Coolant Void Fraction Research Articles

Numerical investigation of rod bundle thermal–hydraulic behavior during reflood transients using COBRA-TF

Accurate prediction of the reflood transients during LOCA has long been a challenging task. One reason for this is the substantial cost incurred to perform large-scale reflood tests, which are formidable in experimental expenditure and intensive in technology as well as program management. Another reason preventing a comprehensive understanding of reflood transients is the high-level complexity involved in the two-phase flow mass and heat transfer processes, which makes the measurement extremely difficult and inefficient. The situation has been significantly improved with the design and operation of the Nuclear Regulatory Commission (NRC)/Pennsylvania State University (PSU) Rod Bundle Heat Transfer (RBHT) test facility. A variety of advanced instrumentations were developed and used at this facility and very high-resolution data has been obtained, especially for the liquid droplet field. In the current study taking advantage of the unique NRC/PSU RBHT data, an extensive and comprehensive code evaluation and validation is carried out using the thermal–hydraulic sub-channel analysis code COBRA-TF. The system parametric effects investigated include: the system pressure, inlet liquid subcooling temperature, inlet flooding rate and rod bundle power. A variety of thermal–hydraulic quantities predicted by the numerical code are evaluated including: the quench front propagation, cladding, spacer grid and vapor temperature variations, two-phase pressure drop, liquid droplet velocity and coolant void fraction. In addition, the prediction errors are presented for each of the quantities in great detail in order to have a comprehensive understanding of the code performance. In general, COBRA-TF agrees relatively well with the experimental data in terms of the overall quench front propagation, only predicting slightly earlier quench. However, under low inlet subcooling and high flooding rate conditions, significant discrepancies are observed. The comparison with rod bundle thermal–hydraulic parameters indicates that COBRA-TF is able to predict the cladding temperature before quench well within a 15% range. The spacer grid and vapor temperature predictions are within 20% error. While the overall droplet velocity prediction is found to be within 30% error, COBRA-TF is able to capture the decreasing trend for droplet velocity during reflood. In addition, the prediction of the dispersed flow film boiling void fraction shows that the code always under-predicts the void fraction. While the most prediction discrepancy is found to be the two-phase flow pressure drop after quench, which involves more than 50% under-prediction for the bulk liquid boiling regime. The present study clarifies the effect of different system parameters on the various two-phase flow quantities during reflood. The results obtained provide answers for many existing code modeling issues and thus can be instructive and useful for future model development and code upgrading.

Prediction of time series of NPP operating parameters using dynamic model based on BP neural network

A dynamic model was developed using two back-propagation neural networks of the same structure, one for online training and the other for prediction, and proposed for continuous dynamic prediction of the time series of NPP operating parameters. The proposed prediction model was validated by predicting such time series of NPP operating parameters as coolant void fraction, water level in SG and pressurizer. Validation results indicated the proposed model could be used to achieve a stable prediction effect with high prediction accuracy for the prediction of fluctuating data.

Hafnium clad fuels for fast spectrum BWRs

We show that by use of hafnium cladding, a fast neutron spectrum is achievable in the top of uprated BWRs. Monte Carlo calculations have been made for Hf clad inert matrix nitride and low fertile MOX fuels, with fuel segments located in the upper part of an uprated BWR, where the coolant void fraction exceeds 70%. The nitride fuel results in the hardest neutron spectrum, but the low fertile MOX fuel still yields fission probabilities for even neutron number nuclides similar to those of sodium cooled reactors. The inert matrix nitride fuel configuration yields high burning rates, permitting to stabilise TRU inventories with less than 50% BWR cores of the here suggested type in the power park. The core with low fertile MOX fuel is less efficient, but still a zero net producer of TRU. Fuel and coolant temperature feedbacks are affected by introduction of absorbing elements in the fuel, but remain within acceptable ranges for the low fertile MOX fuel. Although control rod worths are reduced, shutdown margins are sufficient to ensure sub-criticality in cold conditions. From a materials point of view, the behaviour of hafnium clad MOX fuel would be similar to zircalloy clad MOX fuel already used extensively in nuclear industry. Thus, if dynamic stability of the core can be ensured, the here proposed fuel may be considered as a low cost solution for transmutation of minor actinides on industrial scale.

