[en] Monte Carlo neutron transport codes are widely used in various reactor physics applications, traditionally related to criticality safety analyses, radiation shielding problems, detector modelling and validation of deterministic transport codes. The main advantage of the method is the capability to model geometry and interaction physics without major approximations. The disadvantage is that the modelling of complicated systems is very computing-intensive, which restricts the applications to some extent. The importance of Monte Carlo calculation is likely to increase in the future, along with the development in computer capacities and parallel calculation. An interesting near-future application for the Monte Carlo method is the generation of input parameters for deterministic reactor simulator codes. These codes are used in coupled LWR full-core analyses and typically based on few-group nodal diffusion methods. The input data consists of homogenised few-group constants, presently generated using deterministic lattice transport codes. The task is becoming increasingly challenging, along with the development in nuclear technology. Calculations involving high-burnup fuels, advanced MOX technology and next-generation reactor systems are likely to cause problems in the future, if code development cannot keep up with the applications. A potential solution is the use of Monte Carlo based lattice transport codes, which brings all the advantages of the calculation method. So far there has been only a handful of studies on group constant generation using the Monte Carlo method, although the interest has clearly increased during the past few years. The homogenisation of reaction cross sections is simple and straightforward, and it can be carried out using any Monte Carlo code. Some of the parameters, however, require the use of special techniques that are usually not available in general-purpose codes. The main problem is the calculation of neutron diffusion coefficients, which have no continuous-energy counterparts in the Monte Carlo calculation. This study is focused on the development of an entirely new Monte Carlo neutron transport code, specifically intended for reactor physics calculations at the fuel assembly level. The PSG code is developed at VTT Technical Research Centre of Finland and one of the main applications is the generation of homogenised group constants for deterministic reactor simulator codes. The theoretical background on general transport theory, nodal diffusion calculation and the Monte Carlo method are discussed. The basic methodology used in the PSG code is introduced and previous studies related to the topic are briefly reviewed. PSG is validated by comparison to reference results produced by MCNP4C and CASMO-4E in infinite two-dimensional LWR lattice calculations. Group constants generated by PSG are used in ARES reactor simulator calculations and the results compared to reference calculations using CASMO-4E data. (orig.)

[en] Fortum has been supplying solar electricity systems for various applications since the beginning of the 1980s. The use of solar electricity has been growing at 15 - 20% a year and prices have dropped to one twentieth of their original level, and ever-more efficient products have been introduced. Major progress is also being made in the seasonal storage of solar energy due to developments in hydrogen technology

[en] In nuclear engineering several reactor physics problems can be approached using Monte Carlo neutron transport techniques, which usually give reliable results when properly used. The quality of the results is largely determined by the accuracy of the geometry model and the statistical uncertainty of the Monte Carlo calculation. There is, however, another potential source of error, namely the cross section data used with the Monte Carlo codes. It has been shown in several studies that there may be significant discrepancies between results calculated using cross section libraries based on different evaluated nuclear data files. These discrepancies are well known to the evaluators of nuclear data but less acknowledged by reactor physicists, who often rely on a single cross section library in their calculations. In this study, discrepancies originating from base nuclear data were investigated in a systematic manner using the MCNP4C code. Calculations on simplified UOX and MOX fuelled LWR lattices were carried out using cross section libraries based on ENDF/B-VI.8, JEFF-3.0, JENDL-3.3, JEF-2.2 and JENDL-3.2 evaluated data files. The neutron spectrum of the system was varied over a wide range by changing the ratio of hydrogen to heavy metal atoms. The essential isotopes underlying the discrepancies were identified and the roles of fission and absorption cross sections of the most important nuclides assessed. The results confirm that there are large systematic differences up to a few per cent in the multiplication factors of LWR lattices. The discrepancies are strongly dependent on material compositions and neutron spectra, and largely originate from U-238 and the primary fissile isotopes. It is concluded that these discrepancies should be taken into account in all reactor physics calculations, and that reactor physicists should not rely on results based on a single cross section library. (author)

[en] Discrepancies in fundamental nuclear data pose a potential source of uncertainty in neutron transport calculation. These discrepancies are often overlooked in reactor physics calculations. There is evidence that differences in the order of 1% in the multiplication factor are encountered when criticality calculations are carried out using different neutron cross section libraries. Monte Carlo transport calculation codes are especially prone to such discrepancies, since the base evaluated data can be used without modifications. This paper presents some results of a study, in which cross section library-based discrepancies in light water reactor criticality calculations were investigated in a systematic manner. Point-wise cross section libraries were generated from the ENDFB-VI.8, JEFF-3.0, JENDL-3.3, JEF-2.2 and JENDL-3.2 evaluated nuclear data files using the NJOY-99 nuclear data processing system. The comparison calculations were carried out using the MCNP4C Monte Carlo transport calculation code. A systematic method based on the neutron balance of the system was developed in order to study the origin of the reactivity discrepancies. The energy dependence of the cross section data was taken into account by dividing the neutron flux spectrum into four energy groups. The comparison calculations cover the most typical LWR operating conditions. The basic geometry is an infinite pin-cell lattice. The main free parameter in the system is the fuel-to-moderator ratio. Several variations of the basic geometry were studied, including lattices with burnable absorber and control pins, a finite lattice with leakage and lattices with low- and high-burnup fuel pins. (orig.)

