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Sample reactivity measurements of CaH2 samples as the integral experiments were conducted in the Kinki University Reactor (UTR-KINKI). The sample reactivity worth was determined from a difference between two excess reactivities of the respective reactors with and without the sample. The reactivity worth measurements were repeated for 21.20 g, 41.00 g, 72.78 g and 164.4 g of CaH2 sample. Furthermore, the sample reactivity worth was calculated using the continuous energy Monte Carlo codes MVP3.0 with the nuclear library JEFF-3.1, where each cross section of H and Ca constituting a molecule of CaH2 was taken into the thermal neutron scattering law. The ratios of calculated to experimental values (C/E) were 1.119±0.587, 0.806±0.305, 0.874±0.165 and 0.894±0.099 for 21.29 g, 41.00 g, 78.72 g and 164.4 g of the samples, respectively. [ABSTRACT FROM AUTHOR]

During the wintertime, concrete pavements experience harsh exposure conditions due to the presence of both the freezing–thawing and wetting–drying cycles. Airport concrete pavements are commonly de-iced using chloride-free organic salts such as potassium formate or potassium acetate. However, these materials contain alkali ions which can have harmful effects on both the cement matrix and the aggregate. Specifically, there is an increased risk of occurrence of the alkali–silica reaction (ASR). The goal of this research was to estimate the influence of potassium formate on the potential of causing alkali–silica reaction in aggregates with different categories of reactivity (R0, R1, R2). The accelerated mortar bar test and its modification (which involves replacing sodium hydroxide solution with a potassium formate solution) were used. Detailed SEM-EDS examinations were performed to confirm the presence of alkali–silica reaction and to analyze the influence of potassium formate on the microstructure of mortar. [ABSTRACT FROM AUTHOR]

The CABRI experimental pulse reactor, located at the Cadarache nuclear research center, southern France, is devoted to the study of Reactivity Initiated Accidents (RIA). When the thermal-hydraulic conditions representative of a pressurized water nuclear reactor are reached (mostly temperature, pressure and flowrate conditions), a fuel test rod is submitted to a power excursion, triggered by the specific 3He reactivity injection system, in order to simulate a control rod ejection accident. The experiment, managed by IRSN, aims at studying both the fuel and cladding behavior of the test rod placed into the center of the reactor during the power excursion. Several test rods pre-irradiated in pressurized nuclear power plants and with different characteristics (burn up, cladding material, fuel type, Zirconia thickness) are considered for the programs performed in the CABRI reactor. Physical phenomena occurring during the power transient are monitored by various measuring systems designed or operated by IRSN. Each system provides information linked with the different phases of the experiment. Three main measuring systems will be considered in this paper: • The test devices, a sample holder that is also implemented with almost fifty sensors used to monitor the environmental parameters in the test channel such as temperature and pressure, and to control the rod behavior during the test sequence; • The Hodoscope, an online fuel motion measurement system, which aims at analyzing the fuel motion deduced from the detection of fast neutrons emitted by the test rod, with a time step of 1ms during the transient; • The IRIS facility, conceived to perform X-ray radiography and tomography imaging before and after a power transient thanks to a linear electron accelerator, as well as quantitative gamma scanning analyses. During the experimental sequence, different events are recorded. This paper focuses on a cladding failure that occurred during the transient and that can be revealed by the three systems mentioned above. The test device instrumentation allows to determine the timing of the failure and to analyze its consequences in the vicinity of the test rod from a thermal and hydraulic point of view, while the hodoscope measures fuel elongation and relocation during the power excursion. The IRIS facility, then, helps to confirm the failure, its location and its extent. These three systems are complementary and they allow the analysis of the same event from different perspectives. Their combination will ease the interpretation of the events in the next steps for the test results analysis. The study case taken into account in this paper concerns nuclear fuel after three irradiation cycles in a commercial PWR. [ABSTRACT FROM AUTHOR]

Criticality experiments are the foundation of the criticality safety validation, the reactor parameter prediction and the nuclear data validation. Criticality experiments have been used in the field of nuclear data adjustment in the last decades. In applications like criticality safety validations and nuclear data adjustments, many criticality experiments are used together in one application. In practice, experts found that some experiments have bad influence in nuclear data adjustments, and they excluded them in these applications. But the reason why these experiments should be excluded is not clear. To give these exclusion a clear physical explanation, we have developed the cross-evaluation method, which could evaluate the random biases of the experimental results by analyzing the C − E (Calculation result - Experiment result) values of similar experiments. In this paper, we use the cross-evaluation method to assess the random biases of some highly enriched metal uranium fast criticality experiments. By the cross-evaluation method, experts could choose criticality experiments which should be used in the applications of criticality safety validations or nuclear data adjustments, and might find the reason why some experiments should be excluded in applications of nuclear data adjustments. [ABSTRACT FROM AUTHOR]

The criticality safety control technique is required for the fuel debris removal from the Fukushima Dai-Ichi Nuclear Power Station which experienced a severe accident. The subcriticality estimation is expected to be done with only limited information about the fuel debris while the primary containment vessel internal survey work is ongoing. The purpose of this study is to develop the subcriticality estimation method called the virtual neutron capture method. The neutrons from the surface of the fuel debris represent a major portion of detector counts. The method consists of two evaluations: the evaluation at the surface of the fuel debris for which the isotope compositions are known by fuel debris sampling and the evaluation at the region of the fuel debris for which these compositions are unknown. For the unknown composition region, the average isotope composition with arbitrary water content is given. The method surveys the relationship with the detector count and the neutron multiplication factor with any size of the unknown composition region and any ratio of the water content before the on-site evaluation. The method is verified by experiments done in the Kyoto University Critical Assembly. The method shows that the maximum difference from the reference neutron multiplication factor is 4.5 %dk. As a result, the virtual neutron capture method can be adopted to the subcriticality monitoring if the method includes the estimation margin of 4.5 %dk within the neutron multiplication factor range from 0.70 to 0.95. [ABSTRACT FROM AUTHOR]

Higher enrichment of nuclear fuel along the manufacturing limit of boron content in steel and aluminum alloys represents a significant challenge in designing spent fuel transport and storage facilities. One possible solution for spent fuel pools and casks is the burnup credit method that allows for decreasing very high safety margins associated with fresh fuel assumption in spent fuel facilities. An alternative solution based on placing neutron absorber material directly into the fuel assembly is proposed here. A neutron absorber permanently fixed in guide tubes decreases system reactivity more efficiently than absorber sheets between the assemblies. The efficiency of the newly proposed concept is demonstrated on the criticality safety analysis of the GBC-32 spent fuel cask. Absorber rods from 8 different elements are placed within Westinghouse OFA 17x17 guide tubes. Currently used boron is a good option because of high absorption cross section, low atomic mass and chemical compatibility with various alloys. Alternative options (e.g., Sm, Eu, Gd, Dy, Hf, Re, Ir) are based on very good absorbers that do not require alloy compatibility since the absorbers can be placed inside zirconium or steel cladding. Because of high efficiency of the newly proposed absorber concept, boron content in BORAL sheets can be decreased to more competitive economics. Moreover, fuel assembly pitch is investigated in order to change cask wall inner diameter that will result in lower material consumption for the cask wall with the same shielding thickness. [ABSTRACT FROM AUTHOR]

