Void Reactivity Coefficient Research Articles

Void reactivity aspect and fuel conversion potential of heavy water cooled thorium reactor

Design study on heavy water cooled thorium breeding reactor has been investigated by adopting a nuclear equilibrium state model. Conversion ratio, as an important index, has been evaluated to estimate the breeding capability of the reactors. Void reactivity coefficient has also been investigated to evaluate performance index of safety aspect, which is based on the criticality performance of the reactors during voided condition. In addition, moderator-to-fuel ratio has also been employed to analyze its effect to the required enrichment, conversion ratio, and void reactivity coefficient as well as different burnups and fuel pin diameter effects. Void reactivity coefficient indicates a criticality condition of the reactor when some coolants are lost. If the negative value of void reactivity is achieved, it means that the reactor has less reactivity condition as well as less power production when lost of coolant occurred. Higher fuel conversion capability, that more nuclear fuel are produced, those additional fuel productions can be used for next operation or for other reactors. The results show that higher fuel conversion ratio can be achieved for less moderator-to-fuel ratio because of the harder neutron spectrum effect, while it requires more fissile content of 233U to maintain the reactor operation from fission reaction. Higher burnup gives less conversion ratio because some fissile materials are used to maintain longer reactor operation, and at the same time, it requires more initial required fissile 233U for higher burnup. In addition, it requires less fissile 233U for thicker fuel pin diameter, while its conversion ratio becomes higher, and void reactivity coefficient is more negative for thicker fuel pin diameter. The results also show that thorium utilization on heavy water cooled reactor gives all negative void reactivity values, which means that the system has a safety condition in terms of void reactivity condition. At the same time, it shows some feasible conditions for obtaining fuel breeding to increase the sustainability of nuclear fuel. Copyright © 2016 John Wiley & Sons, Ltd.

The influence of void-reactivity feedback on the bifurcation phenomena and nonlinear characteristics of a single nuclear-coupled boiling channel

This study employs the nonlinear dynamic model developed previously by the authors to investigate the effect of void-reactivity feedback on the bifurcation phenomena and nonlinear characteristics of a single nuclear-coupled boiling channel in an advanced boiling water reactor. The parametric effects on the system stability are evaluated by nonlinear analyses as well. The effects of void-reactivity feedback and outlet flow resistance will both destabilize the system, while the inlet flow resistance will generate a stable effect on the system. The strength of void-reactivity feedback has a great influence on the bifurcation phenomena and nonlinear dynamics of the single nuclear-coupled boiling channel. The system can evolve from stable state, limit cycle, periodic to chaotic oscillation through supercritical Hopf and period-doubled bifurcations as the increase in the absolute value of void-reactivity coefficient. In addition, the system can experience a series of period-doubled bifurcation to present various types of oscillations, from periodic ones to chaos, in the unstable region as the controlling parameter of steady phase change number or subcooling number under some specific conditions with a relatively strong void-reactivity coefficient of Cα=-0.38$/%. The appearance of a periodic cycle of three suggests that there may be immeasurable periodic types of nonlinear oscillations in the limited unstable space of this system.

The effect of external vertical acceleration on the dynamic behaviors of a single nuclear-coupled boiling channel

This study integrates a single nuclear-coupled boiling channel model previously developed by the authors with the external force method to illustrate the qualitative dynamic behaviors of a two-phase flow system subject to external vertical accelerations. With the normal operating condition of an ABWR as a reference case, the effect of external vertical acceleration on the system transients may depend on their imposed amplitudes and frequencies. Periodic and quasi-periodic oscillations can appear in the system as the frequency ratios between two vertical waves. The parametric effects on the seismic-induced oscillations are performed in the present system without nuclear coupling. The resonance oscillations could be triggered if the external vibration frequency is equal to the system natural frequency. The strength of resonance effect may depend on the inherent stability characteristics of the initial states for both the thermal-hydraulic and nuclear-coupled boiling systems. Complex seismic-induced oscillations can be predicted in the nuclear-coupled boiling system with strong void-reactivity feedback. The seismic-induced oscillations can experience various types of oscillations, from periodic ones to chaos, as the inlet subcooling is increased with the reference void-reactivity coefficient being doubled.

