A. dos Santos

The Inversion Point of the Isothermal Reactivity Coefficient of the IPEN/MB-01 Reactor - 1: Experimental Procedure

A new experimental quantity is presented to serve as a benchmark to verify the adequacy of the newly released 235 U thermal and subthermal cross sections for the determination of the reactivity coefficients of light water reactors. Such a quantity is denominated the inversion point, and by definition it is the temperature for which the isothermal reactivity coefficient of a reactor system becomes positive. The experimental bases for its determination are discussed. The experiment has been performed in the IPEN/ MB-01 reactor facility. Instead of heating the reactor system as usual in experiments considering temperature variations, the reactor system is cooled to ∼8.5°C. By means ofa heating/cooling system, the temperature is allowed to increase slowly in a stepwise manner. For each step, the control bank critical position is recorded, and by analyzing its behavior as a function of temperature, the inversion point is inferred. The inversion point has been found to be an adequate experimental quantity to validate the thermal and subthermal 235 U cross section because it does not require any sort of calculated correction factors or any quantity that comes either from the calculational methodologies or from another experiment. In addition, the inversion point is an experimental quantity that can be measured with an excellent level of accuracy due mainly to the very precise characteristics of the control bank system of the IPEN/MB-01 reactor. The final value obtained for the IPEN/MB-01 reactor is 14.99 ± 0.15°C.

The Inversion Point of the Isothermal Reactivity Coefficient of the IPEN/MB-01 Reactor - II: Theoretical Analysis

Abstract TORT, an SN three-dimensional transport code, is employed for the analysis of the inversion point of the isothermal reactivity coefficient of the IPEN/MB-01 reactor. The analyses are performed in companion NJOY, AMPX-II, and TORT systems considering the data libraries ENDF/B-VI.8, JENDL3.3, and JEF3.0. The analyses reveal that for this peculiar problem, there is a need to convert all the computer codes to DOUBLE-PRECISION as well as to increase to seven the number of digits of the ANISN library generated by XSDRNPM. Contrary to the traditional diffusion theory codes, TORT keff results are very sensitive to the number of both fine and broad groups. For instance, the traditional and very well known two- and four-group structure, largely utilized in several diffusion codes, produced simply unacceptable keff results. The highest deviation between calculated and experimental values found for the inversion point was –4.48°C. At first glance, there appears to be a significant discrepancy. However, in terms of reactivity coefficient, this discrepancy means a deviation of –0.90 ± 0.05 pcm/°C, which indicates that the calculational methodology and related nuclear data libraries meet the desired accuracy (–1.0 pcm/°C) for the determination of this parameter for thermal reactors.

A proposal for benchmarking 235U nuclear data

Abstract Two experiments performed at the IPEN/MB-01 reactor are suggested to serve as a benchmark problem to verify mainly the adequacy of the 235U nuclear data for criticality analyses and for the isothermal reactivity coefficient determination of thermal reactors. The experiments are claimed to be well-defined, and they are suitable for a benchmark problem partially due to their small uncertainties and partially due to the lack of any sort of calculated correction factors or any quantity that comes either from the calculational methodologies or from another experiment. The isothermal experiment fulfills a specific need to introduce a reactor response that is sensitive to the 235U cross-section shape below 5 meV. This feature could be accomplished due mainly to the very precise control bank system characteristics of the IPEN/MB-01 reactor. The MCNP-4B calculational analyses reveal that the most recent 235U evaluation (Leal-Derrien-Larson’s evaluation) incorporated in ENDF/B-VI release 5 performs well in the theory-experiment result comparisons of the aforementioned experiments. Particularly in the isothermal experiment, ENDF/B-VI release 5 produces results that even considering the deviations inherent to the Monte Carlo method meet the desired accuracy (±1.0 pcm/°C) for the isothermal reactivity coefficient determination in contrast to the JEF-2.2 and JENDL-3.2 libraries, which produce unacceptably high keff results. The main reasons are the 235U nuclear data in the case of JEF-2.2 and the nuclear data of both 235U and iron in the case of JENDL-3.2.

Spallation Product Distributions and Neutron Multiplicities for Accelerator-Driven System Using the CRISP Code

Abstract Neutron multiplicities for several targets and spallation products of proton-induced reactions in thin targets of interest to an accelerator-driven system obtained with the CRISP code have been reported. This code is a Monte Carlo calculation that simulates the intranuclear cascade and evaporation/fission competition processes. Results are compared with experimental data, and agreement between each other can be considered quite satisfactory in a very broad energy range of incitant particles and different targets.

