Coolant Void 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.

Extension and preliminary validation of the Polaris lattice physics code for CANDU analysis

Polaris is a 2D lattice physics code available for the study of light water reactors (LWRs). It includes a method of characteristics (MoC) transport solver and the embedded self-shielding method (ESSM). Its advantages include a simplified user input structure, quick performance, and integration into the SCALE package, utilizing its nuclear data libraries, along with other codes such as the ORIGEN depletion code and the Sampler stochastic uncertainty analysis sequence. This work extends the Polaris code to support CANDU lattice transport calculations and performs some preliminary verification and validation, comparing the code to NEWT and KENO, along with to benchmark experimental data. Heavy-water specific self-shielding factors were also compared to those distributed with SCALE. In general, some differences are observed, but these are generally comparable to other code-to-code differences and smaller than nuclear data uncertainties. The generation of self-shielding factors has a noticeable effect, particularly on the benchmark coolant void reactivity calculation. Heavy-water specific self-shielding factors, while not required, can potentially improve the accuracy of ESSM-based calculations, but the choice of parameters can significantly affect the results. Overall, the Polaris transport and self-shielding calculations should be suitable for CANDU analysis.

THE EFFECTS OF TYPICAL THORIUM IMPURITIES ON THORIUM-BASED NUCLEAR FUELS

Natural thorium contains impurities of numerous isotopes. A study was performed to examine the influence of naturally occurring impurities in thorium-based fuels on a few parameters of interest such as: exit burnup, coolant void reactivity (CVR), fuel temperature coefficients (FTC), and the radiotoxicity of the used fuel. Two different fuel bundle designs were modeled: a 43-element bundle containing 2.25% U-233, and a 35-element bundle containing 1.45% U-233. Naturally occurring thorium fuel impurities were applied to both fuel bundle models at various concentrations, from 0% to 100% of the expected maximum. For burnup-averaged k-infinity (k∞) values of 1.050 and 1.030, exit burnup, burnup-weighted CVR, and burnup-weighted FTC were calculated using the neutron transport code WIMS-AECL, and plotted against fraction of full impurity concentration to determine how the impurity levels affect these reactor physics parameters of interest. For the most-realistic (for CANDU) burnup-averaged k∞ of 1.050, both the inhalation radiotoxicity and the production of U-232 were calculated using the fuel depletion code WOBI. Up to the maximum impurity concentrations considered, no effects on the investigated fuel performance parameters were found to be greater than a few percent.

Controller design for operation of a 700 MWe PHWR with limited voiding

Enhancement of available power output from a 540 MWe Pressurized Heavy Water Reactor (PHWR) with limited coolant voiding has been attempted in the recent past. Coolant voiding in such a core usually occurs in the high power regime and introduce positive reactivity in the core which makes a reactor increasingly unstable. This paper presents a scheme for a 700 MWe PHWR, which is capable of controlling the reactor from 60% to 92% of its Full Power (FP) using the Liquid Zone Control System (LZCS) and again from 92% to 100%FP in conjunction with a pressurizer which is activated to change the pressure in the coolant channel by a small amount, thus keeping the void fraction constraint. This paper presents a dynamic model for estimation of void fraction in the PHWR core as a function of reactor power and then proposes an optimal controller for pressurizer control using Genetic Algorithm (GA) to keep void fraction limited. The controller performance has been evaluated using credible Hardware-in-Loop (HiL) simulation.

Potential of FAST (floating absorber for safety at transient) as a solution for positive coolant temperature coefficient in sodium-cooled FAST reactors

Floating absorber for safety at transient (FAST) is a passive safety device for sodium-cooled fast reactors. It is designed to insert negative reactivity in case of coolant temperature rise or coolant voiding in an inherently passive way. This study investigates behavior of FAST device during the anticipated transients without scram (ATWS) in a sodium-cooled breed-and-burn fast reactor (B&BR). Unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS) and unprotected transient overpower (UTOP) scenarios are simulated at both MOL (middle of life) and EOL (end of life) conditions of the B&BR to investigate the impacts of FAST on the reactor safety. It is confirmed that FAST can very successfully mitigate the consequences of ULOF, ULOHS and UTOP even in the B&BR core with strongly positive coolant temperature coefficient. In addition, this paper provides key aspects of FAST design optimization for practical applications.