Conceptual Design Study of 180 MWt Small-Sized Reduced-Moderation Water Reactor Core

Conceptual design of a Small-sized Reduced-Moderation Water Reactor (S-RMWR) core, which has the thermal output of 180 MW, the conversion ratio of 1.0 and the void reactivity coefficient of negative value, has been constructed. S-RMWR is a technology demonstration reactor which also conducts material and fuel testing for commercial use of Reduced-Moderation Water Reactor (RMWR) in large-scale power plants. It has a very tight triangular fuel rod lattice and a high coolant void fraction. The RMWR core axially has two short and flat uranium plutonium mixed oxide (MOX) regions with an internal blanket region in between, in order to avoid a positive void reactivity coefficient. The MOX regions are sandwiched between upper and lower blanket regions, in order to increase a conversion ratio. In this small reactor core, leakage of neutrons is expected to be larger than in a large core. Therefore, a core design concept different from that for a large core is necessary. Core burnup calculations and nuclear and thermal-hydraulic coupled calculations were performed in the present study with SRAC and MOSRA codes. MVP code was also used to obtain control rod worth. Because of its large neutron leakage, keeping the void reactivity coefficient negative is easier for S-RMWR than RMWR. Thus, the heights of MOX region can be taller and the plutonium enrichment can be lower than in RMWR. On the other hand, to achieve the conversion ratio of 1.0, radial blanket and stainless steel reflector assemblies are necessary, whereas they are not needed for RMWR.

Conceptual Design Study of 180 MWt Small-Sized Reduced-Moderation Water Reactor Core

Conceptual design of a Small-sized Reduced-Moderation Water Reactor (S-RMWR) core, which has the thermal output of 180 MW, the conversion ratio of 1.0 and the void reactivity coefficient of negative value, has been constructed. S-RMWR is a technology demonstration reactor which also conducts material and fuel testing for commercial use of Reduced-Moderation Water Reactor (RMWR) in large-scale power plants. It has a very tight triangular fuel rod lattice and a high coolant void fraction. The RMWR core axially has two short and flat uranium plutonium mixed oxide (MOX) regions with an internal blanket region in between, in order to avoid a positive void reactivity coefficient. The MOX regions are sandwiched between upper and lower blanket regions, in order to increase a conversion ratio. In this small reactor core, leakage of neutrons is expected to be larger than in a large core. Therefore, a core design concept different from that for a large core is necessary. Core burnup calculations and nuclear and thermal-hydraulic coupled calculations were performed in the present study with SRAC and MOSRA codes. MVP code was also used to obtain control rod worth. Because of its large neutron leakage, keeping the void reactivity coefficient negative is easier for S-RMWR than RMWR. Thus, the heights of MOX region can be taller and the plutonium enrichment can be lower than in RMWR. On the other hand, to achieve the conversion ratio of 1.0, radial blanket and stainless steel reflector assemblies are necessary, whereas they are not needed for RMWR.

Application of the Post-Boiling-Transition Standard to Licensing Analysis

To ensure fuel integrity, light water reactor cores are designed to avoid the onset of boiling transition (BT) inside the fuel assembly that leads to a deterioration of the heat transfer characteristics and subsequent excessive rise of the fuel-cladding temperature in the anticipated operational occurrences (AOOs). However, some boiling water reactors’ AOO events result in immediate scram or suppression of the reactor power due to an increase in the reactor coolant void fraction. Recent studies show that a short duration of dryout inside the fuel assembly only leads to a small rise in the fuel-cladding temperature and thus does not pose a threat to fuel integrity. Many tests on BT and an improved comprehension of its mechanism have led to the development of a methodology to appropriately assess the fuel-cladding temperature after BT has been reached. The Standards Committee of the Atomic Energy Society of Japan has therefore proposed a cladding temperature criterion after BT. Applying the post-BT standard enables the value of the operating limit minimum critical power ratio (OLMCPR) to be decreased by allowing for a short duration of dryout. We calculated the fuel-cladding temperature and dryout duration in the load rejection condition without a bypass event. The calculated results show that both the fuel-cladding temperature and dryout duration meet the post-BT standard in the case of a small OLMCPR, which is determined by the loss of feedwater heating. This enables a more efficient reactor core to be designed by applying the post-BT standard to licensing analysis. The possibility of applying a post-BT standard is demonstrated from the results of this work.

