The COVID-19 induced restrictions have prevented reactor physics students from attending in-person reactor physics exercises which are a vital part of their education. Jo.zef Stefan Institute has organized remote exercises with the help of off-the-shelf technology, including multiple videoconferencing setups, remote desktop software, portable cameras, a dome camera, shared spreadsheets, and a common whiteboard. The students were encouraged to actively participate in the exercises by giving instructions to the reactor operator, asking and answering questions, logging data, operating digital acquisition systems, and performing analysis during the exercise. The first remote exercises were organized as a five-day course of experimental reactor physics for students from Uppsala University. The feedback was collected after the course using an anonymous online form. The aspects being evaluated included the technical content, quality of material, performance of the individual lecturers and the quality of the remote session. In general, the response was overwhelmingly positive, with most questions with a rating scale answered with "excellent".

The COVID-19 induced restrictions have prevented reactor physics students from attending in-person reactor physics exercises which are a vital part of their education. Jožef Stefan Institute has organized remote exercises with the help of off-the-shelf technology, including multiple videoconferencing setups, remote desktop software, portable cameras, a dome camera, shared spreadsheets, and a common whiteboard. The students were encouraged to actively participate in the exercises by giving instructions to the reactor operator, asking and answering questions, logging data, operating digital acquisition systems, and performing analysis during the exercise. The first remote exercises were organized as a five-day course of experimental reactor physics for students from Uppsala University. The feedback was collected after the course using an anonymous online form and was generally positive but has revealed some problems with sound quality which were resolved later. The Jožef Stefan Institute can also organize a remote course during a full lockdown when the reactor is not able to operate using the in-house developed Research Reactor Simulator based on a point kinetics approximation and a simple thermohydraulic module.

This work aims at measuring prompt-fission neutrons at different excitation energies of the nucleus. Two independent techniques, the 2E-2v and the 2E techniques, are used to map the characteristics of the mass-dependent prompt fission neutron multiplicity, v (A), when the excitation energy is increased. The VERDI 2E-2v spectrometer is being developed at JRC-GEEL. The Fission Fragment (FF) energies are measured using two arrays of 16 silicon (Si) detectors each. The FFs velocities are obtained by time-of-flight, measured between micro-channel plates (MCP) and Si detectors. With MCPs placed on both sides of the fission source, VERDI allows for independent timing measurements for both fragments. 252 Cf(sf) was measured and the present results revealed particular features of the 2E-2v technique. Dedicated simulations were also performed using the GEF code to study important aspects of the 2E-2v technique. Our simulations show that prompt neutron emission has a non-negligible impact on the deduced fragment data and affects also the shape of v (A). Geometrical constraints lead to a total-kinetic energy-dependent detection efficiency. The 2E technique utilizes an ionization chamber together with two liquid scintillator detectors. Two measurements have been performed, one of 252 Cf(sf) and another one of thermal-neutron induced fission in 235 U(n,f). Results from 252 Cf(sf) are reported here.

The nuclear de-excitation process of fission fragments (FF) provides fundamental information for the understanding of nuclear fission and nuclear structure in neutron-rich isotopes. The variation of the prompt-neutron multiplicity, ν(A), as a function of the incident neutron energy (En) is one of many open questions. It leads to significantly different treatments in various fission models and implies that experimental data are analyzed based on contradicting assumptions. One critical question is whether the additional excitation energy (Eexc) is manifested through an increase of ν(A) for all fragments or for the heavy ones only. A systematic investigation of ν(A) as a function of En has been initiated. Correlations between prompt-fission neutrons and fission fragments are obtained by using liquid scintillators in conjunction with a Frisch-grid ionization chamber. The proof-of-principle has been achieved on the reaction 235U(nth,f) at the Van De Graff (VdG) accelerator of the JRC-Geel using a fully digital data acquisition system. Neutrons from 252Cf(sf) were measured separately to quantify the neutron-scattering component due to surrounding shielding material and to determine the intrinsic detector efficiency. Prelimenary results on ν(A) and spectrum in correlation with FF properties are presented.