A preliminary comparative study between oxide and metallic fuelled ASTRID-like reactor under a B&B strategy

The objective of this paper is to study the reactivity behaviour and the isotopic fuel performance of an ASTRID-like reactor under a breed and burn strategy for two core designs: with oxide and with metallic fuel. The proposed reshuffling scheme was simulated with MCNP6 and the JEFF-3.2 cross-sections library to extend the fuel life by two more cycles. Our findings showed that the implementation of the reshuffling scheme enhanced the fuel utilisation, obtaining cycle extensions of 805 days and 1305 days, over the first 365 days, for the oxide and metallic fuelled designs, respectively. For the metallic-fuelled design the breeding of Pu-239 achieved a production of 577.7 kg, which represents a production rate of 126.3 kg/EFPY. The conversion rate for the metallic-fuelled design was 1.06 and 0.96 for the oxide-fuelled design. Regarding the coolant void reactivity worth, for the oxide-fuelled design it becomes positive after the second fuel reshuffling.

A preliminary comparative study between oxide and metallic fuelled ASTRID-like reactor under a B&B strategy

The objective of this paper is to study the reactivity behaviour and the isotopic fuel performance of an ASTRID-like reactor under a breed and burn strategy for two core designs: with oxide and with metallic fuel. The proposed reshuffling scheme was simulated with MCNP6 and the JEFF-3.2 cross-sections library to extend the fuel life by two more cycles. Our findings showed that the implementation of the reshuffling scheme enhanced the fuel utilisation, obtaining cycle extensions of 805 days and 1305 days, over the first 365 days, for the oxide and metallic fuelled designs, respectively. For the metallic-fuelled design the breeding of Pu-239 achieved a production of 577.7 kg, which represents a production rate of 126.3 kg/EFPY. The conversion rate for the metallic-fuelled design was 1.06 and 0.96 for the oxide-fuelled design. Regarding the coolant void reactivity worth, for the oxide-fuelled design it becomes positive after the second fuel reshuffling.

Study of the homogeneous and heterogeneous Am transmutation in an ELFR-like reactor loaded with nitride fuel

In this paper a neutronic study of the homogeneous and heterogeneous core configuration for americium transmutation in an ELFR-like reactor loaded with nitride fuel was performed by using the reactor physics code Serpent. For this, the initial Am content was range as follows: 1, 3, 5, 7 and 9 wt%, and some neutronic and safety parameters, such as the Doppler constant, the coolant void worth and the effective delayed neutron fraction were analyzed. Based on these parameters, the maximum allowable Am content at BoC was 7 wt% for the homogeneous core due to penalties in the coolant void worth. In the heterogeneous core no safety penalties were found but a lowest Am consumption was obtained compared to the homogeneous configuration due to changes in the neutron spectrum. In addition, the axial and the radial power profiles were obtained at three different burnup steps: BoC, MoC and EoC. The homogeneous configuration showed a quite smooth radial and axial power distributions, while an increment in the radial power form factor was observed at the center of the core in the heterogeneous fuel configuration, that deserves further studies.

Neutronic study of utilization of discrete thorium-uranium fuel pins in CANDU-6 reactor

Abstract Targeting at simulating the application of thorium-uranium (TU) fuel in the CANDU-6 reactor, this paper analyzes the process using the code DRAGON/DONJON where the discrete TU fuel pins are applied in the CANDU-6 reactor under the time-average equilibrium refueling. The results show that the coolant void reactivity of the assembly analyzed in this paper is lower than that of 37-element bundle cell with natural uranium and 37-element bundle cell with mixed TU fuel pins; that the max time-average channel/bundle power of the core meets the limits - less than 6700kW/860 kW; that the fuel conversion ratio is higher than that of the CANDU-6 reactor with natural uranium; and that the exit burnup increases to 13400 MWd/tU. Thus, the simulation in this paper with the fuel in the 37-element bundle cell using discrete TU fuel pins can be considered to be applied in CANDU-6 reactor with adequate modifications of the core structure and operating modes.