State-of-the-art computer code RELAP5 validation with RBMK-related separate phenomena data

Validation of the RBMK model, developed by employing best estimate system computer code RELAP5 is performed by employing the data from NPPs operation or from integral and separate effects facilities. Validation of the models on the basis of separate phenomena is necessary to perform due to the fact that RELAP5 code has been developed for PWRs, which operate at different conditions (pressure, temperature, coolant void fraction, etc.) from RBMKs. In addition to that, there is a number of phenomena specific for RBMK type reactors (oscillatory flow rate behaviour in parallel channels, flow stagnation in channels, stratification in long horizontal piping, etc.), which have not been studied during RELAP5 validation for PWRs. In the paper, RELAP5 models for separate effects related to RBMK-1500 are presented and modelling of transients is performed. Obtained results are compared with experimental data.

Effect of a 0.3-eV Resonance Cross Section for Plutonium on the Coolant Void Reactivity in a Heavy Water Lattice

Plutonium fuel could be utilized in the entire core of a heavy water-moderated, boiling light water-cooled pressure-tube-type reactor (HWR). The coolant void reactivity, however, depends on the various parameters of the lattice. It is especially significant to clarify the effect of plutonium nuclides on the void reactivity.The void reactivities in the infinite HWR lattices have been parametrically analyzed to clarify the effects of changes in the lattice parameters on the void reactivity using the WIMS-D4 code with the JENDL-3.1 nuclear data. At present, it is known that the behavior of the void reactivity can be clarified by separating the components of fuel nuclides, neutron cross sections, energy groups, and regions in the lattice cell from the global reactivity effect, using the important reaction rates.If the fuels are the same in the macroscopic absorption cross section for the 2200 m/s neutron, it has been shown that the void reactivity shifts further to a negative direction in a narrower pitch lattice and in the plutonium-fueled lattice with a higher content of 239Pu rather than in the uranium one. The effect of reducing the void reactivity to the negative by fissile plutonium is caused mainly by the presence of the resonance cross section at ~0.3 eV of 239Pu. The higher the content of 239Pu, the less the recovery of dipped neutron flux within the resonance energy width due to a decrease in the thermal neutron scattering of hydrogen with an increase in coolant void fraction, so that the decreased resonance fission rate for 239Pu contributes to the more negative direction for the void reactivity.On the other hand, resonance at ~0.3 eV for 241Pu does not have an important role for the void reactivity because its resonance cross section is smaller than that of 239Pu.

Axial Dependence of Partial Void Reactivity in a Light Water-Cooled, Heavy Water-Moderated, Pressure-Tube Reactor

The effect on reactivity of changes in the coolant levels in the pressure tubes of a pressure-tube heavy water reactor is experimentally studied to clarify the effect of an axial coolant void fraction distribution. The coolant void fraction distribution is simulated by stepwise changes in the coolant levels in the Deuterium Critical Assembly (DCA). The reactivity is measured for a 25.0-cm-pitch square-lattice core with a positive coolant void reactivity. The reactivity changes resulting from changes in the coolant levels are measured as changes in the critical heavy water levels. The axial distribution of the thermal neutron flux is also measured by the copper activation method.In these measurements, the critical heavy water levels show a sinuous curve having a maximum and a minimum, and a positive reactivity larger than those of both the 0% and the 100% void uniform cores is introduced at certain coolant levels by stepwise changes in the coolant levels.An experimental analysis is performed with the coupled WIMS-ATR/CITATION code system, whose analytical method was established through DCA critical experiments. Agreement between experiment and analysis is fairly good. Furthermore, the peculiar reactivity behavior resulting from changes in the coolant levels is analyzed using a simplified model to take note of typical reactor physics parameters. It is clarified that this anomalous phenomenon is caused by the combined effect of the flattened S curve change in the thermal neutron absorption and the even flatter S curve change in the neutron leakage caused by the changes in the coolant levels. Useful information is obtained regarding reactivity behavior with an axial coolant void fraction distribution.

