Moderator Temperature Research Articles

STUDY ON CRITICALITY AND NEUTRONIC SAFETY PARAMETERS OF NUSCALE FUEL ASSEMBLY

STUDY ON CRITICALITY AND NEUTRONIC SAFETY PARAMETERS OF NUSCALE FUEL ASSEMBLY. NuScale, a typical Pressurized Water Reactor (PWR) Small Modular Reactors (SMRs), offers a new opportunity for the future of nuclear industry. With 160 MW thermal power, NuScale has several advantages such as flexibility due to its modularity in construction. This work is focused on the study of criticality and neutronic safety parameters of NuScale fuel assembly using MCNP6 code and ENDF/B-VII library. The calculation results shows that criticality of fuel assembly type D is the highest among other assembly types because it has a fuel pin with pure UO2 without Gd2O3 concentration. The Doppler temperature coefficient (DTC) of fuel assembly type C is the most negative among other assemblies due to Doppler broadening effect on resonance region of capture cross section of 238U which is the highest concentration. The moderator temperature coefficient (MTC) of fuel assembly type D is the most negative among the other assembly types. The effective delayed neutron fraction (βeff) does not reflect a consistent trend among fuel assembly types. Fuel assembly type D shows the highest prompt neutron lifetime (ℓ) while the highest neutron generation time (Ʌ) is shown in assembly type C. It can be concluded that this study provides adequate results that can be used as a first step to carry out the neutronic computation and analysis of the NuScale full core.Keywords: Criticality, safety parameters, NuScale fuel assembly, MCNP6, ENDF/B-VII.

Study on the Concept of Organic-Cooled Microreactor

In order to study the concept of organic-cooled reactor, this paper firstly analyzes the important characteristics and key technical problems of the organic fluid used as the reactor coolant and moderator, as well as the main technical schemes for organic-cooled reactors. On this basis, this paper studies the neutronic characteristics of the 5 MW microreactor core. As shown in the study, under the same core layout conditions, the absolute value of the moderator temperature coefficient of the organic-cooled reactor is below that of the pressurized water reactor (PWR), and the power distribution is more flat. The study results from this paper can provide an important reference for the determination of the technology roadmap for multi-purpose microreactor power supplies and heating systems, which are rapidly developed in China.

Enhancement of sub-thermal neutron flux through cold polyethylene

Total thermal neutron cross section measurements serve as the primary means of validation for thermal neutron scattering kernels, an important quantity for neutron transport calculations. In an effort to improve the quality of thermal neutron scattering kernels, researchers at Rensselaer Polytechnic Institute (RPI) designed and constructed a polyethylene based cold moderation system to enhance neutron flux below 10 meV when coupled with the Enhanced Thermal Target (ETT) at the RPI Gaerttner LINAC. The final design yielded an increase in sub-thermal neutron flux (below 10 meV) by a factor of 4.5 for a moderator temperature of 37.5 K relative to the ETT alone. A further increase to a factor of 6 is expected after a minor geometry modification and decrease in polyethylene temperature to 25 K. This novel capability will be used to conduct total thermal neutron cross section measurements from 0.0005–10 eV for different materials including moderator materials.

Core calculations for small modular reactor during burnup cycle

ABSTRACT Nuclear, modular water (AVB) reactor is a small modular reactor with special capabilities and it can be used as a source of energy for various applications. In this research, the core calculations of this reactor have been performed during the burnup cycle with deterministic and Monte Carlo approaches by the computational codes WIMSD, PARCS, and MCNPX. Due to the temperature and pressure limits for coolant in this reactor, the PMAXS generator and the developed PARCS code, PARCS-MOD, have been used in a deterministic approach. Quantities include the effective multiplication factor, the relative power distribution, the fuel, and moderator temperature distribution, reactivity feedbacks, and concentrations of some isotopes during the cycle are calculated. In addition, to comparing the results of the two approaches, some general results were compared with the results published in the documents, which we see an acceptable difference in this regard. The maximum value of for states BOC, MOC and EOC are −1.39, −1.85, and −1.87 (pcm.K−1), respectively. The relative difference between the calculated relative power is less than 2.6%. The relative difference between the coolant average temperature and the fuel average temperature calculated with the reference values is 0.21%, and 0.14%, respectively. The maximum axial coolant density is 650 (kg.m−3). Also in this research, the dead time of the ABV reactor has been calculated which is zero.

