be avoided by careful windowing of refraction events.

  11. Iterative participatory design

    DEFF Research Database (Denmark)

    Simonsen, Jesper; Hertzum, Morten

    2010-01-01

    iterative process of mutual learning by designers and domain experts (users), who aim to change the users’ work practices through the introduction of information systems. We provide an illustrative case example with an ethnographic study of clinicians experimenting with a new electronic patient record......The theoretical background in this chapter is information systems development in an organizational context. This includes theories from participatory design, human-computer interaction, and ethnographically inspired studies of work practices. The concept of design is defined as an experimental...... system, focussing on emergent and opportunity-based change enabled by appropriating the system into real work. The contribution to a general core of design research is a reconstruction of the iterative prototyping approach into a general model for sustained participatory design....

  12. Study of the neutronic activation of the stainless steel in a nuclear reactor; Estudios de la activacion neutronica del acero inoxidable en un reactor nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro Roche, I.; Rodenas Diago, J.; Marques, J. G.

    2013-07-01

    During operation of a nuclear reactor, various components can be activated by neutron reactions. The activity thus generated produces a dose that is a potential risk to workers and environment. Was simulated using the MCNP and CINDER'90 such activation codes on a piece of steel and the values obtained compared with experimental measurements. The equivalence of both methods is verified to calculate neutron activation and evolution of the dose rate with the cooling time.

  13. MCNP modelling of scintillation-detector gamma-ray spectra from natural radionuclides

    NARCIS (Netherlands)

    Hendriks, Peter; Maucec, M; de Meijer, RJ

    2002-01-01

    gamma-ray spectra of natural radionuclides are simulated for a BGO detector in a borehole geometry using the Monte Carlo code MCNP. All gamma-ray emissions of the decay of K-40 and the series of Th-232 and U-238 are used to describe the source. A procedure is proposed which excludes the time-consumi

  14. Using NJOY to Create MCNP ACE Files and Visualize Nuclear Data

    Energy Technology Data Exchange (ETDEWEB)

    Kahler, Albert Comstock [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-10-14

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

  15. Simulasi MCNP5 dalam Eksperimen Kritikalitas Larutan Plutonium Uranium Nitrat Dengan Reflektor Air dan Polyethelene

    Directory of Open Access Journals (Sweden)

    Dinan A.

    2011-12-01

    Full Text Available Banyak perangkat kritik dibangun untuk memenuhi kebutuhan studi fenomena kecelakaan kritikalitas pada larutan fisil di fasilitas daur bahan bakar nuklir. Salah satu diantaranya adalah perangkat kritik SCAMP. Di perangkat ini dikerjakan eksperimen kritikalitas menggunakan bejana silindris stainless steel berisi larutan plutonium uranium nitrat (Pu ditambah U nitrat. Sebanyak 7 eksperimen didemonstrasikan dengan reflektor air di semua sisi permukaan bejana larutan kecuali di bagian atas bejana. Makalah ini membahas simulasi transport Monte Carlo MCNP5 dalam eksperimen kritikalitas larutan Pu ditambah U nitrat dengan reflektor air dan polyethylene. Simulasi MCNP5 dengan pustaka ENDF/BVI memberikan hasil yang paling dekat dengan data eksperimen terutama pada kasus A untuk varian geometri 4. Dibandingkan pustaka ENDF/BV, perhitungan kritikalitas dengan pustaka ENDF/B-VI memberikan hasil lebih dekat dengan perhitungan MONK dimana bias perhitungannya kurang dari 0,44%, khususnya pada kasus A namun pada kasus B dan C simulasi MCNP5dengan pustaka ENDF/BV memberikan hasil dengan kecenderungan lebih baik dibandingkan pustaka ENDFB/VI dengan bias perhitungan kurang dari 2,67% dan kurang dari 1,13%. Secara keseluruhan dapat disimpulkan bahwa MCNP5 telah menunjukkan reliabilitasnya dalam simulasi kritikalitas larutan Pu ditambah U nitrat.

  16. Certification of MCNP Version 4A for WHC computer platforms. Revision 7

    Energy Technology Data Exchange (ETDEWEB)

    Carter, L.L.

    1995-05-03

    MCNP is a general-purpose Monte Carlo code that can be used for neutron, photon, or coupled neutron/photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces, and some special fourth-degree surfaces (elliptical tori).

  17. The Iterate Manual

    Science.gov (United States)

    1990-10-01

    is probably a bad idea. A better versica would use a temporary: (defmacro sum-of-squares (expr) (let ((temp ( gensym ))) ’(lot (,temp ,expr)) (sum...val ( gensym )) (tempi ( gensym )) (temp2 ( gensym )) (winner (or var iterate::*result-var*))) ’(progn (with ,max-val - nil) (with ,winner = nil) (cond ((null...the elements of a vector (disregards fill-pointer)" (let ((vect ( gensym )) (end ( gensym )) (index ( gensym ))) ’(progn (with ,vect - v) (with ,end = (array

  18. Iterative initial condition reconstruction

    Science.gov (United States)

    Schmittfull, Marcel; Baldauf, Tobias; Zaldarriaga, Matias

    2017-07-01

    Motivated by recent developments in perturbative calculations of the nonlinear evolution of large-scale structure, we present an iterative algorithm to reconstruct the initial conditions in a given volume starting from the dark matter distribution in real space. In our algorithm, objects are first moved back iteratively along estimated potential gradients, with a progressively reduced smoothing scale, until a nearly uniform catalog is obtained. The linear initial density is then estimated as the divergence of the cumulative displacement, with an optional second-order correction. This algorithm should undo nonlinear effects up to one-loop order, including the higher-order infrared resummation piece. We test the method using dark matter simulations in real space. At redshift z =0 , we find that after eight iterations the reconstructed density is more than 95% correlated with the initial density at k ≤0.35 h Mpc-1 . The reconstruction also reduces the power in the difference between reconstructed and initial fields by more than 2 orders of magnitude at k ≤0.2 h Mpc-1 , and it extends the range of scales where the full broadband shape of the power spectrum matches linear theory by a factor of 2-3. As a specific application, we consider measurements of the baryonic acoustic oscillation (BAO) scale that can be improved by reducing the degradation effects of large-scale flows. In our idealized dark matter simulations, the method improves the BAO signal-to-noise ratio by a factor of 2.7 at z =0 and by a factor of 2.5 at z =0.6 , improving standard BAO reconstruction by 70% at z =0 and 30% at z =0.6 , and matching the optimal BAO signal and signal-to-noise ratio of the linear density in the same volume. For BAO, the iterative nature of the reconstruction is the most important aspect.

  19. Reactor Neutrinos

    OpenAIRE

    Soo-Bong Kim; Thierry Lasserre; Yifang Wang

    2013-01-01

    We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very ...

  20. BOILING REACTORS

    Science.gov (United States)

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  1. MCNP5 CRITICALITY VALIDATION AND BIAS FOR INTERMEDIATE ENRICHED URANIUM SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    FINFROCK SH

    2009-12-10

    The purpose of this analysis is to validate the Monte Carlo N-Particle 5 (MCNP5) code Version 1.40 (LA-UR-03-1987, 2005) and its cross-section database for k-code calculations of intermediate enriched uranium systems on INTEL{reg_sign} processor based PC's running any version of the WINDOWS operating system. Configurations with intermediate enriched uranium were modeled with the moderator range of 39 {le} H/Fissile {le} 1438. See Table 2-1 for brief descriptions of selected cases and Table 3-1 for the range of applicability for this validation. A total of 167 input cases were evaluated including bare and reflected systems in a single body or arrays. The 167 cases were taken directly from the previous (Version 4C [Lan 2005]) validation database. Section 2.0 list data used to calculate k-effective (k{sub eff}) for the 167 experimental criticality benchmark cases using the MCNP5 code v1.40 and its cross section database. Appendix B lists the MCNP cross-section database entries validated for use in evaluating the intermediate enriched uranium systems for criticality safety. The dimensions and atom densities for the intermediate enriched uranium experiments were taken from NEA/NSC/DOC(95)03, September 2005, which will be referred to as the benchmark handbook throughout the report. For these input values, the experimental benchmark k{sub eff} is approximately 1.0. The MCNP validation computer runs ran to an accuracy of approximately {+-} 0.001. For the cases where the reported benchmark k{sub eff} was not equal to 1.0000 the MCNP calculational results were normalized. The difference between the MCNP validation computer runs and the experimentally measured k{sub eff} is the MCNP5 v1.40 bias. The USLSTATS code (ORNL 1998) was utilized to perform the statistical analysis and generate an acceptable maximum k{sub eff} limit for calculations of the intermediate enriched uranium type systems.

  2. Development of Prototype Neutron Flux Monitor for ITER

    Institute of Scientific and Technical Information of China (English)

    Yang Jinwei; Song Xianying; Zhang Wei; Li Xu; Lee Wenzhong; Wang Shiqing; Xiao Gongshan; Yang Bo; Lu Shuangtong

    2005-01-01

    The prototype neutron flux monitor consists of a high purity 235U fission chamberdetector and a "blank" detector, which is a fissile material free detector with the same dimensionas the fission chamber detector to identify noise issues such as noise coming from gamma rays. Themain parameters of the fission chamber assembly that have been measured in the laboratory areconfirmed to approach the technological level of the International Thermonuclear ExperimentalReactor (ITER) in the near future. This prototype neutron flux monitor will be further developedto become a neutron flux monitor suitable for the operation phase of D-D fusion on the ITER.

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

    Science.gov (United States)

    Jaradat, Safwan Qasim Mohammad

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

  4. Japanese nuclear fusion comes out of the shadows of ITER; Japanse kernfusie uit schaduw van ITER

    Energy Technology Data Exchange (ETDEWEB)

    Stroeks, R.

    2010-06-15

    The National Institute for Fusion Science (NIFS) is located in the Japanese province of Gifu. It is a bold and as of yet successful project that must result in a workable fusion reactor. Parallel to that Japan is involved in the international fusion project ITER, which is based on a fundamentally different concept that seems to overshadow NIFS somewhat. Both concepts are complementary and together they provide the required insights for commercial nuclear fusion in the long term. [Dutch] In de Japanse provincie Gifu staat het National Institute for Fusion Science (NIFS). Het is een gedurfd en vooralsnog succesvol project dat moet leiden tot een werkbare fusiereactor. Parallel doet Japan mee aan het internationale fusieproject ITER, dat gebaseerd is op een fundamenteel ander concept en dat NIFS een beetje in de schaduw lijkt te zetten. Beide concepten zijn complementair en leveren samen het noodzakelijk inzicht voor commerciele kernfusie op lange termijn.

  5. Membrane reactor. Membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shindo, Y.; Wakabayashi, K. (National Chemical Laboratory for Industry, Tsukuba (Japan))

    1990-08-05

    Many reaction examples were introduced of membrane reactor, to be on the point of forming a new region in the field of chemical technology. It is a reactor to exhibit excellent function, by its being installed with membrane therein, and is generally classified into catalyst function type and reaction promotion type. What firstly belongs to the former is stabilized zirconia, where oxygen, supplied to the cathodic side of membrane with voltage, impressed thereon, becomes O {sup 2 {minus}} to be diffused through the membrane and supplied, as variously activated oxygenous species, on the anodic side. Examples with many advantages can be given such as methane coupling, propylene oxidation, methanating reaction of carbon dioxide, etc. Apart, palladium film and naphion film also belong to the former. While examples of the latter comprise, among others, decomposition of hydrogen sulfide by porous glass film and dehydrogenation of cyclohexane or palladium alloy film, which are expected to be developed and materialized in the industry. 33 refs., 8 figs.

  6. Calculation of the reactor neutron time of flight spectrum by convolution technique

    Institute of Scientific and Technical Information of China (English)

    Cheng Jin-Xing; Ouyang Xiao-Ping; Zheng Yi; Zhang An-Hui; Ouyang Mao-Jie

    2008-01-01

    It is a very complex and tlme-consuming process to simulate the nuclear reactor neutron spectrum from the reactor core to the export channel by applying a Monte Carlo program. This paper presents a new method to calculate the neutron spectrum by using the convolution technique which considers the channel transportation as a linear system and the transportation scattering as the response function. It also applies Monte Carlo Neutron and Photon Transport Code (MCNP) to simulate the response function numerically. With the application of convolution technique to calculate thespectrum distribution from the core to the channel, the process is then much more convenient only with the simple numerical integral numeration. This saves computer time and reduces some trouble in re-writing of the MCNP program.

  7. ERGODIC THEOREM FOR INFINITE ITERATED FUNCTION SYSTEMS

    Institute of Scientific and Technical Information of China (English)

    O Hyong-chol; Ro Yong-hwa; Kil Won-gun

    2005-01-01

    A set of contraction maps of a metric space is called an iterated function systems.Iterated function systems with condensation can be considered infinite iterated function systems. Infinite iterated function systems on compact metric spaces were studied. Using the properties of Banach limit and uniform contractiveness, it was proved that the random iterating algorithms for infinite iterated function systems on compact metric spaces satisfy ergodicity. So the random iterating algorithms for iterated function systems with condensation satisfy ergodicity, too.

  8. Runaway electrons and ITER

    Science.gov (United States)

    Boozer, Allen H.

    2017-05-01

    The potential for damage, the magnitude of the extrapolation, and the importance of the atypical—incidents that occur once in a thousand shots—make theory and simulation essential for ensuring that relativistic runaway electrons will not prevent ITER from achieving its mission. Most of the theoretical literature on electron runaway assumes magnetic surfaces exist. ITER planning for the avoidance of halo and runaway currents is focused on massive-gas or shattered-pellet injection of impurities. In simulations of experiments, such injections lead to a rapid large-scale magnetic-surface breakup. Surface breakup, which is a magnetic reconnection, can occur on a quasi-ideal Alfvénic time scale when the resistance is sufficiently small. Nevertheless, the removal of the bulk of the poloidal flux, as in halo-current mitigation, is on a resistive time scale. The acceleration of electrons to relativistic energies requires the confinement of some tubes of magnetic flux within the plasma and a resistive time scale. The interpretation of experiments on existing tokamaks and their extrapolation to ITER should carefully distinguish confined versus unconfined magnetic field lines and quasi-ideal versus resistive evolution. The separation of quasi-ideal from resistive evolution is extremely challenging numerically, but is greatly simplified by constraints of Maxwell’s equations, and in particular those associated with magnetic helicity. The physics of electron runaway along confined magnetic field lines is clarified by relations among the poloidal flux change required for an e-fold in the number of electrons, the energy distribution of the relativistic electrons, and the number of relativistic electron strikes that can be expected in a single disruption event.

  9. Bayesian calibration of reactor neutron flux spectrum using activation detectors measurements: Application to CALIBAN reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cartier, J. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, DAM, DIF, F-91297 Arpajon (France); Casoli, P. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, DAM, Valduc, F-21120 Is sur Tille (France); Chappert, F. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, DAM, DIF, F-91297 Arpajon (France)

    2013-07-01

    In this paper, we present calibration methods in order to estimate reactor neutron flux spectrum and its uncertainties by using integral activation measurements. These techniques are performed using Bayesian and MCMC framework. These methods are applied to integral activation experiments in the cavity of the CALIBAN reactor. We estimate the neutron flux and its related uncertainties. The originality of this work is that these uncertainties take into account measurements uncertainties, cross-sections uncertainties and model error. In particular, our results give a very good approximation of the total flux and indicate that neutron flux from MCNP simulation for energies above about 5 MeV seems to overestimate the 'real flux'. (authors)

  10. Iterative participatory design

    DEFF Research Database (Denmark)

    2010-01-01

    The theoretical background in this chapter is information systems development in an organizational context. This includes theories from participatory design, human-computer interaction, and ethnographically inspired studies of work practices. The concept of design is defined as an experimental...... iterative process of mutual learning by designers and domain experts (users), who aim to change the users’ work practices through the introduction of information systems. We provide an illustrative case example with an ethnographic study of clinicians experimenting with a new electronic patient record...

  11. Quantum iterated function systems.

    Science.gov (United States)

    Łoziński, Artur; Zyczkowski, Karol; Słomczyński, Wojciech

    2003-10-01

    An iterated function system (IFS) is defined by specifying a set of functions in a classical phase space, which act randomly on an initial point. In an analogous way, we define a quantum IFS (QIFS), where functions act randomly with prescribed probabilities in the Hilbert space. In a more general setting, a QIFS consists of completely positive maps acting in the space of density operators. This formalism is designed to describe certain problems of nonunitary quantum dynamics. We present exemplary classical IFSs, the invariant measure of which exhibits fractal structure, and study properties of the corresponding QIFSs and their invariant states.

  12. Iterative Magnetometer Calibration

    Science.gov (United States)

    Sedlak, Joseph

    2006-01-01

    This paper presents an iterative method for three-axis magnetometer (TAM) calibration that makes use of three existing utilities recently incorporated into the attitude ground support system used at NASA's Goddard Space Flight Center. The method combines attitude-independent and attitude-dependent calibration algorithms with a new spinning spacecraft Kalman filter to solve for biases, scale factors, nonorthogonal corrections to the alignment, and the orthogonal sensor alignment. The method is particularly well-suited to spin-stabilized spacecraft, but may also be useful for three-axis stabilized missions given sufficient data to provide observability.

  13. ITER LIDAR performance analysis.

    Science.gov (United States)

    Beurskens, M N A; Giudicotti, L; Kempenaars, M; Scannell, R; Walsh, M J

    2008-10-01

    The core LIDAR Thomson scattering for ITER is specified for core profile measurements with a spatial resolution of 7 cm (a/30) for the range of 500 eV3x10(19) m(-3) at an accuracy of system can meet its spatial and accuracy specifications for higher temperatures of T(e)>5 keV with a combination of a neodymium-doped yttrium aluminum garnet (Nd:YAG) laser (lambda(0)=1064 nm, Delta lambdanear infrared detectors.

  14. Preliminary Neutronics Design of Breed Blanket for Fusion-fission Hybrid Reactor%聚变-裂变增殖堆包层的初步中子学设计

    Institute of Scientific and Technical Information of China (English)

    赵奉超; 栗再新

    2012-01-01

    基于国际热核实验堆ITER的堆芯参数和套管结构,对聚变-裂变增殖堆包层进行了初步中子学设计.基于国际热核实验堆的堆芯参数提出了采用套管结构,以天然金属铀为燃料和硅酸锂为氚增殖剂的快裂变-增殖堆包层的初步中子学设计方案.使用FENDL 2.1核数据库及MCNP程序自带的核数据库,用MCNP程序对套管结构快裂变-增殖堆包层进行一维的方案筛选及三维中子学的计算分析.计算分析包层内的一维功率密度分布、产氚率、钚增殖率分布,通过优化设计分析给出合理的包层设计方案,并计算氚增殖率TBR、能量放大倍数M、有效增值系数(Keff)、裂变增殖比等参数.%A preliminary neutronics design of breed blanket for fusion-fission hybrid reactor has been carried out based on the plasma parameters of International Thermonuclear Experimental Reactor (ITER) and casing structure. In the design of fast-fission breed blanket, the natural Uranium pebble bed is used as fuel and neutron multiplication and the Lithium silicate pebble bed is used as tritium breed material. By using FENDL2.1 nuclear database cross section library with native cross section library of MCNP nuclear database, the calculation and analysis are carried out with MCNP program. Through one-dimension calculation and analysis on different design proposals, a proper design proposal has been screened and then the three-dimension calculation and analysis have been implemented with the parameters of ITER. The calculation shows that the TBR of fusion-fission hybrid reactor is 1.13, it indicates that the design of breed blanket is able to meet self-sustaining of tritium and the calculation also indicates that the energy enlargement of fusion-ission hybrid reactor is 6.5 and Polonium breeding rate is 1.35, it means that the reactor is able to also product large quantities energy and Polonium and they could be used by light water reactor. Meanwhile, fission

  15. Confirmation of a realistic reactor model for BNCT dosimetry at the TRIGA Mainz

    Energy Technology Data Exchange (ETDEWEB)

    Ziegner, Markus, E-mail: Markus.Ziegner.fl@ait.ac.at [AIT Austrian Institute of Technology GmbH, Vienna A-1220, Austria and Institute of Atomic and Subatomic Physics, Vienna University of Technology, Vienna A-1020 (Austria); Schmitz, Tobias; Hampel, Gabriele [Institut für Kernchemie, Johannes Gutenberg-Universität, Mainz DE-55128 (Germany); Khan, Rustam [Pakistan Institute of Engineering and Applied Sciences (PIEAS), Islamabad PK-44000 (Pakistan); Blaickner, Matthias [AIT Austrian Institute of Technology GmbH, Vienna A-1220 (Austria); Palmans, Hugo [Acoustics and Ionising Radiation Division, National Physical Laboratory, Teddington TW11 0LW, United Kingdom and Medical Physics Group, EBG MedAustron GmbH, Wiener Neustadt A-2700 (Austria); Sharpe, Peter [Acoustics and Ionising Radiation Division, National Physical Laboratory, Teddington TW11 0LW (United Kingdom); Böck, Helmuth [Institute of Atomic and Subatomic Physics, Vienna University of Technology, Vienna A-1020 (Austria)

    2014-11-01

    Purpose: In order to build up a reliable dose monitoring system for boron neutron capture therapy (BNCT) applications at the TRIGA reactor in Mainz, a computer model for the entire reactor was established, simulating the radiation field by means of the Monte Carlo method. The impact of different source definition techniques was compared and the model was validated by experimental fluence and dose determinations. Methods: The depletion calculation code ORIGEN2 was used to compute the burn-up and relevant material composition of each burned fuel element from the day of first reactor operation to its current core. The material composition of the current core was used in a MCNP5 model of the initial core developed earlier. To perform calculations for the region outside the reactor core, the model was expanded to include the thermal column and compared with the previously established ATTILA model. Subsequently, the computational model is simplified in order to reduce the calculation time. Both simulation models are validated by experiments with different setups using alanine dosimetry and gold activation measurements with two different types of phantoms. Results: The MCNP5 simulated neutron spectrum and source strength are found to be in good agreement with the previous ATTILA model whereas the photon production is much lower. Both MCNP5 simulation models predict all experimental dose values with an accuracy of about 5%. The simulations reveal that a Teflon environment favorably reduces the gamma dose component as compared to a polymethyl methacrylate phantom. Conclusions: A computer model for BNCT dosimetry was established, allowing the prediction of dosimetric quantities without further calibration and within a reasonable computation time for clinical applications. The good agreement between the MCNP5 simulations and experiments demonstrates that the ATTILA model overestimates the gamma dose contribution. The detailed model can be used for the planning of structural

  16. Monte Carlo modelling of TRIGA research reactor

    Science.gov (United States)

    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.

  17. Obtaining of primary rays of spectrum X codes Penelope and MCNP5; Obtencion del espectro primario de Rayos X con los codigos Penelope y MCNP5

    Energy Technology Data Exchange (ETDEWEB)

    Pozuelo, F.; Querol, A.; Gallardo, S.; Rodenas, J.; Verdu, G.

    2012-07-01

    In this case, used codes PENELOPE MCNP5, based on the Monte Carlo method for x-ray spectrum taking into account the characteristics of the x-ray tube. In order to achieve a greater fit of simulated by the theoretical spectrum. It carried out a sensitivity analysis of the parameters available in both codes. The obtaining of the simulated spectrum could lead to an improvement in quality control of the x-ray tube to incorporate it as a method complementary to techniques.

  18. Iterative guided image fusion

    Directory of Open Access Journals (Sweden)

    Alexander Toet

    2016-08-01

    Full Text Available We propose a multi-scale image fusion scheme based on guided filtering. Guided filtering can effectively reduce noise while preserving detail boundaries. When applied in an iterative mode, guided filtering selectively eliminates small scale details while restoring larger scale edges. The proposed multi-scale image fusion scheme achieves spatial consistency by using guided filtering both at the decomposition and at the recombination stage of the multi-scale fusion process. First, size-selective iterative guided filtering is applied to decompose the source images into approximation and residual layers at multiple spatial scales. Then, frequency-tuned filtering is used to compute saliency maps at successive spatial scales. Next, at each spatial scale binary weighting maps are obtained as the pixelwise maximum of corresponding source saliency maps. Guided filtering of the binary weighting maps with their corresponding source images as guidance images serves to reduce noise and to restore spatial consistency. The final fused image is obtained as the weighted recombination of the individual residual layers and the mean of the approximation layers at the coarsest spatial scale. Application to multiband visual (intensified and thermal infrared imagery demonstrates that the proposed method obtains state-of-the-art performance for the fusion of multispectral nightvision images. The method has a simple implementation and is computationally efficient.

  19. Runaway electrons and ITER

    Science.gov (United States)

    Boozer, Allen

    2016-10-01

    ITER planning for avoiding runaway damage depends on magnetic surface breakup in fast relaxations. These arise in thermal quenches and in the spreading of impurities from massive gas injection or shattered pellets. Surface breakup would prevent a runaway to relativistic energies were it not for non-intercepting flux tubes, which contain magnetic field lines that do not intercept the walls. Such tubes persist near the magnetic axis and in the cores of islands but must dissipate before any confining surfaces re-form. Otherwise, a highly dangerous situation arises. Electrons that were trapped and accelerated in these flux tubes can fill a large volume of stochastic field lines and serve as a seed for the transfer of the full plasma current to runaways. If the outer confining surfaces are punctured, as by a drift into the wall, then the full runaway inventory will be lost in a short pulse along a narrow flux tube. Although not part of ITER planning, currents induced in the walls by the fast magnetic relaxation could be used to passively prevent outer surfaces re-forming. If magnetic surface breakup can be avoided during impurity injection, the plasma current could be terminated in tens of milliseconds by plasma cooling with no danger of runaway. Support by DoE Office of Fusion Energy Science Grant De-FG02-03ER54696.

  20. Bias estimates used in lieu of validation of fission products and minor actinides in MCNP Keff calculations for PWR burnup credit casks

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Don [ORNL; Marshall, William BJ J [ORNL; Wagner, John C [ORNL; Bowen, Douglas G [ORNL

    2015-09-01

    The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the bias due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.

  1. Carbon fiber composites application in ITER plasma facing components

    Science.gov (United States)

    Barabash, V.; Akiba, M.; Bonal, J. P.; Federici, G.; Matera, R.; Nakamura, K.; Pacher, H. D.; Rödig, M.; Vieider, G.; Wu, C. H.

    1998-10-01

    Carbon Fiber Composites (CFCs) are one of the candidate armour materials for the plasma facing components of the International Thermonuclear Experimental Reactor (ITER). For the present reference design, CFC has been selected as armour for the divertor target near the plasma strike point mainly because of unique resistance to high normal and off-normal heat loads. It does not melt under disruptions and might have higher erosion lifetime in comparison with other possible armour materials. Issues related to CFC application in ITER are described in this paper. They include erosion lifetime, tritium codeposition with eroded material and possible methods for the removal of the codeposited layers, neutron irradiation effect, development of joining technologies with heat sink materials, and thermomechanical performance. The status of the development of new advanced CFCs for ITER application is also described. Finally, the remaining R&D needs are critically discussed.

  2. Background Studies for the MINER Coherent Neutrino Scattering Reactor Experiment

    CERN Document Server

    Agnolet, G; Barker, D; Beck, R; Carroll, T J; Cesar, J; Cushman, P; Dent, J B; De Rijck, S; Dutta, B; Flanagan, W; Fritts, M; Gao, Y; Harris, H R; Hays, C C; Iyer, V; Jastram, A; Kadribasic, F; Kennedy, A; Kubik, A; Ogawa, I; Lang, K; Mahapatra, R; Mandic, V; Martin, R D; Mast, N; McDeavitt, S; Mirabolfathi, N; Mohanty, B; Nakajima, K; Newhouse, J; Newstead, J L; Phan, D; Proga, M; Roberts, A; Rogachev, G; Salazar, R; Sander, J; Senapati, K; Shimada, M; Strigari, L; Tamagawa, Y; Teizer, W; Vermaak, J I C; Villano, A N; Walker, J; Webb, B; Wetzel, Z; Yadavalli, S A

    2016-01-01

    The proposed Mitchell Institute Neutrino Experiment at Reactor (MINER) experiment at the Nuclear Science Center at Texas A&M University will search for coherent elastic neutrino-nucleus scattering within close proximity (about 2 meters) of a 1 MW TRIGA nuclear reactor core using low threshold, cryogenic germanium and silicon detectors. Given the Standard Model cross section of the scattering process and the proposed experimental proximity to the reactor, as many as 5 to 20 events/kg/day are expected. We discuss the status of preliminary measurements to characterize the main backgrounds for the proposed experiment. Both in situ measurements at the experimental site and simulations using the MCNP and GEANT4 codes are described. A strategy for monitoring backgrounds during data taking is briefly discussed.

  3. Recriticality calculations for uraniumdioxide-water systems with MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Kumpf, H.