Assessment of Contributions by Void Reactivity Coefficient and End Effect of Control Rods into ChNPP-4 Accident Progression

The paper analyzes physical peculiarities of RBMK leading to catastrophic progression of the accident at ChNPP-4 on 26 April 1986. The modeled transient, which preceded the accident, and modeling of its first phase showed that introduction of absorbing rods into the core played a decisive role in the accident progression.

Analysis of Void Reactivity Coefficient for 3D BWR Assembly Model

The effect of BWR fuel assembly 3D model on void reactivity coefficient (VRC) estimation is investigated. VRC values were calculated for different BWR assembly models applying deterministic T-NEWT and Monte Carlo KENO-VI functional modules of SCALE 6.1 code package. The difference between deterministic T-NEWT and Monte Carlo KENO-VI simulations is negligible (0.18 pcm/%). The influence of the assumed more detailed coolant density profile was estimated as well. VRC increases with the application of a larger number of coolant density values across fuel assembly height. It was shown that the coolant density profile described by 6 values per height could be considered sufficient from prospect of VRC estimation, as a more detailed density profile has impact below 1% on total assembly void effects. VRC values were decomposed to values for individual nodes and isotopes, since decomposition provides useful insights to describe the overall behaviour of VRC in detail.

Analysis of Local Void Coefficients of Reactivity in the Reduced-Moderation BWR

An expression is derived for attributing the reactivity response due to perturbations to spectral, spatial, and isotopic effects. It is shown to be consistent at a global level with similar expressions derived in previous work but can provide more detailed information on the physics phenomena contributing to the reactivity response of the perturbation. Using this expression, the reactivity effect of local coolant density perturbations [local void coefficient of reactivity (VCR)] is studied for two reduced-moderation boiling water reactor (RBWR) core designs—the thorium-fueled RBWR (RBWR-Th) and the uranium-fueled RBWR (RBWR-AC)—as well as for a standard advanced boiling water reactor (ABWR). The RBWR core designs feature large axial variation in their neutron spectra.The axial distribution of local VCR along the RBWR-Th seed and along the ABWR core were found to have the same general shape: negative throughout but most negative near the bottom and asymptotically approaching zero toward the top. However, the RBWR-Th VCR is roughly four times more negative. The RBWR-AC local VCR axial distribution varies greatly: it is very close to zero in the seed regions and has a significant positive component in the central blanket.Three effects were identified as contributing to the VCR due to a local water density change in the lower part of the RBWR-Th seed: local spectrum hardening that tends to increase the local reproduction factor (ηr) of each of the fuel isotopes; a redistribution of the local neutron absorption between the fuel isotopes resulting in a shift of absorptions from higher to lower isotopic reproduction factors and, hence, to a reactivity loss; and an axial flux tilt across the core from axial zones of higher ηr to axial zones of lower ηr, which makes another negative contribution to the reactivity worth of the perturbation.

Experimental study of neutronic parameters in Tehran research reactor mixed-core

In this work, general characteristics of a typical mixed core, including HEU & LEU fuel is studied. The study is performed in the Tehran research reactor (TRR). In this study the neutronic parameters, reactivity feedback coefficients and kinetic parameters are investigated. The reference core designated for such study is the equilibrium core (No. 61) with an average bun-up of 27% & 36% for SFE's & CFE's, respectively. The MTR_PC package is used for neutronic analysis. In this research, experimental and computational results for the reference and mixed core are compared. Meantime, the obtained values for neutronic parameters are mostly below the adopted safety criteria and they are in good agreement with the experimental results. However βeff and ℓp are a little bit higher in the mixed core with respect to the reference core, but in practice, these small changes will not cause substantial impacts on the dynamic behaviour of the reactor core. The absolute values of the fuel temperature, moderator density and void coefficients of reactivity, are less in the mixed core and only the moderator temperature coefficient is higher. The calculated values of power defect, based on the reactivity coefficients; in both core configurations are in good agreement with the experimental values.