The ANSI/ANS-8.1 standard, "Safety Standard for Operations with Fissionable Materials Outside Reactors," has been available since 1964 as ASA N6.1-1964. In 1969, this standard was revised as ANSI N16.1-1969, "Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors." This version of the standard includes a variety of subcritical limits (SCLs) for uniform aqueous solutions and metals containing fissile nuclides for 233U, 235U, and 239Pu. Furthermore, SCLs are also included for uranium-water lattices. In the 1983 version of ANSI/ANS-8.1 (a revision of ANSI N16.1-1975), the suite of SCLs in the standard expanded to include 235U enrichment limits for homogeneous uranium-water mixtures and dry/damp oxides, uniform aqueous solutions of low-enriched uranium, and uniform aqueous mixtures of Pu(NO3)4 containing 240Pu, in addition to the SCLs included in ANSI N16.1-1969. The SCLs have changed little in subsequent revisions (ANSI/ANS-8.1-1998 and ANSI/ANS-8.1-2014). The ANSI/ANS-8.1-2014 standard is currently being revised to include new SCLs (uranium metal and compounds with enrichments up to 20 wt. % 235U) and possible updates to the current SCLs already in the standard, although these SCLs will not be available to the nuclear criticality safety community for a number of years. The bases for these SCLs were documented in journal articles such as Nuclear Science and Engineering, and the American Nuclear Society's meeting transactions; however, the bases were ambiguous enough that sites and regulators in the United States are reluctant to endorse them for safety purposes. The purpose of this paper is to present the results of a comparison study for the SCLs in the ANSI/ANS-8.1-2014 standard using modern codes and cross sections (SCALE/ENDF/B-VIII) to provide some assurance about their quality (bias and bias uncertainty) for use in nuclear criticality safety applications. [ABSTRACT FROM AUTHOR]

We investigated the effect of β on effective multiplication factor(keff) in the 1/fβ spectrum random system. The random system was generated by the 1/fβ noise model. The model is a continuous space model based on the Randomized Weierstrass function and describes the component spatial distribution with a power spectrum of 1/fβ, where f and β are the frequency domain variable and the characteristic parameter related to randomness, respectively. In this work, the two-group Monte Carlo calculations were performed to obtain the keff for a simple cubic geometry that consisted of two materials (fuel burned to 12 GWd/t and concrete). A large number of replicas having different spatial distribution and characterized by the representative β values were generated using the model, and the distribution on keff was analyzed. We found the dependency on β of standard deviation, skewness, and kurtosis of keff distribution. This result is expected to help to predict the keff distribution due to the randomizing model. [ABSTRACT FROM AUTHOR]

The CASMO/SIMULATE/MCNP/FISPACT-II calculation route has been established at the Paul Scherrer Institute (PSI) for reactor dosimetry and activation studies. Furthermore, the in-house tool NUSS is in use at PSI for nuclear data (ND) related uncertainties quantifications with Monte Carlo neutron transport calculations. The use of randomly sampled ACE-formatted ND files not only allows propagation of the ND uncertainties, but also can serve for assessing the applicability of different types of experimental data for validation of calculation predictions of parameters of interest. In the present work an application of the PSI calculation scheme for analysis of activation reaction rates and the fast neutron fluence (FNF), throughout the Swiss pressurised water reactor (PWR) and simplified containment building models, is demonstrated. As particular examples of potentially available experimental data, two kinds of the neutron flux monitors are considered: a) the reactor pressure vessel scraping samples and detectors placed in the dosimetry channels, mounted at the core barrel and designed for validation of FNF calculations, and b) the ex-vessel dosimeters, specifically used by the Swiss waste management organisation (NAGRA) for validation of bio-shield activation predictions. The calculations were done with the ENDF/B-VII.1 library. The obtained results demonstrate importance of the ND uncertainties for the dosimetry evaluations. The assessment of the applicability of the selected experimental information for validation of the bio-shield irradiation calculations was done based on evaluation of the ND-related Pearson correlation coefficients. [ABSTRACT FROM AUTHOR]

Virtual Environmental for Reactor Analysis (VERA) benchmark was released by the Consortium for Advanced Simulation of Light water reactors (CASL) project in 2012. VERA benchmark includes more than ten problems at different levels, from 2D fuel pin case to 2D fuel assembly case to 3D core refuelling case, in addition, reference results and experimental measured data of some problems were provided by CASL. Fuel assemblies in VERA benchmark are various, including control rod assemblies, Pyrex assembly, IFBA assembly, WABA assembly and gadolinium poison assembly, and so on. In this paper, various fuel assembly models in the VERA benchmark have been built by using KYIIN-V2.0 code to verify its calculation ability from 2D fuel pin case to 2D fuel assembly case to 2D 3x3 fuel assembly case, and making a comparative analysis on the reference results in VERA benchmark, as well as the calculation results of the Monte Carlo code RMC. KYLIN-V2.0 is an advanced neutron transport lattice code developed by Nuclear Power Institute of China (NPIC). The subgroup resonance calculation method is used in KYIIN-V2.0 to obtain effective resonance selfshielding cross section, method of modular characteristics (MOC) is adopted to solve the neutron transport equation, and CRAM method and PPC method is adopted to solve the depletion equation. The numerical results show that KYLIN-V2.0 code has the reliable capability of direct heterogeneous calculation of 2D fuel assembly, and the effective multiplication factor, assembly power distribution, rod power distribution and control rod reactivity worths of various fuel assemblies that are calculated by KYLIN-V2.0 are in better agreement with the reference. [ABSTRACT FROM AUTHOR]

The UK fleet of Advanced Gas-Cooled Reactors (AGRs) are moderated with graphite and cooled with carbon dioxide. Each reactor consists of over 300 fuel channels with a cluster of UO2 fuel pins surrounded by graphite. Primary shutdown is achieved with control rods, which are at interstitial locations at six AGR stations, and in vacated fuel channels in the seventh. Nitrogen injection into the coolant is used for diverse shutdown or hold down. The reactivity worth of nitrogen was measured during cold commissioning at each AGR during the 1970s and 1980s. This measurement involved inducing a flux transient through a known change in nitrogen pressure, with conversion into a reactivity worth using kinetic data. Noting the time elapsed since this work (over 40 years) it is sensible to review this validation and reconsider these measurements using a consistent source of kinetic data. The lattice code WIMS and the reactor code PANTHER can be used to analyse the commissioning cores, in particular to determine the core-wide kinetic data. Furthermore the calculated worth of nitrogen can be compared with measurement. For the six interstitial AGR reactor designs nitrogen worth is under-predicted by about 10% and by about 5% for the inchannel design. This consistent under-prediction is investigated further with SERPENT. [ABSTRACT FROM AUTHOR]