The Swedish Centre for Nuclear Technology (Svenskt Karntekniskt Centrum, SKC) is a national initiative to perform industry relevant research at Swedish universities, and to support dedicated education of direct use to the Swedish nuclear industry. SKC has been the meeting point between industry and academia for 25 years, and has coped with varying needs from industry and political situations. The present situation in the Nordic countries is split: Sweden plans to shut down four out of ten reactors by 2020, while Finland is planning and constructing new reactors. Even without a strong signal to construct new reactors in Sweden, the need for nuclear competence will stay, as we have challenges in front of us to operate and dismantle power plants, operate the intermediate storage facility CLAB in Oskarshamn, and to build and fill the final repositories in Forsmark. The education supported by the SKC at the selected universities will facilitate possible recruitment for nuclear installations. The funding body of SKC consists of all the Swedish Nuclear Power Plants (NPP) situated in Forsmark (three BWRs), Oskarshamn (three BWRs) and Ringhals (one BWR and three PWRs), and the nuclear fuel manufacturer (Westinghouse), while the main research and education is carried out at Chalmers, KTH Royal Institute of Technology and Uppsala University with corresponding in-kind contribution. The research activities cover highly requested studies for today’s nuclear fleet: material embrittlement, stress-corrosion cracking, accident-tolerant fuel development and ageing management for long-term operation (LTO), while the educational part consists of Bachelor and Master programmes as well as elective courses for students outside the main nuclear programmes and contract education. In the master programmes, focus is on e-learning platforms for courses and examination. Examples from such development are to be presented in the full contribution to the conference. Another success story is project based courses in industry, especially within the Bachelor programmes. This is highly appreciated by students, providing a direct contact with future employers, The success of SKC originates from close contact between the funding bodies and academia on many levels: base funding for course preparation, project support in annual calls, and dedicated long-term research funding. But also dedicated industrial experts, and an enthusiasm from academia to enlighten present research issues, as well as strong presence at universities during student fairs and career days. That is LTO of SKC!

Criticality, reactor physics and shielding benchmarks are expected to play important roles in GEN-IV design, safety analysis and in the validation of analytical tools used to design these reactors. For existing reactor technology, benchmarks are used for validating computer codes and for testing nuclear data libraries. Given the large number of benchmarks available, selecting these benchmarks for specic applications can be rather tedious and dicult. Until recently, the selection process has been based usually on expert judgement which is dependent on the expertise and the experience of the user and thereby introducing a user bias into the process. This approach is also not suitable for the Total Monte Carlo methodology which lays strong emphasis on automation, reproducibility and quality assurance. In this paper a method for selecting these benchmarks for reactor calculation and for nuclear data uncertainty reduction based on the Total Monte Carlo (TMC) method is presented. For reactor code validation purposes, similarities between a real reactor application and one or several benchmarks are quantied using a similarity index while the Pearson correlation coecient is used to select benchmarks for nuclear data uncertainty reduction. Also, a correlation based sensitivity method is used to identify the sensitivity of benchmarks to particular nuclear reactions. Based on the benchmark selection methodology, two approaches are presented for reducing nuclear data uncertainty using integral benchmark experiments as an additional constraint in the TMC method: a binary accept/reject and a method of assigning le weights using the likelihood function. Finally, the methods are applied to a full lead-cooled fast reactor core and a set of criticality benchmarks. Signicant reductions in 239 Pu and 208 Pb nuclear data uncertainties were obtained after