Self-sustaining Thorium-fueled Reduced Moderation BWR Feasibility Study

Abstract This study assesses the feasibility of a thorium-based fuel-self-sustaining Reduced moderation BWR (RBWR-Th) core. This core features a hexagonal tight-lattice fuel, high exit coolant quality, axial segregation of seed and blanket regions, and compatibility with the ABWR reactor vessel inherited from the RBWR-AC proposed by Hitachi. The RBWR-Th differs from the RBWR-AC by eliminating the internal blanket, eliminating absorbers from the axial reflectors, replacing depleted urania with thoria as the primary fertile fuel and elongating the fissile region. When the Hitachi assumptions and correlations for thermal-hydraulic analysis are used, the RBWR-Th obtains an average core discharge burnup of 61 GWd/t versus 45 GWd/t for the RBWR-AC. However, when more conservative thermal-hydraulic correlations are used, the average discharge burnup drops to 25 GWd/t. The RBWR-Th maintains negative coolant void coefficients of reactivity throughout the cycle – between -155 and -131 pcm/% – but currently does not have sufficient margin for cold shutdown. Future studies will address this shutdown margin deficit and will account for coolant pressure drop constraint.

Neutronics evaluation of a super-deep-burn with TRU Fully Ceramic Microencapsulated (FCM) fuel in CANDU

One of the potential approaches for dealing with spent nuclear fuel in order to ensure the sustainable development of the nuclear energy is to fully utilize (transmute) the transuranics (TRU) in nuclear reactors. The CANDU reactor has an extremely high neutron economy which is necessary for an efficient TRU burning. The current CANDU fuel elements are unable to achieve the high burnup required to transmute the TRUs. In this paper, a special nuclear fuel called FCM (Fully Ceramic Microencapsulated) is introduced to achieve a super-deep-burn of the TRUs and to tackle the issue of fuel element integrity. In the FCM fuel, the conventional TRISO fuel particles are randomly dispersed in a fully dense SiC matrix and the FCM fuel is expected to be able to accommodate an extremely high burnup. The core performances and characteristics were analysed from the neutronics perspective through the fuel lattice analysis. The fuel depletion calculations were conducted using the Serpent Monte Carlo code and the fuel discharge burnup was determined based on the non-linear reactivity theory. Kinetic and major safety parameters including the fuel temperature coefficient, void reactivity, and coolant temperature coefficient were also evaluated.

The use of zirconium hydride blankets in a minor actinide/thorium burner sodium-cooled reactor for void coefficient control with particular reference to UK's plutonium disposition problem

The use of zirconium hydride (Th–ZrH1.6) blankets in a thorium-fuelled sodium-cooled reactor for void reactivity control with particular reference to UK's plutonium disposition problem is proposed and considered. It is shown that, with the use of such blankets, a mild moderation effect is produced during voiding which compensates for the general hardening of the spectrum, enabling a net negative void coefficient at pin level to be attained without the need to rely on traditional neutron leakage enhancement techniques or neutron poisons, and with negligible impact on transmutation capabilities. One important difference in comparison with the traditional methods is that the void coefficient is obtained at the pin level, eliminating or mitigating substantially the spatial dependencies on the location of the void. Combining the use of such blankets with a suitable n-batch fuelling scheme yields a negative void reactivity coefficient throughout the life of fuel. Additional research and development are required to explore further this concept's potential.

Nonlinear dynamic analysis of a two-phase natural circulation loop with multiple nuclear-coupled boiling channels

This study integrates the models developed previously by the authors to explore the stabilities, nonlinear dynamics and oscillation modes of a two-phase natural circulation loop with multiple nuclear-coupled boiling channels. The results indicate that both the pure thermal-hydraulic and nuclear-coupled boiling systems indeed have two instability regions, i.e. type-I and type-II instabilities, respectively. The pure thermal-hydraulic system tends to present in-phase mode of oscillations at the type-I boundary states and, in general, out-of-phase oscillations along with the type-II stability boundary. The oscillation modes may be affected by the configuration of parallel channels. By introducing the void-reactivity feedback together with neutron interaction, the coupling thermal-hydraulic and nuclear effects would induce complex influences on the system stability as well as the nonlinear oscillation modes. The in-phase mode, instead of out-of-phase mode, majorly dominates over the type-II boundary as the neutronic feedback is increased through void-reactivity coefficient. The complex nonlinear phenomena, such as complex periodic and chaotic oscillations, may appear in this multi-channel nuclear-coupled boiling natural circulation loop subject to a strong void-reactivity feedback (3Cα) coupled with a weak subcore-to-subcore neutron interaction (ɛij=7.0).