In the update from ENDF/B-VII.1 to ENDF/B-VIII.0, copper cross sections were significantly altered in the intermediate and fast spectrum of the ENDF-VIII.0 library. Performance of this ENDF data requires validation to determine whether recent evaluation has proven beneficial. To examine the performance of the new library, particularly new copper data, critical benchmarks from the ICSBEP handbook were chosen for their sensitivity to copper cross section changes and modeled using SCALE continuous energy Monte Carlo simulations. Selected benchmarks were modeled in ENDF-VII.1 and ENDF-VIII.0 to compute keff within a statistical uncertainty of 10 pcm and compared in reference to the benchmark experimental criticality. Due to spectrum choices in selection based on the changes to cross section data, the set of benchmarks consist of intermediately enriched uranium, highly enriched uranium, or plutonium systems. 11 separate benchmark evaluations containing 32 individual configurations highly sensitive to copper were selected, modelled, and compared to benchmark experimental criticality. This work demonstrates a significant decrease in the deviation between calculated and experimental criticality as a result of the ENDF-VIII.0 library; a decrease in absolute mean deviation from 522.5±39.3 to 249.6±39.3, and a decrease in root mean square deviation from 630.8±46.1 to 338.1±74.9. Additionally, the role of recently evaluated copper data in this improved agreement is presented, confirming the benefit of reaffirming cross section data. [ABSTRACT FROM AUTHOR]

The cross section is a fundamental quantity which affects the accuracy of Monte Carlo simulations widely used in nuclear applications. A new dosimetry library IRDFF-II that contains cross section evaluations that include full uncertainty quantification is being developed by the International Atomic Energy Agency and expected to be released in January 2020; a preliminary version IRDFF-1.05 was released in 2014 and is being tested in this work. Validation of the cross-section evaluations proposed for this library is a high priority task. The validation can be realized using integral cross sections measured in standard and/or reference neutron benchmark fields. Integral quantities feature significantly lower uncertainties than differential nuclear data. If the neutron spectrum where the cross section is measured is well characterized, then the Spectrum Averaged Cross Section can be used for validating of existing evaluations. [ABSTRACT FROM AUTHOR]

Benchmarks are needed to validate methods to account for temperature-dependence of nuclear data. An evaluation of 37 KRITZ-1-Mk critical water height measurements, together with associated iso-reactivity temperature effects and coefficients, is released with the 2019 Handbook of the International Reactor Physics Experiment Evaluation Project (IRPhEP). The KRITZ zero-power research reactor, operated between 1969 and 1975 in Studsvik (Sweden), was contained in a pressure vessel, allowing full size fuel assemblies or fuel rods in light water at temperatures up to 250 °C without boiling. Preliminary results were published in 1971 and 1972 for four series of altogether 37 measurements with Marviken (Boiling Heavy Water Reactor) UO2 fuel rods, each containing a 235U isotopic mass fraction of 1.35 %. Temperature was the predictor variable, while critical water height was the response variable. Each series was characterized by the fuel rod lattice design and by the soluble boron concentration in water. The KRITZ measurements were focused on temperature-dependence (differences). High measurement correlations reduced the ?k uncertainties, typically from 195 pcm to 40 pcm for a large temperature change. Thermal expansion of fuel and reactor components was not measured. Detailed and simple benchmarks include estimated thermal expansion as a simplification. Benchmark calculation results using JEFF-3.3 nuclear data reduce the large biases observed for older libraries but a remarkable positive temperature trend is observed for series 4. In 2019, Studsvik Nuclear released information on KRITZ-1-Mk and on other KRITZ-1 and KRITZ-2 critical measurements with Boiling Water Reactor fuel assemblies and fuel rod clusters. [ABSTRACT FROM AUTHOR]

Silicon cross-sections, being important for a criticality safety of final spent fuel disposals, were recently reevaluated within the IAEA INDEN project. Similarly, the thermal Scattering Law matrix for silicon dioxide, which is also important for criticality safety, was also reevaluated in the ENDF/B-VIII.0 library. Due to these reasons, a series of validation experiments with silica sand were performed at the LR-0 reactor. This paper describes these validation experiments, which used two different amounts of silica sand placed in the core. The first part of validation was carried out as critical experiments in order to benefit from the suitability of integral experiments for validation. The second part -fast neutron spectrum measurement in the sand -was performed in order to obtain knowledge of its characteristics and its agreement with a calculation. The results showed significant improvement of the Thermal Scattering Law matrix for silicon dioxide available in the ENDF/B-VIII.0 library. They also showed that the new INDEN evaluation of silicon cross-sections, together with its description in the ENDF/B-VIII.0 gives disagreement rate closest to the experiments carried out without silica sand insertions. The spectrum measurement showed that the calculations of fast neutron spectra in the sand show only slight differences between different evaluations of silicon cross-sections. However, the fast neutron spectrum is not dependent on the Thermal Scattering Law. The calculated spectra show relatively good agreement with the measurement. [ABSTRACT FROM AUTHOR]

Molybdenum is being considered as a potential material for future nuclear fusion experiments and power plants. It has good thermo-mechanical properties and can be easily fabricated, making it attractive as an alternative first wall material to tungsten, which is the current leading candidate. Unfortunately, according to simulations, fusion-neutron irradiated Mo may become too activated during reactor operation to maintain the objective of fusion of avoiding the generation of long-term, higher-level radioactive waste. However, these simulated predictions rely heavily on having high-quality, accurate nuclear reaction data. For example, reliable simulations of the radiological response of isotopically-tailored Mo would be required by reactor designers and eventually nuclear regulators to assess if it is a viable low-activation fusion solution (in comparison to pure Mo). In recent years, UKAEA has developed benchmarks to test and validate the FISPACT-II inventory code and the input international nuclear data libraries against experimental measurements. This paper presents nuclear code prediction comparisons to new data acquired from γ-spectroscopy measurements of Mo irradiated in the ASP 14 MeV facility at AWE in the UK. Results demonstrate that FISPACT-II predictions for Mo are remarkably accurate; particularly for activity generated from the shorter-lived radionuclides explored by these laboratory experiments, such as 91Mo and 97Nb, etc., and their metastable isomeric states. [ABSTRACT FROM AUTHOR]