In the Total Monte Carlo (TMC) method [1] developed at the Nuclear Research and Consultancy Group for nuclear data uncertainty propagation, model calculations are compared with differential experimental data and a specific a priori uncertainty is assigned to each model parameter. By varying the model parameters all together within model parameter uncertainties, a full covariance matrix is obtained with its off diagonal elements if desired [1]. In this way, differential experimental data serve as a constraint for the model parameters used in the TALYS nuclear reactions code for the production of random nuclear data files. These files are processed into usable formats and used in transport codes for reactor calculations and for uncertainty propagation to reactor macroscopic parameters of interest. Even though differential experimental data together with their uncertainties are included (implicitly) in the production of these random nuclear data files in the TMC method, wide spreads in parameter distributions have been observed, leading to large uncertainties in reactor parameters for some nuclides for the European Lead cooled Training Reactor [2]. Due to safety concerns and the development of GEN-IV reactors with their challenging technological goals, the present uncertainties should be reduced significantly if the benefits from advances in modelling and simulations are to be utilized fully [3]. In Ref.[4], a binary accept/reject approach and a more rigorous method of assigning file weights based on the likelihood function were proposed and presented for reducing nuclear data uncertainties using a set of integral benchmarks obtained from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP). These methods are depended on the reference nuclear data library used, the combined benchmark uncertainty and the relevance of each benchmark for reducing nuclear data uncertainties for a particular reactor system. Since each nuclear data library normally comes with its own nominal values and covariance matrices, reactor calculations and uncertainties computed with these libraries differ from library to library. In this work, we apply the binary accept/reject approach and the method of assigning file weights based on the likelihood function for reducing a priori 239Pu nuclear data uncertainties for the European Lead Cooled Training Reactor (ELECTRA) using a set of criticality benchmarks. Prior and posterior uncertainties computed for ELECTRA using ENDF/B-VII.1, JEFF-3.2 and JENDL-4.0 are compared after including experimental information from over 10 benchmarks.[1] A.J. Koning and D. Rochman, Modern Nuclear Data Evaluation with the TALYS Code System. Nuclear Data Sheets 113 (2012) 2841-2934. [2] E. Alhassan, H. Sjostrand, P. Helgesson, A. J. Koning, M. Osterlund, S. Pomp, D. Rochman, Uncertainty and correlation analysis of lead nuclear data on reactor parameters for the European Lead Cooled Training reactor (ELECTRA). Annals of Nuclear Energy 75 (2015) 26-37. [3] G. Palmiotti, M. Salvatores, G. Aliberti, H. Hiruta, R. McKnight, P. Oblozinsky, W. Yang, A global approach to the physics validation of simulation codes for future nuclear systems, Annals of Nuclear Energy 36 (3) (2009) 355-361. [4] E. Alhassan, H. Sjostrand, J. Duan, P. Helgesson, S. Pomp, M. Osterlund, D. Rochman, A.J. Koning, Selecting benchmarks for reactor calculations: In proc. PHYSOR 2014 - The Role of Reactor Physics toward a Sustainable Future, kyoto, Japan, Sep. 28 - 3 Oct. (2014).

A common problem for many liquid metal cooled fast reactor designs is the positive void worth of the coolant. In this context, an advantage of lead cooled fast reactors is the high temperature of c ...

Double-differential cross sections of light ions (p, d, t, 3 He, and α) from carbon, oxygen, and silicon induced by 175 MeV quasi mono-energetic neutrons were measured using the Medley setup at the The Svedberg Laboratory (TSL) in order to benchmark evaluated nuclear data and nuclear reaction models.The Medley is a conventional spectrometer system which consists of eight counter telescopes.Each telescope is composed of two silicon surface barrier detectors as the ΔE detectors and a CsI(Tl) scintillator as the E detector for particle identification.The telescopes are placed at angles from 20° to 160° in steps of 20°.The measured double-differential production cross sections of light ions for oxygen and silicon are compared with PHITS calculations using the following nuclear reaction options: the high-energy nuclear data library (JENDL/HE-2007), the quantum molecular dynamics (QMD) model, the modified quantum molecular dynamics (MQMD) model, and the intra-nuclear cascade (INC) model based on the Bertini Model.

In a lead-cooled fast reactor (LFR), small bubbles (in the order of one mm or less) may enter the coolant from a leaking steam generator. If such a leakage is undetected the small bubbles may eventually coalesce into a larger bubble in local stagnation zones under the active core. If such a bubble or void releases and passes through the core, it could drive the reactor into prompt criticality. It is therefore desirable to be able to detect the initial stages of such void formation.In this thesis, a methodology to detect such leaks is presented together with a study on void-induced reactivity effects in various LFR's. The methodology developed is based on information from two fission chambers positioned radially outside the core. The fissile content of the fission chambers consist either of 235U or 242Pu making them sensitive to different parts of the neutron spectrum. It is shown that the information from the fission chambers can be used to obtain an early indication of the presence of a small leak within typically a month. Furthermore, it is shown that for all but the smallest LFR’s, prompt criticality due to voids passing the core cannot be excluded.One conclusion is that the methodology may form an attractive complement to the general monitoring system of future LFR’s but, as is noted, it has potential for further developments.