Study on the recently proposed direct S–CO2 and helium-cooled GCFR: Size of vessel, power conversion, and reactivity coefficients

This article contains a review on the history, progress, and objectives of the GCFR as well as a review on the newly proposed TID-type and block-type fuels of a GCFR, together with a research on their selected physical parameters. We have roughly estimated and compared size of core vessel of a suggested GCFR for S–CO2 and helium cooled as well as power conversion system efficiencies for a direct super-critical CO2 cooled cores. Moreover, the Coolant Void Reactivity (CVR) and Doppler coefficient for a reference GCFR with two different coolant types (supercritical CO2 and helium) and either different matrix materials (blended: fuel + diluents) in the block type core is calculated. MCNP 5.0 was used for our neutronic simulation and MAKXSF module was used to Doppler broaden the cross sections. Finally, the burn up comparison for two coolants and also for suggested diluents is also performed herein.

Void Reactivity Coefficient Analysis during Void Fraction Changes in Innovative BWR Assemblies

The study of the void reactivity variation in innovative BWR fuel assemblies is presented in this paper. The innovative assemblies are loaded with high enrichment fresh UO2and MOX fuels. UO2fuel enrichment is increased above existing design limitations for LWR fuels (>5%). MOX fuel enrichment with fissile Pu content is established to achieve the same burnup level as that of high enrichment UO2fuel. For the numerical analysis, the TRITON functional module of SCALE 6.1 code with the 238-group ENDF/B-VI cross section data library was applied. The investigation of the void reactivity feedback is performed in the entire 0–100% void fraction range. Higher values of void reactivity coefficient for assembly loaded with MOX fuel are found in comparison with values for assembly loaded with UO2fuel. Moreover, coefficient values for MOX fuel are positive over 75% void fraction. The variation of the void reactivity coefficient is explained by the results of the decomposition analysis based on four-factor formula and neutron absorption reactions for main isotopes. Additionally, the impact of the moderation enhancement on the void reactivity coefficient was investigated for the innovative assembly with MOX fuel.

Primary Breeding Ratio Analysis of an Improved Supercritical Water Cooled Fast Reactor

The purpose of the study is to analyze the breeding ratio of a supercritical water cooled fast reactor (SCFR) and to increase the breeding core of SCFR. The sensitivities of assembly parameters, core arrangements and fuel nuclide components to the breeding ratio are analyzed. In assembly parameters, the seed fuel rod diameter has higher sensitivities to the conversion ratio (CR) than the coolant tube diameter in blanket. Increasing heavy metal fraction is good to CR improvement. The CR of SCFR also increases with a reasonable core arrangement and Pu isotope mass fraction reduction in fuel, which can achieve more negative coolant void reactivity coefficient at the same time. The breeding ratio of SCFR is 1.03128 with a new core arrangement. And the coolant void reactivity coefficient is negative, which achieves a fuel breeding in initial fuel cycle.

Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B4C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes.In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly.The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

Nuclear Data Uncertainty Propagation to Reactivity Coefficients of a Sodium Fast Reactor

The assessment of the uncertainty levels on the design and safety parameters for the innovative European Sodium Fast Reactor (ESFR) is mandatory. Some of these relevant safety quantities are the Doppler and void reactivity coefficients, whose uncertainties are quantified. Besides, the nuclear reaction data where an improvement will certainly benefit the design accuracy are identified. This work has been performed with the SCALE 6.1 codes suite and its multigroups cross sections library based on ENDF/B-VII.0 evaluation.

Improvements of two-pass core design for super fast reactor

Improvements of two-pass core design of Super critical-water cooled fast reactor (Super FR) are proposed. The core height has been shortened; assembly pitch is enlarged to take Control Rod (CR) design into account. Well-coupled fuel loading pattern is proposed, number of control rods location are limited only in second flow pass assemblies due to current flow pattern. Fuel shuffling is also confined within each own flow pass because of the different structure of assemblies. CR withdrawal is considered during the operation, satisfactory radial and axial power distribution can be achieved by implementing appropriate CR withdrawal strategy. Results suggest that all the design target, criteria and limits are satisfied in terms of average coolant outlet temperature, maximum linear heat generation rate (MLHGR), maximum cladding surface temperature (MCST) as well as core shutdown margin, negative coolant void reactivity coefficient and positive coolant density reactivity coefficient in all coolant density range.