The Japan Atomic Energy Agency (JAEA) started the Research and Development (R&D) to improve nuclear prediction techniques for High Temperature Gas-cooled Reactors (HTGRs). The objectives are to introduce a generalized bias factor method to avoid full mock-up experiment for the first commercial HTGR and to introduce reactor noise analysis to High Temperature Engineering Test Reactor (HTTR) experiment to observe sub-criticality. To achieve the objectives, the reactor core of graphite-moderation system named B7/4"G2/8"p8EUNU+3/8"p38EU(1) was newly composed in the B-rack of Kyoto University Critical Assembly (KUCA). The core is composed of the fuel assembly, driver fuel assembly, graphite reflector, and polyethylene reflector. The fuel assembly is composed of enriched uranium plate, natural uranium plate and graphite plates to realize the average fuel enrichment of HTTR and it's spectrum. However, driver fuel assembly is necessary to achieve the criticality with the small-sized core. The core plays a role of the reference core of the bias factor method, and the reactor noise was measured to develop the noise analysis scheme. In this study, the overview of the criticality experiments is reported. The reactor configuration with graphite moderation system is rare case in the KUCA experiments, and this experiment is expected to contribute not only for an HTGR development but also for other types of a reactor in the graphite moderation system such as a molten salt reactor development. [ABSTRACT FROM AUTHOR]

Light water is the most important neutron moderator in many reactor applications. Thermal neutron scattering kernels for H bound in H2O are represented by ENDF-format thermal scattering laws (TSLs) evaluated at discrete temperatures T. Nuclear data evaluations are conventionally validated utilizing a combination of critical benchmarks and experimental cross section data. Existing public critical benchmarks may be inadequate for testing water TSLs over the full range of T of interest in reactor applications, and experimental thermal scattering cross section data for water is sparse at elevated T. In this work, MC21 is used to simulate the decay of the equilibrium thermal neutron flux in light-water spheres of several radii. The fundamental-mode time eigenvalue α is calculated at 22 °C and 227 °C (at saturation pressure) for each sphere using three different H-H2O TSLs. Fitting α to a polynomial function of geometric buckling allows calculation of the thermal neutron diffusion length L. Extrapolation lengths are treated as a function of the transport mean free path and geometry. Experimental L data from 25 publications at 49 temperatures (from 10 °C to 295 °C), computed by several different time-dependent and space-dependent decay methods, is used to develop an empirical fit of L vs. T. The MC21-calculated L for the TSLs tested are within ±1% of the predicted values at 22 °C and 227 °C. This validation approach, which may be repeated at arbitrary T, constitutes an integral benchmark specific to and characterizing the detailed physics of the thermal scattering kernel applied. [ABSTRACT FROM AUTHOR]

Yttrium hydride (YHx) is of interest as a high-temperature moderator material in advanced nuclear reactor systems because of its superior ability to retain hydrogen at elevated temperatures. Thermal neutron scattering laws (TSL) for hydrogen bound in yttrium hydride (H-YH2) and yttrium bound in yttrium hydride (Y-YH2) were previously evaluated by Naval Nuclear Laboratory using the ab initio approach and released in ENDF/B-VIII.0. In that work, density functional theory, incorporating the generalized gradient approximation (GGA) for the exchange-correlation energy, was used to simulate the face-centered cubic structure of YH2 and calculate the interatomic Hellmann-Feynman forces for a 2×2×2 supercell containing 96 atoms. Lattice dynamics calculations using PHONON were used to determine the phonon density of states. The calculated phonon density of states for H and Y in YH2 were then used to prepare H-YH2 and Y-YH2 TSL evaluations, in the incoherent approximation, using the LEAPR module of NJOY. In addition elastic scattering was assumed to be incoherent for both H and Y. While the incoherent elastic scattering approximation is appropriate for H-YH2, it introduces an undesirable approximation for Y-YH2. In this work, we re-evaluate the TSL for Y-YH2 using FLASSH (Full Law Analysis Scattering System Hub). Y-YH2 is evaluated using the FLASSH generalized coherent elastic scattering capability in order to capture the Bragg peaks associated with the YH2 crystal structure which were neglected in the prior NJOY-based evaluation due to limitations in LEAPR. An experimental approach to validate the Y-YH2 TSL using neutron transmission measurements is discussed. [ABSTRACT FROM AUTHOR]

In order to perform integral evaluation of 232Th capture cross section, a series of critical experiments for thorium-loaded and solid-moderated cores in KUCA had been carried out. In these experimental cores, H/235U nuclide ratio ranged about from 150 to 315, and 232Th/235U nuclide ratio ranged about from 13 to 19. In this study, a new critical experiment with Th loaded core in KUCA, which had about 70 of the H/235U ratio and 12.7 of 232Th/235U ratio, was carried out. As results, the excess reactivity was 0.086 ± 0.003 (% dk/k) and the keff was 1.0009 ± 0.0003, where the effective delayed neutron fraction was 7.656E-3. The keff was also calculated by MVP3.0 with different nuclear libraries. The respective calculations with JENDL-4.0, JENDL-3.3 and ENDF/B-VII.0 lead to 1.0056 ± 0.0086 (%), 1.0048 ± 0.0085 (%) and 1.0056 ± 0.0086 (%).On the other hand, the further MVP3.0 calculations, where only the 232Th cross sections were taken from JENDL-4.0, JENDL-3.3 or ENDF/B-VII.0 but all other nuclides were done from JENDL-4.0, were carried out to examine an impact of the difference of 232Th cross section among these nuclear libraries to the keff. The keff calculated with respective 232Th cross sections from JENDL-3.3 and ENDF/B-VII.0 was 1.0038 ± 0.0086 (%) and 1.0040 ± 0.0086 (%). [ABSTRACT FROM AUTHOR]

The PETALE experimental program in the CROCUS reactor intends to provide integral measurements on reactivity worth and dosimetry measurement to constrain nuclear data relative to stainless steel heavy reflectors. The experimental setup consists in eight successive plates of pure iron, pure nickel, pure chromium, or nuclear-grade stainless steel set at the close periphery of the core. The plates are interleaved with up to nine dosimeters that consist of thin activation foils with different possible materials to be sensitive to different ranges of the neutron spectrum. A precise measurement with a good estimation of the uncertainties and correlations is required, especially when comparing reaction rates, e.g. transmission measurement and/or spectral indices. The present work focuses on the validation of the dosimetry technics developed in preparation of this experimental program. Different aspects are discussed: monitors, efficiency calibration, self-absorption correction, self-shielding and nuclear data uncertainties. The different sources of uncertainties for the experiment-calculation comparisons are characterized, taking into account all the correlation between the different dosimeters. These correlations are a mandatory element for the aimed Bayesian assimilation in order to avoir overfitting when considering dosimeter providing a similar information. [ABSTRACT FROM AUTHOR]