    1998-10-01

    Investigations of severe accidents in power reactors will hardly produce data on the geometry, composition and density distributions of fuel mixtures in such detail as demanded for criticality calculations. In view of this rather sloppy formulation of the task one might consider as an objective the search for the `worst case`, i.e. the composition and structure of arrangements with maximum multiplication. The fuel geometry with maximum k{sub {infinity}} is a hexagonal close package of spheres with a certain radius, immersed in water. But this arrangement is mechanically unstable. Furthermore, the collapsed hexagonal close package with touching spheres is by no means optimal with respect to k{sub {infinity}}. Thus, mechanical stability is a necessary additional condition in the search for the worst case. The main part of the report deals with the determination of such a structure. In view of the complexity of the task rigorous mathematical demonstration is not expected to be successful. Instead one adheres to heuristic reasoning. (orig.)

  4. Neutronic design of the ITER radial neutron camera

    Energy Technology Data Exchange (ETDEWEB)

    Petrizzi, L. [Associazione EURATOM-ENEA sulla Fusione, ENEA Centro Ricerche, C.P. 65, 00044 Frascati, Rome (Italy)], E-mail: petrizzi@frascati.enea.it; Barnsley, R. [EFDA CSU-Garching (Germany); Bertalot, L.; Esposito, B. [Associazione EURATOM-ENEA sulla Fusione, ENEA Centro Ricerche, C.P. 65, 00044 Frascati, Rome (Italy); Haskell, H. [ITER International Team, Garching (Germany); Mainardi, E.; Marocco, D.; Podda, S. [Associazione EURATOM-ENEA sulla Fusione, ENEA Centro Ricerche, C.P. 65, 00044 Frascati, Rome (Italy); Walker, C. [ITER International Team, Garching (Germany); Villari, S. [Associazione EURATOM-ENEA sulla Fusione, ENEA Centro Ricerche, C.P. 65, 00044 Frascati, Rome (Italy)

    2007-10-15

    This paper summarizes the work, performed in the frame of various EFDA contracts during 2004-2005, on the design review and upgrade of the ITER radial neutron camera (RNC). The RNC, which should provide information on the spatial distribution and energy spectrum of the neutron emission, consists of an ex-vessel system (fan-like collimator with 12 x 3 lines of sights) and an in-vessel system with further 9 lines for a full coverage of the plasma. A Monte Carlo code (MCNP) has been used for the neutronic calculations. The basic ITER model has been developed from the CATIA drawings to include the RNC with all details relevant for the neutronic analysis. In the model the collimator diameters have been set to 2 and 4 cm, respectively, for the ex-vessel and in-vessel systems. A detailed space dependent fusion neutron source (DD and DT phases in various plasma scenarios) has been used with a consistent ion temperature radial profile. A special variance reduction treatment has been developed so that neutrons reach the far regions in the high collimated neutron beam and score with a satisfying statistical error. Neutron and photon fluxes and spectra have been calculated. Approximately, one neutron out of 10{sup 11} emitted in all the plasma reaches a single ex-vessel detector. Therefore, for an emission rate of 1.8 x 10{sup 20} n/s (corresponding to 500 MW fusion power) the flux on the detectors is in the range (1-5) x 10{sup 8} n/(cm{sup 2} s) depending on the poloidal orientation. The fraction of scattered neutrons (>1 MeV) is lower than few % of the total. A measurement simulation software tool (MSST) performing asymmetric Abel inversion of simulated measured neutron signals has also been developed for line of sight and design optimization. Combining information from MCNP calculations and MSST, it has been possible to evaluate the performance of the RNC, check whether the present design of the RNC meets the measurement requirements and optimize the RNC design.

  5. Coupling MCNP-DSP and LAHET Monte Carlo codes for designing subcriticality monitors for accelerator-driven systems

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.; Perez, R. [Oak Ridge National Lab., TN (United States); Rugama, Y.; Munoz-Cobo, J.L. [Poly. Tech. Univ. of Valencia (Spain). Chemical and Nuclear Engineering Dept.

    2001-07-01

    The design of reactivity monitoring systems for accelerator-driven systems must be investigated to ensure that such systems remain subcritical during operation. The Monte Carlo codes LAHET and MCNP-DSP were combined together to facilitate the design of reactivity monitoring systems. The coupling of LAHET and MCNP-DSP provides a tool that can be used to simulate a variety of subcritical measurements such as the pulsed neutron, Rossi-{alpha}, or noise analysis measurements. (orig.)

  6. Coupling MCNP-DSP and LAHET Monte Carlo Codes for Designing Subcriticality Monitors for Accelerator-Driven Systems

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.E.; Rugama, Y. Munoz-Cobos, J.; Perez, R.

    2000-10-23

    The design of reactivity monitoring systems for accelerator-driven systems must be investigated to ensure that such systems remain subcritical during operation. The Monte Carlo codes LAHET and MCNP-DSP were combined together to facilitate the design of reactivity monitoring systems. The coupling of LAHET and MCNP-DSP provides a tool that can be used to simulate a variety of subcritical measurements such as the pulsed neutron, Rossi-{alpha}, or noise analysis measurements.

  7. Power distributions in fresh and depleted LEU and HEU cores of the MITR reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, E.H.; Horelik, N.E.; Dunn, F.E.; Newton, T.H., Jr.; Hu, L.; Stevens, J.G. (Nuclear Engineering Division); (2MIT Nuclear Reactor Laboratory and Nuclear Science and Engineering Department)

    2012-04-04

    The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (UMo) is expected to allow the conversion of U.S. domestic high performance reactors like the MITR-II reactor. Toward this goal, core geometry and power distributions are presented. Distributions of power are calculated for LEU cores depleted with MCODE using an MCNP5 Monte Carlo model. The MCNP5 HEU and LEU MITR models were previously compared to experimental benchmark data for the MITR-II. This same model was used with a finer spatial depletion in order to generate power distributions for the LEU cores. The objective of this work is to generate and characterize a series of fresh and depleted core peak power distributions, and provide a thermal hydraulic evaluation of the geometry which should be considered for subsequent thermal hydraulic safety analyses.

  8. Implementation of a tree algorithm in MCNP code for nuclear well logging applications

    Energy Technology Data Exchange (ETDEWEB)

    Li Fusheng, E-mail: fusheng.li@bakerhughes.com [Baker Hughes Incorporated, 2001 Rankin Rd. Houston, TX 77073-5101 (United States); Han Xiaogang [Baker Hughes Incorporated, 2001 Rankin Rd. Houston, TX 77073-5101 (United States)

    2012-07-15

    The goal of this paper is to develop some modeling capabilities that are missing in the current MCNP code. Those missing capabilities can greatly help for some certain nuclear tools designs, such as a nuclear lithology/mineralogy spectroscopy tool. The new capabilities to be developed in this paper include the following: zone tally, neutron interaction tally, gamma rays index tally and enhanced pulse-height tally. The patched MCNP code also can be used to compute neutron slowing-down length and thermal neutron diffusion length. - Highlights: Black-Right-Pointing-Pointer Tree structure programming is suitable for Monte-Carlo based particle tracking. Black-Right-Pointing-Pointer Enhanced pulse height tally is developed for oilwell logging tool simulation. Black-Right-Pointing-Pointer Neutron interaction tally and gamma ray index tally for geochemical logging.

  9. MCNP6 Simulation of Light and Medium Nuclei Fragmentation at Intermediate Energies

    Energy Technology Data Exchange (ETDEWEB)

    Mashnik, Stepan Georgievich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kerby, Leslie Marie [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Univ. of Idaho, Moscow, ID (United States)

    2015-08-24

    Fragmentation reactions induced on light and medium nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below are studied with the Los Alamos transport code MCNP6 and with its CEM03.03 and LAQGSM03.03 event generators. CEM and LAQGSM assume that intermediate-energy fragmentation reactions on light nuclei occur generally in two stages. The first stage is the intranuclear cascade (INC), followed by the second, Fermi breakup disintegration of light excited residual nuclei produced after the INC. CEM and LAQGSM account also for coalescence of light fragments (complex particles) up to sup>4He from energetic nucleons emitted during INC. We investigate the validity and performance of MCNP6, CEM, and LAQGSM in simulating fragmentation reactions at intermediate energies and discuss possible ways of further improving these codes.

  10. MCNP6 simulation of light and medium nuclei fragmentation at intermediate energies

    Directory of Open Access Journals (Sweden)

    Mashnik Stepan G.

    2016-01-01

    Full Text Available Fragmentation reactions induced on light and medium nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below are studied with the Los Alamos transport code MCNP6 and with its CEM03.03 and LAQGSM03.03 event generators. CEM and LAQGSM assume that intermediate-energy fragmentation reactions on light nuclei occur generally in two stages. The first stage is the intranuclear cascade (INC, followed by the second, Fermi breakup disintegration of light excited residual nuclei produced after the INC. CEM and LAQGSM account also for coalescence of light fragments (complex particles up to 4He from energetic nucleons emitted during INC. We investigate the validity and performance of MCNP6, CEM, and LAQGSM in simulating fragmentation reactions at intermediate energies and discuss possible ways of further improving these codes.

  11. MCNP6 simulation of light and medium nuclei fragmentation at intermediate energies

    CERN Document Server

    Mashnik, Stepan G

    2015-01-01

    Fragmentation reactions induced on light and medium nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below are studied with the Los Alamos transport code MCNP6 and with its CEM03.03 and LAQGSM03.03 event generators. CEM and LAQGSM assume that intermediate-energy fragmentation reactions on light nuclei occur generally in two stages. The first stage is the intranuclear cascade (INC), followed by the second, Fermi breakup disintegration of light excited residual nuclei produced after the INC. CEM and LAQGSM account also for coalescence of light fragments (complex particles) up to 4He from energetic nucleons emitted during INC. We investigate the validity and performance of MCNP6, CEM, and LAQGSM in simulating fragmentation reactions at intermediate energies and discuss possible ways of further improving these codes

  12. MCNP6 simulation of light and medium nuclei fragmentation at intermediate energies

    Science.gov (United States)

    Mashnik, Stepan G.; Kerby, Leslie M.

    2016-05-01

    Fragmentation reactions induced on light and medium nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below are studied with the Los Alamos transport code MCNP6 and with its CEM03.03 and LAQGSM03.03 event generators. CEM and LAQGSM assume that intermediate-energy fragmentation reactions on light nuclei occur generally in two stages. The first stage is the intranuclear cascade (INC), followed by the second, Fermi breakup disintegration of light excited residual nuclei produced after the INC. CEM and LAQGSM account also for coalescence of light fragments (complex particles) up to 4He from energetic nucleons emitted during INC. We investigate the validity and performance of MCNP6, CEM, and LAQGSM in simulating fragmentation reactions at intermediate energies and discuss possible ways of further improving these codes.

  13. SABR fusion-fission hybrid transmutation reactor design concept

    Science.gov (United States)

    Stacey, Weston

    2009-11-01

    A conceptual design has been developed for a sub-critical advanced burner reactor (SABR) consisting of i) a sodium cooled fast reactor fueled with the transuranics (TRU) from spent nuclear fuel, and ii) a D-T tokamak fusion neutron source based on ITER physics and technology. Subcritical operation enables more efficient transmutation fuel cycles in TRU fueled reactors (without compromising safety), which may be essential for significant reduction in high-level waste repository requirements. ITER will serve as the prototype for the fusion neutron source, which means SABRs could be implemented to help close the nuclear fuel cycle during the 2^nd quarter of the century.

  14. Iterated crowdsourcing dilemma game

    Science.gov (United States)

    Oishi, Koji; Cebrian, Manuel; Abeliuk, Andres; Masuda, Naoki

    2014-02-01

    The Internet has enabled the emergence of collective problem solving, also known as crowdsourcing, as a viable option for solving complex tasks. However, the openness of crowdsourcing presents a challenge because solutions obtained by it can be sabotaged, stolen, and manipulated at a low cost for the attacker. We extend a previously proposed crowdsourcing dilemma game to an iterated game to address this question. We enumerate pure evolutionarily stable strategies within the class of so-called reactive strategies, i.e., those depending on the last action of the opponent. Among the 4096 possible reactive strategies, we find 16 strategies each of which is stable in some parameter regions. Repeated encounters of the players can improve social welfare when the damage inflicted by an attack and the cost of attack are both small. Under the current framework, repeated interactions do not really ameliorate the crowdsourcing dilemma in a majority of the parameter space.

  15. UW MCNP source patch for the EPFL Haefely source. EPFL (Swiss) fusion-fission hybrid experiment

    Energy Technology Data Exchange (ETDEWEB)

    McKinney, G; Woodruff, G L

    1986-06-01

    The development of a source patch which describes the Haefely neutron source for use in the MCNP Monte Carlo code has been described in progress reports of the EPFL (Swiss) Fusion Blanket Project at the University of Washington. The most recent of these reports dealing with the source patch was Progress Report No. 14. This report reviews some of the physical description included in the report, and also includes additional details of the patch as well as a listing of the patch itself.

  16. Development of Multi-physics (Multiphase CFD + MCNP) simulation for generic solution vessel power calculation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Jun [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Buechler, Cynthia Eileen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-07-17

    The current study aims to predict the steady state power of a generic solution vessel and to develop a corresponding heat transfer coefficient correlation for a Moly99 production facility by conducting a fully coupled multi-physics simulation. A prediction of steady state power for the current application is inherently interconnected between thermal hydraulic characteristics (i.e. Multiphase computational fluid dynamics solved by ANSYS-Fluent 17.2) and the corresponding neutronic behavior (i.e. particle transport solved by MCNP6.2) in the solution vessel. Thus, the development of a coupling methodology is vital to understand the system behavior at a variety of system design and postulated operating scenarios. In this study, we report on the k-effective (keff) calculation for the baseline solution vessel configuration with a selected solution concentration using MCNP K-code modeling. The associated correlation of thermal properties (e.g. density, viscosity, thermal conductivity, specific heat) at the selected solution concentration are developed based on existing experimental measurements in the open literature. The numerical coupling methodology between multiphase CFD and MCNP is successfully demonstrated, and the detailed coupling procedure is documented. In addition, improved coupling methods capturing realistic physics in the solution vessel thermal-neutronic dynamics are proposed and tested further (i.e. dynamic height adjustment, mull-cell approach). As a key outcome of the current study, a multi-physics coupling methodology between MCFD and MCNP is demonstrated and tested for four different operating conditions. Those different operating conditions are determined based on the neutron source strength at a fixed geometry condition. The steady state powers for the generic solution vessel at various operating conditions are reported, and a generalized correlation of the heat transfer coefficient for the current application is discussed. The assessment of multi

  17. Radioactivity measurements of ITER materials using the TFTR D-T neutron field

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, A.; Abdou, M.A. [California Univ., Los Angeles, CA (United States). School of Engineering and Applied Science; Barnes, C.W. [Los Alamos National Lab., NM (United States); Kugel, H.W. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Loughlin, M.J. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

    1994-08-01

    The availability of high D-T fusion neutron yields at TFTR has provided a useful opportunity to directly measure D-T neutron-induced radioactivity in a realistic tokamak fusion reactor environment for materials of vital interest to ITER. These measurements are valuable for characterizing radioactivity in various ITER candidate materials. for validating complex neutron transport calculations, and for meeting fusion reactor licensing requirements. The radioactivity measurements at TFTR involve potential ITER materials including stainless steel 316, vanadium, titanium, chromium, silicon, iron, cobalt, nickel, molybdenum, aluminum, copper, zinc. zirconium, niobium, and tungsten. Small samples of these materials were irradiated close to the plasma and just outside the vacuum vessel wall of TFTR, locations of different neutron energy spectra. Saturation activities for both threshold and capture reactions were measured. Data from dosimetric reactions have been used to obtain preliminary neutron energy spectra. Spectra from the first wall were compared to calculations from ITER and to measurements from accelerator-based tests.

  18. Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises

    Energy Technology Data Exchange (ETDEWEB)

    El Bakkari, B. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco)], E-mail: bakkari@gmail.com; El Bardouni, T.; Merroun, O.; El Younoussi, Ch.; Boulaich, Y. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco); Chakir, E. [EPTN-LPMR, Faculty of Sciences Kenitra (Morocco)

    2009-05-15

    The aim of this study is to evaluate the capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5, through the use of the MCNP tally information. BUCAL1 uses the fourth order Runge Kutta method with the predictor-corrector approach as the integration method to determine the fuel composition at a desired burnup step. Validation of BUCAL1 was done by code vs. code comparison. Results of two different kinds of codes are employed. The first one is CASMO-4, a deterministic multi-group two-dimensional transport code. The second kind is MCODE and MOCUP, a link MCNP-ORIGEN codes. These codes use different burnup algorithms to solve the depletion equations system. Eigenvalue and isotope concentrations were compared for two PWR uranium and thorium benchmark exercises at cold (300 K) and hot (900 K) conditions, respectively. The eigenvalue comparison between BUCAL1 and the aforementioned two kinds of codes shows a good prediction of the systems'k-inf values during the entire burnup history, and the maximum difference is within 2%. The differences between the BUCAL1 isotope concentrations and the predictions of CASMO-4, MCODE and MOCUP are generally better, and only for a few sets of isotopes these differences exceed 10%.

  19. Voxel2MCNP: software for handling voxel models for Monte Carlo radiation transport calculations.

    Science.gov (United States)

    Hegenbart, Lars; Pölz, Stefan; Benzler, Andreas; Urban, Manfred

    2012-02-01

    Voxel2MCNP is a program that sets up radiation protection scenarios with voxel models and generates corresponding input files for the Monte Carlo code MCNPX. Its technology is based on object-oriented programming, and the development is platform-independent. It has a user-friendly graphical interface including a two- and three-dimensional viewer. A row of equipment models is implemented in the program. Various voxel model file formats are supported. Applications include calculation of counting efficiency of in vivo measurement scenarios and calculation of dose coefficients for internal and external radiation scenarios. Moreover, anthropometric parameters of voxel models, for instance chest wall thickness, can be determined. Voxel2MCNP offers several methods for voxel model manipulations including image registration techniques. The authors demonstrate the validity of the program results and provide references for previous successful implementations. The authors illustrate the reliability of calculated dose conversion factors and specific absorbed fractions. Voxel2MCNP is used on a regular basis to generate virtual radiation protection scenarios at Karlsruhe Institute of Technology while further improvements and developments are ongoing.

  20. Current status of ACE format libraries for MCNP at nuclear date center of KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Gil, Choong Sup; Lee, Young Ouk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-09-15

    The current status of ACE format MCNP/MCNPX libraries by NDC of KAERI is presented with a short description of each library. Validation calculations with recent nuclear data evaluations ENDF/B-VII.0, ENDF/B-VII.1, JEFF-3.2, and JENDL-4.0 have been carried out by the MCNP5 code for 119 criticality benchmark problems taken from the expanded criticality validation suite supplied by LANL. The overall performances of the ACE format KN-libraries have been analyzed in comparison with the results calculated with the ENDF/B-VII.0-based ENDF70 library of LANL. It was confirmed that the ENDF/B-VII.1-based KNE71 library showed better performances than the others by comparing the RMS errors and χ2 values for five benchmark categories as well as whole benchmark problems. ENDF/B-VII.1 and JEFF-3.2 have a tendency to yield more reliable MCNP calculation results within certain confidence intervals regarding the total uncertainties for the keff values. It is found that the adoption of the latest evaluated nuclear data might ensure better outcomes in various research and development areas.

  1. Validation of neutron flux redistribution factors in JSI TRIGA reactor due to control rod movements.

    Science.gov (United States)

    Kaiba, Tanja; Žerovnik, Gašper; Jazbec, Anže; Štancar, Žiga; Barbot, Loïc; Fourmentel, Damien; Snoj, Luka

    2015-10-01

    For efficient utilization of research reactors, such as TRIGA Mark II reactor in Ljubljana, it is important to know neutron flux distribution in the reactor as accurately as possible. The focus of this study is on the neutron flux redistributions due to control rod movements. For analyzing neutron flux redistributions, Monte Carlo calculations of fission rate distributions with the JSI TRIGA reactor model at different control rod configurations have been performed. Sensitivity of the detector response due to control rod movement have been studied. Optimal radial and axial positions of the detector have been determined. Measurements of the axial neutron flux distribution using the CEA manufactured fission chambers have been performed. The experiments at different control rod positions were conducted and compared with the MCNP calculations for a fixed detector axial position. In the future, simultaneous on-line measurements with multiple fission chambers will be performed inside the reactor core for a more accurate on-line power monitoring system.

  2. Multifunctional reactors

    NARCIS (Netherlands)

    Westerterp, K.R.

    1992-01-01

    Multifunctional reactors are single pieces of equipment in which, besides the reaction, other functions are carried out simultaneously. The other functions can be a heat, mass or momentum transfer operation and even another reaction. Multifunctional reactors are not new, but they have received much

  3. Experimental Simulation of the Behaviour of Diagnostic First Mirrors Fabricated of Different Metals for ITER Conditions

    OpenAIRE

    Voitsenya, V.; Bardamid, A. F.; Donne, A. J. H.

    2016-01-01

    In the experimental fusion reactor ITER, the plasma-facing component of each optical and/or laser diagnostic needs to be based on reflective optics with at least one mirror (first mirror) facing the thermonuclear plasma. The different kinds of radiation emanating from the burning plasma (neutrons, neutral atoms, electromagnetic radiation) create hostile operating conditions for the first mirrors. Therefore, a special program has been set up under the ITER framework aimed at solving the first ...

  4. BENCHMARK EVALUATION OF THE START-UP CORE REACTOR PHYSICS MEASUREMENTS OF THE HIGH TEMPERATURE ENGINEERING TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    John Darrell Bess

    2010-05-01

    The benchmark evaluation of the start-up core reactor physics measurements performed with Japan’s High Temperature Engineering Test Reactor, in support of the Next Generation Nuclear Plant Project and Very High Temperature Reactor Program activities at the Idaho National Laboratory, has been completed. The evaluation was performed using MCNP5 with ENDF/B-VII.0 nuclear data libraries and according to guidelines provided for inclusion in the International Reactor Physics Experiment Evaluation Project Handbook. Results provided include updated evaluation of the initial six critical core configurations (five annular and one fully-loaded). The calculated keff eigenvalues agree within 1s of the benchmark values. Reactor physics measurements that were evaluated include reactivity effects measurements such as excess reactivity during the core loading process and shutdown margins for the fully-loaded core, four isothermal temperature reactivity coefficient measurements for the fully-loaded core, and axial reaction rate measurements in the instrumentation columns of three core configurations. The calculated values agree well with the benchmark experiment measurements. Fully subcritical and warm critical configurations of the fully-loaded core were also assessed. The calculated keff eigenvalues for these two configurations also agree within 1s of the benchmark values. The reactor physics measurement data can be used in the validation and design development of future High Temperature Gas-cooled Reactor systems.

  5. V&V of MCNP 6.1.1 Beta Against Intermediate and High-Energy Experimental Data

    Energy Technology Data Exchange (ETDEWEB)

    Mashnik, Stepan G [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-09-08

    This report presents a set of validation and verification (V&V) MCNP 6.1.1 beta results calculated in parallel, with MPI, obtained using its event generators at intermediate and high-energies compared against various experimental data. It also contains several examples of results using the models at energies below 150 MeV, down to 10 MeV, where data libraries are normally used. This report can be considered as the forth part of a set of MCNP6 Testing Primers, after its first, LA-UR-11-05129, and second, LA-UR-11-05627, and third, LA-UR-26944, publications, but is devoted to V&V with the latest, 1.1 beta version of MCNP6. The MCNP6 test-problems discussed here are presented in the /VALIDATION_CEM/and/VALIDATION_LAQGSM/subdirectories in the MCNP6/Testing/directory. README files that contain short descriptions of every input file, the experiment, the quantity of interest that the experiment measures and its description in the MCNP6 output files, and the publication reference of that experiment are presented for every test problem. Templates for plotting the corresponding results with xmgrace as well as pdf files with figures representing the final results of our V&V efforts are presented. Several technical “bugs” in MCNP 6.1.1 beta were discovered during our current V&V of MCNP6 while running it in parallel with MPI using its event generators. These “bugs” are to be fixed in the following version of MCNP6. Our results show that MCNP 6.1.1 beta using its CEM03.03, LAQGSM03.03, Bertini, and INCL+ABLA, event generators describes, as a rule, reasonably well different intermediate- and high-energy measured data. This primer isn’t meant to be read from cover to cover. Readers may skip some sections and go directly to any test problem in which they are interested.

  6. A proposal for the ITER remote participation system in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Nagayama, Y., E-mail: nagayama.yoshio@nifs.ac.j [National Institute for Fusion Science, 322-6 Oroshi, Toki 509-5292 (Japan); Emoto, M.; Kozaki, Y.; Nakanishi, H.; Sudo, S.; Yamamoto, T. [National Institute for Fusion Science, 322-6 Oroshi, Toki 509-5292 (Japan); Hiraki, K. [Graduate School of Information Science and Technology, University of Tokyo, 7-3-1 Hongo, Tokyo 113-8656 (Japan); Urushidani, S. [National Institute of Informatics, 2-1-2 Hitotsubashi, Chiyoda-ku, Tokyo 101-8430 (Japan)

    2010-07-15

    This paper presents a proposal of the remote participation system for the international thermonuclear experimental reactor (ITER). The object of this paper is to clarify technical issues to analyze the ITER data safely and conveniently. The Japanese case is considered as an example, but technologies presented here can be used worldwide. Major technical issues are as follows: (1) the long distance data transfer; (2) the massive data server; (3) the secure network; (4) the convenient and fast data analysis system. Raw data of ITER can be transferred from France to Japan in a short time by optimizing TCP/IP parameters. The virtual private network (VPN) technology provides a secure environment of the data mirroring and the distributed computation. The analysis server with the WEB user interface enables physicists to analyze the ITER data from the Internet. Streaming data, such as plasma parameters in the steady state, video and sound of the ITER plasma and the status of experiment, which provides feeling of reality, are delivered by using the multi-cast technology. These technologies are being developed in SNET, which is a virtual laboratory for Japanese fusion community. International collaboration is required to develop a global distributed file system and a data analysis system further.

  7. ITER perspective on fusion reactor diagnostics - A spectroscopic view

    DEFF Research Database (Denmark)

    De Bock, M. F. M.; Barnsley, R.; Bassan, M.

    2016-01-01

    challenges to the development of spectroscopic (but also other) diagnostics. This contribution presents an overview of recent achievements in 4 topical areas: First mirror protection and cleaning, Nuclear confinement, Radiation mitigation strategy for optical and electronic components and Calibration...

  8. Dynamic behavior of the HTR-10 reactor: Dual temperature feedback model

    Directory of Open Access Journals (Sweden)

    Hosseini Seyed Ali

    2015-01-01

    Full Text Available The current work aims at presenting a simple model for PBM-type reactors' dynamic behavior analysis. The proposed model is based on point kinetics equations coupled with feedbacks from fuel and moderator temperatures. The temperature reactivity coefficients were obtained through MCNP code and via available experimental data. Parameters such as heat capacity and heat conductivity were carefully analyzed and the final system of equations was numerically solved. The obtained results, while in partial agreement with previously proposed models, suggest lower sensitivity to step reactivity insertion as compared to other reactor designs and inherent safety of the design.

  9. Fuel burnup analysis for Thai research reactor by using MCNPX computer code

    Science.gov (United States)

    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.

  10. Determination of the Sensitivity of the Antineutrino Probe for Reactor Core Monitoring

    Science.gov (United States)

    Cormon, S.; Fallot, M.; Bui, V.-M.; Cucoanes, A.; Estienne, M.; Lenoir, M.; Onillon, A.; Shiba, T.; Yermia, F.; Zakari-Issoufou, A.-A.

    2014-06-01

    This paper presents a feasibility study of the use of the detection of reactor-antineutrinos (νbare) for non proliferation purpose. To proceed, we have started to study different reactor designs with our simulation tools. We use a package called MCNP Utility for Reactor Evolution (MURE), initially developed by CNRS/IN2P3 labs to study Generation IV reactors. The MURE package has been coupled to fission product beta decay nuclear databases for studying reactor antineutrino emission. This method is the only one able to predict the antineutrino emission from future reactor cores, which don't use the thermal fission of 235U, 239Pu and 241Pu. It is also the only way to include off-equilibrium effects, due to neutron captures and time evolution of the fission product concentrations during a reactor cycle. We will present here the first predictions of antineutrino energy spectra from innovative reactor designs (Generation IV reactors). We will then discuss a summary of our results of non-proliferation scenarios involving the latter reactor designs, taking into account reactor physics constraints.

  11. Determination of the Sensitivity of the Antineutrino Probe for Reactor Core Monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Cormon, S.; Fallot, M., E-mail: fallot@subatech.in2p3.fr; Bui, V.-M.; Cucoanes, A.; Estienne, M.; Lenoir, M.; Onillon, A.; Shiba, T.; Yermia, F.; Zakari-Issoufou, A.-A.

    2014-06-15

    This paper presents a feasibility study of the use of the detection of reactor-antineutrinos (ν{sup ¯}{sub e}) for non proliferation purpose. To proceed, we have started to study different reactor designs with our simulation tools. We use a package called MCNP Utility for Reactor Evolution (MURE), initially developed by CNRS/IN2P3 labs to study Generation IV reactors. The MURE package has been coupled to fission product beta decay nuclear databases for studying reactor antineutrino emission. This method is the only one able to predict the antineutrino emission from future reactor cores, which don't use the thermal fission of {sup 235}U, {sup 239}Pu and {sup 241}Pu. It is also the only way to include off-equilibrium effects, due to neutron captures and time evolution of the fission product concentrations during a reactor cycle. We will present here the first predictions of antineutrino energy spectra from innovative reactor designs (Generation IV reactors). We will then discuss a summary of our results of non-proliferation scenarios involving the latter reactor designs, taking into account reactor physics constraints.