Convergence analysis of fixed-point iteration with Anderson Acceleration on a simplified neutronics/thermal-hydraulics system

In-depth convergence analyses for neutronics/thermal-hydraulics (T/H) coupled calculations are performed to investigate the performance of nonlinear methods based on the Fixed-Point Iteration (FPI). A simplified neutronics-T/H coupled system consisting of a single fuel pin is derived to provide a testbed. The xenon equilibrium model is considered to investigate its impact during the nonlinear iteration. A problem set is organized to have a thousand different fuel temperature coefficients (FTC) and moderator temperature coefficients (MTC). The problem set is solved by the Jacobi and Gauss-Seidel (G-S) type FPI. The relaxation scheme and the Anderson acceleration are applied to improve the convergence rate of FPI. The performances of solution schemes are evaluated by comparing the number of iterations and the error reduction behavior. From those numerical investigations, it is demonstrated that the number of FPIs is increased as the feedback is stronger regardless of its sign. In addition, the Jacobi type FPIs generally shows a slower convergence rate than the G-S type FPI. It also turns out that the xenon equilibrium model can cause numerical instability for certain conditions. Lastly, it is figured out that the Anderson acceleration can effectively improve the convergence behaviors of FPI, compared to the conventional relaxation scheme.

Neutronic characteristics and feasibility analysis of micro-heterogeneous duplex ThO2-UO2 fuel pin in PWR

The shortage of natural uranium resources, nuclear waste disposal and nuclear proliferation issues are important constraints to the development of nuclear energy. The development and utilization of thorium fuel is an effective way to solve these problems. This paper analyses the neutron physical properties of micro-heterogeneous duplex fuel and compares them with UO2 fuel in terms of burnup depth, reactivity coefficients and radioactive waste. Additionally, the best grid of duplex fuel is explored as well.Results from this study show that the ThO2-UO2 duplex fuel can match the discharge burnup of UO2 fuel with the 235U enrichment greater than 5% but less than 20%. The reactivity was calculated at different levels of thorium dioxide fuel, and the rate of decrease in reactivity is lower for duplex fuel than for uranium dioxide fuel. Fuel temperature coefficient (FTC) and moderator temperature coefficient (MTC) are within the acceptable range specified for pressurized water reactors. Production of both weapons-grade plutonium and minor actinides are greatly reduced and the effect of fission nuclides breeding is significant. When the volume ratio of moderator to fuel is 3.27, the duplex fuel performs the best in burnup and the cycle burnup is greatly increased. It also has good performance on safety and nonproliferation issues under such grid condition.Physical calculations of assembly size were performed for 20% ThO2 duplex fuel at a volume ratio of moderator to fuel of 3.27 and compared them with the UO2 assembly calculations. Duplex fuel has higher cycle burnup and economy. Moreover, The negative FTC and MTC and the high control rod value of duplex fuel enable the reactor to be safe. Duplex fuel has a lower power peaking factor and a more uniform power distribution which are good for reactor operation.

Assessment of the burnup characteristics of UO2 and MOX fuel in the mixed solid and annular rod configuration

Modeling and burnup analysis of Uranium dioxide (UO2) and Mixed Oxide (MOX) fuel in square assemblies comprising both solid and annular rods was carried out. The lattice pitch of the models was optimized first in terms of infinite multiplication factor (k∞) using OpenMC. The burnup analysis was performed at the optimized pitch by lattice physics code Dragon for 1620 EFPD. The burnup dependent fission rate density was studied, and the thermal fission rate was found higher for the models as compared to reference designs. The values of k∞ and reactivity were determined and higher values were found for the models at low fuel burnup period. However, models showed a higher criticality period than the reference design consisting of only the annular rod. The fuel and moderator temperature sensitivity of reactivity was investigated for different fuel burnup periods. The burnup-dependent atomic concentrations of fissionable materials, minor actinides, and neutron poisons were also studied. The production of neutron poison, namely Xe-135 and Sm-149 and minor actinides viz. Np-237, Am-241, Cm-242 and Cm-244 were found lower in models as compared to the reference designs.