Time-Independent Neutronic Analysis of the Chernobyl Accident

Estimates are made of the positive reactivity introduced through the growth of the coolant void fraction in the Chernobyl reactor at both the average burnup value given by the Soviets and the maximum value. Using Monte Carlo models, various possible axial burnup distributions, displacer models, conditions in the control channels, and control rod positions are considered in calculating the insertion of positive reactivity by the manual and emergency control rods, that is, the positive scram. Two possible scenarios are examined for a second reactivity peak: (a) creation of a mixture of fuel, water, and cladding in a number of central fuel channels, resulting in the explosion of these channels, and (b) uniform vaporization throughout the entire reactor, resulting in reactor depressurization. From the data presented in this paper, it can be concluded that vaporization of the cooling water in the fuel channel gave the highest reactivity contribution to the Chernobyl accident. The positive reactivity due to insertion of the manual and emergency control rods played only a minor role in the reactivity balance of the accident.

Study on Coolant Void Reactivity of Pressure-Tube-Type Heavy Water Lattice by the Substitution Method

Using the substitution method combined with the pulsed neutron techniques, coolant void reactivities of PuO/sub 2/-UO/sub 2/ fuel lattices in pressure-tube-type heavy water reactors have been determined as functions of PuO/sub 2/ enrichment in PuO/sub 2/-UO/sub 2/(0.54 and 0.87 wt%), fissile content of plutonium (91 and 75% fissile plutonium), lattice pitch (V/sub m//V/sub f/: 7.4 and 9.9), and coolant void fraction (0, 30, 70, 87 and 100%). The reference loading of 1.2 wt% enriched UO/sub 2/ clusters was progressively replaced by PuO/sub 2/-UO/sub 2/ test clusters. The void reactivities were obtained from Simmons King's formula in which correction was made for a change of the prompt generation time. As decay constants can be maintained invariable due to substitution, buckling differences were analyzed by the first-order perturbation method, on the assumption that lattices are homogeneous and no difference in diffusion coefficients exists between the two lattices. Void reactivities of test lattices were determined with an accuracy of --10% when the minimum number of test fuel clusters was -- 5% of the total. The void reactivity shifted farther to the negative side as the proportion of fissile plutonium was increasingly in the PuO/sub 2/-UO/sub 2/ fuel of the same enrichment of plutonium.

Sensitivity Analysis of Cell Neutronic Parameters in High-Conversion Light-Water Reactors

The sensitivity coefficients of neutronic performance parameters in high-conversion LWR cells have been calculated by means of the SAINT code. In order to show the specific features of the sensitivity coefficients in the HCLWR cells, the differences between sensitivities were investigated for cells with different moderator to fuel volume ratios and different Pu enrichments. The burnup dependence of the sensitivities was also discussed with an emphasis on the effect of fission products on the cell parameters. We have performed the sensitivity analysis for the PROTEUS cores. Group constants of main heavy nuclides were compared for the different cell calculational methods; SRAC and VIM, and the different cross section libraries; JENDL-2 and ENDF/B-IV. The effect of the differences in group constants was estimated for the cell parameters k ∞, reaction rate ratio and coolant void worth. The differences in the 238U capture and 239Pu fission group constants in the resolved and unresolved resonance range produced 0.3∼1.0% change in k ∞ and 1∼6% change in the coolant void worth. These effect was largely dependent on the coolant void fraction.

