Implementation of the CEM-Code into the MCNPX-Code

In the development stage of improving the physics abilities of MCNPX, the CEM code has been implemented into MCNPX as the nuclear reaction model for nuclei and pion induced reactions up to 5 GeV of kinetic energy. The CEM code includes all the reaction stages of intra:nuclear cascade, preequilibrium and equilibrium and is an alternative for the existing MCNPX models. The preliminary implementation fixes the uncompleted fission model of CEM by applying the RAL fission fragmentation model and provides a parameterized formula for the CEM input parameter a{sub f}/a{sub n}, the ratio of level density parameters at the fission saddle point shape and the compound shape of the excited nucleus. All the deficiencies are addressed in an improved CEM code that will replace the existing implementation when finished. Thick target calculations employing lead and tungsten targets have been performed to validate the implementation and test the CEM models against the old MCNPX models

CEM03.03 and LAQGSM03.03 Event Generators for the MCNP6, MCNPX, and MARS15 Transport Codes

A description of the IntraNuclear Cascade (INC), preequilibrium, evaporation, fission, coalescence, and Fermi breakup models used by the latest versions of our CEM03.03 and LAQGSM03.03 event generators is presented, with a focus on our most recent developments of these models. The recently developed "S" and "G" versions of our codes, that consider multifragmentation of nuclei formed after the preequilibrium stage of reactions when their excitation energy is above 2A MeV using the Statistical Multifragmentation Model (SMM) code by Botvina et al. ("S" stands for SMM) and the fission-like binary-decay model GEMINI by Charity ("G" stands for GEMINI), respectively, are briefly described as well. Examples of benchmarking our models against a large variety of experimental data on particle-particle, particle-nucleus, and nucleus-nucleus reactions are presented. Open questions on reaction mechanisms and future necessary work are outlined.Comment: 94 pages, 51 figures, 5 tables, invited lectures presented at the Joint ICTP-IAEA Advanced Workshop on Model Codes for Spallation Reactions, February 4-8, 2008, ICTP, Trieste, Italy; corrected typos and reference

Ess target performance for different beam pulses

Last trends in the design of linear accelerators for high power spallation sources point to the use of ion beams of larger energies and shorter pulse lengths in order to enhance the reliability of the system. In this sense the recommendations for ESS are to increase the energy of the proton beam from 1.3GeV to 2-2.5GeV and to reduce the length of the beam pulse from 2ms to 1-1.5ms, keeping the source average power at 5MW. Different values for the repetition rate are also being discussed (16 2/3, 20, 25 Hz). ESS Bilbao is analyzing the impact of these modifications on the design of the target system. In this paper the effects of the different beam energies on the target disc thermohydraulics and the neutron performance of the source are discussed. Initial calculations were performed for a rotating target with ESS 2002 parameters. During the development of the work –that are being performed in collaboration with SNS– the decision was made to use the SNS-STS Target-Moderator-Reflector Assembly (TMRA) –slightly modified to accommodate the target design being studied for ESS– which presents a state of the art design with a cylindrical liquid para-hydrogen moderator in wing configuration aimed to enhance cold neutron productio

General purpose photoneutron production in MCNP4A

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

Neutronenphysikalische Optimierung von experimentellen Einbauten am Forschungsreaktor FRM-II mit Monte Carlo Methoden

The dissertation reports optimization activities for the experimental structural components (beam holes, cold and hot source) in the D_2O moderator surrounding the reactor core. The spectral neutron flux at the far end of the particular beam holes was optimized taking into account the reactivity decrease and the coolability of the components. The 3D transport equation for neutrons and photons was derived using the codes MORSE-K and MORSE-SGC. (orig.)Optimiert wurden die experimentellen Einbauten (Strahlrohre, kalte und heisse Quelle) in dem den Reaktorkern umgebenden D_2O-Moderator. Der spektrale Neutronenfluss am aeusseren Ende der entsprechenden Strahlrohre unter Beruecksichtigung der Beschraenkung der Reaktivitaetsminderung und der Kuehlbarkeit der Einbauten wurde maximiert. Zur Loesung der 3-dimensionalen Transportgleichung fuer Neutronen und Photonen wurden die Rechencodes MORSE-K und MORSE-SGC benutzt. (orig.)SIGLEAvailable from TIB Hannover: H93B4807 / FIZ - Fachinformationszzentrum Karlsruhe / TIB - Technische InformationsbibliothekBundesministerium fuer Forschung und Technologie (BMFT), Bonn (Germany)DEGerman

A photoneutron production option for MCNP4A

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. Subroutines were developed to calculate the probability of the photoneutron production at the photon collision sites and the energy and flight direction of the created photoneutrons with the help of user supplied data. These subroutines are accessed through subroutine colidp which processes the photon collisions

Validation of multigroup neutron cross sections and calculational methods for the advanced neutron source against the FOEHN critical experiments measurements

The FOEHN critical experiment was analyzed to validate the use of multigroup cross sections and Oak Ridge National Laboratory neutronics computer codes in the design of the Advanced Neutron Source. The ANSL-V 99-group master cross section library was used for all the calculations. Three different critical configurations were evaluated using the multigroup KENO Monte Carlo transport code, the multigroup DORT discrete ordinates transport code, and the multigroup diffusion theory code VENTURE. The simple configuration consists of only the fuel and control elements with the heavy water reflector. The intermediate configuration includes boron endplates at the upper and lower edges of the fuel element. The complex configuration includes both the boron endplates and components in the reflector. Cross sections were processed using modules from the AMPX system. Both 99-group and 20-group cross sections were created and used in two-dimensional models of the FOEHN experiment. KENO calculations were performed using both 99-group and 20-group cross sections. The DORT and VENTURE calculations were performed using 20-group cross sections. Because the simple and intermediate configurations are azimuthally symmetric, these configurations can be explicitly modeled in R-Z geometry. Since the reflector components cannot be modeled explicitly using the current versions of these codes, three reflector component homogenization schemes were developed and evaluated for the complex configuration. Power density distributions were calculated with KENO using 99-group cross sections and with DORT and VENTURE using 20-group cross sections. The average differences between the measured values and the values calculated with the different computer codes range from 2.45 to 5.74%. The maximum differences between the measured and calculated thermal flux values for the simple and intermediate configurations are {approx} 13%, while the average differences are < 8%

Moderators at LENS: Performance and Development Research

AbstractThe Target/Moderator/Reflector (TMR) system at the Low Energy Neutron Source has a flexible design in order to accommodate research into the performance of neutron moderators in general and small-scale accelerator-driven neutron sources in particular. Since producing its first cold neutron beam in April of 2005, the LENS TMR has undergone a number of design changes, and has been used to investigate a number of novel moderator ideas. In this paper we summarize the impact of some of these design changes on moderator performance as well as some recent results from a novel inhomogeneous moderator design that combines traditional moderating material (polyethylene) with single crystals of silicon