Truly-optimized PWR lattice for innovative soluble-boron-free small modular reactor

A novel re-optimization of fuel assembly and new innovative burnable absorber (BA) concepts are investigated in this paper to pursue a high-performance soluble-boron-free (SBF) small modular reactor (SMR), named autonomous transportable on-demand reactor module (ATOM). A truly optimized PWR (TOP) lattice concept has been introduced to maximize the neutron economy while enhancing the inherent safety of an SBF pressurized water reactor. For an SBF SMR design, the 3-D centrally-shielded BA (CSBA) design is utilized and another innovative 3-D BA called disk-type BA (DiBA) is proposed in this study. Both CSBA and DiBA designs are investigated in terms of material, spatial self-shielding effects, and thermo-mechanical properties. A low-leakage two-batch fuel management is optimized for both conventional and TOP-based SBF ATOM cores. A combination of CSBA and DiBA is introduced to achieve a very small reactivity swing (< 1000 pcm) as well as a long cycle length and high fuel burnup. For the SBF ATOM core, safety parameters are evaluated and the moderator temperature coefficient is shown to remain sufficiently and similarly negative throughout the whole cycle. It is demonstrated that the small excess reactivity can be well managed by mechanical shim rods with a marginal increase in the local power peaking, and a cold-zero shutdown is possible with a pseudo checker-board control rod pattern. In addition, a thermal–hydraulic-coupled neutronic analysis of the ATOM core is discussed.

Safety analysis TMSR-500 in terms of the temperature reactivity coefficient of the fuel and the moderator using the MCNP6 software

TMSR-500 is a type of fourth-generation reactor that is technologically advanced with a full actinide recycling system and produces low fuel waste. This research focuses on calculating and analyzing the temperature reactivity coefficient of fuel and moderator in the TMSR-500 core designed by Martingale, Inc. All calculations performed using the MCNP6 software and the ENDF / B-VII.I continuous energy nuclear data library. The fuel material for TMSR-500 is liquid salt NaF-BeF2-ThF4-UF4 with enrichment of U-235 of 19.75%, and the moderator material is graphite. The variation of temperature increase in fuel and moderator starts from 293.6 K to 2500 K. As the reactor temperature increases, the value of fuel reactivity decreased, the fuel temperature reactivity coefficient is negative (-2.4879 ± 0.42) pcm/K. At the stage of calculating the moderator temperature reactivity coefficient, the calculation results show a positive value (0.0019 ± 0.0016) pcm/K as the moderator temperature increases. Based on the study of the temperature reactivity coefficient of the fuel and the moderator, then the reactor is in a safe state in terms of the overall calculation results, which indicate the total temperature reactivity coefficient is negative (-2.4883 ± 0.42) pcm/K.

Neutronic Performance of the VVER-1000 Reactor Using Thorium Fuel with ENDF Library

In this paper, neutronic calculations and the core analysis of the VVER-1000 reactor were performed using MCNP6 code together with both ENDF/B-VII.1 and ENDF/B-VIII libraries. The effect of thorium introduction on the neutronic parameters of the VVER-1000 reactor was discussed. The reference core was initially filled with enriched uranium oxide fuel and then fueled with uranium-thorium fuel. The calculations determine the delayed neutron fraction βeff, the temperature reactivity coefficients, the fuel consumption, and the production of the transuranic elements during reactor operation. βeff and the Doppler coefficient (DC) are found to be in agreement with the design values. It is found that the core loaded with uranium and thorium has lower delayed neutron fraction than the uranium oxide core. The moderator temperature coefficients of the uranium-thorium core are found to be higher than those of the uranium core. Results indicated that thorium has lower production of minor actinides (MAs) and transuranic elements (mainly plutonium isotopes) compared with the relatively large amounts produced from the uranium-based fuel UO2.