  12. Verification and Validation of Monte Carlo n-Particle Code 6 (MCNP6) with Neutron Protection Factor Measurements of an Iron Box

    Science.gov (United States)

    2014-03-27

    want to express my sincere love, respect, and admiration for my wife, who motivated and supported me throughout this long endeavor; this document ...widely utilized radiation transport code is MCNP. First created at Los Alamos National Laboratory ( LANL ) in 1957, the code simulated neutral...explanation of the current capabilities of MCNP will occur within the next chapter of this document ; however, it is important to note that MCNP

  13. Membrane pumping technology for helium and hydrogen isotope separation in the fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pistunovich, V.I. [Kurchatov Inst., Moscow (Russian Federation). NFI RRC; Pigarov, A.Yu. [Kurchatov Inst., Moscow (Russian Federation). NFI RRC; Busnyuk, A.O. [Bonch-Bruyevich University, St. Petersburg (Russian Federation); Livshits, A.I. [Bonch-Bruyevich University, St. Petersburg (Russian Federation); Notkin, M.E. [Bonch-Bruyevich University, St. Petersburg (Russian Federation); Samartsev, A.A. [Bonch-Bruyevich University, St. Petersburg (Russian Federation); Borisenko, K.L. [Efremov Institute, St. Petersburg (Russian Federation); Darmogray, V.V. [Efremov Institute, St. Petersburg (Russian Federation); Ershov, B.D. [Efremov Institute, St. Petersburg (Russian Federation); Filippova, L.V. [Efremov Institute, St. Petersburg (Russian Federation); Mudugin, B.G. [Efremov Institute, St. Petersburg (Russian Federation); Odintsov, V.N. [Efremov Institute, St. Petersburg (Russian Federation); Saksagansky, G.L. [Efremov Institute, St. Petersburg (Russian Federation); Serebrennikov, D.V. [Efremov Institute, St. Petersburg (Russian Federation)

    1995-03-01

    A gas pumping system for ITER, improved by implementation of superpermeable membranes for selective hydrogen isotope exhaust, is considered. A study of the pumping capability of a niobium membrane for a hydrogen-helium mixture has been performed.Monte Carlo simulations of gas behaviour for the experimental facility and fusion reactor have been done.The scheme of the ITER pumping system with the membranes and membrane pumping technology was considered. The conceptual study the membrane pump for the ITER was done. This work gives good prospects for the membrane pumping use in ITER to reduce the total inventory of tritium necessary for reactor operation. (orig.).

  14. Reactor vessel

    OpenAIRE

    Makkee, M.; Kapteijn, F.; Moulijn, J.A

    1999-01-01

    A reactor vessel (1) comprises a reactor body (2) through which channels (3) are provided whose surface comprises longitudinal inwardly directed parts (4) and is provided with a catalyst (6), as well as buffer bodies (8, 12) connected to the channels (3) on both sides of the reactor body (2) and comprising connections for supplying (9, 10, 11) and discharging (13, 14, 15) via the channels (3) gases and/or liquids entering into a reaction with each other and substances formed upon this reactio...

  15. Accuracy of the electron transport in mcnp5 and its suitability for ionization chamber response simulations: A comparison with the egsnrc and penelope codes

    Energy Technology Data Exchange (ETDEWEB)

    Koivunoro, Hanna; Siiskonen, Teemu; Kotiluoto, Petri; Auterinen, Iiro; Hippelaeinen, Eero; Savolainen, Sauli [Department of Physics, University of Helsinki, P.O. Box 64, FI-00014 Helsinki University (Finland) and Department of Oncology, Helsinki University Central Hospital, FI-00029 HUS (Finland); STUK-Radiation and Nuclear Safety Authority, P.O. Box 14, FI-00881 Helsinki (Finland); VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland); Department of Physics, University of Helsinki, P.O. Box 64, FI-00014 Helsinki University (Finland); HUS Medical Imaging Centre, Helsinki University Central Hospital, FI-00029 HUS (Finland)

    2012-03-15

    Purpose: In this work, accuracy of the mcnp5 code in the electron transport calculations and its suitability for ionization chamber (IC) response simulations in photon beams are studied in comparison to egsnrc and penelope codes. Methods: The electron transport is studied by comparing the depth dose distributions in a water phantom subdivided into thin layers using incident energies (0.05, 0.1, 1, and 10 MeV) for the broad parallel electron beams. The IC response simulations are studied in water phantom in three dosimetric gas materials (air, argon, and methane based tissue equivalent gas) for photon beams ({sup 60}Co source, 6 MV linear medical accelerator, and mono-energetic 2 MeV photon source). Two optional electron transport models of mcnp5 are evaluated: the ITS-based electron energy indexing (mcnp5{sub ITS}) and the new detailed electron energy-loss straggling logic (mcnp5{sub new}). The electron substep length (ESTEP parameter) dependency in mcnp5 is investigated as well. Results: For the electron beam studies, large discrepancies (>3%) are observed between the mcnp5 dose distributions and the reference codes at 1 MeV and lower energies. The discrepancy is especially notable for 0.1 and 0.05 MeV electron beams. The boundary crossing artifacts, which are well known for the mcnp5{sub ITS}, are observed for the mcnp5{sub new} only at 0.1 and 0.05 MeV beam energies. If the excessive boundary crossing is eliminated by using single scoring cells, the mcnp5{sub ITS} provides dose distributions that agree better with the reference codes than mcnp5{sub new}. The mcnp5 dose estimates for the gas cavity agree within 1% with the reference codes, if the mcnp5{sub ITS} is applied or electron substep length is set adequately for the gas in the cavity using the mcnp5{sub new}. The mcnp5{sub new} results are found highly dependent on the chosen electron substep length and might lead up to 15% underestimation of the absorbed dose. Conclusions: Since the mcnp5 electron

  16. Some geometrical iteration methods for nonlinear equations

    Institute of Scientific and Technical Information of China (English)

    LU Xing-jiang; QIAN Chun

    2008-01-01

    This paper describes geometrical essentials of some iteration methods (e.g. Newton iteration,secant line method,etc.) for solving nonlinear equations and advances some geomet-rical methods of iteration that are flexible and efficient.

  17. The European contribution to the ITER Remote Maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Damiani, C., E-mail: carlo.damiani@f4e.europa.eu [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Annino, C.; Balagué, S.; Bates, P.; Ceccanti, F.; Di Mascio, T.; Dubus, G.; Esqué, S.; Gonzalez, C.; Lewczanin, M.; Locke, D.; Mont, L.; Olajos, K.; Ranz, R.; Shuff, R.; Puiu, A.; Van Hille, C.; Van Uffelen, M. [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Choi, C.H.; Friconneau, J.P. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); and others

    2014-10-15

    Highlights: •The article introduces the needs for remote maintenance in ITER. •It also discusses some of the issues related to the cultural transition from tokamaks as plasma physics to nuclear reactors. •It highlights the related cultural change and the implications on plant topology and maintenance. •Then, it presents those remote handling systems that will be procured by Europe. •The article emphasises the need of a major involvement of industries from now on. -- Abstract: For a first-of-a-kind nuclear fusion reactor like ITER, remote maintainability of neutron-activated components is one of the key aspects of plant design and operations, and a fundamental ingredient for the demonstration of long-term viability of fusion as energy source. The European Domestic Agency (EU DA, i.e. Fusion for Energy, F4E) is providing important support to the ITER Organisation (IO) in specifying the functional requirements of the Remote Handling (RH) Procurement Packages (i.e. the subsystems allocated to EU DA belonging to the overall ITER Remote Maintenance Systems IRMS), and in performing design and R and D activities – with the support of national laboratories and industries – in order to define a sound concept for these packages. Furthermore, domestic industries are being involved in the subsequent detailed design, validation, manufacturing and installation activities, in order to actually fulfil our procurement-in-kind obligations. After an introduction to ITER Remote Maintenance, this paper will present status and next stages for the RH systems allocated to EU DA, and will also illustrate complementary aspects related to cross cutting technologies like radiation tolerant components and RH control systems. Finally, the way all these efforts are coordinated will be presented together with the overall implementation scenario and key milestones.

  18. Fuel burnup analysis of the TRIGA Mark II Reactor at the University of Pavia

    CERN Document Server

    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...

  19. Simulation of radiation transport using MCNP for a teletherapy machine; Simulacion del transporte de radiacion usando MCNP para una maquina de teleterapia

    Energy Technology Data Exchange (ETDEWEB)

    Flores O, F.E.; Mireles G, F.; Davila R, J.I.; Pinedo V, J.L.; Risorios M, C.; Lopez del Rio, H. [UAZ, Unidad Academica de Estudios Nucleares, 98068 Zacatecas (Mexico)

    2008-07-01

    The MCNP code is used to simulate the radiation transport taking as tools the transport physics of each particle, either photon, neutron or electron, and the generation of random numbers. Developed in the Los Alamos National Laboratory, this code has been used thoroughly with great success, because the results of the simulations are broadly validated with representative experiments. In the one present work the room of radiotherapy of the Institute Zacatecano of the Tumor it is simulated, located in the city of Zacatecas where one is Theratron 780C machine manufactured by MSD Nordion, with the purpose of estimating the contribution to the dose that would be received in different points of the structure, included three directly under the source. Three results of analytical calculations for points located at different distances from the source are presented, and they are compared against those obtained by the simulation. Its are also presented results for the simulation of 10 points more distributed around the source. (Author)

  20. PICARD ITERATION FOR NONSMOOTH EQUATIONS

    Institute of Scientific and Technical Information of China (English)

    Song-bai Sheng; Hui-fu Xu

    2001-01-01

    This paper presents an analysis of the generalized Newton method, approximate Newton methods, and splitting methods for solving nonsmooth equations from Picard iteration viewpoint. It is proved that the radius of the weak Jacobian (RGJ) of Picard iteration function is equal to its least Lipschitz constant. Linear convergence or superlinear convergence results can be obtained provided that RGJ of the Picard iteration function at a solution point is less than one or equal to zero. As for applications, it is pointed out that the approximate Newton methods, the generalized Newton method for piecewise C1problems and splitting methods can be explained uniformly with the same viewpoint.

  1. Iterative optimization in inverse problems

    CERN Document Server

    Byrne, Charles L

    2014-01-01

    Iterative Optimization in Inverse Problems brings together a number of important iterative algorithms for medical imaging, optimization, and statistical estimation. It incorporates recent work that has not appeared in other books and draws on the author's considerable research in the field, including his recently developed class of SUMMA algorithms. Related to sequential unconstrained minimization methods, the SUMMA class includes a wide range of iterative algorithms well known to researchers in various areas, such as statistics and image processing. Organizing the topics from general to more

  2. Chemical Reactors.

    Science.gov (United States)

    Kenney, C. N.

    1980-01-01

    Describes a course, including content, reading list, and presentation on chemical reactors at Cambridge University, England. A brief comparison of chemical engineering education between the United States and England is also given. (JN)

  3. Reactor Neutrinos

    Directory of Open Access Journals (Sweden)

    Soo-Bong Kim

    2013-01-01

    Full Text Available We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very recently the most precise determination of the neutrino mixing angle θ13. This paper provides an overview of the upcoming experiments and of the projects under development, including the determination of the neutrino mass hierarchy and the possible use of neutrinos for society, for nonproliferation of nuclear materials, and geophysics.

  4. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  5. Reactor Engineering

    Science.gov (United States)

    Lema, Juan M.; López, Carmen; Eibes, Gemma; Taboada-Puig, Roberto; Moreira, M. Teresa; Feijoo, Gumersindo

    In this chapter, the engineering aspects of processes catalyzed by peroxidases will be presented. In particular, a discussion of the existing technologies that utilize peroxidases for different purposes, such as the removal of recalcitrant compounds or the synthesis of polymers, is analyzed. In the first section, the essential variables controlling the process will be investigated, not only those that are common in any enzymatic system but also those specific to peroxidative reactions. Next, different reactor configurations and operational modes will be proposed, emphasizing their suitability and unsuitability for different systems. Finally, two specific reactors will be described in detail: enzymatic membrane reactors and biphasic reactors. These configurations are especially valuable for the treatment of xenobiotics with high and poor water solubility, respectively.

  6. Reactor Neutrinos

    OpenAIRE

    Lasserre, T.; Sobel, H.W.

    2005-01-01

    We review the status and the results of reactor neutrino experiments, that toe the cutting edge of neutrino research. Short baseline experiments have provided the measurement of the reactor neutrino spectrum, and are still searching for important phenomena such as the neutrino magnetic moment. They could open the door to the measurement of coherent neutrino scattering in a near future. Middle and long baseline oscillation experiments at Chooz and KamLAND have played a relevant role in neutrin...

  7. Characterization of control rod worths and fuel rod power peaking factors in the university of Utah TRIGA Mark I reactor

    OpenAIRE

    Alroumi Fawaz; Kim Donghoon; Schow Ryan; Jevremovic Tatjana

    2016-01-01

    Control rod reactivity (worths) for the three control rods and fuel rod power peaking factors in the University of Utah research reactor (100 kW TRIGA Mark I) are characterized using the AGENT code system and the results described in this paper. These values are compared to the MCNP6 and existing experimental measurements. In addition, the eigenvalue, neutron spatial flux distributions and reaction rates are analyzed and discussed. The AGENT code system is ...

  8. Simulations of neutron multiplicity measurements of a weapons-grade plutonium sphere with MCNP-PoliMi.

    Energy Technology Data Exchange (ETDEWEB)

    Mattingly, John K.; Pozzi, Sara A. (University of Michigan, Ann Arbor, MI); Clarke, Shaun D. (University of Michigan, Ann Arbor, MI); Dennis, Ben D. (University of Michigan, Ann Arbor, MI); Miller, Eric C. (University of Michigan, Ann Arbor, MI); Padovani, E. (Polytechnic of Milan, Italy)

    2010-06-01

    With increasing concern over the ability to detect and characterize special nuclear materials, the need for computer codes that can successfully predict the response of detector systems to various measurement scenarios is extremely important. These computer algorithms need to be benchmarked against a variety of experimental configurations to ensure their accuracy and understand their limitations. The Monte Carlo code MCNP-PoliMi is a modified version of the MCNP-4c code. Recently these modifications have been ported into the new MCNPX 2.6.0 code, which gives the new MCNPX-PoliMi a wider variety of options and abilities, taking advantage of the improvements made to MCNPX. To verify the ability of the MCNPX-PoliMi code to simulate the response of a neutron multiplicity detector simulated results were compared to experimental data. The experiment consisted of a 4.5-kg sphere of alpha-phase plutonium that was moderated with various thicknesses of polyethylene. The results showed that our code system can simulate the multiplicity distributions with relatively good agreement with measured data. The enhancements made to MCNP since the release of MCNP-4c have had little to no effect on the ability of the MCNP-PoliMi to resolve the discrepancies observed in the simulated neutron multiplicity distributions when compared experimental data.

  9. Validation and Verification of MCNP6 Against Intermediate and High-Energy Experimental Data and Results by Other Codes

    CERN Document Server

    Mashnik, Stepan G

    2010-01-01

    MCNP6, the latest and most advanced LANL transport code representing a recent merger of MCNP5 and MCNPX, has been Validated and Verified (V&V) against a variety of intermediate and high-energy experimental data and against results by different versions of MCNPX and other codes. In the present work, we V&V MCNP6 using mainly the latest modifications of the Cascade-Exciton Model (CEM) and of the Los Alamos version of the Quark-Gluon String Model (LAQGSM) event generators CEM03.02 and LAQGSM03.03. We found that MCNP6 describes reasonably well various reactions induced by particles and nuclei at incident energies from 18 MeV to about 1 TeV per nucleon measured on thin and thick targets and agrees very well with similar results obtained with MCNPX and calculations by CEM03.02, LAQGSM03.01 (03.03), INCL4 + ABLA, and Bertini INC + Dresner evaporation, EPAX, ABRABLA, HIPSE, and AMD, used as stand alone codes. Most of several computational bugs and more serious physics problems observed in MCNP6/X during our V...

  10. Development of liquid metal type TBM technology for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Bong Guen; Kwak, J. G.; Kim, Y. (and others)

    2008-03-15

    The objectives of the ITER project for the construction and operation are to perform the test related to the neutronics, blanket module, tritium treatment technology, advanced plasma technology, and to test the heat extraction and tritium breeding in the test blanket for the fusion reactor. Other parties have been developing the Test Blanket Module (TBM) for testing in the ITER for these purposes. Through this project, we can secure the TBM design and related technology, which will be used as the core technology for the DEMO construction, our own fusion reactor development. In 1st year, the optimized design procedure was established with the existing tools, which have been used in nuclear reactor design, and the optimized HCML TBM design was obtained through iteration method according to the developed design procedure. He cooling system as a TBM auxiliary system was designed considering the final design of the KO HCML TBM such as coolant capacity and operation pressure. Layout for this system was prepared to be installed in the ITER TCWS vault. MHD effect of liquid Li breeder by magnetic flux in ITER such as much higher pressure drop was evaluated with CFD-ACE and it was concluded that the Li breeder should have a slow velocity to reduce this effect. Most results were arranged in the form of DDD including preliminary safety analysis report. In 2nd year, the optimized design procedure was complemented and updated. In performance analysis on thermal-hydraulic and thermo-mechanical one, full 3D meshes were generated and used in this analysis in order to obtain the more exact temperature, deformation, and stress solution. For liquid Li breeder system, design parameters were induced before the detailed design of the system and were used in the design of the liquid Li test loop. LOCA analysis, activation analysis in LOCA, EM analysis were performed as a preliminary safety analysis. In order to develop the manufacturing technology, Be+FMS and FMS to FMS joining conditions

  11. 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.

  12. Rollout sampling approximate policy iteration

    NARCIS (Netherlands)

    Dimitrakakis, C.; Lagoudakis, M.G.

    2008-01-01

    Several researchers have recently investigated the connection between reinforcement learning and classification. We are motivated by proposals of approximate policy iteration schemes without value functions, which focus on policy representation using classifiers and address policy learning as a

  13. Iterative solution of linear systems

    Science.gov (United States)

    Freund, Roland W.; Golub, Gene H.; Nachtigal, Noel M.

    1992-01-01

    Recent advances in the field of iterative methods for solving large linear systems are reviewed. The main focus is on developments in the area of conjugate gradient-type algorithms and Krylov subspace methods for nonHermitian matrices.

  14. Updated safety analysis of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, Neill, E-mail: neill.taylor@iter.org [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Baker, Dennis; Ciattaglia, Sergio; Cortes, Pierre; Elbez-Uzan, Joelle; Iseli, Markus; Reyes, Susana; Rodriguez-Rodrigo, Lina; Rosanvallon, Sandrine; Topilski, Leonid [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

    2011-10-15

    An updated version of the ITER Preliminary Safety Report has been produced and submitted to the licensing authorities. It is revised and expanded in response to requests from the authorities after their review of an earlier version in 2008, to reflect enhancements in ITER safety provisions through design changes, to incorporate new and improved safety analyses and to take into account other ITER design evolution. The updated analyses show that changes to the Tokamak cooling water system design have enhanced confinement and reduced potential radiological releases as well as removing decay heat with very high reliability. New and updated accident scenario analyses, together with fire and explosion risk analyses, have shown that design provisions are sufficient to minimize the likelihood of accidents and reduce potential consequences to a very low level. Taken together, the improvements provided a stronger demonstration of the very good safety performance of the ITER design.

  15. Cooperation between CERN and ITER

    CERN Multimedia

    2008-01-01

    CERN and the International Fusion Organisation ITER have just signed a first cooperation agreeement. Kaname Ikeda, the Director-General of the International Fusion Energy Organisation (ITER) (on the right) and Robert Aymar, Director-General of CERN, signing the agreement.The Director-General of the International Fusion Energy Organization, Mr Kaname Ikeda, and CERN Director-General, Robert Aymar, signed a cooperation agreement at a meeting on the Meyrin site on Thursday 6 March. One of the main purposes of this agreement is for CERN to give ITER the benefit of its experience in the field of technology as well as in administrative domains such as finance, procurement, human resources and informatics through the provision of consultancy services. Currently in its start-up phase at its Cadarache site, 70 km from Marseilles (France), ITER will focus its research on the scientific and technical feasibility of using fusion energy as a fu...

  16. MCNP/MCNPX几何栅元划分方法对精确放疗剂量计算的影响研究%Effect of Different Voxel-uniting Methods on the Dose Calculation of MCNP/MCNPX

    Institute of Scientific and Technical Information of China (English)

    赵攀; 陈义学; 林辉; 郑善良; 吴宜灿

    2006-01-01

    复杂几何模型的建立是Monte Carlo粒子输运程序MCNP/MCNPX在放疗领域广泛应用的关键与难点, 发展了基于医学CT影像的MCNP/MCNPX自动建模软件, 提出并实现了3种几何柵元划分的方法.根据临床实例数据, 分别建立了3种MCNP几何模型. 在此基础上, 研究分析了3种几何柵元划分方法及重复结构描述方法对计算结果的影响, 为MCNP/MCNPX在放疗中的应用提供基础.

  17. ITER leader to head CERN

    CERN Multimedia

    Feder, Toni

    2003-01-01

    After successfully chairing an external review committee for CERN last year, Robert Aymar will leave ITER to become director general of the European particle physics laboratory rom 2004. Before ITER he also successfully managed the startup or Tore Supra. He will attempt to ensure that the LHC begins operating in 2007 - two years late - and is paid for by 2010 and will also start the planning for life after the LHC (1 page)

  18. The ITER project construction status

    Science.gov (United States)

    Motojima, O.

    2015-10-01

    The pace of the ITER project in St Paul-lez-Durance, France is accelerating rapidly into its peak construction phase. With the completion of the B2 slab in August 2014, which will support about 400 000 metric tons of the tokamak complex structures and components, the construction is advancing on a daily basis. Magnet, vacuum vessel, cryostat, thermal shield, first wall and divertor structures are under construction or in prototype phase in the ITER member states of China, Europe, India, Japan, Korea, Russia, and the United States. Each of these member states has its own domestic agency (DA) to manage their procurements of components for ITER. Plant systems engineering is being transformed to fully integrate the tokamak and its auxiliary systems in preparation for the assembly and operations phase. CODAC, diagnostics, and the three main heating and current drive systems are also progressing, including the construction of the neutral beam test facility building in Padua, Italy. The conceptual design of the Chinese test blanket module system for ITER has been completed and those of the EU are well under way. Significant progress has been made addressing several outstanding physics issues including disruption load characterization, prediction, avoidance, and mitigation, first wall and divertor shaping, edge pedestal and SOL plasma stability, fuelling and plasma behaviour during confinement transients and W impurity transport. Further development of the ITER Research Plan has included a definition of the required plant configuration for 1st plasma and subsequent phases of ITER operation as well as the major plasma commissioning activities and the needs of the accompanying R&D program to ITER construction by the ITER parties.

  19. ITER diagnostic system: Vacuum interface

    Energy Technology Data Exchange (ETDEWEB)

    Patel, K.M., E-mail: Kaushal.Patel@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Udintsev, V.S.; Hughes, S.; Walker, C.I.; Andrew, P.; Barnsley, R.; Bertalot, L. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Drevon, J.M. [Bertin Technologies, BP 22, 13762 Aix-en Provence cedex 3 (France); Encheva, A. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Kashchuk, Y. [Institution “PROJECT CENTER ITER”, 1, Akademika Kurchatova pl., Moscow (Russian Federation); Maquet, Ph. [Bertin Technologies, BP 22, 13762 Aix-en Provence cedex 3 (France); Pearce, R.; Taylor, N.; Vayakis, G.; Walsh, M.J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France)

    2013-10-15

    Diagnostics play an essential role for the successful operation of the ITER tokamak. They provide the means to observe control and to measure plasma during the operation of ITER tokamak. The components of the diagnostic system in the ITER tokamak will be installed in the vacuum vessel, in the cryostat, in the upper, equatorial and divertor ports, in the divertor cassettes and racks, as well as in various buildings. Diagnostic components that are placed in a high radiation environment are expected to operate for the life of ITER. There are approx. 45 diagnostic systems located on ITER. Some diagnostics incorporate direct or independently pumped extensions to maintain their necessary vacuum conditions. They require a base pressure less than 10{sup −7} Pa, irrespective of plasma operation, and a leak rate of less than 10{sup −10} Pa m{sup 3} s{sup −1}. In all the cases it is essential to maintain the ITER closed fuel cycle. These directly coupled diagnostic systems are an integral part of the ITER vacuum containment and are therefore subject to the same design requirements for tritium and active gas confinement, for all normal and accidental conditions. All the diagnostics, whether or not pumped, incorporate penetration of the vacuum boundary (i.e. window assembly, vacuum feedthrough etc.) and demountable joints. Monitored guard volumes are provided for all elements of the vacuum boundary that are judged to be vulnerable by virtue of their construction, material, load specification etc. Standard arrangements are made for their construction and for the monitoring, evacuating and leak testing of these volumes. Diagnostic systems are incorporated at more than 20 ports on ITER. This paper will describe typical and particular arrangements of pumped diagnostic and monitored guard volume. The status of the diagnostic vacuum systems, which are at the start of their detailed design, will be outlined and the specific features of the vacuum systems in ports and extensions

  20. New Tools to Prepare ACE Cross-section Files for MCNP Analytic Test Problems

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Codes Group

    2016-06-17

    Monte Carlo calculations using one-group cross sections, multigroup cross sections, or simple continuous energy cross sections are often used to: (1) verify production codes against known analytical solutions, (2) verify new methods and algorithms that do not involve detailed collision physics, (3) compare Monte Carlo calculation methods with deterministic methods, and (4) teach fundamentals to students. In this work we describe 2 new tools for preparing the ACE cross-section files to be used by MCNP® for these analytic test problems, simple_ace.pl and simple_ace_mg.pl.

  1. Simulations of X-ray spectrum and HVL for mammographic equipment using MCNP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Rafael Toledo F. de; Alvarez, Matheus; Velo, Alexandre F.; Oliveira, Marcela de; Miranda, Jose Ricardo A. [Universidade Estadual Paulista Julio de mesquita Filho (UNESP), Botucatu, SP (Brazil). Inst. de Biociencias de Botucatu. Dept. de Fisica e Biofisica; Pina, Diana R. [Universidade Estadual Paulista Julio de mesquita Filho (UNESP), Botucatu, SP (Brazil). Fac. de Medicina. Dept. de Doencas Tropicais e Diagnostico por Imagem

    2012-07-01

    Full text: The main goal of mammography is early detection of breast cancer. Thus, the mammograph should be designed so that the X-ray photons are emitted within an appropriate energy range, to distinguish the normal breast tissue and cancerous tissue. The distribution of the photons amount of X-ray beam, with their respective energies, is called the spectrum. From the spectrum it is possible to estimate the quality of the X-ray beam from the Half Value Layer (HVL). Objectives: This study aims to simulate the Senographe 600T mammography unit, manufactured by General Electric (GE), using the MCNP5 Monte Carlo code, to obtain its spectrum and HVL, and compare the HVL of the simulated model with experimental data. Method: the mammography unit was simulated using a simplified model which a beam of 2x10{sup 8} electrons focuses on a Mo target angled 12 degrees, within a capsule filled with vacuum. The incident electrons were converted into photons. The capsule has a beryllium window, allowing the passage of the X-ray beam. The beam is detected by an air cylinder with 1 cm thickness placed 60 cm from the target. On the path of X-ray beam, is inserted a 0.03 mm Mo filter located 1.6 cm after the beryllium window. The space between the capsule and the detector cylinder was filled with air. The quality of X-ray beam was verified from the HVL using the MCNP5 code and the experimental method for the voltage range typically used in clinical routine (26-31 kVp). Results and discussion: the X-ray spectrum of the mammography device is satisfactorily simulated by MCNP5, showing the characteristic radiation peaks of molybdenum at 17.479 keV and 19.602 keV, the filtered spectrum generated by Bremsstrahlung, and reducing the total number of photons with the decrease in applied tension (kVp). The HVL obtained by MCNP5 and experimental measurements show a maximum difference of 5.31% (for 31 kVp). The result of both methods are within acceptable limits established by national

  2. A system of materials composition and geometry arrangement for fast neutron beam thermalization: An MCNP study

    Science.gov (United States)

    Uhlář, Radim; Alexa, Petr; Pištora, Jaromír

    2013-03-01

    Compact deuterium-tritium neutron generators emit fast neutrons (14.2 MeV) that have to be thermalized for neutron activation analysis experiments. To maximize thermal neutron flux and minimize epithermal and fast neutron fluxes across the output surface of the neutron generator facility, Monte Carlo calculations (MCNP5; Los Alamos National Laboratory) for different moderator types and widths and collimator and reflector designs have been performed. A thin lead layer close to the neutron generator as neutron multiplier followed by polyethylene moderator and surrounded by a massive lead and nickel collimator and reflector was obtained as the optimum setup.