Reactor noise experiments at high frequencies (up to 100 kHz) in subcritical configurations have been performed at the IPEN/MB-01 research reactor facility. The core configuration considered a short version of the IPEN/MB-01 core in a 24 × 26 rectangular array of fuel rods. The subcritical configurations considered the control banks totally withdrawn and the moderator poisoned with H3BO3. The boron concentrations were: 286.8 ± 10 and 578.6 ± 10 ppm. The pulses of two 3He detectors in the reflector region were summed and inserted into the IPEN/MB-01 Correlator and the APSD (Auto Power Spectral Density) was inferred through a mathematical model. The analysis reveals that the APSD in this frequency range is best described by a four-mode decay model. According to the two-region two-group (thermal and fast) kinetic model developed in this work, the first two decay modes describe the thermal group and the other two describe the fast group. After a tedious and severe analyses of the least-square fit of the experimental data, it was concluded that the kinetic behavior of the thermal and fast neutrons can be considered uncoupled. The analysis of the experimental data is still in progress and only the thermal group of the case of 286.8 ppm of boron was analyzed and some parameters could be inferred. The most important one so far is the prompt neutron generation time in the core region that could be inferred with a good level of accuracy. [ABSTRACT FROM AUTHOR]

Fuel burnup of the JSI TRIGA was calculated by simulating complete operational history consisting of 240 different core configurations from 1966 to 2020. At the moment we are unable to perform burnup measurements, e.g. gamma spectroscopy on burned fuel elements, hence we used weekly measured excess reactivity as a reference point of different core configurations to verify the calculated core reactivity. Changes in reactivity due to burnup were assumed to be linear and this assumption was verified for burnup intervals smaller than 3 MWd/kg(HM). The comparison was performed on 46 different core configurations with different type of fuel elements. The Serpent-2 calculations decently predict the rate of reactivity change on different cases, as 52 % of calculations are withing 1σ and 86.9 % within 2σ of the measurements for total number of 46 cases. Additional analysis was performed by comparing unit cell calculations of different fuel types. Four different types of TRIGA fuel were used to analyse burnup changes in LEU and HEU fuel, where positive reactivity feedback on burnup was observed for HEU fuel due to burnable absorbers. Serpent-2 and WIMSD-5B were compared on unit-cell basis where good agreement within 200 pcm of reactivity change for large burnup was observed. In addition neutron spectrum changes due to burnup were investigated using unit-cell calculations where 4 % increase of the thermal peak and 1 % decrease of fast peak of the spectrum was observed for typical fuel burnups of 20 MWd/kg(HM), which approximately represents JSI TRIGA burnup at this moment. [ABSTRACT FROM AUTHOR]

An evaluation of validity of theoretical pulse models (Fuchs-Hansen and Nordheim-Fuchs model) in reactor pulse modelling was performed by developing so called Improved Pulse Model. The effect of each of the five assumptions on the most important pulse physical parameters, maximal power, total released energy and full width at half maximum was studied. In the Improved Pulse Model the assumptions are disposed out with the improvements, where to account the delayed neutrons the six point kinetic equations are solved, the temperature dependences of the temperature reactivity coefficient of fuel and specific heat are taken into account, also the final ejected time of transient control rod from reactor core, whose value of reactivity varies in height and the heat dissipation from the fuel are considered. It is found that the theoretical models predict a higher maximum power, lower total released energy and full width at half maximum than the Improved Pulse Model. [ABSTRACT FROM AUTHOR]

The Modified Source Multiplication method is used to determine an unknown reactivity level of a reactor from a known one if one has access to the detector counting for both levels when the reactor is fed by a constant neutron source like an Am-Be source. When available, an accelerator driven source, in continuous mode, can be useful as its intensity can be tunable and then adapted to the experimental conditions. However, in that case, the MSM technique must be extended to account for an external source whose intensity, energy and angular distributions can vary from one measurement to another. In this paper, this Modified Multi-Source Multiplication (MMSM) method is applied to measurements done during the FREYA project in the GUINEVERE facility, operated with the GENEPI-3C accelerator providing a mixture of (D,T) and (D,D) neutrons. The monitoring of these sources through the detection of the associated charged particles allows the calculation of the MMSM factors and the estimate of the reactivity values. The results are compared in different configurations with the reactivity obtained with an Am-Be source or in dynamic measurements performed with GENEPI-3C. Their excellent agreement shows the possibility of using such accelerator-based neutron sources for MSM measurements when they are correctly monitored. This is of great interest for deep sub-critical level characterization for which detector count rates per source neutrons might be low. [ABSTRACT FROM AUTHOR]

Reactor core design is inherently a multi-objective problem which spans a large design space, and potentially larger objective space. This process relies on high-fidelity models to probe the design space, and sophisticated computer codes to calculate the important physics occurring in the reactor. In the past, the design space has been reduced by individuals with extensive knowledge of reactor core design; however, this approach is not always available. In this paper, we utilize a set of high-fidelity models to generate a reduced-order model, and couple this with a genetic algorithm to quickly and effectively optimize a preliminary design for a prototypical sodium fast reactor. We also examine augmenting the genetic algorithm with physical programming to generate the fitness function(s) that evaluates the degree to which a core has been optimized. Physical programming is used in two variations of multi-objective optimization and is compared with a traditional weighting scheme to examine the solutions present on the Pareto front. Optimization on the reduced-order model produces a set of solutions on the Pareto front for a designer to examine. The uncertainty for the objective functions examined in the reduced-order model is less than 7% for the given designs, and improves as additional data points are employed. Utilizing a reduced-order model can significantly reduce the computation time and storage to perform preliminary optimization. Physical programming was shown to reduce the objective space when compared with a traditional weighting scheme. It also provides an intuitive and computationally efficient way to produce a Pareto front that meets the designer's objectives. [ABSTRACT FROM AUTHOR]

Several methods have recently been developed to solve multiphysics transients using the Improved Quasi-Static method and its derivatives. In order to address some perceived drawbacks of these methods, we have developed a new method for solving multiphysics transient calculations using the Predictor-Corrector Quasi-Static method. Our method involves computing reactivity feedback parameters during each transport timestep in order to enable reactivity feedback on the small timescale used by the point kinetics phase of the calculation. The advantage of this approach is that the transport solver does not need to store local flux information between time steps, potentially making it more appropriate for use with a Monte Carlo solver. We have demonstrated the accuracy of our method by solving a simple model problem that exhibits difficult multiphysics behavior. Additionally, we have compared our results against another recently published multiphysics coupling scheme, confirming that our approach does not negatively affect the accuracy of the transient solution. [ABSTRACT FROM AUTHOR]

