The European Journal of Physics N (EPJ-N)
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Effect of heat transfer correlations on the fuel temperature prediction of SCWRs
No full textIn this paper, we present a numerical analysis of the effect of different heat transfer correlations on the prediction of the cladding wall temperature in a supercritical water reactor at nominal operating conditions. The neutronics process with temperature feedback effects, the heat transfer in the fuel rod, and the thermal-hydraulics in the core were simulated with a three-pass core design
The choice of the fuel assembly for VVER-1000 in a closed fuel cycle based on REMIX-technology
No full textThis paper shows basic features of different fuel assembly (FA) application for VVER-1000 in a closed fuel cycle based on REMIX-technology. This investigation shows how the change in the water–fuel ratio in the VVER FA affects on the fuel characteristics produced by REMIX technology during multiple recycling
Resonance parameter and covariance evaluation for
No full textA resolved resonance evaluation was performed for 16O in the energy range 0 eV to 6 MeV using the computer code SAMMY resulting in a set of resonance parameters (RPs) that describes well the experimental data used in the evaluation. A RP covariance matrix (RPC) was also generated. The RP were converted to the evaluated nuclear data file format using the R-Matrix Limited format and the compact format was used to represent the RPC. In contrast to the customary use of RP, which are frequently intended for the generation of total, capture, and scattering cross sections only, the present RP evaluation permits the computation of angle dependent cross sections. Furthermore, the RPs are capable of representing the (n, α) cross section from the energy threshold (2.354 MeV) of the (n, α) reaction to 6 MeV. The intent of this paper is to describe the procedures used in the evaluation of the RP and RPC, the use of the RPC in benchmark calculations and to assess the impact of the 16O nuclear data uncertainties in the calculate dkeff for critical benchmark experiments
Statistical sampling applied to the radiological characterization of historical waste
No full textThe evaluation of the activity of radionuclides in radioactive waste is required for its disposal in final repositories. Easy-to-measure nuclides, like γ-emitters and high-energy X-rays, can be measured via non-destructive nuclear techniques from outside a waste package. Some radionuclides are difficult-to-measure (DTM) from outside a package because they are α- or β-emitters. The present article discusses the application of linear regression, scaling factors (SF) and the so-called “mean activity method” to estimate the activity of DTM nuclides on metallic waste produced at the European Organization for Nuclear Research (CERN). Various statistical sampling techniques including simple random sampling, systematic sampling, stratified and authoritative sampling are described and applied to 2 waste populations of activated copper cables. The bootstrap is introduced as a tool to estimate average activities and standard errors in waste characterization. The analysis of the DTM Ni-63 is used as an example. Experimental and theoretical values of SFs are calculated and compared. Guidelines for sampling historical waste using probabilistic and non-probabilistic sampling are finally given
Physical and economical aspects of Pu multiple recycling on the basis of REMIX reprocessing technology in thermal reactors
No full textThe basic strategy of Russian nuclear energy is propagation of a closed fuel cycle on the basis of fast breeder and thermal reactors, as well as the solution of the spent nuclear fuel accumulation and resource problems. The three variants of multiple Pu and U recycling in Russian pressurized water reactor concept reactors on the basis of REgenerated MIXture of U, Pu oxides (REMIX) reprocessing technology are considered in this work. The REMIX fuel is fabricated from an unseparated mixture of uranium and plutonium obtained during spent fuel reprocessing with further makeup by enriched natural U or reactor grade Pu. This makes it possible to recycle several times the total amount of Pu obtained from the spent fuel. The main difference in Pu recycling is the concept of 100% or partial fuel loading of the core. The third variant is heterogeneous composition of enriched uranium and uranium–plutonium mixed oxide fuel pins in one fuel assembly. It should be noted that all fuel assemblies with Pu require the involvement of expensive technologies during manufacturing. These three variants of the full core loadings can be balanced on zero Pu accumulation in the cycle. The various physical and economical aspects of Pu and U multiple recycling in selected variants are observed in the given work
PWR circuit contamination assessment tool. Use of OSCAR code for engineering studies at EDF
No full textNormal operation of PWR generates corrosion and wear products in the primary circuit which are activated in the core and constitute the major source of the radiation field. In addition, cases of fuel failure and alpha emitter dissemination in the coolant system could represent a significant radiological risk. Radiation field and alpha risks are the main constraints to carry out maintenance and to handle effluents. To minimize these risks and constraints, it is essential to understand the behavior of corrosion products and actinides and to carry out the appropriate measurements in PWR circuits and loop experiments. As a matter of fact, it is more than necessary to develop and use a reactor contamination assessment code in order to take into account the chemical and physical mechanisms in different situations in operating reactors or at design stage. OSCAR code has actually been developed and used for this aim. It is presented in this paper, as well as its use in the engineering studies at EDF. To begin with, the code structure is described, including the physical, chemical and transport phenomena considered for the simulation of the mechanisms regarding PWR contamination. Then, the use of OSCAR is illustrated with two examples from our engineering studies. The first example of OSCAR engineering studies is linked to the behavior of the activated corrosion products. The selected example carefully explores the impact of the restart conditions following a reactor mid-cycle shutdown on circuit contamination. The second example of OSCAR use concerns fission products and disseminated fissile material behavior in the primary coolant. This example is a parametric study of the correlation between the quantity of disseminated fuel and the variation of Iodine 134 in the primary coolant