  3. Isodose distributions and dose uniformity in the Portuguese gamma irradiation facility calculated using the MCNP code

    CERN Document Server

    Oliveira, C

    2001-01-01

    A systematic study of isodose distributions and dose uniformity in sample carriers of the Portuguese Gamma Irradiation Facility was carried out using the MCNP code. The absorbed dose rate, gamma flux per energy interval and average gamma energy were calculated. For comparison purposes, boxes filled with air and 'dummy' boxes loaded with layers of folded and crumpled newspapers to achieve a given value of density were used. The magnitude of various contributions to the total photon spectra, including source-dependent factors, irradiator structures, sample material and other origins were also calculated.

  4. Retrieval of gamma cell 220 irradiator isodose curves with MCNP simulations and experimental measurements

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, R.R.; Grynberg, S.E.; Ferreira, A.V.; Belo, L.C.M.; Squair, P.L.; Ribeiro, M.A. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Sousa, R.V.; Sebastiao, R.C.O. [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Quimica

    2010-03-15

    Gamma irradiator facilities can be used in a wide range of applications such as biological and chemical researches, sterilization of medical devices and products. Dose mapping must be performed in these equipment in order to establish plant operational parameters, as dose uniformity, source utilization efficiency and maximum and minimum dose positions. The isodoses curves are measured using dosimeters or computer simulations. This work evaluates the absorbed dose in the CDTN/CNEN Gamma Cell Irradiation Facility, using the Monte Carlo N-Particles (MCNP) code. (author)

  5. CRDIAC: Coupled Reactor Depletion Instrument with Automated Control

    Energy Technology Data Exchange (ETDEWEB)

    Steven K. Logan

    2012-08-01

    When modeling the behavior of a nuclear reactor over time, it is important to understand how the isotopes in the reactor will change, or transmute, over that time. This is especially important in the reactor fuel itself. Many nuclear physics modeling codes model how particles interact in the system, but do not model this over time. Thus, another code is used in conjunction with the nuclear physics code to accomplish this. In our code, Monte Carlo N-Particle (MCNP) codes and the Multi Reactor Transmutation Analysis Utility (MRTAU) were chosen as the codes to use. In this way, MCNP would produce the reaction rates in the different isotopes present and MRTAU would use cross sections generated from these reaction rates to determine how the mass of each isotope is lost or gained. Between these two codes, the information must be altered and edited for use. For this, a Python 2.7 script was developed to aid the user in getting the information in the correct forms. This newly developed methodology was called the Coupled Reactor Depletion Instrument with Automated Controls (CRDIAC). As is the case in any newly developed methodology for modeling of physical phenomena, CRDIAC needed to be verified against similar methodology and validated against data taken from an experiment, in our case AFIP-3. AFIP-3 was a reduced enrichment plate type fuel tested in the ATR. We verified our methodology against the MCNP Coupled with ORIGEN2 (MCWO) method and validated our work against the Post Irradiation Examination (PIE) data. When compared to MCWO, the difference in concentration of U-235 throughout Cycle 144A was about 1%. When compared to the PIE data, the average bias for end of life U-235 concentration was about 2%. These results from CRDIAC therefore agree with the MCWO and PIE data, validating and verifying CRDIAC. CRDIAC provides an alternative to using ORIGEN-based methodology, which is useful because CRDIAC's depletion code, MRTAU, uses every available isotope in its

  6. Molten salt reactor: Deterministic safety evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Merle-Lucotte, Elsa; Heuer, Daniel; Mathieu, Ludovic; Le Brun, Christian [Laboratory for Subatomic Physics and Cosmology (LPSC), 53, Avenue des Marthyrs, F-38026 Grenoble (France)

    2006-07-01

    Molten Salt Reactors (MSRs) are one of the systems retained by Generation IV as a candidate for the next generation of nuclear reactors. This type of reactor is particularly well adapted to the thorium fuel cycle (Th- {sup 233}U) which has the advantage of producing less minor actinides than the uranium-plutonium fuel cycle ({sup 238}U- {sup 239}Pu). In the frame of a major re-evaluation of the MSR concept and concentrating on some major constraints such as feasibility, breeding capability and, above all, safety, we have considered a particular reactor configuration that we call the 'unique channel' configuration in which there is no moderator in the core, leading to a quasi fast neutron spectrum. This reactor is presented in the first section. MSRs benefit from several specific advantages which are listed in a second part of this work. Beyond these advantages of the MSR, the level of the deterministic safety in such a reactor has to be assessed precisely. In a third section, we first draw up a list of the reactivity margins in our reactor configuration. We then define and quantify the parameters characterizing the deterministic safety of any reactor: the fraction of delayed neutrons, and the system's feedback coefficients that are here negative. Finally, using a simple point-kinetic evaluation, we analyze how these safety parameters impact the system when the total reactivity margins are introduced in the MSR. The results of this last study are discussed, emphasizing the satisfactory behavior of the MSR and the excellent level of deterministic safety which can be achieved. This work is based on the coupling of a neutron transport code called MCNP with a materials evolution code. The former calculates the neutron flux and the reaction rates in all the cells while the latter solves the Bateman equations for the evolution of the materials composition within the cells. These calculations take into account the input parameters (power released

  7. Preliminary Study for Inventories of Minor Actinides in Thorium Molten Salt Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Choong Wie; Kim, Hee Reyoung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2015-05-15

    It has different characteristic with the conventional reactors which use a solid fuel. It can continually supply the fuel by online refueling and reprocessing of minor actinides so that those can be separated and eliminated from the reactor. The MSR maintains steady state except initial stage and the reactor becomes stable. In this research, considering online refueling, bubbling and reprocessing, the basic concept for evaluation of the inventory of minor actinide in the molten salt reactor is driven using the Bateman equation. The simulation results, where REM and MCNP code from CNRS (Centre National de la Recherche Scientifique) applied to the concept equation are analyzed. The analysis of the basic concept was carried out for evaluation of the inventory of the minor actinides in MSR. It was thought that the inventories of the minor actinides should be evaluated by solving the modified Bateman equation due to the MSR characteristic of online refueling, chemical reprocessing and bubbling.

  8. Improvements in the simulation of the efficiency of a HPGe detector with Monte Carlo code MCNP5; Mejoras en la simulacion de la eficiencia de un detector HPGe con el codigo Monte Carlo MCNP5

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo, S.; Querol, A.; Rodenas, J.; Verdu, G.

    2014-07-01

    in this paper we propose to perform a simulation model using the MCNP5 code and a registration form meshing to improve the simulation efficiency of the detector in the range of energies ranging from 50 to 2000 keV. This meshing is built by FMESH MCNP5 registration code that allows a mesh with cells of few microns. The photon and electron flow is calculated in the different cells of the mesh which is superimposed on detector geometry. It analyzes the variation of efficiency (related to the variation of energy deposited in the active volume). (Author)

  9. Preliminary RAMI analysis of DFLL TBS for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Dagui [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230031 (China); Yuan, Run [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Wang, Jiaqun, E-mail: jiaqun.wang@fds.org.cn [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Wang, Fang; Wang, Jin [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China)

    2016-11-15

    Highlights: • We performed the functional analysis of the DFLL TBS. • We performed a failure mode analysis of the DFLL TBS. • We estimated the reliability and availability of the DFLL TBS. • The ITER RAMI approach was applied to the DFLL TBS for technical risk control in the design phase. - Abstract: ITER is the first fusion machine fully designed to prove the physics and technological basis for next fusion power plants. Among the main technical objectives of ITER is to test and validate design concepts of tritium breeding blankets relevant to the fusion power plants. To achieve this goal, China has proposed the dual functional lithium-lead test blanket module (DFLL TBM) concept design. The DFLL TBM and its associated ancillary system were called DFLL TBS. The DFLL TBS play a key role in next fusion reactor. In order to ensure reliable and available of DFLL TBS, the risk control project of DFLL TBS has been put on the schedule. As the stage of the ITER technical risk control policy, the RAMI (Reliability, Availability, Maintainability, Inspectability) approach was used to control the technical risk of ITER. In this paper, the RAMI approach was performed on the conceptual design of DFLL TBS. A functional breakdown was prepared on DFLL TBS, and the system was divided into 3 main functions and 72 basic functions. Based on the result of functional breakdown of DFLL TBS, the reliability block diagrams were prepared to estimate the reliability and availability of each function under the stipulated operating conditions. The inherent availability of the DFLL TBS expected after implementation of mitigation actions was calculated to be 98.57% over 2 years based on the ITER reliability database. A Failure Modes Effects and Criticality Analysis (FMECA) was performed with criticality charts highlighting the risk level of the different failure modes with regard to their probability of occurrence and their effects on the availability.

  10. Development of a Scale Model for High Flux Isotope Reactor Cycle 400

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Dan [ORNL

    2012-03-01

    The development of a comprehensive SCALE computational model for the High Flux Isotope Reactor (HFIR) is documented and discussed in this report. The SCALE model has equivalent features and functionality as the reference MCNP model for Cycle 400 that has been used extensively for HFIR safety analyses and for HFIR experiment design and analyses. Numerical comparisons of the SCALE and MCNP models for the multiplication constant, power density distribution in the fuel, and neutron fluxes at several locations in HFIR indicate excellent agreement between the results predicted with the two models. The SCALE HFIR model is presented in sufficient detail to provide the users of the model with a tool that can be easily customized for various safety analysis or experiment design requirements.

  11. ITER Central Solenoid Module Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Smith, John [General Atomics, San Diego, CA (United States)

    2016-09-23

    The fabrication of the modules for the ITER Central Solenoid (CS) has started in a dedicated production facility located in Poway, California, USA. The necessary tools have been designed, built, installed, and tested in the facility to enable the start of production. The current schedule has first module fabrication completed in 2017, followed by testing and subsequent shipment to ITER. The Central Solenoid is a key component of the ITER tokamak providing the inductive voltage to initiate and sustain the plasma current and to position and shape the plasma. The design of the CS has been a collaborative effort between the US ITER Project Office (US ITER), the international ITER Organization (IO) and General Atomics (GA). GA’s responsibility includes: completing the fabrication design, developing and qualifying the fabrication processes and tools, and then completing the fabrication of the seven 110 tonne CS modules. The modules will be shipped separately to the ITER site, and then stacked and aligned in the Assembly Hall prior to insertion in the core of the ITER tokamak. A dedicated facility in Poway, California, USA has been established by GA to complete the fabrication of the seven modules. Infrastructure improvements included thick reinforced concrete floors, a diesel generator for backup power, along with, cranes for moving the tooling within the facility. The fabrication process for a single module requires approximately 22 months followed by five months of testing, which includes preliminary electrical testing followed by high current (48.5 kA) tests at 4.7K. The production of the seven modules is completed in a parallel fashion through ten process stations. The process stations have been designed and built with most stations having completed testing and qualification for carrying out the required fabrication processes. The final qualification step for each process station is achieved by the successful production of a prototype coil. Fabrication of the first

  12. Calculation of Absorbed Dose in Target Tissue and Equivalent Dose in Sensitive Tissues of Patients Treated by BNCT Using MCNP4C

    Science.gov (United States)

    Zamani, M.; Kasesaz, Y.; Khalafi, H.; Pooya, S. M. Hosseini

    Boron Neutron Capture Therapy (BNCT) is used for treatment of many diseases, including brain tumors, in many medical centers. In this method, a target area (e.g., head of patient) is irradiated by some optimized and suitable neutron fields such as research nuclear reactors. Aiming at protection of healthy tissues which are located in the vicinity of irradiated tissue, and based on the ALARA principle, it is required to prevent unnecessary exposure of these vital organs. In this study, by using numerical simulation method (MCNP4C Code), the absorbed dose in target tissue and the equiavalent dose in different sensitive tissues of a patiant treated by BNCT, are calculated. For this purpose, we have used the parameters of MIRD Standard Phantom. Equiavelent dose in 11 sensitive organs, located in the vicinity of target, and total equivalent dose in whole body, have been calculated. The results show that the absorbed dose in tumor and normal tissue of brain equal to 30.35 Gy and 0.19 Gy, respectively. Also, total equivalent dose in 11 sensitive organs, other than tumor and normal tissue of brain, is equal to 14 mGy. The maximum equivalent doses in organs, other than brain and tumor, appear to the tissues of lungs and thyroid and are equal to 7.35 mSv and 3.00 mSv, respectively.

  13. Comparison of TG-43 dosimetric parameters of brachytherapy sources obtained by three different versions of MCNP codes.

    Science.gov (United States)

    Zaker, Neda; Zehtabian, Mehdi; Sina, Sedigheh; Koontz, Craig; Meigooni, Ali S

    2016-03-01

    Monte Carlo simulations are widely used for calculation of the dosimetric parameters of brachytherapy sources. MCNP4C2, MCNP5, MCNPX, EGS4, EGSnrc, PTRAN, and GEANT4 are among the most commonly used codes in this field. Each of these codes utilizes a cross-sectional library for the purpose of simulating different elements and materials with complex chemical compositions. The accuracies of the final outcomes of these simulations are very sensitive to the accuracies of the cross-sectional libraries. Several investigators have shown that inaccuracies of some of the cross section files have led to errors in  125I and  103Pd parameters. The purpose of this study is to compare the dosimetric parameters of sample brachytherapy sources, calculated with three different versions of the MCNP code - MCNP4C, MCNP5, and MCNPX. In these simulations for each source type, the source and phantom geometries, as well as the number of the photons, were kept identical, thus eliminating the possible uncertainties. The results of these investigations indicate that for low-energy sources such as  125I and  103Pd there are discrepancies in gL(r) values. Discrepancies up to 21.7% and 28% are observed between MCNP4C and other codes at a distance of 6 cm for  103Pd and 10 cm for  125I from the source, respectively. However, for higher energy sources, the discrepancies in gL(r) values are less than 1.1% for  192Ir and less than 1.2% for  137Cs between the three codes. PACS number(s): 87.56.bg.

  14. Code development incorporating environmental, safety, and economic aspects of fusion reactors (FY 89--91)

    Energy Technology Data Exchange (ETDEWEB)

    Ho, S.K.; Fowler, T.K.; Holdren, J.P. (eds.)

    1991-11-01

    This report discusses the following aspects of Fusion reactors.: Activation Analysis; Tritium Inventory; Environmental and Safety Indices and Their Graphical Representation; Probabilistic Risk Assessment (PRA) and Decision Analysis; Plasma Burn Control -- Application to ITER; and Other Applications.

  15. Benchmark Evaluation of the NRAD Reactor LEU Core Startup Measurements

    Energy Technology Data Exchange (ETDEWEB)

    J. D. Bess; T. L. Maddock; M. A. Marshall

    2011-09-01

    The Neutron Radiography (NRAD) reactor is a 250-kW TRIGA-(Training, Research, Isotope Production, General Atomics)-conversion-type reactor at the Idaho National Laboratory; it is primarily used for neutron radiography analysis of irradiated and unirradiated fuels and materials. The NRAD reactor was converted from HEU to LEU fuel with 60 fuel elements and brought critical on March 31, 2010. This configuration of the NRAD reactor has been evaluated as an acceptable benchmark experiment and is available in the 2011 editions of the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) and the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Significant effort went into precisely characterizing all aspects of the reactor core dimensions and material properties; detailed analyses of reactor parameters minimized experimental uncertainties. The largest contributors to the total benchmark uncertainty were the 234U, 236U, Er, and Hf content in the fuel; the manganese content in the stainless steel cladding; and the unknown level of water saturation in the graphite reflector blocks. A simplified benchmark model of the NRAD reactor was prepared with a keff of 1.0012 {+-} 0.0029 (1s). Monte Carlo calculations with MCNP5 and KENO-VI and various neutron cross section libraries were performed and compared with the benchmark eigenvalue for the 60-fuel-element core configuration; all calculated eigenvalues are between 0.3 and 0.8% greater than the benchmark value. Benchmark evaluations of the NRAD reactor are beneficial in understanding biases and uncertainties affecting criticality safety analyses of storage, handling, or transportation applications with LEU-Er-Zr-H fuel.

  16. Wall reflection modeling for charge exchange recombination spectroscopy (CXRS) measurements on Textor and ITER

    NARCIS (Netherlands)

    Banerjee, S.; Vasu, P.; von Hellermann, M.; Jaspers, R. J. E.

    2010-01-01

    Contamination of optical signals by reflections from the tokamak vessel wall is a matter of great concern. For machines such as ITER and future reactors, where the vessel wall will be predominantly metallic, this is potentially a risk factor for quantitative optical emission spectroscopy. This is, i

  17. Wall reflection modeling for charge exchange recombination spectroscopy (CXRS) measurements on Textor and ITER

    NARCIS (Netherlands)

    Banerjee, S.; Vasu, P.; von Hellermann, M.; Jaspers, R. J. E.

    2010-01-01

    Contamination of optical signals by reflections from the tokamak vessel wall is a matter of great concern. For machines such as ITER and future reactors, where the vessel wall will be predominantly metallic, this is potentially a risk factor for quantitative optical emission spectroscopy. This is, i

  18. Experimental Simulation of the Behaviour of Diagnostic First Mirrors Fabricated of Different Metals for ITER Conditions

    NARCIS (Netherlands)

    Voitsenya, V.; Bardamid, A. F.; Donne, A. J. H.

    2016-01-01

    In the experimental fusion reactor ITER, the plasma-facing component of each optical and/or laser diagnostic needs to be based on reflective optics with at least one mirror (first mirror) facing the thermonuclear plasma. The different kinds of radiation emanating from the burning plasma (neutrons, n

  19. Development of the superconductors for ITER magnet system

    Science.gov (United States)

    Shikov, A.; Nikulin, A.; Silaev, A.; Vorobieva, A.; Pantsyrnyi, V.; Vedernikov, G.; Salunin, N.; Sudiev, S.

    1998-10-01

    A review is given of the present status of the development and production of Nb 3Sn and Nb-Ti superconductors for the Model Coils and the real Magnet System of the International Thermonuclear Experimental Reactor (ITER) in the Russian Federation Home Team. It is shown that Nb 3Sn bronze processed superconductors produced for the Model Central Solenoid Coil insert meet the ITER joint Central Team requirements. In particular, the critical current density, measued in non-Cu area is not less than 550 A/mm 2 for 12 T at 4.2 K, the level of hysteresis losses is not in excess of 200 mJ/cm 3, and the Cu-stabilizing shell resistivity ratio of Cr-plated wire is 150. Internal tin Nb 3Sn superconductor development and test results are presented, confirming the possibility of their application for the ITER Magnet System winding. Nb-Ti superconductors for PF coils properties have also been considered. The possibility of Nb 3Sn and Nb-Ti superconductor manufacture with the use of large composite billets up to 300 mm in dia is shown, creating the possibility for large scale industrial production (several tens of tons/year) of these materials for the ITER Magnet System.

  20. Bioconversion reactor

    Science.gov (United States)

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  1. A recursion identity for formal iterated logarithms and iterated exponentials

    CERN Document Server

    Robinson, Thomas J

    2010-01-01

    We prove a recursive identity involving formal iterated logarithms and formal iterated exponentials. These iterated logarithms and exponentials appear in a natural extension of the logarithmic formal calculus used in the study of logarithmic intertwining operators and logarithmic tensor category theory for modules for a vertex operator algebra. This extension has a variety of interesting arithmetic properties. We develop one such result here, the aforementioned recursive identity. We have applied this identity elsewhere to certain formal series expansions related to a general formal Taylor theorem and these series expansions in turn yield a sequence of combinatorial identities which have as special cases certain classical combinatorial identities involving (separately) the Stirling numbers of the first and second kinds.

  2. Development of thick wall welding and cutting tools for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Nakahira, Masataka; Takahashi, Hiroyuki; Akou, Kentaro; Koizumi, Koichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    The Vacuum Vessel, which is a core component of International Thermonuclear Experimental Reactor (ITER), is required to be exchanged remotely in a case of accident such as superconducting coil failure. The in-vessel components such as blanket and divertor are planned to be exchanged or fixed. In these exchange or maintenance operations, the thick wall welding and cutting are inevitable and remote handling tools are necessary. The thick wall welding and cutting tools for blanket are under developing in the ITER R and D program. The design requirement is to weld or cut the stainless steel of 70 mm thickness in the narrow space. Tungsten inert gas (TIG) arc welding, plasma cutting and iodine laser welding/cutting are selected as primary option. Element welding and cutting tests, design of small tools to satisfy space requirement, test fabrication and performance tests were performed. This paper reports the tool design and overview of welding and cutting tests. (author)

  3. Total reaction cross sections in CEM and MCNP6 at intermediate energies

    Science.gov (United States)

    Kerby, Leslie M.; Mashnik, Stepan G.

    2015-08-01

    Accurate total reaction cross section models are important to achieving reliable predictions from spallation and transport codes. The latest version of the Cascade Exciton Model (CEM) as incorporated in the code CEM03.03, and the Monte Carlo N-Particle transport code (MCNP6), both developed at Los Alamos National Laboratory (LANL), each use such cross sections. Having accurate total reaction cross section models in the intermediate energy region (∼ 50 MeV to ∼ 5 GeV) is very important for different applications, including analysis of space environments, use in medical physics, and accelerator design, to name just a few. The current inverse cross sections used in the preequilibrium and evaporation stages of CEM are based on the Dostrovsky et al. model, published in 1959. Better cross section models are available now. Implementing better cross section models in CEM and MCNP6 should yield improved predictions for particle spectra and total production cross sections, among other results. Our current results indicate this is, in fact, the case.

  4. Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system

    Energy Technology Data Exchange (ETDEWEB)

    Iga, Kiminori [Kyushu Univ., Fukuoka (Japan); Takada, Hiroshi; Nagao, Tadashi

    1998-01-01

    In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B{sub 4}C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)

  5. Total Reaction Cross Sections in CEM and MCNP6 at Intermediate Energies

    CERN Document Server

    Kerby, Leslie M

    2015-01-01

    Accurate total reaction cross section models are important to achieving reliable predictions from spallation and transport codes. The latest version of the Cascade Exciton Model (CEM) as incorporated in the code CEM03.03, and the Monte Carlo N-Particle transport code (MCNP6), both developed at Los Alamos National Laboratory (LANL), each use such cross sections. Having accurate total reaction cross section models in the intermediate energy region ($\\sim$50 MeV to $\\sim$5 GeV) is very important for different applications, including analysis of space environments, use in medical physics, and accelerator design, to name just a few. The current inverse cross sections used in the preequilibrium and evaporation stages of CEM are based on the Dostrovsky {\\it et al.} model, published in 1959. Better cross section models are available now. Implementing better cross section models in CEM and MCNP6 should yield improved predictions for particle spectra and total production cross sections, among other results. Our current...

  6. Physics and Algorithm Enhancements for a Validated MCNP/X Monte Carlo Simulation Tool, Phase VII

    Energy Technology Data Exchange (ETDEWEB)

    McKinney, Gregg W [Los Alamos National Laboratory

    2012-07-17

    Currently the US lacks an end-to-end (i.e., source-to-detector) radiation transport simulation code with predictive capability for the broad range of DHS nuclear material detection applications. For example, gaps in the physics, along with inadequate analysis algorithms, make it difficult for Monte Carlo simulations to provide a comprehensive evaluation, design, and optimization of proposed interrogation systems. With the development and implementation of several key physics and algorithm enhancements, along with needed improvements in evaluated data and benchmark measurements, the MCNP/X Monte Carlo codes will provide designers, operators, and systems analysts with a validated tool for developing state-of-the-art active and passive detection systems. This project is currently in its seventh year (Phase VII). This presentation will review thirty enhancements that have been implemented in MCNPX over the last 3 years and were included in the 2011 release of version 2.7.0. These improvements include 12 physics enhancements, 4 source enhancements, 8 tally enhancements, and 6 other enhancements. Examples and results will be provided for each of these features. The presentation will also discuss the eight enhancements that will be migrated into MCNP6 over the upcoming year.

  7. 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.

  8. Moderator configuration options for a low-enriched uranium fueled Kilowatt-class Space Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    King, Jeffrey C., E-mail: kingjc@mines.edu [Nuclear Science and Engineering Program, Colorado School of Mines (CSM), Golden, CO (United States); Mencarini, Leonardo de Holanda; Guimaraes, Lamartine N. F., E-mail: guimaraes@ieav.cta.br, E-mail: mencarini@ieav.cta.br [Instituto de Estudos Avancados (IEAV), Sao Jose dos Campos, SP (Brazil). Divisao de Energia Nuclear

    2015-07-01

    The Brazilian Air Force, through its Institute for Advanced Studies (Instituto de Estudos Avancados, IEAv/DCTA), and the Colorado School of Mines (CSM) are studying the feasibility of a space nuclear reactor with a power of 1-5 kW{sub e} and fueled with Low-Enriched Uranium (LEU). This type of nuclear reactor would be attractive to signatory countries of the Non-Proliferation Treaty (NPT) or commercial interests. A LEU-fueled space reactor would avoid the security concerns inherent with Highly Enriched Uranium (HEU) fuel. As an initial step, the HEU-fueled Kilowatt Reactor Using Stirling Technology (KRUSTY) designed by the Los Alamos National Laboratory serves as a basis for a similar reactor fueled with LEU fuel. Using the computational code MCNP6 to predict the reactor neutronics performance, the size of the resulting reactor fueled with 19.75 wt% enriched uranium-10 wt% molybdenum alloy fuel is adjusted to match the excess reactivity of KRUSTY. Then, zirconium hydride moderator is added to the core to reduce the size of the reactor. This work presents the preliminary results of the computational modeling, with special emphasis on the comparison between homogeneous and heterogeneous moderator systems, in terms of the core diameter required to meet a specific multiplication factor (k{sub eff} = 1.035). This comparison illustrates the impact of moderator configuration on the size and performance of a LEU-fueled kilowatt-class space nuclear reactor. (author)

  9. Relaxation Criteria for Iterated Traffic Simulations

    Science.gov (United States)

    Kelly, Terence; Nagel, Kai

    Iterative transportation microsimulations adjust traveler route plans by iterating between a microsimulation and a route planner. At each iteration, the route planner adjusts individuals' route choices based on the preceding microsimulations. Empirically, this process yields good results, but it is usually unclear when to stop the iterative process when modeling real-world traffic. This paper investigates several criteria to judge relaxation of the iterative process, emphasizing criteria related to traveler decision-making.

  10. Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR

    Directory of Open Access Journals (Sweden)

    Brovchenko Mariya

    2017-01-01

    Full Text Available The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR. The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

  11. Effect of DUPIC Cycle on CANDU Reactor Safety Parameters

    Directory of Open Access Journals (Sweden)

    Nader M.A. Mohamed

    2016-10-01

    Full Text Available Although, the direct use of spent pressurized water reactor (PWR fuel in CANda Deuterium Uranium (CANDU reactors (DUPIC cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by UO2 enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1 the power distribution amongst the fuel elements of the bundle; (2 the coolant void reactivity; and (3 the reactor point-kinetics parameters.

  12. Effect of DUPIC cycle on CANDU reactor safety parameters

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, Nader M. A. [Atomic Energy Authority, ETRR-2, Cairo (Egypt); Badawi, Alya [Dept. of Nuclear and Radiation Engineering, Alexandria University, Alexandria (Egypt)

    2016-10-15

    Although, the direct use of spent pressurized water reactor (PWR) fuel in CANda Deuterium Uranium (CANDU) reactors (DUPIC) cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by UO{sub 2} enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1) the power distribution amongst the fuel elements of the bundle; (2) the coolant void reactivity; and (3) the reactor point-kinetics parameters.

  13. Time variability of α from realistic models of Oklo reactors

    Science.gov (United States)

    Gould, C. R.; Sharapov, E. I.; Lamoreaux, S. K.

    2006-08-01

    We reanalyze Oklo Sm149 data using realistic models of the natural nuclear reactors. Disagreements among recent Oklo determinations of the time evolution of α, the electromagnetic fine structure constant, are shown to be due to different reactor models, which led to different neutron spectra used in the calculations. We use known Oklo reactor epithermal spectral indices as criteria for selecting realistic reactor models. Two Oklo reactors, RZ2 and RZ10, were modeled with MCNP. The resulting neutron spectra were used to calculate the change in the Sm149 effective neutron capture cross section as a function of a possible shift in the energy of the 97.3-meV resonance. We independently deduce ancient Sm149 effective cross sections and use these values to set limits on the time variation of α. Our study resolves a contradictory situation with previous Oklo α results. Our suggested 2σ bound on a possible time variation of α over 2 billion years is stringent: -0.11≤Δα/α≤0.24, in units of 10-7, but model dependent in that it assumes only α has varied over time.

  14. Sonochemical Reactors.

    Science.gov (United States)

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation.

  15. A Linear Iterative Unfolding Method

    CERN Document Server

    Laszlo, Andras

    2011-01-01

    A frequently faced task in experimental physics is to measure the probability distribution of some quantity. Often this quantity to be measured is smeared by a non-ideal detector response or by some physical process. The procedure of removing this smearing effect from the measured distribution is called unfolding, and is a delicate problem in signal processing. Due to the numerical ill-posedness of this task, various methods were invented which, given some assumptions on the initial probability distribution, try to regularize the problem. Most of these methods definitely introduce bias on the estimate of the initial probability distribution. We propose a linear iterative method (motivated by the Neumann series / Landweber iteration known in functional analysis), which has the advantage that no assumptions on the initial probability distribution is needed, and the only regularization parameter is the stopping order of the iteration. Convergence is proved under certain quite general conditions, which hold for p...