Uncertainties of PWR spent nuclear fuel isotope inventory for back-end cycle analysis with STREAM/RAST-K

This paper presents a study of uncertainty quantification (UQ) of spent nuclear fuel isotope inventory. Takahama-3 realistic assays were used for calculation to quantify the impact of three different parameter groups, namely, design parameters (i.e., fuel pellet radius, clad outer radius, UO2 density, and UO2 enrichment), operation conditions (i.e., power density, fuel temperature, moderator temperature, and boron concentration), and covariance of nuclear data. The UQ study is focused on analysis of fission products and actinides used in criticality analysis during the back-end cycle. Stochastic sampling method was used for the UQ and Lagrange interpolation is used for the prediction of isotope inventory. In-house two step code system, STREAM/RAST-K was used for calculation with ENDF/B-VII.1 cross section library and ENDF/B-VII.0 decay library. Most of the actinides and fission products are calculated within ±5% of experiment. In addition, 238U and 239Pu have the largest contribution on keff uncertainty. Moreover, 245Cm and 246Cm have the largest isotope inventory uncertainty compared with other isotopes and the covariance of nuclear data has the most dominant effect on the uncertainty of isotope composition.

Bayesian method and polynomial chaos expansion based inverse uncertainty quantification of spent fuel using decay heat measurements

This work presents a new application of decay heat measurement based on the calibration and inverse uncertainty quantification (IUQ) of modeling parameters of pressurized water reactor (PWR) fuel assemblies. This work (i) solves the problem encountered in forward UQ i.e., the lack of fuel vendor proprietary information (manufacturing tolerances of fuel assembly design) and operating condition uncertainties which are based on ad-hoc expert judgement or personal opinion (ii) calibrates the parameters of a fuel assembly model for improved code calculation-to-experiment agreement. The IUQ is conducted under the Bayesian framework and finds the fuel assembly model parameters that are consistent with decay heat measurements. The forward model implements a polynomial chaos expansion (PCE) based surrogate model for a computationally efficient inverse analysis. The approach introduced is then tested with the decay heat calculations of the deterministic code STREAM. Eight input parameters: fuel density, enrichment, pellet radius, clad outer radius, fuel temperature, power density, moderator temperature and boron concentration, are considered. The outcomes of this work include quantification of uncertainties and determination of probability distribution function (PDF) of these parameters, in addition to reducing code calculation-to-experiment discrepancies. These outcomes are important for future studies on the forward propagation of model parameters uncertainties.

Full-core reactor physics analysis for accident tolerant cladding in a VVER-1000 reactor

Advanced accident tolerant cladding materials have brought up the potential to delay the deleterious consequences of loos of coolant accidents related to slowing down hydrogen formation from reaction of zirconium with steam in order to minimize the additional heat generation and improve fuel and cladding retention of fission products. The performance improvement offered by these advanced materials may expand the operating envelope of existing light water reactors. This paper examines the neutronic performance of the VVER-1000 light water reactor for the application of accident tolerant cladding in order to realize the endurance of severe accident conditions. This study includes a detailed analysis of the control rod worth, reactivity coefficient, fuel cycle length, and power distribution for three accident tolerant cladding candidates of Ferritic-based alloy (FeCrAl), silicon carbide (SiC), and chromium coating application on zirconium claddings (ZrCr). The analysis was performed using diffusion-based core code PARCS, and lattice physics code DRAGON, including a developed package for regeneration of cross-sections in PMAX format.The reactor performance analysis indicates that the SiC cladding would have improved performance in terms of fuel cycle length, fuel and moderator temperature reactivity coefficients, and control rod worth. The cycle length would significantly decrease in magnitude for FeCeAl. Therefore, decreasing the cladding thickness by half and increasing the fuel enrichment by factor of 1.1875 made it possible to satisfy the required cycle length. A higher enrichment is also necessary for ZrCr cladding to increase the fuel burnup limits at nominal operating conditions.

Study the neutronic feasibility of using Zr as an energy regulator instead of traditional methods

The control of the reactor energy increases the reactor fuel cycle time. In this work, a new method has been investigated to control the reactor energy instead of traditional methods. SERPENT2 version 2.1.30 was used to investigate the possibility of using Zirconium rods as an energy regulator in the VVER-1000. Zirconium rods were used as water displacers. The effect of different Zr rod diameter on the infinite multiplication factor (k∞) of the reactor and the fuel concentration at different fuel burnup steps has been analyzed. Larger fuel pitches have been investigated to increase the Zr rod diameter. The main safety parameters such as the void volume reactivity coefficients, the Doppler reactivity coefficients and the Moderator temperature coefficient have been studied at different fuel pitch and different Zr rod diameter. A comparison between the effect of Zr rods and boric acid on the reactor reactivity proved the efficiency of using Zr rods as reactivity regulation.