The distribution of the power inside a reactor core can be described by the time dependent multigroup neutron diffusion equation. One of the approaches to integrate this time-dependent equation is the modal method, that assumes that the solution can be described by the sum of amplitude function multiplied by shape functions of modes. These shape functions can be computed by solving a _-modes problems. The modal method has a great interest when the distribution of the power cannot be well approximated by only one shape function, mainly, when local perturbations are applied during the transient. Usually, the shape functions of the modal methods are updated for the time-dependent equations with a constant time-step size to obtain accurate results. In this work, we propose a modal methodology with an adaptive control time-step to update the eigenfunctions associated with the modes. This algorithm improves efficiency because of time is not spent solving the systems to a level of accuracy beyond relevance and reduces the step size if they detect a numerical instability. Step size controllers require an error estimation. Different error estimations are considered and analyzed in a benchmark problem with a out of phase local perturbation. [ABSTRACT FROM AUTHOR]

Decay heat (DH) is the heat produced through a radioactive decay of fission products during or after a reactor operation. It is known as the second largest source of power in the core after fission. Being such a strong contributor to reactor power, it should be accurately determined at any time of reactor operation. Currently, there are two main approaches for DH estimation used in reactor simulation codes. One approach is based on careful inventorying of all produced target nuclides and their individual contributions to total power. Alternatively, the other popular approach is based on collapsing all target fission products into a small number of groups similar to delayed neutron estimation techniques. However, the last (multigroup) method currently has limitations when used in some transient scenarios such as transients occurred in fresh fuel. In this study, the multigroup method was further developed for reducing limitations while retaining the advantage in computation speed. Then, it was implemented into Reactor Analysis code for Steady state and Transient (RAST-K) and tested against other codes. As a result, the improved method was found capable of determining DH power at all tested stages of reactor operation under any tested operation scenario. In particular, the test simulations using the improved method showed better results in those scenarios that were under accuracy limitations of the original multigroup method. Overall, the quality of transient calculations in RAST-K was improved when using the newly implemented DH module. [ABSTRACT FROM AUTHOR]

In recent years, there has been renewed interest in Molten Salter Reactors (MSRs) for their potential advantages compared to reactors that rely on solid fuel. In response to such interest, the System Analysis Module (SAM) was enhanced to include MSR-specific modeling features including a delayed neutron precursor drift model and a modified point kinetics model. This paper discusses the validation of these features using the experiments conducted in the Molten Salt Reactor Experiment (MSRE). These experiments include the pump start-up and coast-down tests at zero power and a thermal convection test. For the zero power tests, the change in pump speeds induces flow rate changes in the core that impact the precursor concentrations. This introduces a neutron imbalance and requires the adjustment of the control rods to counter-balance this effect. SAM was used to evaluate the precursor concentration in the core as a function of time, and the resulting changes in reactivity were evaluated through the modified point kinetics equation. The results show good agreement with the experimental data. It should be noted that the pump performance curve used in this analysis was re-constructed based on the initial water test data of the fuel pump. The steady-state pump curve is assumed to be applicable to transient flow operations. The thermal convection test was conducted by shutting off the pumps, reducing the inlet core temperature for 360 minutes, and allowing the power to be adjusted by the inherent feedbacks of the system. The power level during this transient was evaluated by SAM as a function of time. [ABSTRACT FROM AUTHOR]

This paper describes the development of a 3-d high-fidelity multi-physics pin-by-pin neu-tronic and sub-channel thermal hydraulic model of a SVEA-96 Optima2 fuel assembly for the coupled code system TORT-TD/CTF. No pin power reconstruction is applied. Detailed descriptions are given on (a) the generation of few-group parameterized pin cell-homogenized cross section data using HELIOS, (b) the 3-d neutronic pin-by-pin and ther-mal-hydraulic sub-channel model of a single SVEA-96 Optima2 assembly and its qualifica-tion by Monte Carlo models using Serpent, (c) coupled steady-state results for a single as-sembly and (d) coupled steady-state and transient simulations of a 4-by-4 mini core model of SVEA-96 Optima2 assemblies in a partially rodded state. [ABSTRACT FROM AUTHOR]

The OECD Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC) established the Expert Group on Multi-physics Experimental Data, Benchmarks and Validation (EGMPEBV) in 2014 to bridge the gap between advanced, multi-physics simulation capabilities and the relatively low availability of dedicated, high-fidelity experimental data and benchmarks specifically for multi-physics modelling and simulation tools. The EGMPEBV was mandated to establish mechanisms for the certification of experimental data and benchmark models and to establish the processes and procedures for the validation of multi-physics modelling and simulation tools. The EGMPEBV oversees three task forces, covering (1) experimental data qualification and benchmark evaluation, (2) validation guidelines and needs and (3) example application of validation experiments. These have generated numerous reports surveying the state-of-the-art in multi-physics validation, challenge areas and recommendations for the evaluation of multi-physics benchmarks, while in parallel developing the specifications for multi-physics benchmarks. Three benchmark specifications are in active development, including a reactivity compensation scenario in the Rostov Unit 2 VVER-1000, multi-cycle depletion of the TVA Watts Bar Unit 1 and study of pellet cladding mechanical interaction within ramp tests performed at the Studsvik R2 reactor. We provide an overview of the recent progress in these areas and a summary of the future activities of the EGMPEBV in establishing international multi-physics benchmarks. [ABSTRACT FROM AUTHOR]

Most safety criteria for Sodium cooled Fast Reactors (SFR) are local core parameters. Thus, application of 3-D neutron kinetic and thermal-hydraulic coupled codes including detailed modelling of core expansion effects is mandatory for best estimate evaluations of safety margins. A recently published benchmark based on measurements performed at the Superph´enix (SPX) reactor offers the opportunity to validate codes and methods for SFR safety assessment. In this paper, the SPX core is modelled in ATHLET and PARCS. Explicit models for axial and radial core expansion effects for 3-D coupled calculations were recently implemented in PARCS. In ATHLET, axial thermal expansion of structures (strongback, vessel, control rod drive line) influencing the relative position of the control rod in the active core is modelled. The newly implemented models are tested on a transient initiated by a reactivity insertion of -50 pcm. A point kinetic simulation is also performed to compare with the 3D solution. Both ATHLET-point kinetic model and ATHLET-PARCS simulations deliver similar power responses during the transient but with an offset. By analysing the different feedbacks in the point kinetic model, it can be concluded that models of expansion of the different structures are well implemented. In the future further analysis of different transients of the benchmark are planned. [ABSTRACT FROM AUTHOR]

Kraken is a new computational reactor analysis framework under development at VTT Technical Research Center of Finland Ltd. The framework builds heavily on the new generation of Finnish simulation codes such as the Serpent Monte Carlo code, the Ants nodal neutronics solver and the FINIX fuel behavior module. This paper describes the application of Kraken to its first realistic full core problem, a small modular reactor (SMR) core in its fresh state in order to evaluate control rod worths, shutdown margins and reactivity coefficients for the core. [ABSTRACT FROM AUTHOR]

