Reactivity Worth Research Articles

Reactivity Worth Measurement of Calcium Hydride Sample in UTR-KINKI Reactor

Small modular reactors (SMR) using calcium hydride (CaH2) as a moderator have attracted attention due to their passive safety and economic efficiency. In order to design the nuclear design of the SMR, the integral validation of the cross section considering the thermal neutron scattering law(S(α,β)) of CaH2 is necessary. However, there are no integral experiments using CaH2 in the reactor. As the first integral experiment in nuclear reactors, we performed reactivity worth measurements of CaH2 samples in a university training and research reactor (UTR-KINKI) of Kindai University. The measured sample reactivity worth was compared with the results calculated by the continuous energy Monte Carlo code MVP3 and JEFF-3.1, and integral validation of the CaH2 cross section stored in JEFF-3.1 was performed. To obtain the sample reactivity worth components, we carried out perturbation calculations.

Reactivity modulation experiments for nuclear data in CROCUS within a CEA-EPFL collaboration

In nuclear research reactors, integral experiments are powerful tools to measure integral core parameters, such as the delayed neutron fraction. Within the scope of the point kinetic approximation, reactivity modulation experiments can be used for probing the reactor transfer function and then infer integral parameters of the core. In this context, Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA) and Ecole Polytechnique Fédérale de Lausanne (EPFL) have been collaborating for developing a probe device (PISTIL) and measurement setup adapted to the CROCUS zero power research reactor operated by EPFL (Lausanne, Switzerland). Despite some mechanical limitations of PISTIL, its maximum reactivity worth was measured with a good precision and repeatability using different methods (8.82 ±0.07 pcm), and its value is found rather close to the simulated one using TRIPOLI-4 (9.4 ± 0.4 pcm with JEFF-3.3). Above 1 Hz, the shape of the used modulation is pseudo-sinusoidal, with only a few well defined harmonics of excitation. The strongest harmonic only was analyzed using standard signal processing algorithms such as the Fourier transform and the Bartlett estimator. Twelve data points were produced in the range 0.5 Hz to 200 Hz, with uncertainty ranging from 1 % to 15 %. The prompt decay constant was measured at 150 ± 3 rad/s. Below 1 Hz, stepwise modulations were used with pseudo-random time sequences, which allowed exciting at once a large number of frequencies. Around 150 data points were produced in this particularly interesting frequency domain, between 1.6 mHz and 0.75 Hz, thanks to the use of three distinctive sequences with different base frequencies and overlapping ranges. The amplitude and phase of the RTF were measured satisfactorily, with uncertainties below 1 % for the strongest harmonics. The shape of the RTF was found consistent with the predictions of both JEFF-3.3 and ENDF/B-VII.1 libraries.

Measurement of 233U-HU Substitution Reactivity Worth in KUCA for Integral Validation of 233U Nuclear Data

In order to perform an integral evaluation of 233U nuclear data, measurements of substitution reactivity worth between 233U sample and high enriched uranium (HU) sample in the Kyoto University Critical Assembly (KUCA) were carried out. In this study, the experimental core was consisted of 24 fuel elements and 1 special fuel element with the 233U or HU sample. The fuel element was consisted of 31 unit cells and sandwiched by a upper and a lower polyethylene reflector. The unit has one enriched uranium plate of 1/16” thickness, three polyethylene plates of 1/8” thickness. The special fuel element was loaded at the center of the core. The substitution reactivity worth was defined as difference between the excess reactivity of the core in which the 233U sample or the HU sample inserted the core. The excess reactivities ware measured by the positive period method. As the result, the substation reactivity was obtained 0.0146±0.0006 %∆k/k. A numerical result by MVP3 with JENDL-4.0 was 0.0141±0.0002 %∆k/k and a C/E was 0.964 ± 0.041. On the other hand, the calculated substitution reactivity worth with JENDL-5 was 0.0168±0.0002 %∆k/k and a C/E was 1.151±0.047.

