Volume 70, Issue 3

Uncertainty and sensitivity analyses were performed for the calculation of power and void reactivity coefficients of an RBMK-1500 reactor core state. Both parameters are important for the reactivity feedback behaviour and are of special safety relevance. The reactivity coefficients were calculated using the 3D-reactor core model QUABOX/CUBBOX-HYCA and the uncertainty and sensitivity analyses were performed by the GRS methodology applying the SUSA software package (Software System for Uncertainty and Sensitivity Analysis). A total number of 20 uncertain input parameters and their probability distributions were identified as possibly relevant for the size of the reactivity effects and taken into account in the SUSA analysis. The method helps to identify the main contributors to the uncertainty of calculated results, as well as to determine two-sided (95 %, 95 %) statistical tolerance intervals of the investigated reactivity coefficients. The approach to quantify uncertain parameters for RBMK-1500 core calculations using the SUSA code package and QUABOX/CUBBOX-HYCA was successfully performed. The results provide guidance where model improvements should be performed to increase the accuracy of the calculations efficiently.

The Advanced Heavy Water Reactor (AHWR) being designed in India is a vertical pressure tube type boiling water-cooled and heavy water moderated reactor. A key feature of the AHWR is the use of simple passive systems to respond to any postulated accidental condition. Passive Containment Cooling System (PCCS), is one such system, which provides long term cooling for the reactor containment following a Loss Of Coolant Accident (LOCA). The system consists of Immersed Condensers immersed in a gravity driven water pool. One of the important phenomena related to PCCS functioning is the effect of thermal stratification in the water pool. The heat transfer from IC to water pool takes place by natural convection. In this process, the pool water can get stratified. Towards the top of the water pool, the temperature of water may reach almost the boiling point while at the bottom of the pool, the water is likely to remain at the initial temperature. Hence, in order to study the integral system response of PCCS, the phenomenon of stratification requires to be investigated. Towards this objective, the case of a rectangular enclosure heated symmetrically on both the vertical sides was considered and a two-dimensional analysis was done by numerically solving the mass, momentum and energy equations. Theoretical results were generated by the use of a computer code developed for this purpose and the results are then compared with the experimental results available in literature.

A 232 Th activation sample was exposed directly to the spallation neutrons produced by 1GeV proton colliding with a lead target and other samples of 197 Au, 209 Bi, 59 Co, 115 In and 181 Ta were irradiated by moderated neutrons after passing through 6 cm of paraffin. Different (n, γ) and (n, xn) reaction products in 232 Th, 197 Au and 209 Bi samples are analyzed by gamma-ray spectrometry to measure the neutron fluence at the positions of the samples. For this purpose a histogram of ‘flat’ spectrum averaged neutron cross-sections in over thirty given energy groups was determined from the PREPRO-2000 software. Results of measurements are compared with Monte-Carlo simulations by the Dubna CASCADE code. Experimental values for both spallation and moderated neutron fluences are in reasonable agreement with the CASCADE code. Thus, the experimentally measured fluence of moderated neutrons was used to infer the ‘one group’ reaction cross-sections of 209 Bi (n, xn) reactions for x = 6 to 9.

Exact discrete ordinates interface conditions were derived for neutron reflection and leakage for one-dimensional nuclear reactor computations with anisotropic scattering. The key to our derivation is a suitable reformulation of exact discretized equations – spectral nodal equations defined in the nonactive surrounding regions of nuclear reactor cores. As a result, interface conditions are obtained that can advantageously replace the nonactive surrounding regions of a nuclear reactor core in reactor computations. Numerical experiments with a representative one-dimensional code and an axial core model with anisotropic scattering were performed in order to illustrate the improvements that can be experienced by most one-dimensional discrete ordinates nuclear reactor codes with our interface conditions.

