[en] We present a methodology to estimate thermal conductivity, heat capacity and linear expansion of MA-MCX fuel based on existing correlations and new experimental results of MCX and minor actinide oxides, We validate this method against experimental measurements with different concentrations of MA.

[en] Publicly available in the open literature are different friction factor correlations for wirewrapped fuel bundles based on a particular set of experimental data. These correlations usually are very good for the prediction of friction factors for wire-wrapped fuel bundle within the parameter range for which they were derived based on certain fluid and certain fuel bundle parameters. But when extrapolated to another fluid (coolant) or different fuel bundle parameters, these friction factor correlations provide us with predictions of the friction factor that are not always co-relatable to the experimental data. So an important question arises, which friction factor correlation should one use in order to obtain reliable prediction of the friction factor for any coolant and any set of fuel bundle parameters. This paper tries to address this very important issue, based on the qualitative evaluation of the most commonly used friction factor correlations provided to us by different authors, while analyzing more than ten different sets of experimental data that are available electronically today. These experiments were conducted using different coolants (water, sodium, air), different sets of fuel bundle parameters, by different scientists in different countries and organizations. (orig.)

[en] Benchmarking validation of SIM-MS transient code to Zero Power Experiments performed at ORNL 10 MW MSRE during 1965. Zero-Power means: no temperature feedbacks, only neutron kinetics taken into account. It allows testing of the neutronic characteristics (delayed neutron precursor physics) of circulating fuel if correctly simulated in the various computer codes models. Modelling of Circulating Fuel in SIM-MS covers neutron kinetics. Application of transient code system SIM-MS to compare experimental measurement results with calculational SIM-MS results. Experiments performed at MSRE were as follows, Criticality experiments: critical U-235 concentrations and control rod reactivity calibrations; determination of reactivity effects of circulating fuel: neutron precursors in core and external loop flow reactivity beta-lost; Flow Transients: Pump Start-up and Pump Coast-down. The experimental data on MSRE as documented in ORNL-4233 is ideal for validating the modelling (i.e. computer models) of the neutron characteristics of circulating fuels (neutron kinetics only; no temperature feedbacks). The neutron kinetic part of the transient code system SIM-MSRE has been validated for circulating fuel type reactors by comparing results to experimental MSRE data (no temperature feedbacks). No experimental data in the power range of MSRE (i.e. flow transients) could be found so far to validate the codes in the power range (with temperature feedbacks)

[en] Three-dimensional space-tem dynamic computer code for ADS is a coupled code system consisting of SIMS-ADS and Citation. It applies a coarse group library and involves special features of ADS dynamics It has been tested on FZK-3 source benchmark problem. It has also been benchmarked on other transient code systems for critical and subcritical configurations by comparing reactivity insertion transients; LOF and ULOF transients; LOH and ULOH transients. It was tested on LWR; sodium cooled, lead cooled and CO₂ cooled FBRs, as well as molten salt reactors

[en] The basic kinetic behavior of an accelerator driven system (ADS) is demonstrated using a point kinetic reactor model. The model was expanded to include also effects of fuel and moderator temperature changes on the dynamic response of a sub-critical system with and without an external source. This dynamic module allows also a description of protected and unprotected loss of flow (LOF) transients. By the unprotected case, non of the safety rods are inserted and in particular for sub-critical devices the source(s) remain(s) active. The entire spectrum of spatial dynamic effects including mechanical and material damage are to be simulated with the coupled codes SAS4A and CITATION. Preliminary results indicate that a relatively low power density will be required, so that the mechanical integrity of the core configuration can be assured under severe transient conditions. (orig.)

[en] Transients analysed for MSBR and AMSTER Molten Salt Reactor Concept are: 1. ULOF: loss of flow (i.e. loss of a primary pumping); 2. ULOHS : loss of heat sinks (i.e. HXs); 3. UOC : overcooling of primary fuel because of secondary side malfunction 4. UTOP : sudden reactivity insertion (actinide fuel agglomerated in piping system. Dislodged fuel particle then sweeping through core). Observations are made as regards the AMSTERIncinerator Molten Salt concepts as follows: 1. Above 800 deg. C the graphite reactivity coefficient of the AMSTERIncinerator becomes negative without Er-167. 2. Adding Er-167, the graphite coefficient becomes negative > 600 deg. C. 3. Negative graphite coefficient will stabilize system long-term. Thus recommend that Er-167 is also added in the AMSTER-Incinerator. 4. The large thermal inertia associated with graphite makes the transient behaviour very 'sluggish' of this reactor concept. 5. This 'sluggish' transient behaviour allows sufficient grace time for effective operator intervention