  16. ITER CS Model Coil and CS Insert Test Results

    Energy Technology Data Exchange (ETDEWEB)

    Martovetsky, N; Michael, P; Minervina, J; Radovinsky, A; Takayasu, M; Thome, R; Ando, T; Isono, T; Kato, T; Nakajima, H; Nishijima, G; Nunoya, Y; Sugimoto, M; Takahashi, Y; Tsuji, H; Bessette, D; Okuno, K; Ricci, M

    2000-09-07

    The Inner and Outer modules of the Central Solenoid Model Coil (CSMC) were built by US and Japanese home teams in collaboration with European and Russian teams to demonstrate the feasibility of a superconducting Central Solenoid for ITER and other large tokamak reactors. The CSMC mass is about 120 t, OD is about 3.6 m and the stored energy is 640 MJ at 46 kA and peak field of 13 T. Testing of the CSMC and the CS Insert took place at Japan Atomic Energy Research Institute (JAERI) from mid March until mid August 2000. This paper presents the main results of the tests performed.

  17. Rollout Sampling Approximate Policy Iteration

    CERN Document Server

    Dimitrakakis, Christos

    2008-01-01

    Several researchers have recently investigated the connection between reinforcement learning and classification. We are motivated by proposals of approximate policy iteration schemes without value functions which focus on policy representation using classifiers and address policy learning as a supervised learning problem. This paper proposes variants of an improved policy iteration scheme which addresses the core sampling problem in evaluating a policy through simulation as a multi-armed bandit machine. The resulting algorithm offers comparable performance to the previous algorithm achieved, however, with significantly less computational effort. An order of magnitude improvement is demonstrated experimentally in two standard reinforcement learning domains: inverted pendulum and mountain-car.

  18. MCNP - transport calculations in ducts using multigroup albedo coefficients; Calculos de transporte em dutos utilizando coeficientes de albedo multigrupo no codigo MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Ono, Shizuca; Vieira, Wilson J.; Garcia, Roberto D.M. [Centro Tecnico Aeroespacial (CTA-IEAv), Sao Jose dos Campos, SP (Brazil). Inst. de Estudos Avancados

    2000-07-01

    In this work, the use of multigroup albedo coefficients in Monte Carlo calculations of particle reflection and transmission by ducts is investigated. The procedure consists in modifying the MCNP code so that an albedo matrix computed previously by deterministic methods or Monte Carlo is introduced into the program to describe particle reflection by a surface. This way it becomes possible to avoid the need of considering particle transport in the duct wall explicitly, changing the problem to a problem of transport in the duct interior only and reducing significantly the difficulty of the real problem. The probability of particle reflection at the duct wall is given, for each group, as the sum of the albedo coefficients over the final groups. The calculation is started by sampling a source particle and simulating its reflection on the duct wall by sampling a group for the emerging particle. The particle weight is then reduced by the reflection probability. Next, a new direction and trajectory for the particle is selected. Numerical results obtained for the model are compared with results from a discrete ordinates code and results from Monte Carlo simulations that take particle transport in the wall into account. (author)

  19. The First Benchmarking of ITER BR Nb3Sn Strand of CNDA

    Institute of Scientific and Technical Information of China (English)

    龙风; 刘方; 武玉; 倪志鹏

    2012-01-01

    According to the International Thermonuclear Experimental Reactor (ITER) Pro- curement Arrangement (PA) of Cable-In-Conduit Conductor (CICC) unit lengths for the Toroidal Field (TF) and Poloidal Field (PF) magnet systems of ITER, at the start of process qualification, the Domestic Agency (DA) shall be required to conduct a benchmarking of the room and low tem- perature acceptance tests carried out at the Strand Suppliers and/or at its Reference Laboratories designated by the ITER Organization (IO). The first benchmarking was carried out successfully in 2009. Nineteen participants from six DAs (China, European Union, Japan, South Korea, Russia, and the United States) participated in the first benchmarking. Bronze-route (BR) Nb3Sn strand and samples prepared by the ITER reference lab (CERN) were sent out to each participant by CERN. In this paper, the test facility and test results of the first benchmarking by the Chinese DA (CNDA) are presented.

  20. 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.

  1. Bounded Fixed-Point Iteration

    DEFF Research Database (Denmark)

    Nielson, Hanne Riis; Nielson, Flemming

    1992-01-01

    they obtain a quadratic bound. These bounds are shown to be tight. Specializing the case of strict and additive functions to functionals of a form that would correspond to iterative programs they show that a linear bound is tight. This is related to several analyses studied in the literature (including...

  2. Iterative method for interferogram processing

    Science.gov (United States)

    Kotlyar, Victor V.; Seraphimovich, P. G.; Zalyalov, Oleg K.

    1994-12-01

    We have developed and numerically evaluated an iterative algorithm for interferogram processing including the Fourier-transform method, the Gerchberg-Papoulis algorithm and Wiener's filter-based regularization used in combination. Using a signal-to-noise ratio not less than 1, it has been possible to reconstruct the phase of an object field with accuracy better than 5%.

  3. Energetic ions in ITER plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Pinches, S. D. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul-lez-Durance Cedex (France); Chapman, I. T.; Sharapov, S. E. [CCFE, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Lauber, Ph. W. [Max-Planck-Institut für Plasmaphysik, EURATOM-Association, Boltzmanstraße 2, D-85748 Garching (Germany); Oliver, H. J. C. [H H Wills Physics Laboratory, University of Bristol, Royal Fort, Tyndall Avenue, Bristol BS8 1TL (United Kingdom); CCFE, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Shinohara, K. [Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan); Tani, K. [Nippon Advanced Technology Co., Ltd, Naka, Ibaraki 311-0102 (Japan)

    2015-02-15

    This paper discusses the behaviour and consequences of the expected populations of energetic ions in ITER plasmas. It begins with a careful analytic and numerical consideration of the stability of Alfvén Eigenmodes in the ITER 15 MA baseline scenario. The stability threshold is determined by balancing the energetic ion drive against the dominant damping mechanisms and it is found that only in the outer half of the plasma (r/a>0.5) can the fast ions overcome the thermal ion Landau damping. This is in spite of the reduced numbers of alpha-particles and beam ions in this region but means that any Alfvén Eigenmode-induced redistribution is not expected to influence the fusion burn process. The influence of energetic ions upon the main global MHD phenomena expected in ITER's primary operating scenarios, including sawteeth, neoclassical tearing modes and Resistive Wall Modes, is also reviewed. Fast ion losses due to the non-axisymmetric fields arising from the finite number of toroidal field coils, the inclusion of ferromagnetic inserts, the presence of test blanket modules containing ferromagnetic material, and the fields created by the Edge Localised Mode (ELM) control coils in ITER are discussed. The greatest losses and associated heat loads onto the plasma facing components arise due to the use of the ELM control coils and come from neutral beam ions that are ionised in the plasma edge.

  4. Energetic ions in ITER plasmas

    Science.gov (United States)

    Pinches, S. D.; Chapman, I. T.; Lauber, Ph. W.; Oliver, H. J. C.; Sharapov, S. E.; Shinohara, K.; Tani, K.

    2015-02-01

    This paper discusses the behaviour and consequences of the expected populations of energetic ions in ITER plasmas. It begins with a careful analytic and numerical consideration of the stability of Alfvén Eigenmodes in the ITER 15 MA baseline scenario. The stability threshold is determined by balancing the energetic ion drive against the dominant damping mechanisms and it is found that only in the outer half of the plasma ( r / a > 0.5 ) can the fast ions overcome the thermal ion Landau damping. This is in spite of the reduced numbers of alpha-particles and beam ions in this region but means that any Alfvén Eigenmode-induced redistribution is not expected to influence the fusion burn process. The influence of energetic ions upon the main global MHD phenomena expected in ITER's primary operating scenarios, including sawteeth, neoclassical tearing modes and Resistive Wall Modes, is also reviewed. Fast ion losses due to the non-axisymmetric fields arising from the finite number of toroidal field coils, the inclusion of ferromagnetic inserts, the presence of test blanket modules containing ferromagnetic material, and the fields created by the Edge Localised Mode (ELM) control coils in ITER are discussed. The greatest losses and associated heat loads onto the plasma facing components arise due to the use of the ELM control coils and come from neutral beam ions that are ionised in the plasma edge.

  5. Iterative Specialisation of Horn Clauses

    DEFF Research Database (Denmark)

    Nielsen, Christoffer Rosenkilde; Nielson, Flemming; Nielson, Hanne Riis

    2008-01-01

    We present a generic algorithm for solving Horn clauses through iterative specialisation. The algorithm is generic in the sense that it can be instantiated with any decidable fragment of Horn clauses, resulting in a solution scheme for general Horn clauses that guarantees soundness and terminatio...

  6. Cooperation between CERN and ITER

    CERN Multimedia

    CERN Audiovisual Service

    2008-01-01

    CERN and the International Fusion Organisation ITER have just signed a first cooperation agreeement. The Director-General of the International Fusion Energy Organization, Mr Kaname Ikeda, and CERN Director-General, Robert Aymar, signed a cooperation agreement at a meeting on the Meyrin site on Thursday 6 March.

  7. Shielding optimisation of the ITER ICH&CD antenna for shutdown dose rate

    Energy Technology Data Exchange (ETDEWEB)

    Turner, Andrew, E-mail: andrew.turner@ccfe.ac.uk [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Leichtle, Dieter [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Lamalle, Philippe; Levesy, Bruno [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St., Paul-lez-Durance (France); Meunier, Lionel [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Polunovskiy, Eduard [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St., Paul-lez-Durance (France); Sartori, Roberta [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Shannon, Mark [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

    2015-10-15

    Highlights: • Neutronics analysis on the ITER ICH&CD system conducted to reduce shutdown dose rate. • Several designs for shielding the port plug gaps were modelled. • Shielding significantly reduced interspace dose rate but still exceed project requirements. • Design optimisation of the ICH port is continuing. • Significant contributions from other ports require an integrated modelling approach. - Abstract: The Ion Cyclotron Heating and Current Drive (ICH&CD) system will reside in ITER equatorial port plugs 13 and 15. Shutdown dose rates (SDDR) within the port interspace are required to be less than 100 μSv/h at 10{sup 6} s cooling. A significant contribution to the SDDR results from neutrons streaming down gaps around the port frame, and the mitigation of this streaming is the main subject of these analyses. An updated MCNP model of the antenna was created and integrated into an ITER reference model. Shielding plates were defined in the port gaps, and scoping studies conducted to assess their effectiveness in several configurations, based on which a front dog-leg arrangement was selected for high resolution 3-D activation analysis using MCR2S. It was concluded that the selected configuration reduced the SDDR from ∼500 μSv/h to 220 μSv/h but were still in excess of dose rate requirements. Approximately 30% of this was due to cross-talk from neighbouring ports. In addition, increased dose rates were observed in the port interspace along the lines of sight of the removable vacuum transmission lines. Design optimisation is continuing, however an integrated approach is needed with regard to ITER port plug design and the shielding of surrounding systems.

  8. Fission and Surface Source Iteration Scheme with Source Splitting in Domain Decomposition Monte Carlo Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Yu Gwon; Cho, Nam Zin [KAIST, Daejeon (Korea, Republic of)

    2014-10-15

    The OLG iteration scheme uses overlapping regions for each local problem solved by continuous-energy MC calculation to reduce errors in inaccurate boundary conditions (BCs) that are caused by discretization in space, energy, and angle. However, the overlapping region increases computational burdens and the discretized BCs for continuous-energy MC calculation result in an inaccurate global p-CMFD solution. On the other hand, there also have been several studies on the direct domain decomposed MC calculation where each processor simulates particles within its own domain and exchanges the particles crossing the domain boundary between processors with certain frequency. The efficiency of this method depends on the message checking frequency and the buffer size. Furthermore, it should overcome the load-imbalance problem for better parallel efficiency. Recently, fission and surface source (FSS) iteration method based on banking both fission and surface sources for the next iteration (i.e., cycle) was proposed to give exact BCs for non overlapping local problems in domain decomposition and tested in one-dimensional continuous-energy reactor problems. In this paper, the FSS iteration method is combined with a source splitting scheme to reduce the load imbalance problem and achieve global variance reduction. The performances are tested on a two dimensional continuous-energy reactor problem with domain-based parallelism and compared with the FSS iteration without source splitting. Numerical results show the improvements of the FSS iteration with source splitting. This paper describes the FSS iteration scheme in the domain decomposition method and proposes the FSS iteration combined with the source splitting based on the number of sampled sources, reducing the load-imbalance problem in domain-based parallelism and achieving global variance reduction.

  9. Active beam spectroscopy for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Hellermann, M.G. von, E-mail: mgvh@jet.u [FOM Institute Rijnhuizen, Euratom Association, 3430BE Nieuwegein (Netherlands); Barnsley, R. [ITER Organization, 13108 St.-Paul-Lez-Durance, Cadarache (France); Biel, W. [Institut fuer Energieforschung, Plasmaphysik, Forschungszentrum Juelich, Euratom Association, 52425 Juelich (Germany); Delabie, E. [FOM Institute Rijnhuizen, Euratom Association, 3430BE Nieuwegein (Netherlands); Hawkes, N. [Culham Centre for Fusion Energy, Euratom Association, Culham OX14 3DB (United Kingdom); Jaspers, R. [FOM Institute Rijnhuizen, Euratom Association, 3430BE Nieuwegein (Netherlands); Johnson, D. [Princeton Plasma Physics Laboratory, Princeton, NJ-08548 (United States); Klinkhamer, F. [TNO Science and Industry, Stieltjesweg 1, 2628CK Delft (Netherlands); Lischtschenko, O. [FOM Institute Rijnhuizen, Euratom Association, 3430BE Nieuwegein (Netherlands); Marchuk, O. [Institut fuer Energieforschung, Plasmaphysik, Forschungszentrum Juelich, Euratom Association, 52425 Juelich (Germany); Schunke, B. [ITER Organization, 13108 St.-Paul-Lez-Durance, Cadarache (France); Singh, M.J. [Institute for Plasma Research, Bhat, Gandhinagar, Gurajat 384828 (India); Snijders, B. [TNO Science and Industry, Stieltjesweg 1, 2628CK Delft (Netherlands); Summers, H.P. [Culham Centre for Fusion Energy, Euratom Association, Culham OX14 3DB (United Kingdom); Thomas, D. [ITER Organization, 13108 St.-Paul-Lez-Durance, Cadarache (France); Tugarinov, S. [TRINITI Troitsk, Moscow Region 142092 (Russian Federation); Vasu, P. [Institute for Plasma Research, Bhat, Gandhinagar, Gurajat 384828 (India)

    2010-11-11

    Since the first feasibility studies of active beam spectroscopy on ITER in 1995 the proposed diagnostic has developed into a well advanced and mature system. Substantial progress has been achieved on the physics side including comprehensive performance studies based on an advanced predictive code, which simulates active and passive features of the expected spectral ranges. The simulation has enabled detailed specifications for an optimized instrumentation and has helped to specify suitable diagnostic neutral beam parameters. Four ITER partners share presently the task of developing a suite of ITER active beam diagnostics, which make use of the two 0.5 MeV/amu 18 MW heating neutral beams and a dedicated 0.1 MeV/amu, 3.6 MW diagnostic neutral beam. The IN ITER team is responsible for the DNB development and also for beam physics related aspects of the diagnostic. The RF will be responsible for edge CXRS system covering the outer region of the plasma (1>r/a>0.4) using an equatorial observation port, and the EU will develop the core CXRS system for the very core (0ITER environment. Additionally, an essential change of the orientation of the DNB injection angle and specification of suitable blanket aperture has been made to avoid trapped particle damage to the first wall.

  10. Shut-Down Dose Rate analysis for ITER Diagnostic Equatorial and Upper Ports

    Energy Technology Data Exchange (ETDEWEB)

    Serikov, Arkady, E-mail: arkady.serikov@kit.edu [Karlsruhe Institute of Technology KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bertalot, Luciano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Fischer, Ulrich [Karlsruhe Institute of Technology KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Pitcher, Charles Spencer; Suarez, Alejandro; Udintsev, Victor S. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Weinhorst, Bastian [Karlsruhe Institute of Technology KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2014-10-15

    Highlights: •Shut-Down Dose Rate (SDDR) analysis for ITER Diagnostic Equatorial and Upper Port Plugs (EPP and UPP). •ALARA principle and minimization of SDDR are used for the optimization of the port plugs shielding. •Activation and radiation shielding analyses with the MCNP5, FISPACT-2007, D1S and R2Smesh codes. •Significance of contribution of ELM/in-vessel coils and blanket manifolds into the port SDDR is shown. •Shielding improvements for EPP, UPP, and adjacent ITER components were proposed. -- Abstract: The Shut-Down Dose Rate (SDDR) is an important criterion of radiation safety for the personnel access for maintenance operations in ITER ports after the cessation of the D-T 14 MeV neutron fusion source. Therefore, the problem of the SDDR calculations attracts the attention of fusion neutronics community because SDDR in such a large and geometrically complicated fusion device as the ITER tokamak is challenging to compute. This challenge has been faced and overcome by applying dedicated methodological approaches explained in this paper. The results of the SDDR analysis allowed us to propose several design solutions for improvement of the radiation shielding of the ITER Generic Diagnostic Equatorial and Upper Port Plugs (EPP and UPP). The SDDR analysis was focused on the interspace area located between the ITER bioshield and port plugs where the personnel access is envisaged at ∼12 days after the ITER shut-down. By this analysis the radiation streaming pathways and dominant sources of decay radiation were revealed and the methods to mitigate the streaming and subsequent activation were found. The optimization of the port shielding was targeted on minimization of the SDDR in the interspace area following the ALARA principle and taking into account the feasibility to implement proposed shielding options with the actual hardware. Among them, wrapping the EPP walls with the B{sub 4}C tiles improves the EPP shielding performance. While void around the ELM

  11. MCNP-DSP calculations of measurements with uranyl nitrate solution system

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.E. [Oak Ridge National Lab., TN (United States)

    1998-09-01

    The {sup 252}Cf-source-driven noise analysis method has been used to determine the subcriticality of various configurations of fissile materials. In the past, the application of this method was limited because point-kinetics models had to be used to interpret the data; however, with the development of the Monte Carlo code MCNP-DSP, the measurements can be analyzed using the more general Monte Carlo models. The results of the Monte carlo calculations will be dependent on the ability to model the experiment accurately and on the nuclear data used to perform the calculations. This paper presents a comparison of the measured and calculated ratio of spectral densities for a subset of measurements performed with a uranyl nitrate solution tank filled to various heights. The results presented are for calculations that were performed with both ENDF/B-IV and ENDF/B-V cross-section data sets.

  12. Mcnp calculation of neutron scatter in the Main Bay of the Chadwick Building, NPL

    Energy Technology Data Exchange (ETDEWEB)

    Naismith, O.F.; Thomas, D.J.

    1996-02-01

    The Monte Carlo neutron transport code MCNP has been used to calculate the room and air scattered neutron component at 75 cm from a radionuclide source located at the center of the low-scatter area in the Chadwick Building, Bldg. 47, at National Physical Laboratory (NPL). This is the standard distance used for calibrating personal dosemeters, and the calculation provides information for correcting the response of dosemeters to the scattered radiation. Calculations were performed for both an Am-Be and a (252)Cf source. These measurements revealed that the model used for features within the low-scatter area needs to be refined for calculating scatter at distances further from the source than 75 cm.

  13. Comparison of a laboratory spectrum of Eu-152 with results of simulation using the MCNP code

    Energy Technology Data Exchange (ETDEWEB)

    Rodenas, J. [Departamento de Ingenieria Quimica y Nuclear, Universidad Politecnica de Valencia, Apartado 22012, E-46071 Valencia (Spain); Gallardo, S. [Departamento de Ingenieria Quimica y Nuclear, Universidad Politecnica de Valencia, Apartado 22012, E-46071 Valencia (Spain)], E-mail: sergalbe@iqn.upv.es; Ortiz, J. [Laboratorio de Radiactividad Ambiental, Universidad Politecnica de Valencia, Apartado 22012, E-46071 Valencia (Spain)

    2007-09-21

    Detectors used for gamma spectrometry must be calibrated for each geometry considered in environmental radioactivity laboratories. This calibration is performed using a standard solution containing gamma emitter sources. Nevertheless, the efficiency curves obtained are periodically checked using a source such as {sup 152}Eu emitting many gamma rays that cover a wide energy range (20-1500 keV). {sup 152}Eu presents a problem because it has a lot of peaks affected by True Coincidence Summing (TCS). Two experimental measures have been performed placing the source (a Marinelli beaker) at 0 and 10 cm from the detector. Both spectra are simulated by the MCNP 4C code, where the TCS is not reproduced. Therefore, the comparison between experimental and simulated peak net areas permits one to choose the most convenient peaks to check the efficiency curves of the detector.

  14. MCNP Simulation to Hard X-Ray Emission of KSU Dense Plasma Focus Machine

    CERN Document Server

    Mohamed, Amgad E

    2015-01-01

    The MCNP program used to simulate the hard x-ray emission from KSU dense plasma focus device, an electron beam spectrum of maximum energy 100 keV was used to hit anode target. The bremsstrahlung radiation was measured using the F2 tally functions on the chamber walls and on a virtual sphere surrounding the machine, the radiation spectrum was recorded for various anode materials like tungsten, stainless steel and molybdenum. It was found that tungsten gives the best and the most intense radiation for the same electron beam. An aluminum filter of thickness 2mm and 4mm was used to cutoff the lower energy band from the x-ray spectrum. It was found that the filters achieved the mission and there is no distinct difference in between.

  15. Image enhancement using MCNP5 code and MATLAB in neutron radiography.

    Science.gov (United States)

    Tharwat, Montaser; Mohamed, Nader; Mongy, T

    2014-07-01

    This work presents a method that can be used to enhance the neutron radiography (NR) image for objects with high scattering materials like hydrogen, carbon and other light materials. This method used Monte Carlo code, MCNP5, to simulate the NR process and get the flux distribution for each pixel of the image and determines the scattered neutron distribution that caused image blur, and then uses MATLAB to subtract this scattered neutron distribution from the initial image to improve its quality. This work was performed before the commissioning of digital NR system in Jan. 2013. The MATLAB enhancement method is quite a good technique in the case of static based film neutron radiography, while in neutron imaging (NI) technique, image enhancement and quantitative measurement were efficient by using ImageJ software. The enhanced image quality and quantitative measurements were presented in this work.

  16. Using MCNP to estimate nuclear energy deposition in a cold neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Lecot, Carlos A.; Hergenreder, Daniel F.; Lovotti, Osvaldo P. [INVAP S.A., San Carlos de Bariloche (Argentina). Nuclear Projects Department. Nuclear Engineering Division

    2002-07-01

    The location of a Cold Neutron Source (CNS) implies a careful cost/benefit balance between neutron performance and heat removal capacity of the required cryogenic equipment. To justify this balance, the calculation of the total heat deposited in the device is a critical parameter. It depends on many different contributions, i.e. neutron and gamma radiation, beta decay, fission product decay gammas, among others. With minor modifications to some standard cross section sets, the Monte Carlo code MCNP offers the possibility to calculate the total heat load in a single calculation, without the utilization of intermediate calculations and/or auxiliary codes. This paper describes the methodology used to modify the cross section sets, to calculate the energy deposited in the CNS and to evaluate the cold neutron flux which is the variable used to compare performance at different locations. (author)

  17. Mechanical properties of type 316L stainless steel welded joint for ITER vacuum vessel (1). Experiment of unirradiated welded joint

    Energy Technology Data Exchange (ETDEWEB)

    Saito, Shigeru; Fukaya, Kiyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ishiyama, Shintaro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Takahashi, Hiroyuki; Koizumi, Kouichi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2001-01-01

    In design activity of ITER, the vacuum vessel (VV) is ranked as one of the most important components in core reactor from the view point of first barrier to tritium release from the reactor. The VV of ITER is designed as double walled structure so that some parts of them are not qualified in the conventional design standards. So it is necessary to prepare the new design standards to be applied them. JAERI has executed the preparation activity of the new design standards and the technical data to support them. In this study, the results of metallographic observation and mechanical properties of unirradiated type 316L stainless steel welded joint were reported. (author)

  18. Assessment of doses caused by electrons in thin layers of tissue-equivalent materials, using MCNP.

    Science.gov (United States)

    Heide, Bernd

    2013-10-01

    Absorbed doses caused by electron irradiation were calculated with Monte Carlo N-Particle transport code (MCNP) for thin layers of tissue-equivalent materials. The layers were so thin that the calculation of energy deposition was on the border of the scope of MCNP. Therefore, in this article application of three different methods of calculation of energy deposition is discussed. This was done by means of two scenarios: in the first one, electrons were emitted from the centre of a sphere of water and also recorded in that sphere; and in the second, an irradiation with the PTB Secondary Standard BSS2 was modelled, where electrons were emitted from an (90)Sr/(90)Y area source and recorded inside a cuboid phantom made of tissue-equivalent material. The speed and accuracy of the different methods were of interest. While a significant difference in accuracy was visible for one method in the first scenario, the difference in accuracy of the three methods was insignificant for the second one. Considerable differences in speed were found for both scenarios. In order to demonstrate the need for calculating the dose in thin small zones, a third scenario was constructed and simulated as well. The third scenario was nearly equal to the second one, but a pike of lead was assumed to be inside the phantom in addition. A dose enhancement (caused by the pike of lead) of ∼113 % was recorded for a thin hollow cylinder at a depth of 0.007 cm, which the basal-skin layer is referred to in particular. Dose enhancements between 68 and 88 % were found for a slab with a radius of 0.09 cm for all depths. All dose enhancements were hardly noticeable for a slab with a cross-sectional area of 1 cm(2), which is usually applied to operational radiation protection.

  19. ACTI - An MCNP Data Library for Prompt Gamma-Ray Spectroscopy.

    Energy Technology Data Exchange (ETDEWEB)

    Frankle, S. C. (Stephanie C.); Reedy, R. C. (Robert C.); Young, P. G. (Phillip Gaffney),

    2002-01-01

    Prompt gamma-ray spectroscopy is used in a wide variety of applications for determining material compositions. High-quality photon-production data from thermal-neutron capture reactions are essential for these applications. Radiation transport codes, such as MCNP{trademark}, are often used to design detector systems, determine minimum detection thresholds, etc. These transport codes rely on evaluated nuclear databases such as ENDF (Evaluated Nuclear Data File) to provide the fundamental data used in the transport calculations. Often the photon-production data from incident neutron reactions in the evaluations are of relatively poor quality. We have compiled the best experimental data for thermal-neutron capture for the naturally occurring isotopes for elements from H through Zn as well as for {sup 70,72,73,74,76}Ge, {sup 149}Sm, {sup 155,157}Gd, {sup 181}Ta and {sup 182,183,184,186}W. This compilation has been used to update the ENDF evaluations for {sup 1}H, {sup 4}He, {sup 9}Be, {sup 14}N, {sup 16}O, {sup 19}F, Na, Mg, {sup 27}Al, {sup 32}S, S, {sup 35,37}Cl, K, Ca, {sup 45}Sc, Ti, {sup 51}V, {sup 50,52,53,54}Cr, {sup 55}Mn, {sup 54,56,57,58}Fe, {sup 58,60,61,62,64}Ni, {sup 63,65}Cu and {sup 182,183,184,186}W. In addition, the inelastic cross sections and corresponding secondary-photon distributions were updated for {sup 16}O. Complete new evaluations were submitted to ENDF for {sup 35,37}Cl. This paper will discuss the evaluation effort and the production of the MCNP data library, ACTI, based on the new evaluations. Data from the ENDF evaluations for {sup 28-30}Si were also included in the ACTI library for completeness. The silicon evaluations were updated in 1997 and include the latest experimental data for radiative capture.

  20. ITERATIVE ALGORITHMS FOR DATA ASSIMILATION PROBLEMS

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    Iterative algorithms for solving the data assimilation problems are considered, based on the main and adjoint equations. Spectral properties of the control operators of the problem are studied, the iterative algorithms are justified.

  1. Existence test for asynchronous interval iterations

    DEFF Research Database (Denmark)

    Madsen, Kaj; Caprani, O.; Stauning, Ole

    1997-01-01

    In the search for regions that contain fixed points ofa real function of several variables, tests based on interval calculationscan be used to establish existence ornon-existence of fixed points in regions that are examined in the course ofthe search. The search can e.g. be performed...... as a synchronous (sequential) interval iteration:In each iteration step all components of the iterate are calculatedbased on the previous iterate. In this case it is straight forward to base simple interval existence and non-existencetests on the calculations done in each step of the iteration. The search can also...... be performed as an asynchronous (parallel) iteration: Only a few components are changed in each stepand this calculation is in general based on components from differentprevious iterates. For the asynchronous iteration it turns out thatsimple tests of existence and non-existence can be based...