Safety analysis of research reactors include a determination of control rod worth and kinetic parameters upon which the dynamic performance of the reactor depends. The core control of the ET-RR-2 reactor is accomplished through six Ag-In-Cd alloy-absorbing plates, which can be inserted into the core with a high velocity. In the present work, the fully and partially inserted control rod configuration in 3D reactor geometry and the effect of reactivity worth of the control rods on the reactor core is studied. The reactivity worth were calculated for each control rod and also for combinations of more than one rod. The total worth of all control rods in the reactor core is about 16.95% k/k from the core reactivity. Other neutronic parameters, i. e. the 3D neutron flux distribution and the change of the power peaking factor as a results of inserted and withdrawing the control rods were calculated.

The conceptual neutronic design of the cold neutron source (CNS) for the Egyptian second research reactor (ETRR-2) was carried out using the MCNP code. A parametric analysis was also performed to choose the type and geometry of the moderator and the required CNS dimensions to maximize the cold neutron production. The moderator cell is a spherical annulus containing liquid hydrogen. The cold neutron gain and brightness are calculated together with the nuclear heat load of the CNS. The effects of void fraction in the moderator cell and the ortho:para ratios on cold neutron gain were calculated.

Investigations of necessary conditions for a self-sustained chain reaction in the Earth inner regions hypothetically containing uranium is presented for the time interval from Earth formation to present time. It is determined that criticality was theoretically possible up to 2.5 Ga before present if uranium concentrated in pure form. In the early geological history (4 Ga before present) the self-sustained criticality could occur even if uranium was diluted up to 1: 20 by the average core material or 1: 60 by the average mantle material. If other metallic materials of similar density as uranium (e. g., Au, W) or similar atomic weight (e. g., Th) concentrated from the primordial mixture in equal proportion as uranium, criticality was not possible for any period in Earth history provided that the basic material contained no light nuclides (H, C). Criticality in the Earth inner regions could have established only if uranium concentrated from the basic material more effectively than elements of similar density or atomic number.

A technique to combine statistical and deterministic approaches in mechanical fuel rod analysis is described. The impact of different fuel rod input parameters on the fuel rod design parameter under consideration is identified by statistical means. It is expressed through rank correlation coefficients. These coefficients are then employed to calculate corresponding quantiles for the input parameters allowing the selection of appropriate input values from the parameter probability distributions. The quantiles are computed with the provision that the fuel rod reactions obtained with the associated input parameter combination cover the reactions in a subset of the analysed fuel rod totality. The described procedure reduces the computational costs in comparison to some pure statistical methods of fuel rod design.

Im Teil 1 dieser Untersuchungen wurde in das Problem der Brennstabauslegung eingeführt sowie die Grundlagen für das Verständnis des zweiten Teils gelegt. Dieser zweite Teil beschäftigt sich mit der mehr technischen Seite der mathematischen Modellierung des Verhaltens der Brennstäbe im Einsatz. Zusätzlich wird auf die in Deutschland aktuelle Diskussion über das sogenannte Ein-Brennstab-Quantil eingegangen.

In 1971 evidence for the production of element 112 via secondary reactions in CERN W targets was obtained. The evidence was mainly based on the observation of fission fragments in Hg sources separated from the W targets, on the measured masses of the fissioning nuclei and on the assumption that element 112 (Eka-Hg) actually behaves like Hg in the chemical separation process. This assumption is analyzed in view of recent relativistic calculations of the electronic structure of element 112. It is shown that in the superheavy element region only the chemistry of element 112 is similar to that of Hg.

There are two independent claims for the discovery of element 112: The claim by Hofmann et al. from 1996 and the older claim from 1971 by Marinov et al. This Comment will not challenge the experimental results of Hofmann et al., but it will discuss one aspect of the claimed discovery of element 112 by Marinov et al., as their experiment has never been reproduced in exactly the form in which the original experiment has been carried out. The reasons for this deficiency may not be found in the field of science, but possibly in radioprotection restrictions for researchers who want to carry out such an experiment. However, such is not a sufficient reason to exclude the original claim from all considerations of the responsible international authorities, who have to settle such priority questions. It may be in agreement with scientific traditions, that when the responsible international committees do not feel to be able to come to a positive decision on the “1971” claim, they could keep the priority problem unsettled for the time being.