Diagnostic methods applied to Esfahan light water subcritical reactor (ELWSCR)

In this work, Esfahan light water subcritical reactor (ELWSCR) is analysed using experimental and theoretical diagnostic methods. Important neutronic parameters of the system such as prompt neutron lifetime, delayed neutron fraction, prompt neutron decay constant, negative reactivity of the core, fuel and moderator temperature coefficient of reactivity, and overall and local void coefficient of reactivity are estimated. Also, neutron flux distribution, reflector saving, water level effect, and lattice pitch of the core including operating point of the facility are studied in details. Theoretical results are calculated by MCNPX and measurements are performed utilizing zero power reactor noise method. Detailed descriptions of the results are explained in the text.

Neutronic performance of fully ceramic microencapsulated of uranium oxycarbide and uranium nitride composite fuel in SMR

The existing commercial nuclear power plants (PWR and BWR) utilize the oxide fuels, i.e., UO2. This fuel selection is not questionable, where the safety and economy are the top priority in the nuclear industry. In this work, the potential advantages of microencapsulated fuels to System Integrated Modular Advanced Reactor “SMART” has been explored. The UN and UCO have been considered as the candidate fuel material for the SMART reactor. The detailed design and fuel assembly configurations with different fuel types of the SMART reactor have been investigated for criticality with depletion (burn up), fuel and moderator temperature coefficients, and power peaking factor using OpenMC. The fissile and fertile fuel elements with coated particles are embedded in graphite matrices. The results are presented with a review of attributes and potential benefits. The nitride fuel has an advantage of mechanical stability, enhanced thermal conductivity, and high fuel density compared to dioxide fuel (UO2). The melting point of standard fuel, UN, and UCO are similar. Thus, they have higher safety margins under NPP's operating conditions. Therefore, nitride (UN) and carbide (UCO) fuel types are more attractive than standard uranium oxides because they are safer and beneficial. The dioxide fuel (UO2) is considered as the reference and compared with candidate material. The neutronic assessment within SMART core specification examined the effective multiplication factor, thermal flux distribution, axial and radial power distribution, and power peaking factor at the beginning, and end of the fuel cycle.

Verification and validation of isotope inventory prediction for back-end cycle management using two-step method

This paper presents the verification and validation (V&V) of a calculation module for isotope inventory prediction to control the back-end cycle of spent nuclear fuel (SNF). The calculation method presented herein was implemented in a two-step code system of a lattice code STREAM and a nodal diffusion code RAST-K. STREAM generates a cross section and provides the number density information using branch/history depletion branch calculations, whereas RAST-K supplies the power history and three history indices (boron concentration, moderator temperature, and fuel temperature). As its primary feature, this method can directly consider three-dimensional core simulation conditions using history indices of the operating conditions. Therefore, this method reduces the computation time by avoiding a recalculation of the fuel depletion. The module for isotope inventory calculates the number densities using the Lagrange interpolation method and power history correction factors, which are applied to correct the effects of the decay and fission products generated at different power levels. To assess the reliability of the developed code system for back-end cycle analysis, validation study was performed with 58 measured samples of pressurized water reactor (PWR) SNF, and code-to-code comparison was conducted with STREAM-SNF, HELIOS-1.6 and SCALE 5.1. The V&V results presented that the developed code system can provide reasonable results with comparable confidence intervals. As a result, this paper successfully demonstrates that the isotope inventory prediction code system can be used for spent nuclear fuel analysis.