Time-of-flight measurements of MINERVE samples containing fission products and neutron absorbing isotopes

The zero-power reactor MINERVE (CEA Cadarache) was designed to perform reactivity worth measurements by the oscillation technique. The various experimental programs, undertaken for the last thirty years, involved cylindrical samples with a diameter of about 1 cm and a height ranging from a few cm to 10 cm. Most of the samples are composed of UO2 pellets mixed with a high neutron absorbing nuclide, i.e. fission product, actinide, in a double-sealed Zry-4 container. An experimental program started in 2015 in collaboration with the Joint Research Centre of Geel to study the MINERVE samples at the time-of-flight facility GELINA by the neutron transmission technique. The two main objectives consist of checking both the composition of the MINERVE samples provided by the manufacturer and the quality of the resonance parameters recommended in the evaluated neutron data library JEFF-3.3. The pioneer experiments on MINERVE samples containing 107Ag and 109Ag revealed a substantial Tungsten contamination that was not reported by the manufacturer. Such a Tungsten contamination is related to the manufacturing process of the sample pellets. The observed Tungsten contaminations lead to non-negligible increases of the C/E ratios up to a few percent. second experimental campaign on MINERVE samples containing 99Tc provided useful insight on the quality of the 99Tc resonance parameters measured at the GELINA facility at the end of the 90s. The ongoing program continuing through 2022 will deliver data for samarium (natural, 147, 149, 152), neodymium (natural,143,145), gadolinium (natural, 155), europium (151, 153), rhodium (103), cesium (133), hafnium (180), dysprosium (160, 161, 162, 163, 164) and erbium (168, 170). The present work focuses on the data analysis technique developed for long cylindrical samples with a diameter smaller than the neutron beam, and on the grain size distribution model implemented in the resonance shape analysis REFIT.

Reactor physics calculation and measurement during first approach to criticality of Apsara-U research reactor

Apsara-U is 2 MW swimming pool type research reactor with low enriched uranium fuelled core. On 10th September 2018, Apsara-U was made critical at 18:41 hrs. After the First Approach to Criticality (FAC) of Apsara-U, low power physics experiments were carried out. The experiments included reactivity worth measurement of all reactivity devices and measurement of the isothermal temperature coefficient of reactivity. Inflated reactivity worth of Shut off Rods as calculated from the power from some of the neutronic channels were obtained on account of shadowing effect of neutron reaching the detector locations. Further the isothermal coefficient of reactivity was observed positive. In this paper, we discuss the mentioned Reactor Physics measurement and related theoretical calculation during FAC that includes solving the benchmark problem for validating the calculation tools, and steady state thermal hydraulic analysis, Reactivity initiated transients for the FAC core configuration for ascertaining the safety of the reactor.

At-power subcritical multiplication in the advanced test reactor

The degree to which a given physical subdivision of a critical reactor acts as a neutron source for other lobes is not analytically characterized for general application. ATR and ATRC both operate at power, and individual power-producing lobes rely on each other as neutron sources in order to maintain constant power, which in general requires either exactly critical multiplication within a reactor or an external neutron source. This work shows that lobe powers in ATR or ATRC can be related with subcritical multiplication theory, demonstrating that the lobes are not individually critical. This explanation is significant for ATR due to the desire to irradiate a large variety of experiments simultaneously, each having its impact on the core neutron population. This work also establishes the lack of relationship between experiment reactivity worth and cross-sections calculated from fundamental nuclear data. For any physical subdivision of any other critical reactor, it is likewise true that the subdivision undergoes only subcritical multiplication.

Control rods reactivity worth calculation using deterministic and Monte Carlo approaches for an MTR type research reactor

Reactivity worth of control rods as a main parameter in different nuclear reactor fields such as safety, design, and operation that could be calculated or measured with different experimental and theoretical methods. Reliable answers in calculations necessitate taking into account different characteristics such as geometries, materials, temperatures, spatial nodes, libraries, and energy groups. Reactivity worth of different core states of a Material Testing Reactor (MTR) is calculated using MCNPX 2.6.0 code and MTR_PC package as Monte Carlo and deterministic approaches respectively. It is seen that the MTR_PC and MCNPX results has considerable differences up to 51%. Therefore, an exhaustive study is done concentrating on different involved parameters. Precise modification of inputs, applying one common library in the two approaches, correcting spatial nodes, and employing more energy groups in the deterministic approach are performed. It is determined that the effect of the spatial nodes is much more important than the other parameters in the deterministic method. Finally, results of two approaches are found to be satisfactory as discrepancy is less than 11%.