  2. Measurement of tritium production rate distribution for a fusion-fission hybrid conceptual reactor

    Institute of Scientific and Technical Information of China (English)

    WANG Xin-Hua; GUO Hai-Ping; MOU Yun-Feng; ZHENG Pu; LIU Rong; YANG Xiao-Fei; YANG Jian

    2013-01-01

    A fusion-fission hybrid conceptual reactor is established.It consists of a DT neutron source and a spherical shell of depleted uranium and hydrogen lithium.The tritium production rate (TPR) distribution in the conceptual reactor was measured by DT neutrons using two sets of lithium glass detectors with different thicknesses in the hole in the vertical direction with respect to the D+ beam of the Cockcroft-Walton neutron generator in direct current mode.The measured TPR distribution is compared with the calculated results obtained by the threedimensional Monte Carlo code MCNP5 and the ENDF/B-Ⅵ data file.The discrepancy between the measured and calculated values can be attributed to the neutron data library of the hydrogen lithium lack S(α,β) thermal scattering model,so we show that a special database of low-energy and thermal neutrons should be established in the physics design of fusion-fission hybrid reactors.

  3. Test problem for thermal-hydraulics and neutronic coupled calculation fore ALFREAD reactor core

    Science.gov (United States)

    Filip, A.; Darie, G.; Saldikov, I. S.; Smirnov, A. D.; Tikhomirov, G. V.

    2017-01-01

    The beginning of a new era of nuclear reactor requires technological advances and also multiples studies. The European Liquid metal cooled Fast breeder Reactor is one of the designs for the generation IV nuclear reactor, selected by ENEA. A pioneer of its time, ELFR needs a demonstrator in order to prove the feasibility of this project and to acquire more data and experience in operating a LFR. For this reason the ALFRED project was started and it is expected to be under operation by the year 2030. This paper has the objective of analyzing the neutronic and thermohydraulics of the ALFRED core by the means of a coupled scheme. The selected code for neutronic simulation is MCNP and the selected code for thermohydraulics is ANSYS.

  4. SEU mitigation exploratory tests in a ITER related FPGA

    Energy Technology Data Exchange (ETDEWEB)

    Batista, Antonio J.N., E-mail: toquim@ipfn.tecnico.ulisboa.pt [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, 1049-001 Lisboa (Portugal); Leong, Carlos [Instituto de Engenharia de Sistemas e Computadores – Investigação e Desenvolvimento (INESC-ID), 1000-029 Lisboa (Portugal); Santos, Bruno; Fernandes, Ana [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, 1049-001 Lisboa (Portugal); Ramos, Ana Rita; Santos, Joana P.; Marques, José G. [Centro de Ciências e Tecnologias Nucleares (C2TN), Instituto Superior Técnico (IST), Universidade de Lisboa - UL, 2695-066 Bobadela (Portugal); Teixeira, Isabel C.; Teixeira, João P. [Instituto de Engenharia de Sistemas e Computadores – Investigação e Desenvolvimento (INESC-ID), 1000-029 Lisboa (Portugal); Sousa, Jorge; Gonçalves, Bruno [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, 1049-001 Lisboa (Portugal)

    2017-05-15

    Data acquisition hardware of ITER diagnostics if located in the port cells of the tokamak, as an example, will be irradiated with neutrons during the fusion reactor operation. Due to this reason the majority of the hardware containing Field Programmable Gate Arrays (FPGA) will be placed after the ITER bio-shield, such as the cubicles instrumentation room. Nevertheless, it is worth to explore real-time mitigation of soft-errors caused by neutrons radiation in ITER related FPGAs. A Virtex-6 FPGA from Xilinx (XC6VLX365T-1FFG1156C) is used on the ATCA-IO-PROCESSOR board, included in the ITER Catalog of Instrumentation & Control (I & C) products – Fast Controllers. The Virtex-6 is a re-programmable logic device where the configuration is stored in Static RAM (SRAM), the functional data is stored in dedicated Block RAM (BRAM) and the functional state logic in Flip-Flops. Single Event Upsets (SEU) due to the ionizing radiation of neutrons cause soft errors, unintended changes (bit-flips) of the logic values stored in the state elements of the FPGA. Real-time SEU monitoring and soft errors repairing, when possible, were explored in this work. An FPGA built-in Soft Error Mitigation (SEM) controller detects and corrects soft errors in the FPGA Configuration Memory (CM). BRAM based SEU sensors with Error Correction Code (ECC) detect and repair the respective BRAM contents. Real-time mitigation of SEU can increase reliability and availability of data acquisition hardware for nuclear applications. The results of the tests performed using the SEM controller and the SEU sensors are presented for a Virtex-6 FPGA (XC6VLX240T-1FFG1156C) when irradiated with neutrons from the Portuguese Research Reactor (RPI), a 1 MW nuclear fission reactor, operated by IST in the neighborhood of Lisbon. Results show that the proposed SEU mitigation technique is able to repair the majority of the detected SEU soft-errors in the FPGA memory.

  5. Final report of shielding calculations performed at ECN Petten for ITER CTA task D4/EU

    Energy Technology Data Exchange (ETDEWEB)

    Hogenbirk, A.

    1995-11-01

    In this report the final results are presented of neutronics calculations performed at ECN Petten in the framework of ITER task D4. It is shown, that self-shielding in the unresolved resonance region, which is not taken into account in MCNP, will increase the fast neutron flux in ITER shielding blanket calculations with at least 10%. A method is presented, with which an approximate calculation of the effect is possible. It is observed, that homogenisation of stainless steel and water in neutronics calculations is not allowed for accurate calculations. Calculations were performed with several sources of cross section data (EFF-1, EFF-2.4 and FENDL-1.0). It is shown, that systematic differences occur between the results from different calculations. However, a fair agreement is observed if results of EFF-2.4 and FENDL-1.0 calculations are compared. Sensitivity and uncertainty studies show, that the uncertainty in the energy integrated flux in a relevant energy range amounts to 15%. This uncertainty is mainly due to uncertainties in the total cross section and the elastic angular distribution of Fe. (orig.).

  6. Determining the effects of microsphere and surrounding material composition on {sup 90}Y dose kernels using egsnrc and mcnp5

    Energy Technology Data Exchange (ETDEWEB)

    Paxton, Adam B.; Davis, Stephen D.; DeWerd, Larry A. [Department of Medical Physics, University of Wisconsin-Madison, Madison, Wisconsin 53705 (United States); Department of Medical Physics, University of Wisconsin-Madison, Madison, Wisconsin 53705 and McGill University Health Centre, Department of Medical Physics, Montreal, Quebec H3G 1A4 (Canada); Department of Medical Physics, University of Wisconsin-Madison, Madison, Wisconsin 53705 (United States)

    2012-03-15

    Purpose: Recent advances in the imaging of {sup 90}Y using positron emission tomography (PET) and improved uncertainty in the branching ratio for the internal pair production component of {sup 90}Y decay allow for a more accurate determination of the activity distribution of {sup 90}Y microspheres within a patient. This improved activity distribution can be convolved with the dose kernel of {sup 90}Y to calculate the dose distribution within a patient. This work investigates the effects of microsphere and surrounding material composition on {sup 90}Y dose kernels using egsnrc and mcnp5 and compares the results of these two transport codes. Methods: Monte Carlo simulations were performed with egsnrc and mcnp5 to calculate the dose rate at multiple radial distances around various {sup 90}Y sources. Point source simulations were completed with mcnp5 to determine the optimal electron transport settings for this work. After determining the optimal settings, point source simulations were completed using egsnrc (user code edknrc) and mcnp5 in water and liver [as defined by the International Commission on Radiation Units and Measurements (ICRU) Report 44]. The results were compared to ICRU Report 72 reference data. Point source simulations were also completed in water with a density of 1.06 g{center_dot}cm{sup -3} to evaluate the effect of the density of the surrounding material. Glass and resin microsphere simulations were performed with average and maximum diameter and density values (based on values given in the literature) in water and in liver. The results were compared to point source simulation results using the same transport code and in the same surrounding material. All simulations had statistical uncertainties less than 1%. Results: The optimal transport settings in mcnp5 for this work included using the energy-and step-specific algorithm (DBCN 17J 2) and ESTEP set to 10. These settings were used for all subsequent simulations with mcnp5. The point source

  7. Modeling Results For the ITER Cryogenic Fore Pump. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Pfotenhauer, John M. [University of Wisconsin, Madison, WI (United States); Zhang, Dongsheng [University of Wisconsin, Madison, WI (United States)

    2014-03-31

    A numerical model characterizing the operation of a cryogenic fore-pump (CFP) for ITER has been developed at the University of Wisconsin – Madison during the period from March 15, 2011 through June 30, 2014. The purpose of the ITER-CFP is to separate hydrogen isotopes from helium gas, both making up the exhaust components from the ITER reactor. The model explicitly determines the amount of hydrogen that is captured by the supercritical-helium-cooled pump as a function of the inlet temperature of the supercritical helium, its flow rate, and the inlet conditions of the hydrogen gas flow. Furthermore the model computes the location and amount of hydrogen captured in the pump as a function of time. Throughout the model’s development, and as a calibration check for its results, it has been extensively compared with the measurements of a CFP prototype tested at Oak Ridge National Lab. The results of the model demonstrate that the quantity of captured hydrogen is very sensitive to the inlet temperature of the helium coolant on the outside of the cryopump. Furthermore, the model can be utilized to refine those tests, and suggests methods that could be incorporated in the testing to enhance the usefulness of the measured data.

  8. Analysis of the mechanical behaviour of the ITER magnet system

    Energy Technology Data Exchange (ETDEWEB)

    Jong, C.T.J.

    1996-03-01

    The International Thermonuclear Experimental Reactor (ITER) is a tokamak fusion device with the objective of demonstrating controlled ignition and an extended burn for a duration sufficient to achieve stationary conditions. The design of ITER will be based on extensive new design work supported by new physical and technological results. As part of the ITER Engineering Design Activities, the mechanical behaviour of the toroidal field coil (TF coil) system during normal operating conditions and fault conditions has to be analyzed. The displacements and/or stresses in the components must be limited to prevent mechanical failure of parts of the overall structure. These Engineering Design Activities are supported by R and D programs in the European Union. This final report describes the work carried out by ECN to develop a finite element model (FE model) of the TF-coil system which is suitable for the analysis of the mechanical behaviour and presents results obtained with this model. For the analysis of the mechanical behaviour, a large three dimensional (3D) non-linear finite element model has been developed. With this FE model a large number of load cases has been analyzed which correspond with several time points during multiple pulses. (orig./WL).

  9. Safety Analysis Results for Cryostat Ingress Accidents in ITER

    Science.gov (United States)

    Merrill, B. J.; Cadwallader, L. C.; Petti, D. A.

    1997-06-01

    Accidents involving the ingress of air, helium, or water into the cryostat of the International Thermonuclear Experimental Reactor (ITER) tokamak design have been analyzed with a modified version of the MELCOR code for the ITER Non-site Specific Safety Report (NSSR-1). The air ingress accident is the result of a postulated breach of the cryostat boundary into an adjoining room. MELCOR results for this accident demonstrate that the condensed air mass and increased heat loads are not a magnet safety concern, but that the partial vacuum in the adjoining room must be accommodated in the building design. The water ingress accident is the result of a postulated magnet arc that results in melting of a Primary Heat Transport System (PHTS) coolant pipe, discharging PHTS water and PHTS water activated corrosion products and HTO into the cryostat. MELCOR results for this accident demonstrate that the condensed water mass and increased heat loads are not a magnet safety concern, that the cryostat pressure remains below design limits, and that the corrosion product and HTO releases are well within the ITER release limits.

  10. An Exploration of Advanced X-Divertors on ITER

    CERN Document Server

    Covele, Brent; Kotschenreuther, Mike; Mahajan, Swadesh

    2013-01-01

    It is found that the X-Divertor (XD) configuration [1-3] can be made with the conventional PF coil set on ITER[4], where all PF coils are outside the TF coils. Desirable configurations are possible where the PF currents are below the present maximum design limits on ITER, and where the baseline divertor cassette is used. It is possible that the XD could be used to assist in high-power operation on ITER, but some further issues need examination. Note that the increased major radius of the Super X-Divertor (SXD) [5-8] is not a feature of the XD geometry. In addition, we present an XD configuration for K-DEMO [9], to demonstrate that it is also possible to attain the XD configuration in advanced tokamak reactors with all PF coils outside the TF coils. The results given here for the XD are far more encouraging than recent calculations by Lackner and Zohm [10] for the Snowflake [11,12], where the required high PF currents represent a major technological challenge. The magnetic field structure in the outboard diver...

  11. REMARK ON STABILITY OF ISHIKAWA ITERATIVE PROCEDURES

    Institute of Scientific and Technical Information of China (English)

    薛志群; 田虹

    2002-01-01

    The stability of the Ishikawa iteration procedures was studied for one class ofcontinuity strong pseudocontraction and continuity strongly accretive operators with boundedrange in real uniformly smooth Banach space. Under parameters satisfying certainconditions, the convergence of iterative sequences was proved. The results improve andextend the recent corresponding results, and supply the basis of theory for further discussingconvergence of iteration procedures with errors.

  12. On One-Point Iterations and DIIS

    DEFF Research Database (Denmark)

    Østerby, Ole; Sørensen, Hans Henrik Brandenborg

    2009-01-01

    We analyze various iteration procedures in many dimensions inspired by the SCF iteration used in first principles electronic structure calculations. We show that the simple mixing of densities can turn a divergent (or slowly convergent) iteration into a (faster) convergent process provided all th...

  13. An Iterative Rejection Sampling Method

    CERN Document Server

    Sherstnev, A

    2008-01-01

    In the note we consider an iterative generalisation of the rejection sampling method. In high energy physics, this sampling is frequently used for event generation, i.e. preparation of phase space points distributed according to a matrix element squared $|M|^2$ for a scattering process. In many realistic cases $|M|^2$ is a complicated multi-dimensional function, so, the standard von Neumann procedure has quite low efficiency, even if an error reducing technique, like VEGAS, is applied. As a result of that, many of the $|M|^2$ calculations go to ``waste''. The considered iterative modification of the procedure can extract more ``unweighted'' events, i.e. distributed according to $|M|^2$. In several simple examples we show practical benefits of the technique and obtain more events than the standard von Neumann method, without any extra calculations of $|M|^2$.

  14. ITER LHCD plans and design

    Energy Technology Data Exchange (ETDEWEB)

    Bibet, Ph.; Beaumont, B.; Delpech, L.; Ekedahl, A.; Kazarian, F.; Litaudon, X.; Prou, M. [CEA Cadarache, Dept. de Recherches sur la Fusion Controlee (DRFC), 13 - Saint-Paul-lez-Durance (France); Belo, J.H.; Bizarro, J.P.S. [Centro de Fusao Nuclear, Associacao Euratom-IST, Instituto Superior Tecnico, Lisboa (Portugal); Granucci, G. [Associazione EURATOM-ENEA sulla Fusione, Milano (Italy); Kuzikov, S. [Institute of Applied Physics, Nizhny Novgorod (Russian Federation); Mailloux, J. [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon (United Kingdom); Mirizzi, F.; Pericoli, V.; Tuccillo, A.A. [Association Euratom-ENEA sulla Fusione, Centro Ricerche Energia Frascati (Italy); Rantamaki, K. [Association Euratom-Tekes, VTT (Finland)

    2005-07-01

    LH waves experimentally exhibit the highest Current Drive efficiency at low plasma temperature, therefore they are the most suitable candidates for controlling the current profile in the off axis part of ITER Steady State plasmas. For this purpose, a 5 GHz, 20 MW CW LH system has been designed, that relies on a generator made of 24 klystrons, 1 MW each, 60 metres long circular oversized transmission lines, one antenna, based on the Passive Active Multi-function (PAM) concept. High reliability of the launcher is achieved, by limiting the power density to 33 MW/m{sup 2}. Together with the overall system description, the present results achieved toward ITER are presented. The different ongoing project are listed. The remaining outstanding problems are depicted. (authors)

  15. Development of an automated core model for nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mosteller, R.D.

    1998-12-31

    This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The objective of this project was to develop an automated package of computer codes that can model the steady-state behavior of nuclear-reactor cores of various designs. As an added benefit, data produced for steady-state analysis also can be used as input to the TRAC transient-analysis code for subsequent safety analysis of the reactor at any point in its operating lifetime. The basic capability to perform steady-state reactor-core analysis already existed in the combination of the HELIOS lattice-physics code and the NESTLE advanced nodal code. In this project, the automated package was completed by (1) obtaining cross-section libraries for HELIOS, (2) validating HELIOS by comparing its predictions to results from critical experiments and from the MCNP Monte Carlo code, (3) validating NESTLE by comparing its predictions to results from numerical benchmarks and to measured data from operating reactors, and (4) developing a linkage code to transform HELIOS output into NESTLE input.

  16. Matlab modeling of ITER CODAC

    Energy Technology Data Exchange (ETDEWEB)

    Pangione, L. [Associazione Euratom/ENEA Ssulla Fusione, Centro Ricerche Frascati, CP 65, 00044 Frascati, Roma (Italy)], E-mail: pangione@frascati.enea.it; Lister, J.B. [CRPP-EPFL, Association EURATOM-Suisse, Station 13, 1015 Lausanne (Switzerland)

    2008-04-15

    The ITER CODAC (COntrol, Data Access and Communication) conceptual design resulted from 2 years of activity. One result was a proposed functional partitioning of CODAC into different CODAC Systems, each of them partitioned into other CODAC Systems. Considering the large size of this project, simple use of human language assisted by figures would certainly be ineffective in creating an unambiguous description of all interactions and all relations between these Systems. Moreover, the underlying design is resident in the mind of the designers, who must consider all possible situations that could happen to each system. There is therefore a need to model the whole of CODAC with a clear and preferably graphical method, which allows the designers to verify the correctness and the consistency of their project. The aim of this paper is to describe the work started on ITER CODAC modeling using Matlab/Simulink. The main feature of this tool is the possibility of having a simple, graphical, intuitive representation of a complex system and ultimately to run a numerical simulation of it. Using Matlab/Simulink, each CODAC System was represented in a graphical and intuitive form with its relations and interactions through the definition of a small number of simple rules. In a Simulink diagram, each system was represented as a 'black box', both containing, and connected to, a number of other systems. In this way it is possible to move vertically between systems on different levels, to show the relation of membership, or horizontally to analyse the information exchange between systems at the same level. This process can be iterated, starting from a global diagram, in which only CODAC appears with the Plant Systems and the external sites, and going deeper down to the mathematical model of each CODAC system. The Matlab/Simulink features for simulating the whole top diagram encourage us to develop the idea of completing the functionalities of all systems in order to finally

  17. Iterative Goal Refinement for Robotics

    Science.gov (United States)

    2014-06-01

    Iterative Goal Refinement for Robotics Mark Roberts1, Swaroop Vattam1, Ronald Alford2, Bryan Auslander3, Justin Karneeb3, Matthew Molineaux3... robotics researchers and practitioners. We present a goal lifecycle and define a formal model for GR that (1) relates distinct disciplines concerning...researchers to collaborate in exploring this exciting frontier. 1. Introduction Robotic systems often act using incomplete models in environments

  18. Truncated States Obtained by Iteration

    Institute of Scientific and Technical Information of China (English)

    W.B.Cardoso; N.G.de Almeida

    2008-01-01

    We introduce the concept of truncated states obtained via iterative processes(TSI)and study its statistical features,making an analogy with dynamical systems theory(DST).As a specific example,we have studied TSI for the doubring and the logistic functions,which are standard functions in studying chaos.TSI for both the doubling and logistic functions exhibit certain similar patterns when their statistical features are compared from the point of view of DST.

  19. Filter/moderator system for a BNCT beam of epithermal neutrons at nuclear reactor MARIA

    Science.gov (United States)

    Tyminska, Katarzyna

    2009-01-01

    Boron Neutron Capture Therapy is a very promising form of cancer therapy, consisting in irradiating a stable isotope of boron (10B) concentrated in tumor cells with a low energy neutron beam. This technique makes it possible to destroy tumor cells, leaving healthy tissues practically unaffected. In order to carry out the therapy in the proper way, the proper range of the neutron beam energy has to be chosen. In this paper we present a filter/moderator system modeled with MCNP code in order to obtain an epithermal neutron beam for BNCT post at MARIA reactor in Swierk.

  20. Monte Carlo optimisation of a BNCT facility for treating brain gliomas at the TAPIRO reactor.

    Science.gov (United States)

    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').

  1. An alternative experimental approach for subcritical configurations of the IPEN/MB-01 nuclear reactor

    Science.gov (United States)

    Gonnelli, E.; Lee, S. M.; Pinto, L. N.; Landim, H. R.; Diniz, R.; Jerez, R.; dos Santos, A.

    2015-07-01

    This work presents an alternative approach for the reactivity worth experiments analysis in the IPEN/MB-01 reactor considering highly subcritical arrays. In order to reach the subcritical levels, the removal of a specific number of fuel rods is proposed. Twenty three configurations were carried out for this purpose. The control bank insertion experiment was used only as reference for the fuel rod experiment and, in addition, the control banks were maintained completely withdrawn during all the fuel rods experiment. The theoretical simulation results using the MCNP5 code and the ENDF/B-VII.0 library neutron data are in a very good agreement to experimental results.

  2. Pressure tube creep impact on the physics parameters for CANDU-6 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, W. Y.; Min, B. J. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Kam, S. C.; Kim, M. E. [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    2004-07-01

    The lattice cell calculations are performed to assess the sensitivity of the reactor physics parameters to pressure tube creep resulting from radiation aging. The physics parameters of the lattice cell are calculated by using WIMSD-5B code, WIMS- AECL code, and MCNP code. The reference model(normal state) and two perturbed models accounting for the pressure tube creep are developed on the basis of CANDU-6 lattice cell. The 2.5% and 5% values of pressure tube diameter creep are considered. Also, The effects of the analyzed lattice parameters which are the coolant void reactivity, the fuel fission density and the atom density of Pu isotopes on the lattice.

  3. Review of past experiments at the FELIX facility and future plans for ITER applications

    Energy Technology Data Exchange (ETDEWEB)

    Hua, T.Q.; Turner, L.R.

    1993-10-01

    FELIX is an experimental test facility at Argonne National Laboratory (ANL) for the study of electromagnetic effects in first wall, blanket, shield systems of fusion reactors. From 1983 to 1986 five major test series, including static and dynamic tests, were conducted and are reviewed in this paper. The dynamic tests demonstrated an important coupling effect between eddy currents and motion in a conducting structure. Recently the US has proposed to the ITER Joint Central Team to use FELIX for testing mock-up components to study electromagnetic effects encountered during plasma disruptions and other off-normal events. The near and long term plans for ITER applications are discussed.

  4. Code development incorporating environmental, safety, and economic aspects of fusion reactors (FY 89--91). Final report

    Energy Technology Data Exchange (ETDEWEB)

    Ho, S.K.; Fowler, T.K.; Holdren, J.P. [eds.

    1991-11-01

    This report discusses the following aspects of Fusion reactors.: Activation Analysis; Tritium Inventory; Environmental and Safety Indices and Their Graphical Representation; Probabilistic Risk Assessment (PRA) and Decision Analysis; Plasma Burn Control -- Application to ITER; and Other Applications.

  5. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

    2011-01-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  6. D and DR Reactors

    Data.gov (United States)

    Federal Laboratory Consortium — The world's second full-scale nuclear reactor was the D Reactor at Hanford which was built in the early 1940's and went operational in December of 1944.D Reactor ran...

  7. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  8. Determination of dosimetric parameters for 125I seed source using MCNP5 and EGSnrc MC codes%MCNP5与EGSnrc比较计算125I种子源剂量参数

    Institute of Scientific and Technical Information of China (English)

    曹振; 阮锡超; 孟贝蒂; 石翠燕

    2014-01-01

    根据AAPM TG43U1的推荐,使用MCNP5与EGSnrc两种蒙特卡罗程序计算6711型125I种子源剂量计算参数,并将两者计算结果和AAPM推荐值比较,得到相对偏差结果如下:剂量率常数,MCNP5为0.62%,EGSnrc为2.07%;径向剂量函数,MNCP5为0.15%-5.12%,EGSnrc为0%-2.18%.两者计算结果均与推荐值符合得很好,而EGSnrc的计算结果更具优势.

  9. Contributions to safety studies for new concepts of nuclear reactors; Contributions aux etudes de surete pour des filieres innovantes de reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Perdu, F

    2003-12-01

    The complete study of molten salt reactors, designed for a massive and durable nuclear energy production, must include neutronics, hydraulics and thermal effects. This coupled study, using the MCNP and Trio{sub U} codes, is undertaken in the case of the MSRE (molten salt reactor experiment) prototype. The obtained results fit very well the experiment. Their extrapolation suggests ways of improving the safety coefficients of power molten salt reactors. A second part is devoted to accelerator driven subcritical reactors, developed to incinerate radioactive waste.We propose a method to measure the prompt reactivity from the decay following a neutron pulse. It relies only on the distribution of times between generations, which is a characteristic of the reactor. This method is implemented on the results of the MUSE 4 experiment, and the obtained reactivity is accurate within 5%. (author)

  10. Kinetic Parameter Measurements in the MINERVE Reactor

    Science.gov (United States)

    Perret, Grégory; Geslot, Benoit; Gruel, Adrien; Blaise, Patrick; Di-Salvo, Jacques; De Izarra, Grégoire; Jammes, Christian; Hursin, Mathieu; Pautz, Andréas

    2017-01-01

    In the framework of an international collaboration, teams of the PSI and CEA research institutes measure the critical decay constant (α0 = β/A), delayed neutron fraction (β) and generation time (A) of the Minerve reactor using the Feynman-α, Power Spectral Density and Rossi-α neutron noise measurement techniques. These measurements contribute to the experimental database of kinetic parameters used to improve nuclear data files and validate modern methods in Monte Carlo codes. Minerve is a zero-power pool reactor composed of a central experimental test lattice surrounded by a large aluminum buffer and four high-enriched driver regions. Measurements are performed in three slightly subcritical configurations (-2 cents to -30 cents) using two high-efficiency 235U fission chambers in the driver regions. Measurement of α0 and β obtained by the two institutes and with the different techniques are consistent for the configurations envisaged. Slight increases of the β values are observed with the subcriticality level. Best estimate values are obtained with the Cross-Power Spectral Density technique at -2 cents, and are worth: β = 716.9±9.0 pcm, α0 = 79.0±0.6 s-1 and A = 90.7±1.4 μs. The kinetic parameters are predicted with MCNP5-v1.6 and TRIPOLI4.9 and the JEFF-3.1/3.1.1 and ENDF/B-VII.1 nuclear data libraries. The predictions for β and α0 overestimate the experimental results by 3-5% and 10-12%, respectively; that for A underestimate the experimental result by 6-7%. The discrepancies are suspected to come from the driven system nature of Minerve and the location of the detectors in the driver regions, which prevent accounting for the full reactor.

  11. Iterative methods for mixed finite element equations

    Science.gov (United States)

    Nakazawa, S.; Nagtegaal, J. C.; Zienkiewicz, O. C.

    1985-01-01

    Iterative strategies for the solution of indefinite system of equations arising from the mixed finite element method are investigated in this paper with application to linear and nonlinear problems in solid and structural mechanics. The augmented Hu-Washizu form is derived, which is then utilized to construct a family of iterative algorithms using the displacement method as the preconditioner. Two types of iterative algorithms are implemented. Those are: constant metric iterations which does not involve the update of preconditioner; variable metric iterations, in which the inverse of the preconditioning matrix is updated. A series of numerical experiments is conducted to evaluate the numerical performance with application to linear and nonlinear model problems.

  12. Design and analysis of a single stage to orbit nuclear thermal rocket reactor engine

    Energy Technology Data Exchange (ETDEWEB)

    Labib, Satira, E-mail: Satira.Labib@duke-energy.com; King, Jeffrey, E-mail: kingjc@mines.edu

    2015-06-15

    Graphical abstract: - Highlights: • Three NTR reactors are optimized for the single stage launch of 1–15 MT payloads. • The proposed rocket engines have specific impulses in excess of 700 s. • Reactivity and submersion criticality requirements are satisfied for each reactor. - Abstract: Recent advances in the development of high power density fuel materials have renewed interest in nuclear thermal rockets (NTRs) as a viable propulsion technology for future space exploration. This paper describes the design of three NTR reactor engines designed for the single stage to orbit launch of payloads from 1 to 15 metric tons. Thermal hydraulic and rocket engine analyses indicate that the proposed rocket engines are able to reach specific impulses in excess of 800 s. Neutronics analyses performed using MCNP5 demonstrate that the hot excess reactivity, shutdown margin, and submersion criticality requirements are satisfied for each NTR reactor. The reactors each consist of a 40 cm diameter core packed with hexagonal tungsten cermet fuel elements. The core is surrounded by radial and axial beryllium reflectors and eight boron carbide control drums. The 40 cm long reactor meets the submersion criticality requirements (a shutdown margin of at least $1 subcritical in all submersion scenarios) with no further modifications. The 80 and 120 cm long reactors include small amounts of gadolinium nitride as a spectral shift absorber to keep them subcritical upon submersion in seawater or wet sand following a launch abort.