This work presents a methodology to coarsen subchannel models to accelerate largescale Serpent2-SUBCHANFLOW calculations. The method is based on first building a fully detailed subchannel model and then combining subchannels into larger channels with condensed hydraulic parameters. To quantify the accuracy and performance of this scheme, a full-core VVER-1000 problem is simulated with Serpent2-SUBCHANFLOW using two types of coarsened models, as well as a standard subchannel model as reference. [ABSTRACT FROM AUTHOR]

This work presents the results for a coupled neutronic-thermalhydraulic-thermomechanic pin-level depletion calculation of a PWR fuel assembly using Serpent2-SUBCHANFLOW-TRANSURANUS. This tool is based on a semi-implicit depletion scheme with pin-by-pin feedback, mesh-based field exchange and an object-oriented software design. The impact of including fuel-performance capabilities is analyzed, with focus on high-burnup effects. The treatment of the Doppler feedback to the neutronics is examined as well, in particular the use of radial fuel-temperature profiles or radially averaged values. [ABSTRACT FROM AUTHOR]

In this paper, the modeling and simulation of the PWRs loaded with hexagonal fuel assemblies has been implemented with the NECP-Bamboo code. NECP-Bamboo, consisting of a 2D lattice code named Bamboo-Lattice and a 3D steady-state core code named Bamboo-Core, was primitively designed for the PWRs loaded with the rectangular fuel assemblies. As the capability extension for PWRs with hexagonal fuel assemblies, four aspects of improvement have been implemented in NECP-Bamboo. Firstly, the Constructive Solid Geometry (CSG) has been implemented in Bamboo-Lattice for the lattice modeling. Secondly, the explicit modeling of the reflector assembly has been applied to provide more reliable few-group constants, compared with the conventional 1D model for the reflector assembly. Thirdly, the assembly-homogenization capability has been extended to the hexagonal assembly. Fourthly, the diffusion solver in Bamboo-Core based on the Variational Nodal Method (VNM) has been extended to handle hexagonal geometry. With application of the capability-extended NECP-Bamboo, the modeling and simulations for the VVER-1000 benchmark loaded with MOX fuel has been implemented. It can be observed that the numerical results provided by NECP-Bamboo can agree well with corresponding results by the Monte-Carlo code. [ABSTRACT FROM AUTHOR]

A new OECD/NEA benchmark entitled "Reactivity compensation with diluted boron by stepwise insertion of control rod cluster" is starting. This benchmark, based on high quality measurements performed at the NPP Rostov Unit 2, aims to validate and assess high fidelity multi-physics simulation code capabilities. The Benchmark is divided in two phases: assembly wise and pin-by-pin resolution of steady-state and transient multi-physics problems. Multi-physics simulation requires the generation of parametrized few-group cross-sections. This task used to be done with deterministic (2-D) lattice codes, but in the past few years the Monte-Carlo code SERPENT has demonstrate its ability to generate accurate few-group homogenized cross-section without approximations, neither on the geometry nor in the nuclear data. Since the whole core SERPENT models for production of such cross-section libraries would be computationally costly (and the standard 2-D approach may introduce unnecessary large approximations), 3-D models of each assembly type in infinite radial lattice configurations have been created. These cross-sections are then used to evaluate effective multiplication factors for different core configurations with the diffusion code PARCS. The results are compared with the reference SERPENT calculations. In the next step, a thermal-hydraulic model with the system code ATHLET applying an assembly-wise description of the core (i.e. one channel per fuel assembly) has been developed for coupled PARCS/ATHLET transient test calculations. This paper describes in detail the models and techniques used for the generation of the few-group parameterized cross section libraries, the PARCS model and the ATHLET model. Additionally, a simple exercise with coupled code system PARCS/ATHLET is presented and analysed. [ABSTRACT FROM AUTHOR]

Fully ceramic micro-encapsulated (FCM) fuel is a kind of fuel that employs tri-structural isotropic (TRISO) particles to enhance safety. The FCM fuel assembly is a double heterogeneous system. The conventional self-shielding calculation methods cannot treat the DH effect. In this paper, three methods based on equivalent homogenization of the TRISO particle and the matrix are studied and compared: the hyper-fine energy group cross sections (XSs) homogenization based hyper-fine energy group method (HHM), the hyper-fine energy group XSs homogenization based subgroup method (HSM) and the subgroup XSs homogenization based subgroup method (SSM). These methods are implemented in a high-fidelity neutronics code NECP-X. Numerical results show that these methods are able to treat the double heterogeneity of the FCM fuel. The precision of the HHM and HSM is higher than that of the SSM. [ABSTRACT FROM AUTHOR]

The topic addressed deals with the determination of adjoint parameters for instrumentation relevance. This is a crucial subject for comprehension of subcritical levels in the frame of safety analysis. Indeed, such states require interpretation and raw data cannot be processed as such. To do so, the transcription of core reactivity through instrumentation located in the reactor periphery is considered with the use of MSM factors [1],[2]. We implement this method inside a TRIPOLI4® [3] sequence in order to establish predictive mapping of MSM factors and figure out optimal position for instrumentation location at the beginning of reactor operations. Firstly, MSM factors are introduced, along with the designer point of view for geometry construction based on ROOT package [4]. At this point, the methodology of TRIPOLI4® calculation is explained in detail, including the sequencing associated to and how the Green Functions are performed within TRIPOLI4®. In this second part and within the verification framework, the previous method is extended to a "fictitious core" developed in TechnicAtome for Monte Carlo [5] calculation and for different core pattern loadings. After the completion of these numerical validations gained on a High Performing Cluster, the method is then expanded to critical mock up [6] and challenged to recent experimental results for validation. The comparisons end up with a good agreement between predictive calculation and experimental values of reactivity worth. Finally the document ends with a mid-term projection for outlooks and improvements, for ensuring an enhancement of the safety approach. Several items are discussed especially, fine tuning for the spatial meshing (regarding instrumentation size) and the impact on TRIPOLI4® Monte Carlo code with the development of new features. Then, the authors focus on sensitivity effect concerning delayed neutron spectrum and kinetics parameters. As a conclusion, this paper proposes to validate the method exposed in the near future, using experimental data coming from many years of critical mock up operations. [ABSTRACT FROM AUTHOR]

The estimation of the neutron fluence at the Reactor Pressure Vessel (RPV) is classically carried out by a two-step approach. The first step is to estimate the full core neutron source term whether the second step of the calculation consists in the transport of neutrons from the core (source term) to the RPV using the neutron fission distribution determined in the previous step. For this purpose, the neutron fission distribution is to be accurately determined at the fuel pin level for the assemblies on the border of the core. To achieve this goal, two methods are evaluated in this study. The first method considered is a full core 2D Monte Carlo calculation using the MNCP6 code. The second method is based on a deterministic approach using the CASMO5 multi-segment option, allowing a full 2D transport calculation at the pin level with an expected accuracy similar to a stochastic method. The comparison of the two methods shows an overall good agreement with differences within the statistical uncertainty for different cores: homogeneous UOX core, mixed UOX-MOX loading and the effect of the hafnium rods used in the assemblies in the periphery of the core. The modelling limitation and the associated calculational time are discussed for the comparison of the two approaches. [ABSTRACT FROM AUTHOR]