Core optimization of UO2 fuelled ALLEGRO reactor

Feasibility of UO2 core is investigated considering neutronic and thermal constraints. Calculations showed that the amount of UO2 fuel must be increased significantly to reach criticality. Irradiation time needed to reach the target value of 75dpa increases by increasing fuel amount and keeping total thermal power constant. Changes in fuel composition and design-related differences of fissile isotope vectors in depleted fuel were studied. Beta effective value, void, Doppler and fuel expansion coefficients were calculated. Compliance to the shutdown criteria is demonstrated since reactivity worth of control rods is affected by the modifications of the core design. Assembly and pin-wise power distribution, as well as axial power distribution is examined for different core designs. Reactivity changes were calculated for the case of water ingress to the core as a result of design a basis event PRISE.

Conceptual design of a 3 MWth yttrium hydride moderated heat pipe cooled micro reactor

Micro reactors are gaining substantial interest worldwide due to their capabilities to deliver power to remote and decentralized areas and their long fuel cycles, making them more attractive than other energy sources. This paper presents a conceptual design and steady-state neutronics and thermal-hydraulics analyses results for a heat pipe cooled yttrium hydride moderated micro reactor. The reactor was designed for both terrestrial and space application purposes; hence, size and weight were important. The neutronics analysis calculations were performed to evaluate key safety parameters such as control devices reactivity worth, temperature reactivity coefficients, and power distribution of the reactor core. Burnup calculations were performed using Monte Carlo code MCNP6.2 and verified against Serpent code results. The thermal-hydraulics analyses calculations were performed to evaluate the temperature distribution in a representative unit cell in the reactor core using the STAR CCM+ multi-physics computational fluid dynamics (CFD) software by utilizing the power distribution obtained from neutronics calculations. It was demonstrated that this reactor could be operated safely for more than 11 years, generating 3 MW of thermal power.

Evaluation of Fuel Reactivity Worth Measurement in the Prototype Fast Reactor Monju

In the prototype fast breeder reactor Monju, fuel reactivity worth was measured at six positions as the reactivity corresponding to the differences of critical control rod positions between cores with and without a dummy fuel subassembly. In this paper, the measurements are evaluated in detail, and their reliability and usefulness as validation data for fast reactor neutronics design methodologies are examined through a comparison with calculations. Calculated-to-experiment values and their uncertainty of fuel reactivity worth were 0.97 to 1.02 and 4% to 6%. Through this study, the consistency and reliability of the measurements and calculations were confirmed.

Processing of sunflower hulls and peat with production of melanin and composite fuel briquettes

The article discusses production of composite peat fuel briquettes from butter-making waste represented by sunflower hulls. The experiments prove manufacturability of briquettes at the quality of the standard first-grade peat fuel briquettes. The specific heats of sunflower hulls and composite briquettes are determined. Furthermore, the experimental tests show that ultrasonic treatment used to extract melanin from sunflower hulls changes the chemical composition of the raw material and its reactivity worth, as well as promotes an increase in the specific heat of sunflower hulls from 17.5 to 19.7 MJ/kg. The ultrasonic treatment degrades lignified tissue matrix of sunflower hulls. The Fouriertransform infrared spectra of lignin from sunflower hulls are analyzed before and after the ultrasonic treatment. Despite the strong resemblance, the spectra display differences in the structure of lignin, as well as in the ranges of fluctuation of methyl and methylene groups. It is shown that the yield of melanin from sunflower hulls in the optimized extraction conditions in the ultrasonic field (temperature 60 0С, extraction time 25 min, density 414 W/cm2, sunflower hulls : extractant ratio 1 : 40) totals 7.6% of the biomass of sunflower hulls. The size of melanin particles is 457–465 nm. The accomplished research proves feasibility of the integrated approach to processing of sunflower hulls to produce bio-active melanin and composite peat fuel briquettes.The study was supported by the Russian Science Foundation, Project No. 20-69-47084.

Comparative analysis between homogeneous and heterogeneous models of gas cooled fast reactor core (GFR-2400)

Abstract The main objective of the presented work is to investigate the effectiveness of homogenization of the GFR-2400 core in simulating its neutronic characteristics. To explore and evaluate the neutronic behavior of the large scale Gas cooled Fast Reactor GFR-2400 core, two computer models (homogeneous and heterogeneous) were designed using the MCNPX code. The designed heterogeneous model has been validated by comparing its results with a previously published paper. The results of both models were compared with each other to study the effect of fuel homogeneity on the radial flux and power distribution through the core. As part of a safety analysis of the reactor core for both designs, the reactivity worth of control rods, Effective delayed neutron fraction (ß eff), Doppler constant and Depressurization effect have been analyzed and compared. The variations of the effective multiplication factor (k eff), the minor actinides concentration and the most important fission products as a function of operation time have been investigated for the presented designs.