Physics characteristics of internally cooled annular fuel for potential application in pressure tube heavy water reactors

This paper summarizes the results of exploratory lattice physics studies of alternative, advanced fuel bundle concepts that could potentially be implemented in pressure tube heavy water reactors (PT-HWRs). The lattice physics code WIMS-AECL was used to analyze the physics performance and operational characteristics of an 18-element and a 12-element internally cooled annular fuel (ICAF) fuel bundle, made with (LEU,Th)O2 fuel, with both low-burnup and high-burnup options. Such fuel bundles with annular fuel elements may be able to operate at higher bundle power levels and with higher linear element (LER) ratings than fuel bundles with conventional solid cylindrical fuel elements. In addition, the use of thorium mixed with LEU can help extend uranium resources, exploit the energy potential in thorium, and also reduce the production of plutonium and minor actinides, due to the smaller fraction of 238U in the fuel. The influence of improvements in the neutron economy of these lattices was also investigated, by incorporating higher-purity heavy water moderator and coolant (99.90 at.% D2O) in the models, along with enriched zirconium (95 wt% 90Zr/Zr) for the zirconium alloys used in the structural components. The results were compared with those for a more conventional 37-element PT-HWR fuel bundle using natural uranium (NU) fuel. Results show that annular fuels could be very attractive, being able to achieve higher burnup, comparable or better fissile utilization, reduced coolant void reactivity, and comparable or more negative fuel temperature coefficients.

Research on the application of burnable poison ZrB2 and Gd2O3 in a small modular lead-based thermal reactor

A lead-based reactor with thermal or epithermal spectra is considered to be a good choice for small modular reactors design. Beside the safe and better thermal property of lead coolant it also takes the advantage of the smaller critical mass characteristics of thermal reactor. However, it needs to control the substantial amounts of excess reactivity at the beginning of core life which is essential to get a long-life cycle. Therefore, choosing the appropriate burnable poison material and optimizing its arrangement is a big concern during the design of a small modular lead-based thermal reactor.In this article, the analysis on use of burnable poison in a small modular lead-based thermal reactor has been carried out to achieve effective control of reactivity. The burnup characteristics of burnable poison ZrB2 and Gd2O3 were investigated and compared. The results show that ZrB2 is better than Gd2O3 for flatting the keff curve, while the depletion rate of Gd2O3 is much faster than ZrB2. Reducing the poison content in poison rods and increasing the number of poison assemblies can speed up the poison depletion. Based on the results, an optimized core (with the use of burnable poison) which can give a more flat keff curve with the burnup reactivity swing less than 3303 pcm was given. And the influences of the burnable poison on the main parameters of the optimized core scheme such as power distribution and reactivity coefficient were analyzed. The results show that with the use of burnable poison, the radial power distribution becomes more flat and the coolant void effect and fuel temperature reactivity become more negative.

Comparisons between RFSP and MCNP for modeling pressure tube heavy water reactor cores with thorium-based fuels

Thorium-based fuels are recognized to hold significant promise as an option for achieving a long-term, sustainable nuclear fuel cycle and energy security. Pressure tube heavy water reactors (PT-HWRs) are well suited to exploit the energy potential of thorium. Deterministic reactor physics codes are often used in exploratory studies to evaluate the performance and operational characteristics of various fuel bundle, lattice and core concepts with thorium based fuels in PT-HWRs. Because of the approximations inherent in deterministic codes, they are often considered less accurate than stochastic codes. In order to enhance confidence in deterministic code-based predictions, these codes are often benchmarked against stochastic codes, when experimental data is not available for code validation. Code-to-code comparisons of core physics calculations were made between the deterministic reactor physics toolset WIMS-AECL/WIMS-Utilities/RFSP and the stochastic neutron transport code MCNP for a series of core configurations with mixed oxide fuels containing thorium in PT-HWRs. The core neutron multiplication factors (keff) appear to have a difference (RFSP-MCNP) ranging between −2.4 mk and +4.0 mk. The MCNP full-core calculations confirm that that thorium-based fuels have a lower coolant void reactivity (CVR), ranging from +8.3 mk to +11.3 mk (versus 14 mk for NU fuel). The core cases with NU fuel have a small difference (RFSP-MCNP) in peak bundle power (ranging between −0.13% and 0.65%). Cores with LEU at 1.2 wt% 235U/U, Pu/Th, and LEU/Th (LEU at 5 wt% 235U/U) fuel have higher differences in peak bundle power (ranging between −7% and −12%). All these core cases have peak channel power differences between −2.1% and −8.9%. Core with 233U fuel has the smallest peak bundle difference (−0.05%) and smallest peak channel differences (−0.58%) which represent the best agreement between MCNP and RFSP simulations. The performed code-to-code comparisons have demonstrated that the core physics parameters estimated by RFSP calculations are consistent with MCNP simulations, especially for fuel where the main fissile component are 235U-based and 233U-based fuel.