On the Use of Reduced-Moderation LWRs for Transuranic Isotope Burning in Thorium Fuel—II: Core Analysis

Multiple recycle of transuranic (TRU) isotopes in thermal reactors results in a degradation of the plutonium (Pu) fissile quality with buildup of higher actinides (e.g., Am, Cm, Cf), some of which are thermal absorbers. These phenomena lead to increasing amounts of Pu feed being required to sustain criticality and accordingly larger TRU content in the multirecycled fuel inventory, ultimately resulting in a positive moderator temperature coefficient (MTC) and void reactivity coefficient (VC). Because of the favorable impact fostered by use of thorium (Th) on these coefficients, the feasibility of Th-TRU multiple recycle in reduced-moderation (RM) pressurized water reactors (PWRs) and RM boiling water reactors (called RMPWRs and RBWRs, respectively) has been investigated. In this paper, Part II of two companion papers, the results of the single-assembly analyses of Part I are developed to investigate full-core feasibility. A large reduction in moderation is necessary to allow full actinide recycle. This increases the core pressure drop, which poses some thermal-hydraulic challenges, which are more pronounced if the design implementation is through retrofitting an existing PWR. For a given reactor cooling pump, the core flow rate is reduced. Despite this, it is possible to achieve feasible inlet and outlet temperatures and minimum departure from nucleate boiling ratio, for the reduction in moderation considered here. Reflood after loss-of-coolant accident is expected to be slower, which may lead to unacceptable peak clad temperatures and/or clad oxidation. Equilibrium cycles are presented for the RMPWR and RBWR, with a negative MTC and VC. However, the RMPWR may have positive reactivity when fully voided, and the hard spectrum makes it difficult to achieve an adequate shutdown margin, such that for the considered fuel designs, additional rod banks would be required.

On the Use of Reduced-Moderation LWRs for Transuranic Isotope Burning in Thorium Fuel—I: Assembly Analysis

Multiple recycle of transuranic (TRU) isotopes in thermal reactors results in degradation of the plutonium (Pu) fissile quality with buildup of higher actinides (e.g., Am, Cm, Cf), some of which are thermal absorbers. These phenomena lead to increasing amounts of Pu feed being required to sustain criticality and accordingly larger TRU content in the multirecycled fuel inventory, ultimately resulting in a positive moderator temperature coefficient (MTC) and void reactivity coefficient. Because of the favorable impact fostered by use of thorium (Th) on these coefficients, the feasibility of Th-TRU multiple recycle in reduced-moderation pressurized water reactors (PWRs) and boiling water reactors (BWRs) has been investigated. In this paper, Part I of two companion papers, the analysis is limited to a single assembly, with full-core models presented in Part II. Spatial separation of TRU from bred uranium is found to greatly improve neutronic performance. A large reduction in moderation is necessary to allow full actinide recycle. This will pose thermal-hydraulic challenges, which are discussed in Part II. In addition, the harder neutron spectrum resulting from the reduced moderation also reduces the control rod worth, while there is a neutronic incentive to use increased mechanical shim to maintain a negative MTC. It may therefore be desirable to increase the number of rod cluster control assemblies. Superior burnup is achievable in a reduced-moderation BWR as a larger reduction in moderation is feasible, although the incineration rate is reduced relative to a PWR due to a higher conversion ratio.

Decomposition Analysis of Void Reactivity Coefficient for Innovative and Modified BWR Assemblies

The decomposition analysis of void reactivity coefficient for innovative BWR assemblies is presented in this paper. The innovative assemblies were loaded with high enrichment UO2and MOX fuels. Additionally the impact of the moderation enhancement on the void reactivity coefficient through a full fuel burnup discharge interval was investigated for the innovative assembly with MOX fuel. For the numerical analysis the TRITON functional module of SCALE code with ENDF/B-VI cross section library was applied. The obtained results indicate the influence of the most important isotopes to the void reactivity behaviour over a fuel burnup interval of 70 GWd/t for both UO2and MOX fuels. From the neutronic safety concern positive void reactivity coefficient values are observed for MOX fuel at the beginning of the fuel irradiation cycle. For extra-moderated assembly designs, implementing 8 and 12 water holes, the neutron spectrum softening is achieved and consequently the lower void reactivity values. Variations in void reactivity coefficient values are explained by fulfilled decomposition analysis based on neutrons absorption reactions for separate isotopes.