  13. Simulation tools and new developments of the molten salt fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Merle-Lucotte, E.; Doligez, X.; Heuer, D.; Allibert, M.; Ghetta, V. [LPSC-IN2P3-CNRS / UJF / Grenoble INP, 53 avenue des Martyrs, F-38026 Grenoble Cedex (France)

    2010-07-01

    Starting from the Molten Salt Breeder Reactor project of Oak-Ridge, we have performed parametric studies in terms of safety coefficients, reprocessing requirements and breeding capabilities. In the frame of this major re-evaluation of the molten salt reactor (MSR), we have developed a new concept called Molten Salt Fast Reactor or MSFR, based on the Thorium fuel cycle and a fast neutron spectrum. This concept has been selected for further studies by the MSR steering committee of the Generation IV International Forum in 2009. Our reactor's studies of the MSFR concept rely on numerical simulations making use of the MCNP neutron transport code coupled with a code for materials evolution which resolves the Bateman's equations giving the population of each nucleus inside each part of the reactor at each moment. Because of MSR's fundamental characteristics compared to classical solid-fuelled reactors, the classical Bateman equations have to be modified by adding two terms representing the reprocessing capacities and the fertile or fissile alimentation. We have thus coupled neutronic and reprocessing simulation codes in a numerical tool used to calculate the extraction efficiencies of fission products, their location in the whole system (reactor and reprocessing unit) and radioprotection issues. (authors)

  14. Meromorphic iterative roots of linear fractional functions

    Institute of Scientific and Technical Information of China (English)

    SHI YongGuo; CHEN Li

    2009-01-01

    Iterative root problem can be regarded as a weak version of the problem of embedding a homeomorphism into a flow. There are many results on iterative roots of monotone functions. However, this problem gets more difficult in non-monotone cases. Therefore, it is interesting to find iterative roots of linear fractional functions (abbreviated as LFFs), a class of non-monotone functions on R. In this paper, iterative roots of LFFs are studied on C. An equivalence between the iterative functional equation for non-constant LFFs and the matrix equation is given. By means of a method of finding matrix roots, general formulae of all meromorphic iterative roots of LFFs are obtained and the precise number of roots is also determined in various cases. As applications, we present all meromorphic iterative roots for functions z and 1/z.

  15. Performance bounds for Lambda Policy Iteration

    CERN Document Server

    Scherrer, Bruno

    2007-01-01

    We consider the discrete-time infinite-horizon discounted stationary optimal control problem formalized by Markov Decision Processes. We study Lambda Policy Iteration, a family of algorithms parameterized by lambda, originally introduced by Ioffe and Bertsekas. Lambda Policy Iteration generalizes the standard algorithms Value Iteration and Policy Iteration, and has some connections with TD(Lambda) introduced by Sutton & Barto. We deepen the original theory developped by Ioffe and Bertsekas by providing convergence rate bounds which generalize standard bounds for Value Iteration described for instance by Puterman. We also develop the theory of this algorithm when it is used in an approximate form. Doing so, we extend and unify the separate analyses developped by Munos for Approximate Value Iteration and Approximate Policy Iteration.

  16. Development of an interface between MCNP and ORIGEN codes for calculations of fuel evolution in nuclear systems. Initial project; Desenvolvimento de uma interface entre os codigos MCNP e ORIGEN para calculos de evolucao de combustiveis em sistemas nucleares. Projeto inicial

    Energy Technology Data Exchange (ETDEWEB)

    Campolina, Daniel de Almeida Magalhaes

    2009-07-01

    In Many situations of nuclear system study, it is necessary to know the detailed particle flux in a geometry. Deterministic 1-D and 2-D methods aren't suitable to represent some strong 3-D behavior configurations, for example in cores where the neutron flux varies considerably in the space and Monte Carlo analysis are necessary. The majority of Monte Carlo transport calculation codes, performs time static simulations, in terms of fuel isotopic composition. This work is a initial project to incorporate depletion capability to the MCNP code, by means of a connection with ORIGEN2.1 burnup code. The method to develop the program proposed followed the methodology of other programs used to the same purpose. Essentially, MCNP data library are used to generate one group microscopic cross sections that override default ORIGEN libraries. To verify the actual implemented part, comparisons which MCNPX (version 2.6.0) results were made. The neutron flux and criticality value of core agree. The neutron flux and criticality value of the core agree, especially in beginning of burnup when the influence of fission products are not very considerable. The small difference encountered was probably caused by the difference in the number of isotopes considered in the transport models (89 MCNPX x 25 GB). Next step of this work is to adapt MCNP version 4C to work with a memory higher than its standard value (4MB), in order to allow a greater number of isotopes in the transport model. (author)

  17. Reactor physics methods, models, and applications used to support the conceptual design of the Advanced Neutron Source

    Energy Technology Data Exchange (ETDEWEB)

    Gehin, J.C.; Worley, B.A.; Renier, J.P. [Oak Ridge National Lab., TN (United States); Wemple, C.A.; Jahshan, S.N.; Ryskammp, J.M. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

    1995-08-01

    This report summarizes the neutronics analysis performed during 1991 and 1992 in support of characterization of the conceptual design of the Advanced Neutron Source (ANS). The methods used in the analysis, parametric studies, and key results supporting the design and safety evaluations of the conceptual design are presented. The analysis approach used during the conceptual design phase followed the same approach used in early ANS evaluations: (1) a strong reliance on Monte Carlo theory for beginning-of-cycle reactor performance calculations and (2) a reliance on few-group diffusion theory for reactor fuel cycle analysis and for evaluation of reactor performance at specific time steps over the fuel cycle. The Monte Carlo analysis was carried out using the MCNP continuous-energy code, and the few- group diffusion theory calculations were performed using the VENTURE and PDQ code systems. The MCNP code was used primarily for its capability to model the reflector components in realistic geometries as well as the inherent circumvention of cross-section processing requirements and use of energy-collapsed cross sections. The MCNP code was used for evaluations of reflector component reactivity effects and of heat loads in these components. The code was also used as a benchmark comparison against the diffusion-theory estimates of key reactor parameters such as region fluxes, control rod worths, reactivity coefficients, and material worths. The VENTURE and PDQ codes were used to provide independent evaluations of burnup effects, power distributions, and small perturbation worths. The performance and safety calculations performed over the subject time period are summarized, and key results are provided. The key results include flux and power distributions over the fuel cycle, silicon production rates, fuel burnup rates, component reactivities, control rod worths, component heat loads, shutdown reactivity margins, reactivity coefficients, and isotope production rates.

  18. Uncertainty analysis in the simulation of X-ray spectra in the diagnostic range using the MCNP5 code.

    Science.gov (United States)

    Gallardo, S; Querol, A; Ródenas, J; Verdú, G

    2011-01-01

    An accurate knowledge of the photonic spectra emitted by X-ray tubes in radiodiagnostics is essential to better estimate the imparted dose to patients and to improve the image quality obtained with these devices. In this work, several X-ray spectra have been simulated using the MCNP5 code to simulate X-ray production in a commercial device. To validate the Monte Carlo results, simulated spectra have been compared to those extracted from the IPEM 78 database. The uncertainty associated to some geometrical features of the tube and its effect on the simulated spectra has been analyzed using the Noether-Wilks formula. This analysis has been focused on the thickness of collimators, filters, shielding and barrel shutter. Furthermore, results show that the uncertainty due to geometrical parameters (0.98% in terms of Root Mean Squared) is higher than the statistical uncertainty associated to the MCNP5 calculations.

  19. ITER neutral beam system US conceptual design

    Energy Technology Data Exchange (ETDEWEB)

    Purgalis, P.

    1990-09-01

    In this document we present the US conceptual design of a neutral beam system for International Thermonuclear Experimental Reactor (ITER). The design incorporates a barium surface conversion D{sup {minus}} source feeding a linear array of accelerator channels. The system uses a dc accelerator with electrostatic quadrupoles for strong focusing. A high voltage power supply that is integrated with the accelerator is presented as an attractive option. A gas neutralizer is used and residual ions exiting the neutralizer are deflected to water-cooled dumps. Cryopanels are located at the accelerator exit to pump excess gas from the source and the neutralizer, and in the ion dump cavity to pump re-neutralized ions and neutralizer gas. All the above components are packaged in compact identical, independent modules which can be removed for remote maintenance. The neutral beam system delivers 75 MW of DO at 1.3 MeV, into three ports with a total of 9 modules arranged in stacks of three modules per port . To increase reliability each module is designed to deliver up to 10 MW; this allows eight modules operating at partial capacity to deliver the required power in the event one module is out of service, and provides 20% excess capacity to improve availability. Radiation protection is provided by shielding and by locating critical components in the source and accelerator 46.5 m from the torus centerline. Neutron shielding in the drift duct and neutralizer provides the added feature of limiting conductance and thus reducing gas flow to and from the torus.

  20. Using MCNP6 to Estimate Fission Neutron Properties of a Reflected Plutonium Sphere

    Energy Technology Data Exchange (ETDEWEB)

    Clark, Alexander Rich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Nelson, Mark Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hutchinson, Jesson D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-08-08

    The purpose of this project was to determine the fission multiplicity distribution, p(v), for the Beryllium Reflected Plutonium (BeRP) ball and to determine whether or not it changed appreciably for various High Density Polyethylene (HDPE) reflected configurations. The motivation for this project was to determine whether or not the average number of neutrons emitted per fission, v, changed significantly enough to reduce the discrepancy between MCNP6 and Robba, Dowdy, Atwater (RDA) point kinetic model estimates of multiplication. The energy spectrum of neutrons that induced fissions in the BeRP ball, NIF (E), was also computed in order to determine the average energy of neutrons inducing fissions, NIF . p(v) was computed using the FMULT card, NIF (E) and NIF were computed using an F4 tally with an FM tally modifier (F4/FM) card, and the multiplication factor, keff, was computed using the KCODE card. Although NIF (E) changed significantly between bare and HDPE reflected configurations of the BeRP ball, the change in p(v), and thus the change in v, was insignificant. This is likely due to a difference between the way that NIF is computed using the FMULT and F4/FM cards. The F4/FM card indicated that NIF (E) was essentially Watt-fission distributed for a bare configuration and highly thermalized for all HDPE reflected configurations, while the FMULT card returned an average energy between 1 and 2 MeV for all configurations, which would indicate that the spectrum is Watt-fission distributed, regardless of the amount of HDPE reflector. The spectrum computed with the F4/FM cards is more physically meaningful and so the discrepancy between it and the FMULT card result is being investigated. It is hoped that resolving the discrepancy between the FMULT and F4/FM card estimates of NIF(E) will provide better v estimates that will lead to RDA multiplication estimates that are in better agreement with MCNP6 simulations.

  1. MCNP6 model of the University of Washington clinical neutron therapy system (CNTS)

    Science.gov (United States)

    Moffitt, Gregory B.; Stewart, Robert D.; Sandison, George A.; Goorley, John T.; Argento, David C.; Jevremovic, Tatjana

    2016-01-01

    A MCNP6 dosimetry model is presented for the Clinical Neutron Therapy System (CNTS) at the University of Washington. In the CNTS, fast neutrons are generated by a 50.5 MeV proton beam incident on a 10.5 mm thick Be target. The production, scattering and absorption of neutrons, photons, and other particles are explicitly tracked throughout the key components of the CNTS, including the target, primary collimator, flattening filter, monitor unit ionization chamber, and multi-leaf collimator. Simulations of the open field tissue maximum ratio (TMR), percentage depth dose profiles, and lateral dose profiles in a 40 cm  ×  40 cm  ×  40 cm water phantom are in good agreement with ionization chamber measurements. For a nominal 10  ×  10 field, the measured and calculated TMR values for depths of 1.5 cm, 5 cm, 10 cm, and 20 cm (compared to the dose at 1.7 cm) are within 0.22%, 2.23%, 4.30%, and 6.27%, respectively. For the three field sizes studied, 2.8 cm  ×  2.8 cm, 10.4 cm  ×  10.3 cm, and 28.8 cm  ×  28.8 cm, a gamma test comparing the measured and simulated percent depth dose curves have pass rates of 96.4%, 100.0%, and 78.6% (depth from 1.5 to 15 cm), respectively, using a 3% or 3 mm agreement criterion. At a representative depth of 10 cm, simulated lateral dose profiles have in-field (⩾10% of central axis dose) pass rates of 89.7% (2.8 cm  ×  2.8 cm), 89.6% (10.4 cm  ×  10.3 cm), and 100.0% (28.8 cm  ×  28.8 cm) using a 3% and 3 mm criterion. The MCNP6 model of the CNTS meets the minimum requirements for use as a quality assurance tool for treatment planning and provides useful insights and information to aid in the advancement of fast neutron therapy.

  2. A CONCEPTUAL DESIGN OF NEUTRON COLLIMATOR IN THE THERMAL COLUMN OF KARTINI RESEARCH REACTOR FOR IN VITRO AND IN VIVO TEST OF BORON NEUTRON CAPTURE THERAPY

    OpenAIRE

    Nina Fauziah; Andang Widiharto; Yohannes Sardjono

    2015-01-01

    Studies were carried out to design a collimator which results in epithermal neutron beam for IN VITRO and IN VIVO of Boron Neutron Capture Therapy (BNCT) at the Kartini research reactor by means of Monte Carlo N-Particle (MCNP) codes. Reactor within 100 kW of thermal power was used as the neutron source. The design criteria were based on recommendation from the International Atomic Energy Agency (IAEA). All materials used were varied in size, according to the value of mean free path for each ...

  3. The application of the Monte-Carlo neutron transport code MCNP to a small "nuclear battery" system

    OpenAIRE

    Puigdellívol Sadurní, Roger

    2009-01-01

    The project consist in calculate the keff to a small nuclear battery. The code Monte- Carlo neutron transport code MCNP is used to calculate the keff. The calculations are done at the beginning of life to know the capacity of the core becomes critical in different conditions. These conditions are the study parameters that determine the criticality of the core. These parameters are the uranium enrichment, the coated particles (TRISO) packing factor and the size of the core. More...

  4. Nuclear reactor neutron shielding

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-09-12

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

  5. Reactor and method of operation

    Science.gov (United States)

    Wheeler, John A.

    1976-08-10

    A nuclear reactor having a flattened reactor activity curve across the reactor includes fuel extending over a lesser portion of the fuel channels in the central portion of the reactor than in the remainder of the reactor.

  6. Cryogenic instrumentation for ITER magnets

    Science.gov (United States)

    Poncet, J.-M.; Manzagol, J.; Attard, A.; André, J.; Bizel-Bizellot, L.; Bonnay, P.; Ercolani, E.; Luchier, N.; Girard, A.; Clayton, N.; Devred, A.; Huygen, S.; Journeaux, J.-Y.

    2017-02-01

    Accurate measurements of the helium flowrate and of the temperature of the ITER magnets is of fundamental importance to make sure that the magnets operate under well controlled and reliable conditions, and to allow suitable helium flow distribution in the magnets through the helium piping. Therefore, the temperature and flow rate measurements shall be reliable and accurate. In this paper, we present the thermometric chains as well as the venturi flow meters installed in the ITER magnets and their helium piping. The presented thermometric block design is based on the design developed by CERN for the LHC, which has been further optimized via thermal simulations carried out by CEA. The electronic part of the thermometric chain was entirely developed by the CEA and will be presented in detail: it is based on a lock-in measurement and small signal amplification, and also provides a web interface and software to an industrial PLC. This measuring device provides a reliable, accurate, electromagnetically immune, and fast (up to 100 Hz bandwidth) system for resistive temperature sensors between a few ohms to 100 kΩ. The flowmeters (venturi type) which make up part of the helium mass flow measurement chain have been completely designed, and manufacturing is on-going. The behaviour of the helium gas has been studied in detailed thanks to ANSYS CFX software in order to obtain the same differential pressure for all types of flowmeters. Measurement uncertainties have been estimated and the influence of input parameters has been studied. Mechanical calculations have been performed to guarantee the mechanical strength of the venturis required for pressure equipment operating in nuclear environment. In order to complete the helium mass flow measurement chain, different technologies of absolute and differential pressure sensors have been tested in an applied magnetic field to identify equipment compatible with the ITER environment.

  7. Computational thermo-fluid exploratory design analysis for complex ITER first wall/shield components

    Energy Technology Data Exchange (ETDEWEB)

    Youchison, Dennis L. [Sandia National Laboratories, Albuquerque, NM 87185 (United States)], E-mail: dlyouch@sandia.gov; Natoni, Greg [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Narula, Manmeet; Ying, Alice [University of California, Los Angeles, CA 90095 (United States)

    2008-12-15

    Engineers in the ITER US Party Team used several computational fluid dynamics codes to evaluate design concepts for the ITER first wall panels and the neutron shield modules. The CFdesign code enabled them to perform design studies of modules 7 and 13 very efficiently. CFdesign provides a direct interface to the CAD program, CATIA v5. The geometry input and meshing are greatly simplified. CFdesign is a finite element code, rather than a finite volume code. Flow experiments and finite volume calculations from SC-Tetra, Fluent and CFD2000 verified the CFdesign results. Several new enhancements allow CFdesign to export temperatures, pressures and convective heat transfer coefficients to other finite element models for further analysis. For example, these loads and boundary conditions directly feed into codes such as ABAQUS to perform stress analysis. In this article, we review the use of 2- and 4-mm flow driver gaps in the shield modules and the use of 1-mm gaps along the tee-vane in the front water header to obtain a good flow distribution in both the first wall and shield modules for 7 and 13. Plasma heat flux as well as neutron heating derived from MCNP calculations is included in the first wall and shield module analyses. We reveal the non-uniformity of the convective heat transfer coefficient inside complex 3D geometries exposed to a one-sided heat flux and non-uniform volumetric heating. Most models consisted of 3-4 million tetrahedron elements. We obtained temperature and velocity distributions, as well as pressure drop information, for models of nearly exact geometry compared to the CATIA fabrication models. We also describe the coupling to thermal stress analysis in ABAQUS. The results presented provide confidence that the preliminary design of these plasma facing components will meet ITER requirements.

  8. MCNP{trademark} simulations for identifying environmental contaminants using prompt gamma-rays from thermal neutron capture reactions

    Energy Technology Data Exchange (ETDEWEB)

    Frankle, S.C.; Conaway, J.G.

    1996-12-31

    The primary purposes of the Multispectral Neutron Logging Project, (MSN Project, funded by the U.S. Department of Energy), were to assess the effectiveness of existing neutron- induced spectral gamma-ray logging techniques for identifying environmental contaminants along boreholes, to further improve the technology, and to transfer that technology to industry. Using a pulsed neutron source with a high-resolution gamma-ray detector, spectra from thermal neutron capture reactions may be used to identify contaminants in the borehole environment. Direct borehole measurements such as this complement physical sampling and are useful in environmental restoration projects where characterization of contaminated sites is required and long-term monitoring may be needed for many years following cleanup or stabilization. In the MSN Project, a prototype logging instrument was designed which incorporated a pulsed 14-MeV neutron source and HPGe detector. Experimental measurements to determine minimum detection thresholds with the prototype instrument were conducted in the variable-contaminant test model for Cl, Cd, Sm, Gd, and Hg. We benchmarked an enhanced version of the Monte Carlo N-Particle computer code MCNP{trademark} using experimental data for Cl provide by our collaborators and experimental data from the variable-contaminant test model. MCNP was then used to estimate detection thresholds for the other contaminants used in the variable-contaminant model with the goal of validating the use of MCNP to estimate detection thresholds for many other contaminants that were not measured.

  9. Use of a Boron Doped Spherical Phantom for the Investigation of Neutron Directional Properties: Comparison Between Experiment and MCNP Simulation

    Energy Technology Data Exchange (ETDEWEB)

    Drake, P.; Kierkegaard, J

    1999-07-01

    A boron doped 19 cm diameter spherical phantom was constructed to give information on the direction of neutrons inside the Ringhals 4 containment. The phantom was made of 40% paraffin and 60% boric acid. 10B contributes 2% of the total phantom weight. The phantom was tested for its angular sensitivity to neutrons. The response was tested with a {sup 252}Cf source and with a Monte Carlo calculation (MCNP) simulating a {sup 252}Cf source. In these investigations the phantom showed a strong directional response. However, there was only a fair correspondence between the experiment and the simulation. The discrepancies are, at least in part, due to the difference in energy and angular response of the dosemeters as compared with the idealised response characteristics in the MCNP calculation. In the MCNP calculation the experimental conditions were not fully simulated. The investigations also showed that the addition of boron to the phantom reduces the leakage of thermalised neutrons from the phantom, and the production of neutron induced photons in the phantom to insignificant levels. (author)

  10. Beryllium in the ITER blanket

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.C.

    1995-01-01

    This paper consists of viewgraphs used in a presentation on the application of beryllium in breeding blankets for ITER and JET. The paper brings together data on the physical, thermal, mechanical, and chemical properties of beryllium and beryllium oxide for this type of application, as well as issues of compatibility with construction materials, and irradiation experience. It includes the results from testing programs carried out to arrive at some of the information, including fabrication work, irradiation experiments, and sample tests performed both in and out of the irradiation piles.

  11. Development of laser-based techniques for in situ characterization of the first wall in ITER and future fusion devices

    NARCIS (Netherlands)

    Philipps, V.; Malaquias, A.; Hakola, A.; Karhunen, J.; Maddaluno, G.; Almaviva, S.; Caneve, L.; Colao, F.; Fortuna, E.; Gasior, P.; Kubkowska, M.; Czarnecka, A.; Laan, M.; Lissovski, A.; Paris, P.; van der Meiden, H. J.; Petersson, P.; Rubel, M.; Huber, A.; Zlobinski, M.; Schweer, B.; Gierse, N.; Xiao, Q.; Sergienko, G.

    2013-01-01

    Analysis and understanding of wall erosion, material transport and fuel retention are among the most important tasks for ITER and future devices, since these questions determine largely the lifetime and availability of the fusion reactor. These data are also of extreme value to improve the understan

  12. Development of laser-based techniques for in situ characterization of the first wall in ITER and future fusion devices

    NARCIS (Netherlands)

    Philipps, V.; Malaquias, A.; Hakola, A.; Karhunen, J.; Maddaluno, G.; Almaviva, S.; Caneve, L.; Colao, F.; Fortuna, E.; Gasior, P.; Kubkowska, M.; Czarnecka, A.; Laan, M.; Lissovski, A.; Paris, P.; van der Meiden, H. J.; Petersson, P.; Rubel, M.; Huber, A.; Zlobinski, M.; Schweer, B.; Gierse, N.; Xiao, Q.; Sergienko, G.

    2013-01-01

    Analysis and understanding of wall erosion, material transport and fuel retention are among the most important tasks for ITER and future devices, since these questions determine largely the lifetime and availability of the fusion reactor. These data are also of extreme value to improve the

  13. IHadoop: Asynchronous iterations for MapReduce

    KAUST Repository

    Elnikety, Eslam Mohamed Ibrahim

    2011-11-01

    MapReduce is a distributed programming frame-work designed to ease the development of scalable data-intensive applications for large clusters of commodity machines. Most machine learning and data mining applications involve iterative computations over large datasets, such as the Web hyperlink structures and social network graphs. Yet, the MapReduce model does not efficiently support this important class of applications. The architecture of MapReduce, most critically its dataflow techniques and task scheduling, is completely unaware of the nature of iterative applications; tasks are scheduled according to a policy that optimizes the execution for a single iteration which wastes bandwidth, I/O, and CPU cycles when compared with an optimal execution for a consecutive set of iterations. This work presents iHadoop, a modified MapReduce model, and an associated implementation, optimized for iterative computations. The iHadoop model schedules iterations asynchronously. It connects the output of one iteration to the next, allowing both to process their data concurrently. iHadoop\\'s task scheduler exploits inter-iteration data locality by scheduling tasks that exhibit a producer/consumer relation on the same physical machine allowing a fast local data transfer. For those iterative applications that require satisfying certain criteria before termination, iHadoop runs the check concurrently during the execution of the subsequent iteration to further reduce the application\\'s latency. This paper also describes our implementation of the iHadoop model, and evaluates its performance against Hadoop, the widely used open source implementation of MapReduce. Experiments using different data analysis applications over real-world and synthetic datasets show that iHadoop performs better than Hadoop for iterative algorithms, reducing execution time of iterative applications by 25% on average. Furthermore, integrating iHadoop with HaLoop, a variant Hadoop implementation that caches

  14. Nonlinear Projective-Iteration Methods for Solving Transport Problems on Regular and Unstructured Grids

    Energy Technology Data Exchange (ETDEWEB)

    Dmitriy Y. Anistratov; Adrian Constantinescu; Loren Roberts; William Wieselquist

    2007-04-30

    This is a project in the field of fundamental research on numerical methods for solving the particle transport equation. Numerous practical problems require to use unstructured meshes, for example, detailed nuclear reactor assembly-level calculations, large-scale reactor core calculations, radiative hydrodynamics problems, where the mesh is determined by hydrodynamic processes, and well-logging problems in which the media structure has very complicated geometry. Currently this is an area of very active research in numerical transport theory. main issues in developing numerical methods for solving the transport equation are the accuracy of the numerical solution and effectiveness of iteration procedure. The problem in case of unstructured grids is that it is very difficult to derive an iteration algorithm that will be unconditionally stable.

  15. Irradiation tests of ITER candidate Hall sensors using two types of neutron spectra.

    Science.gov (United States)

    Ďuran, I; Bolshakova, I; Viererbl, L; Sentkerestiová, J; Holyaka, R; Lahodová, Z; Bém, P

    2010-10-01

    We report on irradiation tests of InSb based Hall sensors at two irradiation facilities with two distinct types of neutron spectra. One was a fission reactor neutron spectrum with a significant presence of thermal neutrons, while another one was purely fast neutron field. Total neutron fluence of the order of 10(16) cm(-2) was accumulated in both cases, leading to significant drop of Hall sensor sensitivity in case of fission reactor spectrum, while stable performance was observed at purely fast neutron spectrum. This finding suggests that performance of this particular type of Hall sensors is governed dominantly by transmutation. Additionally, it further stresses the need to test ITER candidate Hall sensors under neutron flux with ITER relevant spectrum.

  16. Numerical tools for Molten salt reactor simulation

    Energy Technology Data Exchange (ETDEWEB)

    Doligez, X.; Heuer, D.; Merle-Lucotte, E.; Allibert, M.; Ghetta, V. [LPSC-IN2P3-CNRS/Universite Joseph Fourier/Grenoble-INP, 53 Avenue des Martyrs, 38026 Grenoble Cedex (France)

    2009-06-15

    Molten salt reactors (MSR) are basically different from other reactors mainly because the fuel is liquid. It flows through the core, pipes, pumps and heat exchangers. Previous studies showed that a particular configuration of a molten salt reactor perfectly fulfils criteria chosen by the Generation 4 International Forum (GIF). This configuration, called non-moderated Thorium Molten Salt Reactor is a 1000 GW electrical thorium cycle based molten salt reactor with no moderator inside the core. Consequently, the neutron spectrum is fast. The reactor is coupled with a salt control unit, which complicates the studies. Reactors simulation is based on resolving Bateman's equations, which give the population of each nucleus inside the core at each moment. Because of MSR's fundamental characteristics, those equations have to be modified adding two terms: a fertile/fissile alimentation for the reactivity and the salt composition control, and the reprocessing associated term. Equations become: {delta}N{sub i}/{delta}t = {sigma}{sub j{ne}}{sub i} {lambda}{sub j{yields}}{sub i} N{sub j} + X{sub j} <{sigma}{sub j}{phi}> N{sub j} - {lambda}{sub i}N{sub i} - <{sigma}{sub i}{phi}> N{sub i} {lambda}{sub chem} N{sub i} + A where {lambda}{sub chem} represents the reprocessing capacities and A represents the fertile/fissile alimentation. All our studies are made with a homemade code, REM, which is a precision driven code for material evolution. Neutron flux and neutron reactions rate are calculated thanks MCNP and the temporal integration is made thanks a Runge-Kutta fourth order method. This code REM, whose calculation scheme will be described in the paper, does not allow a coupling flexible enough between the reprocessing and the core physics. Indeed, reprocessing terms in the previous equation ({lambda}{sub chem}) are set for the whole evolution that can last several hundreds of years. A new way is to drive chemical needs to keep the core critical. Therefore, we are

  17. Human eye analytical and mesh-geometry models for ophthalmic dosimetry using MCNP6

    Energy Technology Data Exchange (ETDEWEB)

    Angelocci, Lucas V.; Fonseca, Gabriel P.; Yoriyaz, Helio, E-mail: hyoriyaz@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    Eye tumors can be treated with brachytherapy using Co-60 plaques, I-125 seeds, among others materials. The human eye has regions particularly vulnerable to ionizing radiation (e.g. crystalline) and dosimetry for this region must be taken carefully. A mathematical model was proposed in the past [1] for the eye anatomy to be used in Monte Carlo simulations to account for dose distribution in ophthalmic brachytherapy. The model includes the description for internal structures of the eye that were not treated in previous works. The aim of this present work was to develop a new eye model based on the Mesh geometries of the MCNP6 code. The methodology utilized the ABAQUS/CAE (Simulia 3DS) software to build the Mesh geometry. For this work, an ophthalmic applicator containing up to 24 model Amersham 6711 I-125 seeds (Oncoseed) was used, positioned in contact with a generic tumor defined analytically inside the eye. The absorbed dose in eye structures like cornea, sclera, choroid, retina, vitreous body, lens, optical nerve and optical nerve wall were calculated using both models: analytical and MESH. (author)

  18. MCNP modeling of a neutron generator and its shielding at Missouri University of Science and Technology

    Science.gov (United States)

    Sharma, Manish K.; Alajo, Ayodeji Babatunde; Liu, Xin

    2014-12-01

    The shielding of a neutron generator producing fast neutrons should be sufficient to limit the dose rates to the prescribed values. A deuterium-deuterium neutron generator has been installed in the Nuclear Engineering Department at Missouri University of Science and Technology (Missouri S&T). The generator produces fast neutrons with an approximate energy of 2.5 MeV. The generator is currently shielded with different materials like lead, high-density polyethylene, and borated polyethylene. An MCNP transport simulation has been performed to estimate the dose rates at various places in and around the facility. The simulations incorporated the geometric and composition information of these shielding materials to determine neutron and photon dose rates at three central planes passing through the neutron source. Neutron and photon dose rate contour plots at these planes were provided using a MATLAB program. Furthermore, the maximum dose rates in the vicinity of the facility were used to estimate the annual limit for the generator's hours of operation. A successful operation of this generator will provide a convenient neutron source for basic and applied research at the Nuclear Engineering Department of Missouri S&T.