The Total Monte Carlo (TMC) method was used in this study to assess the impact of Pb-206, 207 and 208 nuclear data uncertainties on keff , βeff, coolant temperature coefficient, the coolant void worth for the ELECTRA reactor. Relatively large uncertainties were observed in the keff and the coolant void worth for all the isotopes with significant contribution coming from Pb-208 nuclear data. The large Pb-208 nuclear data uncertainty observed was further investigated by studying the impact of partial channels on the keff and the βeff. Various sections of ENDF file: elastic scattering (n, el), inelastic scattering (n, inl), neutron capture (n, γ), (n, 2n), resonance parameters and the angular distribution were varied randomly and distributions in keff and the βeff obtained. The dominant contributions to the uncertainty in the keff from Pb-208 came from uncertainties in the resonance parameters; however, elastic scattering cross section and the angular distribution also had significant impact. The impact of nuclear data uncertainties on the βeff was observed to be small.

Nuclear data for neutron-induced reactions in the intermediate energy range of 20 to 200 MeV are of importance for several different applications, notably accelerator-driven incineration of nuclear waste. The Medley setup located at The Svedberg Laboratory in Uppsala, Sweden was used for a series of measurements of p, d, t, 3 He and alpha particle production from 175 MeV neutrons on various target nuclei. Medley uses an arrangement of 8 detector telescopes placed at angles from 20 to 160 degrees relative to the neutron beam to register event and discriminate among the particle types. Using the ΔE-ΔE-E technique we were able to measure double-differential cross sections over a wide dynamic range. This paper describes the experimental set-up, summarizes the data analysis and reports on recent changes in the previously reported preliminary data set on bismuth. Experimental data are compared with several different model calculations. In this paper we specifically compare the experimental results with INCL4 intra-nuclear cascade calculations combined with the ABLA de-excitation code. Considering the fact that the experimental data are obtained at an energy normally considered too low for intra-nuclear cascade calculations, we do find some systematic difference but generally reasonable agreement.

Criticality, reactor physics, fusion and shielding benchmarks are expected to play important roles in GENIV design, safety analysis and in the validation of analytical tools used to design these re ...

The lead-cooled fast reactor (LFR) is one of the nuclear reactor technologies proposed by the Generation IV International Forum (GIF). The lead coolant allows for inherent safety properties attractive from a nuclear safety point of view, but issues related to corrosion of structural materials and the possible positive coolant reactivity coefficient must be addressed before LFRs can be commercially viable. As an example, a small crack in e.g. a heat exchanger can generate a more or less homogeneous distribution of bubbles in the coolant (void) which if unnoticed, has the potential to cause criticality issues. This fact motivated an investigation of a methodology to detect such voids.The suggested methodology is based on measurements of the "slow" and "fast" parts of the neutron spectrum because these parts respond in different ways to voiding. For detection, it is tentatively assumed that fission chambers loaded with U-235 and Pu-239, respectively, are deployed. To investigate the methodology according to sensitivity and precision, a number of scenarios have been simulated and analysed using the core simulator Serpent.The results show that the methodology yields a sensitivity of 3% for each per cent unit of void. Assuming typical detection limits of a few per cent this implies the possibility to detect voids down to the order of 1%. From these studies it was also concluded that the positioning of the detectors relative the reactor core is crucial, which may be useful input during the design phase of a reactor in order to achieve an efficient monitoring system. (C) 2013 Elsevier Ltd. All rights reserved.

Boiling water reactor (BWR) bottom reflector calculations in lattice codes such as CASMO are presently used only to produce accurate boundary conditions for core interfaces in nodal diffusion codes. Homogenized cross-section constants and discontinuity factors are calculated in one dimension (1-D) without the explicit presence of the control rod absorber. If the spatial flux in a BWR bottom reflector is required, for example, for depletion calculations of withdrawn control rods, the homogenization of the reflector must be based on a representation of the three-dimensional (3-D) geometry and material composition that is as true as possible.This paper investigates differences in cross-section and discontinuity factors from 1-D calculations in CASMO with 3-D Monte Carlo calculations of a realistic bottom reflector model in MCNP5. The cross-section and discontinuity factors from CASMO and MCNP5 are furthermore implemented in the nodal diffusion code SIMULATES to investigate the effect on the neutron fluxes in the bottom reflector.The results show that for the case investigated, the 1-D homogenization in CASMO5 produces a 26% overestimation of the homogenized thermal absorption cross section in the reflector and a 62% underestimation of the homogenized fast absorption cross section. These cross-section differences have essentially no impact on the neutron flux in the core but cause a 4.5% and 12.3% underestimation of the thermal and fast neutron flux, respectively, in the reflector region.