Substitution Measurements on 28-Fuel-Rod Critical Clusters in D2O and Their Analysis by the Second-Order Perturbation Method

Material bucklings have been determined as functions of 235U enrichment in UO2 (0.7, 1.2, and 1.5 wt% 235U), PuO2 enrichment in PuO2-UO2 (0.54 and 0.87 wt% PuO2), fissile content of plutonium (91 and 75% Pu-fissile), lattice pitch (Vmod/Vfuel: 7.4 and 9.9), and coolant void fraction. The reference loading of 1.2 wt% 235U-enriched UO2 clusters was progressively replaced by the test clusters.Buckling differences resulting from the substitutions were analyzed by the new second-order (iterative) perturbation method, on the assumption that neutron diffusion is isotropic and that no difference in diffusion coefficients exists between the two lattices. This analysis takes into account the effect of distortion in radial neutron flux distribution in the substituted core without any iterative correction procedure that is usually adopted in the first-order perturbation method. Also, it is not necessary in the case of the present analysis to introduce any usual intermediate region for taking into account the effect of spectrum mismatch between the two lattices. The material buckling differences between the test and reference lattices, which are in the range of −10.2 to 9.1 m−2, were determined within 3% of uncertainty.

Reactivity Worths of Annular Control Rods in a Pressure-Tube-Type Heavy Water Lattice

The control rod effect has been experimentally studied in the Deuterium Critical Assembly (DCA) by using annular absorbers that simulate control rods of the FUGEN reactor, a prototype heavy-water-moderated, boiling-light-water-cooled, pressure-tube-type reactor. The DCA cores for this experiment are of the 1.2 percent-/sup 235/U-enriched UO/sub 2/ lattices, and consist of 28-pin fuel clusters arranged in a square array of 22.5-cm lattice pitch. The experiment has been carried out with various control rod patterns and with varying coolant void fraction. Experimental results were analyzed by the absorption method, which was employed in the FUGEN control rod design calculations. The calculated reactivity worth agreed with the experiment within +-10 percent. The calculations somewhat overestimated the absorber worths in the nonvoided core and underestimated them in the voided core. This tendency was found to be greatly improved by considering the anisotropy effect in the migration area of the cluster lattice. The experimental results were also analyzed by the logarithmic derivative method. This method more poorly predicted the worths, but described better the flux shape around the rods.

Lattice Parameter Measurements on Cluster-Type Fuel for Advanced Thermal Reactor

Abstract The behavior of neutrons in a highly heterogeneous unit cell consisting of D2O moderator, H2O coolant and a 28-pin fuel cluster contained in a pressure tube has been studied through lattice parameter measurements covering three different 235U enrichments, four coolant void fractions and two lattice pitches. A single-region core configuration was adopted, with which measurements were made to determine—in relation to coolant void fraction—the critical D2O level, as well as various lattice parameters A strong dependence on coolant void was observed for the critical level and the lattice parameters, in the case of the smaller 22.5 cm pitch lattice, due to the positive effect on core reactivity exerted by the slowing-down faculty of H2O in the epithermal energy region. With the larger 25.0 cm pitch, however, no meaningful dependence on void fraction was shown by any of the measured values, and this was ascribed to a compensating negative effect due to enhanced thermal neutron self-shielding in the fuel region produced by the H2O coolant. The results of cell calculations obtained by means of the METHUSELAH-II code showed generally good agreement with experimentally determined data, for both critical D2O levels and lattice parameters, in the case of coolant-filled lattices (0, 30 and 70% void fractions). For lattices devoid of coolant (100% void fraction), however, discrepancies in lattice parameters—particularly in p 28—produced corresponding deviations in core reactivity amounting to 1% in excess of those incurred with other void fractions.

Lattice Parameter Measurements on Cluster-Type Fuel for Advanced Thermal Reactor

Abstract The behavior of neutrons in a highly heterogeneous unit cell consisting of D2O moderator, H2O coolant and a 28-pin fuel cluster contained in a pressure tube has been studied through lattice parameter measurements covering three different 235U enrichments, four coolant void fractions and two lattice pitches. A single-region core configuration was adopted, with which measurements were made to determine—in relation to coolant void fraction—the critical D2O level, as well as various lattice parameters A strong dependence on coolant void was observed for the critical level and the lattice parameters, in the case of the smaller 22.5 cm pitch lattice, due to the positive effect on core reactivity exerted by the slowing-down faculty of H2O in the epithermal energy region. With the larger 25.0 cm pitch, however, no meaningful dependence on void fraction was shown by any of the measured values, and this was ascribed to a compensating negative effect due to enhanced thermal neutron self-shielding in the fue...