[en] No abstract prepared. Power Point presentation

[en] SERPENT is a three-dimensional continuous-energy Monte Carlo reactor physics burnup calculation code, developed at VTT Technical Research Centre of Finland since 2004. The code is specialized in lattice physics applications, but the universe-based geometry description allows transport simulation to be carried out in complicated three-dimensional geometries as well. The suggested applications of SERPENT include generation of homogenized multi-group constants for deterministic reactor simulator calculations, fuel cycle studies involving detailed assembly-level burnup calculations, validation of deterministic lattice transport codes, research reactor applications, educational purposes and demonstration of reactor physics phenomena. The Serpent code has been publicly distributed by the OECD/NEA Data Bank since May 2009 and RSICC in the U. S. since March 2010. The code is being used in some 35 organizations in 20 countries around the world. This paper presents an overview of the methods and capabilities of the Serpent code, with examples in the modelling of WWER-440 reactor physics. (Author)

[en] PSG is a new Monte Carlo neutron transport code, developed at the Technical Research Centre of Finland (VTT). The code is mainly intended for fuel assembly-level reactor physics calculations, such as group constant generation for deterministic reactor simulator codes. This paper presents the current status of the project and the essential capabilities of the code. Although the main application of PSG is in lattice calculations, the geometry is not restricted in two dimensions. This paper presents the validation of PSG against the experimental results of the three-dimensional MOX fuelled VENUS-2 reactor dosimetry benchmark. (authors)

[en] This paper presents a dynamic neutron transport mode, currently being implemented in the Serpent 2 Monte Carlo code for the purpose of simulating short reactivity transients with temperature feedback. The transport routine is introduced and validated by comparison to MCNP5 calculations. The method is also tested in combination with an internal temperature feedback module, which forms the inner part of a multi-physics coupling scheme in Serpent 2. The demo case for the coupled calculation is a reactivity-initiated accident (RIA) in PWR fuel. (authors)

[en] FiR 1 is a pool-type Triga Mk-II research reactor with 250 kW thermal power. A model of the FiR 1 reactor has been previously generated for the Serpent Monte Carlo reactor physics and burnup calculation code. In the current article, this model is validated by comparing the predicted reaction rates of nickel and manganese at 9 different positions in the reactor to measurements. In addition, track-length estimators are implemented in Serpent 2.1.18 to increase its performance in dosimetry calculations. The usage of the track-length estimators is found to decrease the reaction rate calculation times by a factor of 7-8 compared to the standard estimator type in Serpent, the collision estimators. The differences in the reaction rates between the calculation and the measurement are below 20%

[en] Target Motion Sampling (TMS) is a stochastic on-the-fly temperature treatment technique that is being developed as a part of the Monte Carlo reactor physics code Serpent. The method provides for modeling of arbitrary temperatures in continuous-energy Monte Carlo tracking routines with only one set of cross sections stored in the computer memory. Previously, only the performance of the TMS method in terms of CPU time per transported neutron has been discussed. Since the effective cross sections are not calculated at any point of a transport simulation with TMS, reaction rate estimators must be scored using sampled cross sections, which is expected to increase the variances and, consequently, to decrease the figures-of-merit. This paper examines the effects of the TMS on the statistics and performance in practical calculations involving reaction rate estimation with collision estimators. Against all expectations it turned out that the usage of sampled response values has no practical effect on the performance of reaction rate estimators when using TMS with elevated basis cross section temperatures (EBT), i.e. the usual way. With 0 Kelvin cross sections a significant increase in the variances of capture rate estimators was observed right below the energy region of unresolved resonances, but at these energies the figures-of-merit could be increased using a simple re-sampling technique to decrease the variances of the responses. It was, however, noticed that the usage of the TMS method increases the statistical shifts of all estimators, including the flux estimator, by tens of percents in the vicinity of very strong resonances. This effect is actually not related to the usage of sampled responses, but is instead an inherent property of the TMS tracking method and concerns both EBT and 0 K calculations. (authors)