We have measured double-differential yields of protons, deuterons, and tritons produced from carbon induced by 175 MeV quasi mono-energetic neutrons using the MEDLEY setup at the TSL neutron beam facility.The measured data are used for benchmarking of a high-energy nuclear data file, JENDL/HE-2007, and both intra-nuclear cascade (INC) model and quantum molecular dynamics (QMD) calculations.

We have measured fission fragment-mass yields for neutron-induced fission of $^{232}\mathrm{Th}$ and $^{238}\mathrm{U}$ at energies 32.8, 45.3, and 59.9 MeV. The experiments were done at quasimonoenergetic neutron beams of the Cyclotron Research Center at Louvain-la-Neuve. To detect the fission fragments, a multisection Frisch-gridded ionization chamber was used. The measurement and data analysis techniques are discussed in detail. The obtained mass yields are compared to model calculations with the intermediate-energy nuclear reaction code MCFX. The MCFX code is used to calculate the fraction of fissioning nuclei after cascade, preequilibrium, and statistical reaction stages. The formation of mass distributions is considered as a result of oscillations of the mass-asymmetry degree of freedom in the potential well calculated with the temperature-dependent shell correction method. The experimental results as well as the results of the model calculations demonstrate that the probability of symmetric fission increases with incident neutron energy for both nuclei. The comparison also shows that the symmetric fission is more enhanced for thorium than for uranium with increasing neutron energy. We also compare $^{238}\mathrm{U}$ results with available experimental data; the $^{232}\mathrm{Th}$ data were measured for the first time.

Models of the neutron flux shape in a withdrawn control rod in a boiling water reactor (BWR) bottom reflector have been constructed from simulations with the Monte Carlo code MCNP. These neutron flux models are intended for determining absorber depletion and fast fluence accumulation for withdrawn control rods with nodal codes.So-called G-factors are created for coupling the neutron flux models to a conventional nodal code via the core bottom neutron flux.The neutron flux models and G-factors are created for three different neutron energies, and their dependence on various parameters such as blanket enrichments, Hf and B(4)C control rod absorber, and depletion and reflector geometry is investigated.The neutron flux models and G-factors are found to be very insensitive; the neutron flux models predict the simulated neutron flux in the withdrawn control rod from MCNP over a variety of reflector configurations with an error < 3.0%. This implies that the neutron flux models constructed in this paper are generally applicable for BWR reflectors and control rods not fundamentally deviating from the designs investigated in this paper.

Control rod withdrawal in BWRs induces large power steps in the adjacent fuel assemblies. This paper investigates how well a 2D/3D method, e.g., CASMO5/SIMULATE5 computes axial pin power gradients adjacent to an asymmetrical control-rod tip in a BWR. The ability to predict pin power gradients accurately is important for safety considerations whereas large powers steps induced by control rod withdrawal can cause Pellet Cladding Interaction. The computation of axial pin power gradients axially around a control rod tip is a challenging task for any nodal code. On top of that, asymmetrical control rod handles are present in some BWR designs. The lattice code CASMO requires diagonal symmetry of all control rod parts. This introduces an error in computed pin power gradients that has been evaluated by Monte Carlo calculations.The results show that CASMO5/SIMULATE5, despite the asymmetrical control rod handle, is able to predict the axial pin power gradient within 1%/cm for axial nodal sizes of 15-3.68 cm. However, a nodal size of 3.68 cm still causes underestimations of pin power gradients compared with 1 cm nodes. Furthermore, if conventional node sizes are used, similar to 15 cm, pin power gradients can be underestimated by over 50% compared with 1 cm nodes.The detailed axial pin power profiles from MCNP are corroborated by measured gamma scan data on fuel rods irradiated adjacent to control rods. (C) 2011 Elsevier Ltd. All rights reserved.