The CRISP Code for Nuclear Reactions

The CRISP package performs the intranuclear cascade process and the evaporation/fission competition resulting in a code that represents a good tool to describe complexes characteristics of the nuclear reactions, and opens the opportunity for applications in different fields, such as medical physics, photonuclear reactions, spallation or fission process initiated by different probes and in Accelerator Driven Systems, where precise description of energetic and angular neutron distribution, neutron multiplicity and spallation products information are needed. In the CRISP model, was included the time‐sequence characteristics of the MCMC code and the evaporation/fission competition process model of the MCEF. Also, includes improvements in the code, as the excitation of nucleonic resonances heavier than Delta; the initial nuclear ground state construction according to the Fermi model and Pauli principle; and a more realistic Pauli blocking mechanism. Some consequences of the improvements performed in the code w...

Neutron Damage in the Plasma Chamber First Wall of the GCFTR-2 Fusion-Fission Hybrid Reactor

The successful development of energy-conversion machines based on either nuclear fission or fusion is completely dependent on the behaviour of the engineering materials used to construct the fuel containment and primary heat extraction systems. Such materials must be designed in order to maintain their structural integrity and dimensional stability in an environment involving high temperatures and heat fluxes, corrosive media, high stresses and intense neutron fluxes. However, despite the various others damage issues, such as the effects of plasma radiation and particle flux, the neutron flux is sufficiently energetic to displace atoms from their crystalline lattice sites. It is clear that the understanding of the neutron damage is essential for the development and safe operation of nuclear systems. Considering this context, the work presents a study of neutron damage in the Gas Cooled Fast Transmutation Reactor (GCFTR-2) driven by a Tokamak D-T fusion neutron source of 14.03 MeV. The theoretical analysis was performed by MCNP-5 and the ENDF/B-VII.1 neutron data library. A brief discussion about the determination of the radiation damage is presented, along with an analysis of the total neutron energy deposition in seven points through the material of the plasma source wall (PSW), in which was considered the HT-9 steel. The neutron flux was subdivided into three energy groups and their behaviour through the material was also examined.

Development of the CRISP Package for Spallation Studies and Accelerator-Driven Systems

Abstract Power generation from nuclear reactors provides an almost inexhaustive power source due to the huge quantities of nuclear fuel existent in our planet, which guarantees its utilization for thousands of years. Interest has been shifted to the so-called hybrid reactors [accelerator-driven systems (ADS)] as an alternative technology for power generation and transmutation, thus requiring precise knowledge about nuclear structure and nuclear reaction characteristics. Research groups from Instituto de Fisica, Universidade de São Paulo and Brazilian Center for Research in Physics made a joint effort to develop a computer program, CRISP, to calculate the intranuclear cascade proprieties and the nuclear evaporation process, present in all nuclear reactions with energies above a few tens of mega-electron-volts, using Monte Carlo techniques. Some reaction channels were included in these programs, resulting in a more realistic representation of the processes involved, aiming at reactor physics studies and academic studies about hadron and meson properties in nuclear matter. Some results obtained with this code and a comparison with experimental data are presented. Although all these results are preliminary, they are very consistent with the available experimental data. Since the applicability of the CRISP package has a wide range of options, especially in ADS, some results describing the effectiveness of the code were achieved.

The Intermediate Resonance Parameters in the Multigroup Formalism

A new methodology that is applicable to individual nuclides is developed for the determination of the intermediate resonance (IR) parameters in the multigroup formalism. The method keeps the main steps commonly used for the determination of these parameters and is compatible with the methods utilized for the generation of the multigroup libraries for thermal and epithermal reactors. The proposed method does not impose any restriction on the formalism used to describe the resonances. Use is made of the computational approach used by the GROUPR module of the NJOY system (flux calculator option). A numerical scheme is presented to determine the IR parameters by means of an iterative approach. Numerical results for the IR parameters in a heterogeneous system composed of UO 2 ( 238 U only) and hydrogen as an external moderator are reported as a function of the dilution σ 0 , heterogeneity factor β, and temperature Tfor several epithermal groups of the MUFT structure. The results are consistent, as shown by the consistency checks performed.