  19. Simulation of the BNCT of Brain Tumors Using MCNP Code: Beam Designing and Dose Evaluation

    Directory of Open Access Journals (Sweden)

    Fatemeh Sadat Rasouli

    2012-09-01

    Full Text Available Introduction BNCT is an effective method to destroy brain tumoral cells while sparing the healthy tissues. The recommended flux for epithermal neutrons is 109 n/cm2s, which has the most effectiveness on deep-seated tumors. In this paper, it is indicated that using D-T neutron source and optimizing of Beam Shaping Assembly (BSA leads to treating brain tumors in a reasonable time where all IAEA recommended criteria are met. Materials and Methods The proposed BSA based on a D-T neutron generator consists of a neutron multiplier system, moderators, reflector, and collimator. The simulated Snyder head phantom is used to evaluate dose profiles in tissues due to the irradiation of designed beam. Monte Carlo Code, MCNP-4C, was used in order to perform these calculations.   Results The neutron beam associated with the designed and optimized BSA has an adequate epithermal flux at the beam port and neutron and gamma contaminations are removed as much as possible. Moreover, it was showed that increasing J/Φ, as a measure of beam directionality, leads to improvement of beam performance and survival of healthy tissues surrounding the tumor. Conclusion According to the simulation results, the proposed system based on D-T neutron source, which is suitable for in-hospital installations, satisfies all in-air parameters. Moreover, depth-dose curves investigate proper performance of designed beam in tissues. The results are comparable with the performances of other facilities.

  20. Simulation of dental intensifying screen for intraoral radiographic using MCNP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, Vanessa M.; Oliveira, Renato C.M., E-mail: vanessamachado@ufmg.br [Curso Superior de Tecnologia em Radiologia. Faculdade de Medicina da Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil); Barros, Graiciany P.; Oliveira, Arno H.; Veloso, M. Auxiliadora F. [Departamento de Engenharia Nuclear. Escola de Engenharia. Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil)

    2011-07-01

    One of basic principles for radiological protection is the optimization of techniques for obtain radiographic images, in way that the dose in the patient is kept as low as reasonably achievable (ALARA). Intensifying screens are used in medical radiology, which reduce considerably the dose rates in the production of radiographic images, maintaining the quality of these, while in dental radiology, there is no a intensifying screen available for intraoral examinations. From this technological requirement, this paper evaluates a computational modeling of an intensifying screen for use in intraoral radiography. For this, it was used the Monte Carlo code MCNP5 that allows the radiography simulation through the transport of electrons and photons in the different materials present in this examination. The goal of an intensifying screen is the conversion of X-ray photons to photons in the visible spectrum, knowing that radiographic films are more sensitive to light photons than to X-ray photons. So the screen should be composed of an efficient material for converting x-rays photons in light photons, therefore was made simulations using different materials, thicknesses and positions possible for placing screen in radiographic film in order to find the way more technically feasible. (author)

  1. Comparative dosimetry of prostate brachytherapy with I-125 and Pd-103 seeds via SISCODES/MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Trindade, Bruno Machado; Falcao, Patricia Lima, E-mail: bmtrindade@yahoo.com [Nucleo de Radiacoes Ionizantes - Universidade Federal de Minas Gerais (NRI/UFMG), Belo Horizonte, MG (Brazil); Christovao, Marilia Tavares [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Trindade, Daniela de Fatima Maia [Centro Universitario Una, Belo Horizonte, MG (Brazil); Campos, Tarcisio Passos Ribeiro de [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)

    2012-09-15

    Objective: The present paper is aimed at presenting a comparative dosimetric study of prostate brachytherapy with I-125 and Pd-103 seeds. Materials and Methods: A protocol for both implants with 148 seeds was simulated on a heterogeneous three-dimensional pelvic phantom by means of the SISCODES/MCNP5 codes. Dose-volume histograms on prostate, rectum and bladder, dose indexes D10, D30, D90, D0.5cc, D2cc and D7cc, and representations of the spatial dose distribution were evaluated. Results: For a D90 index equivalent to the prescription dose, the initial activity of each I-125 seed was calculated as 0.42 mCi and of Pd-103 as 0.94 mCi. The maximum dose on the urethra was 90% and 108% of the prescription dose for I-125 and Pd-103, respectively. The D2cc for I-125 was 30 Gy on the rectum and 127 Gy on the bladder; for Pd-103 was 29 Gy on the rectum and 189 Gy on the bladder. The D10 on the pubic bone was 144 Gy for I-125 and 66 Gy for Pd-103. Conclusion: The results indicate that Pd-103 and I-125 implants could deposit the prescribed dose on the target volume. Among the findings of the present study, there is an excessive radiation exposure of the pelvic bones, particularly with the I-125 protocol. (author)

  2. Production of Energetic Light Fragments in CEM, LAQGSM, and MCNP6

    CERN Document Server

    Mashnik, Stepan G; Gudima, Konstantin K; Sierk, Arnold J; Bull, Jeffrey S; James, Michael R

    2016-01-01

    We extend the cascade-exciton model (CEM), and the Los Alamos version of the quark-gluon string model (LAQGSM), event generators of the Monte-Carlo N-particle transport code version 6 (MCNP6), to describe production of energetic light fragments (LF) heavier than 4He from various nuclear reactions induced by particles and nuclei at energies up to about 1 TeV/nucleon. In these models, energetic LF can be produced via Fermi break-up, preequilibrium emission, and coalescence of cascade particles. Initially, we study several variations of the Fermi break-up model and choose the best option for these models. Then, we extend the modified exciton model (MEM) used by these codes to account for a possibility of multiple emission of up to 66 types of particles and LF (up to 28Mg) at the preequilibrium stage of reactions. Then, we expand the coalescence model to allow coalescence of LF from nucleons emitted at the intranuclear cascade stage of reactions and from lighter clusters, up to fragments with mass numbers A < ...

  3. A Validated MCNP(X) Cross Section Library based on JEFF 3.1

    Energy Technology Data Exchange (ETDEWEB)

    Haeck, W.; Verboomen, B.

    2006-10-15

    ALEPH-LIB is a multi-temperature neutron transport library for standard use by MCNP(X) and ALEPH generated with ALEPH-DLG. This is an auxiliary computer code to ALEPH, the Monte Carlo burn-up code under development at SCK-CEN in collaboration with Ghent university. ALEPH-DLG automates the entire process of generating library files with NJOY and takes care of the first requirement of a validated application library: verify the processing. It produces tailor made NJOY input files using data from the original ENDF file (initial temperature, the fact if the nuclide is fissile or if it has unresolved resonances, etc.) When the library files have been generated, ALEPH-DLG will also process the output from NJOY by extracting all messages and warnings. If ALEPH-DLG finds anything out of the ordinary, it will either warn the user or perform corrective actions. The temperatures included in the ALEPH-LIB library are 300, 600, 900, 1200, 1500 and 1800 K. Library files were produced for the JEF 2.2, JEFF 3.0, JEFF 3.1, JENDL 3.3 and ENDF/B-VI.8 nuclear data libraries. This will be extended with ENDF/B-VII when it becomes available. This report deals with the JEFF 3.1 files included in ALEPH-LIB that are now released by the NEA-OECD.

  4. In-vessel tritium retention and removal in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Federici, G. [ITER JWS Garching Co-Center (Germany); Anderl, R.A. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.; Andrew, P. [JET Joint Undertaking, Abingdon (United Kingdom)] [and others

    1998-06-01

    The International Thermonuclear Experimental Reactor (ITER) is envisioned to be the next major step in the world`s fusion program from the present generation of tokamaks and is designed to study fusion plasmas with a reactor relevant range of plasma parameters. During normal operation, it is expected that a fraction of the unburned tritium, that is used to routinely fuel the discharge, will be retained together with deuterium on the surfaces and in the bulk of the plasma facing materials (PFMs) surrounding the core and divertor plasma. The understanding of he basic retention mechanisms (physical and chemical) involved and their dependence upon plasma parameters and other relevant operation conditions is necessary for the accurate prediction of the amount of tritium retained at any given time in the ITER torus. Accurate estimates are essential to assess the radiological hazards associated with routine operation and with potential accident scenarios which may lead to mobilization of tritium that is not tenaciously held. Estimates are needed to establish the detritiation requirements for coolant water, to determine the plasma fueling and tritium supply requirements, and to establish the needed frequency and the procedures for tritium recovery and clean-up. The organization of this paper is as follows. Section 2 provides an overview of the design and operating conditions of the main components which define the plasma boundary of ITER. Section 3 reviews the erosion database and the results of recent relevant experiments conducted both in laboratory facilities and in tokamaks. These data provide the experimental basis and serve as an important benchmark for both model development (discussed in Section 4) and calculations (discussed in Section 5) that are required to predict tritium inventory build-up in ITER. Section 6 emphasizes the need to develop and test methods to remove the tritium from the codeposited C-based films and reviews the status and the prospects of the

  5. In-vessel tritium retention and removal in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Federici, G. [ITER JWS Garching Co-Center (Germany); Anderl, R.A. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.; Andrew, P. [JET Joint Undertaking, Abingdon (United Kingdom)] [and others

    1998-06-01

    The International Thermonuclear Experimental Reactor (ITER) is envisioned to be the next major step in the world`s fusion program from the present generation of tokamaks and is designed to study fusion plasmas with a reactor relevant range of plasma parameters. During normal operation, it is expected that a fraction of the unburned tritium, that is used to routinely fuel the discharge, will be retained together with deuterium on the surfaces and in the bulk of the plasma facing materials (PFMs) surrounding the core and divertor plasma. The understanding of he basic retention mechanisms (physical and chemical) involved and their dependence upon plasma parameters and other relevant operation conditions is necessary for the accurate prediction of the amount of tritium retained at any given time in the ITER torus. Accurate estimates are essential to assess the radiological hazards associated with routine operation and with potential accident scenarios which may lead to mobilization of tritium that is not tenaciously held. Estimates are needed to establish the detritiation requirements for coolant water, to determine the plasma fueling and tritium supply requirements, and to establish the needed frequency and the procedures for tritium recovery and clean-up. The organization of this paper is as follows. Section 2 provides an overview of the design and operating conditions of the main components which define the plasma boundary of ITER. Section 3 reviews the erosion database and the results of recent relevant experiments conducted both in laboratory facilities and in tokamaks. These data provide the experimental basis and serve as an important benchmark for both model development (discussed in Section 4) and calculations (discussed in Section 5) that are required to predict tritium inventory build-up in ITER. Section 6 emphasizes the need to develop and test methods to remove the tritium from the codeposited C-based films and reviews the status and the prospects of the

  6. Status of US ITER Diagnostics

    Science.gov (United States)

    Stratton, B.; Delgado-Aparicio, L.; Hill, K.; Johnson, D.; Pablant, N.; Barnsley, R.; Bertschinger, G.; de Bock, M. F. M.; Reichle, R.; Udintsev, V. S.; Watts, C.; Austin, M.; Phillips, P.; Beiersdorfer, P.; Biewer, T. M.; Hanson, G.; Klepper, C. C.; Carlstrom, T.; van Zeeland, M. A.; Brower, D.; Doyle, E.; Peebles, A.; Ellis, R.; Levinton, F.; Yuh, H.

    2013-10-01

    The US is providing 7 diagnostics to ITER: the Upper Visible/IR cameras, the Low Field Side Reflectometer, the Motional Stark Effect diagnostic, the Electron Cyclotron Emission diagnostic, the Toroidal Interferometer/Polarimeter, the Core Imaging X-Ray Spectrometer, and the Diagnostic Residual Gas Analyzer. The front-end components of these systems must operate with high reliability in conditions of long pulse operation, high neutron and gamma fluxes, very high neutron fluence, significant neutron heating (up to 7 MW/m3) , large radiant and charge exchange heat flux (0.35 MW/m2) , and high electromagnetic loads. Opportunities for repair and maintenance of these components will be limited. These conditions lead to significant challenges for the design of the diagnostics. Space constraints, provision of adequate radiation shielding, and development of repair and maintenance strategies are challenges for diagnostic integration into the port plugs that also affect diagnostic design. The current status of design of the US ITER diagnostics is presented and R&D needs are identified. Supported by DOE contracts DE-AC02-09CH11466 (PPPL) and DE-AC05-00OR22725 (UT-Battelle, LLC).

  7. ITER Port Interspace Pressure Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, Juan J [ORNL; Van Hove, Walter A [ORNL

    2016-01-01

    The ITER Vacuum Vessel (VV) is equipped with 54 access ports. Each of these ports has an opening in the bioshield that communicates with a dedicated port cell. During Tokamak operation, the bioshield opening must be closed with a concrete plug to shield the radiation coming from the plasma. This port plug separates the port cell into a Port Interspace (between VV closure lid and Port Plug) on the inner side and the Port Cell on the outer side. This paper presents calculations of pressures and temperatures in the ITER (Ref. 1) Port Interspace after a double-ended guillotine break (DEGB) of a pipe of the Tokamak Cooling Water System (TCWS) with high temperature water. It is assumed that this DEGB occurs during the worst possible conditions, which are during water baking operation, with water at a temperature of 523 K (250 C) and at a pressure of 4.4 MPa. These conditions are more severe than during normal Tokamak operation, with the water at 398 K (125 C) and 2 MPa. Two computer codes are employed in these calculations: RELAP5-3D Version 4.2.1 (Ref. 2) to calculate the blowdown releases from the pipe break, and MELCOR, Version 1.8.6 (Ref. 3) to calculate the pressures and temperatures in the Port Interspace. A sensitivity study has been performed to optimize some flow areas.

  8. Iterated Stretching of Viscoelastic Jets

    Science.gov (United States)

    Chang, Hsueh-Chia; Demekhin, Evgeny A.; Kalaidin, Evgeny

    1999-01-01

    We examine, with asymptotic analysis and numerical simulation, the iterated stretching dynamics of FENE and Oldroyd-B jets of initial radius r(sub 0), shear viscosity nu, Weissenberg number We, retardation number S, and capillary number Ca. The usual Rayleigh instability stretches the local uniaxial extensional flow region near a minimum in jet radius into a primary filament of radius [Ca(1 - S)/ We](sup 1/2)r(sub 0) between two beads. The strain-rate within the filament remains constant while its radius (elastic stress) decreases (increases) exponentially in time with a long elastic relaxation time 3We(r(sup 2, sub 0)/nu). Instabilities convected from the bead relieve the tension at the necks during this slow elastic drainage and trigger a filament recoil. Secondary filaments then form at the necks from the resulting stretching. This iterated stretching is predicted to occur successively to generate high-generation filaments of radius r(sub n), (r(sub n)/r(sub 0)) = square root of 2[r(sub n-1)/r(sub 0)](sup 3/2) until finite-extensibility effects set in.

  9. Reactor physics studies for the Advanced Fuel Cycle Initiative (AFCI) Reactor-Accelerator Coupling Experiments (RACE) Project

    Science.gov (United States)

    Stankovskiy, Evgeny Yuryevich

    In the recently completed RACE Project of the AFCI, accelerator-driven subcritical systems (ADS) experiments were conducted to develop technology of coupling accelerators to nuclear reactors. In these experiments electron accelerators induced photon-neutron reactions in heavy-metal targets to initiate fission reactions in ADS. Although the Idaho State University (ISU) RACE ADS was constructed only to develop measurement techniques for advanced experiments, many reactor kinetics experiments were conducted there. In the research reported in this dissertation, a method was developed to calculate kinetics parameters for measurement and calculation of the reactivity of ADS, a safety parameter that is necessary for control and monitoring of power production. Reactivity is measured in units of fraction of delayed versus prompt neutron from fission, a quantity that cannot be directly measured in far-subcritical reactors such as the ISU RACE configuration. A new technique is reported herein to calculate it accurately and to predict kinetic behavior of a far-subcritical ADS. Experiments conducted at ISU are first described and experimental data are presented before development of the kinetic theory used in the new computational method. Because of the complexity of the ISU ADS, the Monte-Carlo method as applied in the MCNP code is most suitable for modeling reactor kinetics. However, the standard method of calculating the delayed neutron fraction produces inaccurate values. A new method was developed and used herein to evaluate actual experiments. An advantage of this method is that its efficiency is independent of the fission yield of delayed neutrons, which makes it suitable for fuel with a minor actinide component (e.g. transmutation fuels). The implementation of this method is based on a correlated sampling technique which allows the accurate evaluation of delayed and prompt neutrons. The validity of the obtained results is indicated by good agreement between experimental

  10. (Meeting on fusion reactor materials)

    Energy Technology Data Exchange (ETDEWEB)

    Jones, R.H. (Pacific Northwest Lab., Richland, WA (USA)); Klueh, R.L.; Rowcliffe, A.F.; Wiffen, F.W. (Oak Ridge National Lab., TN (USA)); Loomis, B.A. (Argonne National Lab., IL (USA))

    1990-11-01

    During his visit to the KfK, Karlsruhe, F. W. Wiffen attended the IEA 12th Working Group Meeting on Fusion Reactor Materials. Plans were made for a low-activation materials workshop at Culham, UK, for April 1991, a data base workshop in Europe for June 1991, and a molecular dynamics workshop in the United States in 1991. At the 11th IEA Executive Committee on Fusion Materials, discussions centered on the recent FPAC and Colombo panel review in the United States and EC, respectively. The Committee also reviewed recent progress toward a neutron source in the United States (CWDD) and in Japan (ESNIT). A meeting with D. R. Harries (consultant to J. Darvas) yielded a useful overview of the EC technology program for fusion. Of particular interest to the US program is a strong effort on a conventional ferritic/martensitic steel for fist wall/blanket operation beyond NET/ITER.

  11. Protection sequence of AC/DC converters for ITER PF MAGNET coils

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Byung Hoon; Hwang, Churl Kew; Lee, Kwang Won; Jin, Jeong Tae; Chang, Sae Sik; Oh, Jong Seok; Choi, Jung Wan; Suh, Jae Hak [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Tao, Jun; Song, In Ho [ITER Organization, Paul Lez Durance (France)

    2010-06-15

    The protection sequence of an AC/DC converter for an ITER PF coil system is developed in this study. Possible faults in the coil system are simulated and the results reflected in the design of a sequence to protect the coil and converter. The inductances of the current sharing reactors and thyristor numbers in parallel with the bridge arms are optimized with the designed protection sequence

  12. Analysis of the Temporal Response of Coupled Asymmetrical Zero-Power Subcritical Bare Metal Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Klain, Kimberly L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-06-21

    The behavior of symmetrical coupled-core systems has been extensively studied, yet there is a dearth of research on asymmetrical systems due to the increased complexity of the analysis of such systems. In this research, the multipoint kinetics method is applied to asymmetrical zeropower, subcritical, bare metal reactor systems. Existing research on asymmetrical reactor systems assumes symmetry in the neutronic coupling; however, it will be shown that this cannot always be assumed. Deep subcriticality adds another layer of complexity and requires modification of the multipoint kinetics equations to account for the effect of the external neutron source. A modified set of multipoint kinetics equations is derived with this in mind. Subsequently, the Rossi-alpha equations are derived for a two-region asymmetrical reactor system. The predictive capabilities of the radiation transport code MCNP6 for neutron noise experiments are shown in a comparison to the results of a series of Rossi-alpha measurements performed by J. Mihalczo utilizing a coupled set of symmetrical bare highly-enriched uranium (HEU) cylinders. The ptrac option within MCNP6 can generate time-tagged counts in a cell (list-mode data). The list-mode data can then be processed similarly to measured data to obtain values for system parameters such as the dual prompt neutron decay constants observable in a coupled system. The results from the ptrac simulations agree well with the historical measured values. A series of case studies are conducted to study the effects of geometrical asymmetry in the coupling between two bare metal HEU cylinders. While the coupling behavior of symmetrical systems has been reported on extensively, that of asymmetrical systems remains sparse. In particular, it appears that there has been no previous research in obtaining the coupling time constants for asymmetrically-coupled systems. The difficulty in observing such systems is due in part to the inability to determine the

  13. Study of a fibre optics current sensor for the measurement of plasma current in ITER

    Science.gov (United States)

    Wuilpart, Marc; Vanus, Benoit; Andrasan, Alina; Gusarov, Andrei; Moreau, Philippe; Mégret, Patrice

    2016-05-01

    In this article, we study the feasibility of using a fibre-optics current sensor (FOCS) for the measurement of plasma current in the future fusion reactor ITER. The sensor is based on a classical FOCS interrogator involving the measurement of the state of polarization rotation undergone by the light in presence of a magnetic field (Faraday effect) in an optical fibre surrounding the current and terminated by a Faraday mirror. We considered a uniformly spun optical fibre as the sensing element and we used the Stokes formalism to simulate the sensor. The objective of the simulations is to quantify the ratio LB/SP (beat length over the spun period of the spun fibre) enabling a measurement error in agreement with the ITER specifications. The simulator takes into account the temperature variations undergone by the measurement system under ITER operation. The simulation work showed that a LB/SP ratio of 19.2 is adequate.

  14. Recent Updates to the MELCOR 1.8.2 Code for ITER Applications

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad J

    2007-05-01

    This report documents recent changes made to the MELCOR 1.8.2 computer code for application to the International Thermonuclear Experimental Reactor (ITER), as required by ITER Task Agreement ITA 81-18. There are four areas of change documented by this report. The first area is the addition to this code of a model for transporting HTO. The second area is the updating of the material oxidation correlations to match those specified in the ITER Safety Analysis Data List (SADL). The third area replaces a modification to an aerosol tranpsort subroutine that specified the nominal aerosol density internally with one that now allows the user to specify this density through user input. The fourth area corrected an error that existed in an air condensation subroutine of previous versions of this modified MELCOR code. The appendices of this report contain FORTRAN listings of the coding for these modifications.

  15. Recent Updates to the MELCOR 1.8.2 Code for ITER Applications

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad J

    2007-04-01

    This report documents recent changes made to the MELCOR 1.8.2 computer code for application to the International Thermonuclear Experimental Reactor (ITER), as required by ITER Task Agreement ITA 81-18. There are four areas of change documented by this report. The first area is the addition to this code of a model for transporting HTO. The second area is the updating of the material oxidation correlations to match those specified in the ITER Safety Analysis Data List (SADL). The third area replaces a modification to an aerosol tranpsort subroutine that specified the nominal aerosol density internally with one that now allows the user to specify this density through user input. The fourth area corrected an error that existed in an air condensation subroutine of previous versions of this modified MELCOR code. The appendices of this report contain FORTRAN listings of the coding for these modifications.

  16. Project of law, adopted by the Senate, giving permission to the approval of the agreement between the French government and the international organization for thermonuclear fusion energy ITER, relative to the head office of ITER organization and to the privileges and immunities of ITER organization in the French territory; Projet de loi adopte par le Senat, autorisant l'approbation de l'accord entre le Gouvernement de la Republique francaise et l'Organisation internationale ITER pour l'energie de fusion relatif au siege de l'Organisation ITER et aux privileges et immunites de l'Organisation ITER sur le territoire francais

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-01-15

    The will of building up an international thermonuclear experimental reactor (ITER) gathers since several years the European community of atomic energy (Euratom), Japan, the USA, and Russia, next followed by China, South Korea and, since 2005, by India. The agreement signed in Paris between these seven parties on November 21, 2006 entrusted the international organization ITER with the realization of this project. The implications of the ITER project are enormous both in their scientific and in their economical aspects. France has a particular position in this project since the head office of ITER organisation is sited at Saint-Paul-lez-Durance and the tokamak will be built at Cadarache. Therefore, an agreement has been signed between ITER organization and the French government. The approval of this agreement is the object of this project of law made of a single article. The agreement between the French government and the international organization ITER is attached to the document. It defines the juridical status, the privileges and immunities of the organization itself and of its personnel inside the French territory. An appendix to the agreement precises the cooperation modalities between the French authorities and ITER organization. (J.S.)

  17. Correction Method of Considering Resonance Elastic Scattering in MCNP%在 MCNP 中考虑共振弹性散射的修正方法

    Institute of Scientific and Technical Information of China (English)

    贺清明; 曹良志; 吴宏春; 郑友琦

    2014-01-01

    The free gas model is adopted to consider the thermal scattering effect of the elastic collision between neutron and target .The conventional model assumes that the elastic scattering cross sections are constant at 0 K , w hich neglects the influence of resonance effect .In order to consider the resonance elastic scattering in the free gas model ,the Doppler broadening rejection correction (DBRC ) method was applied to correct the free gas model of MCNP .The Mosteller’s Doppler defect benchmark for LWR pin cell was analyzed .The numerical results show that neglect of resonance elastic scattering effect contributes to overestimation of the infinite multiplicative factor to the extent of 40‐100 pcm and 140‐200 pcm for hot zero power and hot full power cases , respectively . T he fuel temperature coefficients are also overestimated 7%‐15% . T he computational time of the newly developed sampling technique was studied and the influence of the resonance elastic scattering effect on the emergent energy distribution w as analyzed .%蒙特卡罗方法采用自由气体模型来考虑中子与靶核的弹性碰撞中的热效应。传统的模型假设绝对零度下的弹性散射截面是常数,忽略了截面的共振效应所带来的影响。为在自由气体模型中考虑共振弹性散射效应,采用多普勒展宽舍弃修正方法,修正了连续能量蒙特卡罗程序MCNP的自由气体模型,并对Mosteller轻水堆多普勒基准题进行了分析。数值结果表明:对于轻水堆,在热态零功率的情况下,忽略共振弹性散射会高估燃料棒的无限介质增殖因数(k∞)40~100 pcm ,热态满功率下高估140~200 pcm ;忽略共振弹性散射给燃料温度系数带来7%~15%正的偏差。同时分析了新的抽样方法对计算时间的影响,以及共振弹性散射效应对中子出射能量分布的影响。

  18. Investigation of materials for fusion power reactors

    Science.gov (United States)

    Bouhaddane, A.; Slugeň, V.; Sojak, S.; Veterníková, J.; Petriska, M.; Bartošová, I.

    2014-06-01

    The possibility of application of nuclear-physical methods to observe radiation damage to structural materials of nuclear facilities is nowadays a very actual topic. The radiation damage to materials of advanced nuclear facilities, caused by extreme radiation stress, is a process, which significantly limits their operational life as well as their safety. In the centre of our interest is the study of the radiation degradation and activation of the metals and alloys for the new nuclear facilities (Generation IV fission reactors, fusion reactors ITER and DEMO). The observation of the microstructure changes in the reactor steels is based on experimental investigation using the method of positron annihilation spectroscopy (PAS). The experimental part of the work contains measurements focused on model reactor alloys and ODS steels. There were 12 model reactor steels and 3 ODS steels. We were investigating the influence of chemical composition on the production of defects in crystal lattice. With application of the LT 9 program, the spectra of specimen have been evaluated and the most convenient samples have been determined.

  19. A FAST CONVERGENT METHOD OF ITERATED REGULARIZATION

    Institute of Scientific and Technical Information of China (English)

    Huang Xiaowei; Wu Chuansheng; Wu Di

    2009-01-01

    This article presents a fast convergent method of iterated regularization based on the idea of Landweber iterated regularization, and a method for a-posteriori choice by the Morozov discrepancy principle and the optimum asymptotic convergence order of the regularized solution is obtained. Numerical test shows that the method of iterated regu-larization can quicken the convergence speed and reduce the calculation burden efficiently.

  20. Preconditioned iterations to calculate extreme eigenvalues

    Energy Technology Data Exchange (ETDEWEB)

    Brand, C.W.; Petrova, S. [Institut fuer Angewandte Mathematik, Leoben (Austria)

    1994-12-31

    Common iterative algorithms to calculate a few extreme eigenvalues of a large, sparse matrix are Lanczos methods or power iterations. They converge at a rate proportional to the separation of the extreme eigenvalues from the rest of the spectrum. Appropriate preconditioning improves the separation of the eigenvalues. Davidson`s method and its generalizations exploit this fact. The authors examine a preconditioned iteration that resembles a truncated version of Davidson`s method with a different preconditioning strategy.