Inherent safety enhancement by design optimization of a floating absorber for safety at transient (FAST) in an advanced burner reactor

A floating absorber for safety at transient (FAST) is a passive safety system to enhance the inherent safety of sodium-cooled fast reactors (SFRs). This study optimized and improved the original FAST design in terms of the initial position, barycenter, eccentricity, reactivity worth, and diameter ratio of FAST to pin (k) in advanced burner reactor (ABR) cores. In contrast to previous studies based on average assembly analysis, a comprehensive design optimization study was conducted through transient simulations based on multiassembly analysis. Applying a damped FAST, which is an improvement over the original FAST design, enhanced the negative reactivity feedback in the oxide cores depending on the diameter ratio. Oscillatory behaviors were observed when the k of the damped FAST system was within 0.1 ∼ 0.55 for $1 of the FAST reactivity under an unprotected loss of flow event (ULOF). A damped FAST with a k of 0.7 maintained the lowest coolant temperature while suppressing the oscillatory behavior for $1 of the FAST reactivity. The results showed that damped FAST systems support negative reactivity feedback for the ABR oxide core and should be properly designed to avoid power and temperature oscillations depending on the FAST reactivity worth.

Assessment of TMSR-500 Shutdown Capability

The molten salt reactor (MSR) is a generation IV reactor with liquid fuel having nearly zero excess reactivity. Due to the very low excess reactivity, it requires a small number of control rods worth to shut down the reactor. However, as it operates at high temperatures, the core reactivity increases as the fuel temperature cools down during shutdown. In such a case, the control rods might not be able to keep the reactor at a subcritical state, and consequently, the fuel must be removed from the core for long-term shutdown into a fuel drain tank (FDT) below the core. This paper is intended to assess the shutdown capability of the first active shutdown system and fuel drain tank of ThorCon MSR by doing neutronic calculations with MCNP6. The results indicated that the control rods having reactivity worth ‑1.699 %dk/k are unable to maintain the core at a subcritical state as the core excess reactivity increases to +7.760 %dk/k when the fuel reaches room temperature. Therefore, the fuel must be drained to FDT to be cooled down and kept subcritical. Evaluation for various cases of FDT produced the highest multiplication factor of 0.57008 ± 0.00004 at the most conservative condition. The multiplication factor is well below the critical state of 1.0. The evaluations suggest that soon after the control rods shut the reactor down, the fuel has to be drained to FDT to maintain shutdown condition and dissipate the decay heat.

Neutronic studies for the second Egyptian research reactor ET-RR-2

Abstract The start up physics studies SPS of the Second Egyptian Research Reactor is presented in this work. The SPS consists of the approach to criticality, the power and flux distribution of the first critical core, criticality, power and flux distribution of the working core. The working core is the beryllium-reflected core with cobalt loaded for irradiation. Some of the neutronic parameters are determined, these parameters are the power peaking factor, power distribution, the excess reactivity for both the water and beryllium reflector cores, the core excess reactivity with the cobalt loaded, the reactivity worth of the cobalt irradiation box. Flux distribution for one fast, three eipthermal and one thermal energy group are also determined. All the above neutronic parameters are calculated by using the lattice cell code WIMSD4 and the two dimensional diffusion code DIXY2.

A design study on a metal fuel fast reactor core for high efficiency minor actinide transmutation by loading silicon carbide composite material

ABSTRACT A metal fuel fast reactor core for high efficiency minor actinide (MA) transmutation was designed by loading silicon carbide composite material (SiC/SiC) which can improve sodium-cooled fast reactor (SFR) core safety characteristics such as sodium void reactivity worth and Doppler coefficient due to neutron moderation. Based on a 750 MWe metal fueled SFR core concept designed in a prior work, the reactor core loading fuel subassemblies with SiC/SiC wrapper tubes and moderator subassemblies was designed. To improve the reactor core safety characteristics efficiently, three layers of SiC/SiC moderator subassemblies were loaded in the core by replacing 108 out of 393 fuel subassemblies with the moderators. The reactor core with approximately 20 wt% MA-containing metal fuel satisfied all safety design criteria and achieved the MA transmutation amount as high as 420 kg/GWe-y which is twice as high as that of the axially heterogeneous core with inner blanket and upper sodium plenum, and two-thirds of that of the accelerator-driven system.