Testing of high temperature materials within HTR program in Czech Republic
No full textResearch institutes and also industrial companies in Czech Republic are involved in High Temperature Gas Cooled Reactor (HTGR) program and activities related to the study of advanced materials and HTGR technologies. These activities are supported by EC (within international projects, e.g. FP7-ARCHER, ALLIANCE, GoFastR can be mentioned) and also by Technology Agency of Czech Republic. Within these activities, degradation of metallic and ceramic materials in the high temperature helium atmosphere is investigated, and also new experimental facilities for material testing are built. As examples of tested materials, Alloy 800 H, ferritic steel P91, austenitic steel 316, Inconel 713 and 738 and corundum ceramics could be named. The selected results of exposure experiments in the high temperature helium environment are presented in this paper
Cross check of the new economic and mass balance features of the fuel cycle scenario code TR_EVOL
No full textVersatile computational tools with up to date capabilities are needed to assess current nuclear fuel cycles or the transition from the current status of the fuel cycle to the more advanced and sustainable ones. This work is intended to cross check the new capabilities of the fuel cycle scenario code TR_EVOL. This process has been divided in two stages. The first stage is dedicated to check the improvements in the nuclear fuel mass balance estimation using the available data for the Spanish nuclear fuel cycle. The second stage has been focused in verifying the validity of the TR_EVOL economic module, comparing results to data published by the ARCAS EU project. A specific analysis was required to evaluate the back-end cost. Data published by the waste management responsible institutions was used for the validation of the methodology. Results were highly satisfactory for both stages. In particular, the economic assessment provides a difference smaller than 3% regarding results published by the ARCAS project (NRG estimations). Furthermore, concerning the back-end cost, results are highly acceptable (7% difference for a final disposal in a once-through scenario and around 11% for a final disposal in a reprocessing strategy) given the significant uncertainties involved in design concepts and related unit costs
Thermal-hydraulics/thermal-mechanics temporal coupling for unprotected loss of flow accidents simulations on a SFR
No full textIn the frame of ASTRID designing, unprotected loss of flow (ULOF) accidents are considered. As the reactor is not scrammed, power evolution is driven by neutronic feedbacks, among which Doppler effect, linked to fuel temperature, is prominent. Fuel temperature is calculated using thermal properties of fuel pins (we will focus on heat transfer coefficient between fuel pellet and cladding, Hgap, and on fuel thermal conductivity, λfuel) which vary with irradiation conditions (neutronic flux, mass flow and history for instance) and during transient (mainly because of dilatation of materials with temperature). In this paper, we propose an analysis of the impact of spatial variation and temporal evolution of thermal properties of fuel pins on a CFV-like core [M.S. Chenaud et al., Status of the ASTRID core at the end of the pre-conceptual design phase 1, in Proceedings of ICAPP 2013, Jeju Island, Korea (2013)] behavior during an ULOF accident. These effects are usually neglected under some a priori conservative assumptions. The vocation of our work is not to provide a best-estimate calculation of ULOF transient, but to discuss some of its physical aspects. To achieve this goal, we used TETAR, a thermal-hydraulics system code developed by our team to calculate ULOF transients, GERMINAL V1.5, a CEA code dedicated to SFR pin thermal-mechanics calculations and APOLLO3®, a neutronic code in development at CEA
Three-component U-Pu-Th fuel for plutonium irradiation in heavy water reactors
No full textThis paper discusses concepts for three-component fuel bundles containing plutonium, uranium and thorium for use in pressurised heavy water reactors, and cases for and against implementation of such a nuclear energy system in the United Kingdom. Heavy water reactors are used extensively in Canada, and are deploying within India and China, whilst the UK is considering the use of heavy water reactors to manage its plutonium inventory of 140 tonnes. The UK heavy water reactor proposal uses a mixed oxide (MOX) fuel of plutonium in depleted uranium, within the enhanced CANDU-6 (EC-6) reactor. This work proposes an alternative heterogeneous fuel concept based on the same reactor and CANFLEX fuel bundle, with eight large-diameter fuel elements loaded with natural thorium oxide and 35 small-diameter fuel elements loaded with a MOX of plutonium and reprocessed uranium stocks from UK MAGNOX and AGR reactors. Indicative neutronic calculations suggest that such a fuel would be neutronically feasible. A similar MOX may alternatively be fabricated from reprocessed <5% enriched light water reactor fuel, such as the fuel of the AREVA EPR reactor, to consume newly produced plutonium from reprocessing, similar to the DUPIC (direct use of PWR fuel in CANDU) process