Some Considerations on the Energy Deposition During a RIA Transient Based On Monte Carlo Simulations

Specific research reactors are capable of reproducing reactivity injection accidents in order to study the behavior of the nuclear fuel pins in accidental situations. In the CABRI research reactor, the fuel pin to be examined (test pin) is placed in the center of the core in a dedicated test loop. It is then subjected to a power transient, obtained by the fast depressurization of the 3He neutron absorber gas from the transient rods located in the core. One of the central parameters of the experiment is the energy deposition in the test pin, which is currently not measured during a transient. Instead, it is assumed that the relative energy distribution between the core and the test pin is constant regardless the operational state of the reactor. Currently, this correlation is measured in steady state. As such, the impact of the variations in the neutron flux, fuel and moderator temperatures during the transient is assumed equivalent on the energy deposition in the core and in the test pin. The goal of this work is to improve our knowledge on the mechanisms involved in the transient energy deposition. The aim of this paper is to present a methodological approach for the energy deposition estimation during a CABRI transient, based on static Monte Carlo calculations. The results suggest that the transient energy deposition rate is mainly dependent on the helium pressure and the Doppler feedback, and the relative energy distribution between the core and test pin changes during the transient.

VERIFICATION AND VALIDATION OF BACK-END CYCLE SOURCE TERM CALCULATION OF THE NODAL CODE RAST-K

Verification and validation (V&amp;V) results of source term calculation capability implemented in the nodal diffusion code RAST-K are presented in this paper. An isotope inventory prediction method is presented in this work which is implemented with RAST-K and the lattice code STREAM. STREAM generates cross-section and provides number density information by history branch calculations. RAST-K supplies the power history and three history indexes (boron concentration, moderator temperature and fuel temperature). The main feature of the newly implemented spent nuclear fuel (SNF) characterization is the direct consideration of three-dimensional (3D) core simulation conditions by using operation history information. As a result of this, it could reduce the computation time. The implemented SNF analysis capability have two main functions. The first is to predict isotope inventory by Lagrange non-linear interpolation method, using power history correction factors. The second is to calculate the radiological response activity, decay heat, and neutron/gamma source strengths. The V&amp;V of these two functions are thus presented herein. The isotope inventory prediction is validated with measured data from ten SNF samples of Takahama-3 and six samples of Calvert Cliffs-1 pressurized water reactors (PWR). Eighteen decay heat measurements of Ringhals Unit 3 PWR fuel assemblies are then employed to validate the decay heat calculation results. In addition, STREAM is employed in a code-to-code comparison for verification. The fuel assemblies cover the burnup range 14.3 - 47.25 GWd/tU, initial enrichment of 2.1 - 4.11 235U w/o and cooling time of 3.96 to 20.01 years. The comparison to STREAM shows the accuracy of the RAST-K SNF and prediction of the decay heat is within 4%. Overall, this paper demonstrates that RAST-K SNF calculation can be applied to the back-end cycle source term analysis.

Analysis of thermal hydraulic behavior of KONVOI PWR during a design extension condition

In nuclear power plants the design extension conditions are more complex and severe than those postulated as design basis accidents, therefore, they must be taken into account in the safety analyses. In this study, many hypothetical investigated transients are applied on KONVOI pressurized water reactor during a 6 inches (182 cm2) cold leg small break loss-of-coolant-accident to revise the effects of all safety systems ways through their availability/non-availability on the thermal hydraulic behaviour of the reactor. The investigated transients are represented through three cases of small break loss-of-coolant-accident as, Case-1, without scram and all of the safety systems are a failure, Case-2, the normal scram actuation with a failure of all safety systems (non-availability), and finally Case-3, with normal actuation scram sequence and normal sequential actuation of all safety systems (availability). These three investigated transient cases are simulated by creating a model using analysis of thermal-hydraulics of leaks and transient code. In all transient cases, all types of reactivity feed- backs, boron, moderator density, moderator temperature, and fuel temperature are considered. The steady-state results are nearly in agreement with the plant parameters available in previous literature. The results show the importance of the reactivity feedback effects in loss-of-coolant-accident on the fallouts power as they are considered the key parameters for controlling the clad and fuel temperatures to maintain them below their melting point. Moreover, the calculated results in all cases show that the thermal-hydraulic parameters are in acceptable ranges and encounter the safety criterion during loss-of-coolant-accident design extension conditions accidents processes. Furthermore, the results show that the core uncover and fuel heat up do not occur in KONVOI pressurized water reactor the design extension conditions simulations, as all safety systems provide adequate core cooling by sufficient water inventory into the core to cover it.