The neutron flux fluctuation magnitude of KWU-built PWRs shows a hitherto unexplained correlation with the types of loaded fuel assemblies. Also, certain measured long-range neutron flux fluctuation patterns in neighboring core quadrants still lack a closed understanding of their origin. The explanation of these phenomena has recently revived a new interest in neutron noise research. The contribution at hand investigates the idea that a synchronized coolant-driven vibration of major parts of the fuel-assembly ensemble leads to these phenomena. Starting with an assumed mode of such collective vibration, the resulting effects on the time-dependent neutron-flux distribution are analyzed via a DYN3D simulation. A three-dimensional representation of the time-dependent bow of all fuel assemblies is taken into account as a nodal DYN3D feedback parameter by time-dependent variations of the fuel-assembly pitch. The impact of its variation on the cross sections is quantified using a cross-section library that is generated from the output of corresponding CASMO5 calculations. The DYN3D simulation qualitatively reproduces the measured neutron-flux fluctuation patterns. The magnitude of the fluctuations and its radial dependence are comparable to the measured details. The results imply that collective fuel-assembly vibrations are a promising candidate for being the key to understand long-known fluctuation patterns in KWU built PWRs. Further research should elaborate on possible excitation mechanisms of the assumed vibration modes. [ABSTRACT FROM AUTHOR]

The BEAVRS benchmark was analyzed using the CASMO5/SIMULATE5 in order to compare the measurement data and the calculation results based on the JENDL-4.0 and ENDF/B-VII.1 and investigate the difference between those calculation results. For the hot zero power (HZP) physics test, the calculation results showed good agreement with the measurement data for both of cycles 1 and 2. For cycle 1, the calculation results of the isothermal temperature coefficient (ITC) differed from the measurement data by approximately 1 pcm/℉, and the same tendency has been reported in previous studies. For the cycle operation, the calculation results of the boron letdown agreed well with the measurement data. On the other hand, some calculation results of the axial detector signals had a large difference from the measurement data, which is supposedly attributed to the discrepancy of the axial offset (AO) caused by the authors' approximation for the control rod and shutdown bank positions. In terms of the comparison between the JENDL-4.0 and ENDF/B-VII.1, although approximately 15 ppm difference of the boron letdown in the cycle operation was observed, no significant difference was seen for other core parameters, thus, the influence of the two nuclear data library was small on the present results. [ABSTRACT FROM AUTHOR]

The problem of determining the axial velocity profile from the in-core neutron noise measurements is revisited, with the purpose of developing an objective method for the determination of the void fraction. Until now it was assumed that in order to determine a realistic velocity profile which shows an inflection point and hence has to be at least a third order polynomial, one needs four transit times and hence five in-core detectors at various axial elevations. However, attempts to determine a fourth transit time by adding a TIP detector to the existing four LPRMs and cross-correlate it with any of the LPRMs were unsuccessful so far. In this paper we thus propose another approach, where the TIP detector is only used for the determination of the axial position of the onset of boiling. By this approach it is sufficient to use only three transit times. Moreover, with another parametrisation of the velocity profile, it is possible to reconstruct the velocity profile even without knowing the onset point of boiling, in which case the TIP is not needed. In the paper the principles are explained and the strategy is demonstrated by concrete examples. [ABSTRACT FROM AUTHOR]

This paper presents the qualification of the newly developed Westinghouse lattice physics code PARAGON2. PARAGON2 uses high energy resolution in the solution of the transport equation. The objective of this paper is to demonstrate that PARAGON2 accurately predicts the integral and critical experiments. The integral experiments are used to assess PARAGON2 predictions of fine neutronics parameters such as: resonance integrals and radial profiles of reactions rates, isotopics, and burnup for depleted pellets. The integral experiments considered are: the Hellstrand's, TRX, and the PIE experiments. For critical experiments, this paper will focus only on VENUS-2 MOX critical experiment. The results obtained for the integral experiments clearly show the good predictions of PARAGON2 with the resonance scattering model which are close to measurement. The PARAGON2 predicted capture reaction rates, temperature coefficients, burnup and isotopic profiles match the measured values both in shape and magnitude. VENUS-2 reactivity prediction is in excellent agreement with the critical measurement value. Also, the standard deviations of measured minus predicted pin-wise fission reaction rates are very good (i.e. ≤ 2%) for both individual assemblies and the whole core. [ABSTRACT FROM AUTHOR]

Recent researches have become more interested in the feasibility of using Monte Carlo (MC) code to generate multi-group (MG) cross sections (XSs) for fast reactor analysis using nodal diffusion codes. The current study, therefore, presents a brief methodology for MG XSs generation by the in-house UNIST MC code MCS, which can be compatibly utilized in nodal diffusion codes, PARCS and RAST-K. The applicability of the methodology is quantified on the sodium fast reactor (SFR) ABR-1000 design with a metallic fuel from the OECD/NEA SRF benchmark. The few-group XSs generated by MCS with a two-dimensional (2D) fuel assembly geometry are well consistent with those of SERPENT 2. Furthermore, the simulation of beginning-of-cycle (BOC) steady-state three-dimensional (3D) whole-core problem with PARCS and RAST-K is conducted using the generated 24-group XSs by MCS. The nodal diffusion solutions, including the core keff, power profiles and various of reactivity parameters, are compared to reference whole-core results obtained by MC code MCS. Overall, the code-to-code comparison indicates a reasonable agreement between deterministic and stochastic codes, with the difference in keff less than 100 pcm and the root-mean-square (RMS) error in assembly power less than 1.15%. Therefore, it is successfully demonstrated that the employment of the MG XSs generation by MCS for nodal diffusion codes is feasible to accurately perform analyses for fast reactors. [ABSTRACT FROM AUTHOR]

The ESFR-SMART project is the latest iteration of research into the behaviour of a commercial-size SFR core throughout its lifetime. As part of this project the ESFR core has been modelled by a range of different reactor physics simulation codes at its end of cycle state, and the important safety relevant parameters evaluated. These parameters are found to agree well between the different codes, giving good confidence in the results. A detailed mapping of the local sodium void worth is also performed due to the problems associated with the positive void coefficient seen in large SFR designs. The local void worth maps show that the use of zone-wise coefficients replicates the important reactivity feedbacks to a high degree, indicating their suitability for use in SFR simulations. [ABSTRACT FROM AUTHOR]