A perturbation-based hybrid methodology for control drum worth prediction applied to the HOLOS-Quad microreactor concept

Fast and accurate control system reactivity worth estimates are desirable in many applications. Unfortunately, the full core Monte Carlo transport simulations often used to estimate the reactivity worth of control drums are often too costly to run in the time scale needed for many applications, such as model predictive control, reactor design optimization, uncertainty quantification and sensitivity analysis. Therefore, this paper presents a general methodology for control drum worth estimation that is based on a hybrid model consisting of physics-based and statistics-based components. It was found that in the application of the methodology to the HOLOS-Quad Reactor Concept, errors as low as 50 pcm were achieved. Model accuracy was also analyzed as a function of training set size and it was found that the most accurate model required about 70 transport model evaluations to train with further reactivity estimates being calculated on the order of milliseconds.

Preliminary analysis of a zirconium hydride moderated megawatt heat pipe reactor

At present, megawatt-class Heat-Pipe Reactors (HPRs) are under-investigated widely as a candidate for robust, self-contained, and long-term power for special purpose applications. In this work, zirconium hydride is introduced to design a moderated megawatt-class HPR core. The influence of moderators on core parameters is analyzed. According to the calculation, although the use of a moderator will lead to higher burnup reactivity loss, the reactivity decreasing caused by the reduction of fuel can be offset by the proper moderator. The introduction of moderator has no significant influence on radial and axial power distribution and power peak factor. Meanwhile, the moderation of core will decrease the maximum neutron flux as well as the reactivity worth of control devices. The moderated core possesses a stronger doppler effect while the thermal expansion reactivity feedback of monolith and fuel decreases. Results also show that the introduction of moderators is beneficial to reducing the thickness and weight of the shield. Preliminary thermal analysis is conducted to verify the feasibility, the results show that zirconium hydride can stay stable in the moderated megawatt heat pipe reactor, and the yttrium hydride shows better thermal stability.

Reactivity worth measurement of the lead target on VENUS-II light water reactor and validation of evaluated nuclear data

In this paper, the benchmark experiment of the lead reactivity worth was introduced in detail, which has been carried out under a super-critical state at the light water reactor of the VENUS-II experimental facility. The reactivity worth of the cylindrical lead target was measured and processed as 142.0 ± 11.3 pcm by the period method. In simulations, the reactivity worth of the lead target was calculated by MCNP with the five libraries, i.e., ENDF/B-VII.1, ENDF/B-VII.0, JENDL-4.0, CENDL-3.1 and JEFF-3.2. By comparisons, the MCNP simulated results of the lead reactivity worth with ENDF/B-VII.1, ENDF/B-VII.0, CENDL-3.1 and JEFF-3.2 libraries are underestimated, and that of JENDL-4.0 shows slight better agreement with the experiment data within errors. Through the perturbation analysis, the (n, el) reaction channel is most related to the positive reactivity worth of the lead target. Additionally, the reaction rates of the (n, el), (n, inl) and (n, γ) channels were calculated by FM card in MCNP. It is observed that the (n, γ) reaction rates from different libraries show the maximum difference, which is due to the discrepancy of the (n, γ) cross sections in the thermal neutron region. In conclusion, the lead cross sections from the main libraries in the world have been preliminarily validated based on the measurements, which lays the foundation for the engineering design of CIADS.

A critical study on best methodology to perform UQ for RIA transients and application to SPERT-III experiments

The aim of this paper is to assess the reliability and accuracy of the PSI standard method, used in many previous works, for the quantification of ND uncertainties in the SPERT-III RIA transient, by quantifying the discrepancy between the actual inserted reactivity and the original static reactivity worth and their associated uncertainties. The assessment has shown that the inherent S3K neutron source renormalization scheme, introduced before starting the transient, alters the original static reactivity worth of the transient CR and reduces the associated uncertainty due to the ND perturbation. In order to overcome these limitations, two additional methods have been developed based on CR adjustment. The comparative study performed between the three methods has showed clearly the high sensitivity of the obtained results to the selected approach and pointed out the importance of using the right procedure in order to simulate correctly the effect of ND uncertainties on the overall parameters in a RIA transient. This study has proven that the approach that allows matching the original static reactivity worth and starting the transient from criticality is the most reliable method since it conservatively preserves the effect of the ND uncertainties on the inserted reactivity during a RIA transient.