[en] This paper gives a summary of a risk study assuming unprotected loss of flow (ULOF) in the SNR 300. This study was initiated in 1979/80 by the Karlsruhe Nuclear Research Center and performed in close cooperation with Science Applications Inc., Palo Alto, USA, and Interatom Company. Part of the results also was integrated in the 'Risk Related Analysis for the SNR 300' carried out by the Gesellschaft fuer Reactorsicherheit. The character of the study described here is similar to other risk studies like the Reactor Safety Study and the German Risk Study for Nuclear Power Plants. The objectives and the methodology of the analyses are described and its results are discussed. (orig./RW)

[en] Spectrum of transients analysed for MOSART Design was:1. ULOF : loss of pump; 2. ULOHS: HX failure; 3. UOC (overcooling): - 100 deg. C in core inlet temperature in 1 min; 4. UTOP: +500 pcmUsed idealized conditions for the following transient analysis : Some Limitations to be aware of : 1. Core inlet temperatures : Assume constant core inlet temperatures unless specified differently in the table of transients. (i.e. UOC and ULOH transient where core inlet temperatures do change); 2.assume 'idealized' fuel flow pattern inside core region ( no flow eddies which might increase the fuel residence time and may cause jugging of the core power). 3. Assumed Temperature reactivity Coefficients : For reflector graphite: - 0.04 pcm/K Fuel (Doppler + dilitation): - 3.92 pcm /K; 4. Pump run-down characterisitics (used MSRE half-time ∼ 4.2 sec). Observations were made as regards the MOSART Molten Salt concepts are: 1. The critical MOSART design is immune to most unprotected transients except the UTOP transient (fuel agglomeration issue in piping system); 2. If fuel agglomeration issue should be an issue, we need to assess how large (in terms of reactivity) potential fuel particles could be

[en] Open literature data were used to create computer models of the Superphénix sodium-cooled fast reactor (SFR) core and in-vessel structures with two codes: TRACE/FRED (PSI) and SIM-SFR (KIT). In both models 2D heat structures, 1D thermal-hydraulic channels and 0/1D reactor kinetics were used to account for dynamic reactivity feedbacks due to fuel temperature evolution as well as differential thermal expansions of fuel, cladding, sodium, diagrid, strongback, vessel and control rod drivelines. Six Superphénix start-up transient tests were analyzed using dynamic boundary conditions based on the test data for core inlet temperatures, flow rates and control rod position. Good agreement between calculations and measurements for reactor power and outlet temperature were obtained for all six cases using a single reactivity coefficient input data set. This analysis contributes to better understanding of SFR core/primary system dynamic behavior and to validation of the tools currently used for transient analysis of the Generation-IV SFR. (author)

[en] The major characteristics of the Molten Salt Reactor are presented. The basic transient behaviour of these particular reactor designs can be characterised by the mismatch in the temperature response of the fuel (fast acting) and the graphite (slow acting). After the initial transient phase, during which the average fuel temperature dominates the transient response, the graphite temperature catches up and impose its characteristics onto the plant dynamic behaviour thereafter. In general, however, all transients are observed to be very sluggish due to the very large thermal inertia associated with the graphite in the core. The long term dynamic behaviour of the reactor becomes unstable if the total reactivity coefficient of the system should be positive. The long term reactor power level will not stabilize under these conditions. Should the total temperature coefficient be negative, the reactor will stabilize at a certain power level with corresponding temperatures. The total temperature coefficients for the 3 reactors studied have values close to zero, if erbium is not added to the graphite. Moreover, they show considerable variations with temperature, leading to quite complex and unpredictable long term behaviour (if operator does not manually intervene).The sluggish transient behaviour of this reactor design, however, provides sufficiently response time for the reactor operators to counteract the failed control rod system that has been assumed not functional during all of the above transients analyzed. Since the initial phase of all transients is dominated by the negative reactivity coefficient associated with the fuel temperature, the reactor can be basically characterized as safe