Neutronic design study of a small modular IPWR loaded with accident tolerant-fully ceramic micro-encapsulated (AT-FCM) fuel

Small modular reactors (SMRs) offer a green energy option that will be suitable to serve smaller energy markets with less developed infrastructure. In the present work, a neutronic model of a near term deployable SMR of a natural circulation integral pressurized water reactor (IPWR) design is developed using the TRISO type fuel embedded in SiC matrix. The model development is done using the Monte Carlo Code capable of treating the double heterogeneity in implicit random manner. The Low Enriched Uranium Carbide (LEUC) fuel in TRISO form embedded in SiC matrix and cladded with FeCrAl forms a perfect combination for an accident tolerant fuel for light water reactors due to enhanced capability of fission-product retention of TRISO fuel and lower hydrogen production by FeCrAl at elevated temperature. Length of assembly is altered for reduced active height suitable for SMR application. The depletion analysis is performed for the optimization of fuel enrichment and core loading pattern to achieve a uniform core power distribution. The optimized core is capable of producing 160 MW(th) power with a core life time of more than four EFPYs without refueling or shuffling. The Erbia (Er2O3), 80 w/o enriched in 167Er as an integral fuel burnable absorber (IFBA) in QUADRISO form, shown excellent depletion performance and control of power peaking throughout the life cycle. This study showed that Doppler, moderator temperature and coolant void reactivity coefficients are all negative over the core lifetime. Transuranic production is low, with relatively higher 240Pu content in the spent fuel. This leads to a highly proliferation resistant discharge fuel. A new design of RCCA loaded with HfB2 absorber is capable of providing the required shutdown margin in all core conditions. Therefore, the SM-IPWR concept satisfies completely the design criteria and the requirements of an ATF loaded PWR.

Lattice physics evaluation of 35-element mixed oxide thorium-based fuels for use in pressure tube heavy water reactors

A series of 2-D lattice physics calculations with depletion were carried with WIMS-AECL Version 3.1 out as part of exploratory scoping studies to evaluate various thorium-based fuel bundle concepts for potential application in pressure tube heavy water reactors (PT-HWRs). Fuel bundles concepts investigated consisted of a cluster of 35 fuel elements arranged in two rings (14 + 21), and surrounding a central graphite displacer rod. The fuel is comprised of thorium dioxide mixed with a fissile driver of reactor-grade plutonium (∼67 wt% Pufissile/Pu; 3.5–4.5 wt% PuO2/(Pu,Th)O2), low enriched uranium (5 wt% 235U/U; 40–50 wt% LEUO2/(LEU,Th)O2) or uranium–233 (1.8 wt% 233UO2/(233U,Th)O2). Estimates of burnup-averaged fuel temperature coefficients (FTC) and coolant void reactivity (CVR) were found to be lower than those for conventional natural uranium dioxide (NUO2) PT-HWR fuel in a 37-element bundle. A low-burnup option for using (LEU,Th)O2 fuel in a PT-HWR is found to be attractive as a means for extracting energy from thorium, while also generating stockpiles of 233U, and demonstrating enhanced safety characteristics with reduced CVR and FTC relative to NUO2.

Simulations of Pressure-Tube–Heavy-Water Reactor Cores Fueled with Thorium-Based Mixed-Oxide Fuels

Thorium, a fertile nuclear fuel which is nearly three times as abundant as uranium, represents a long-term energy resource that could complement uranium by making the current nuclear fuel cycle more sustainable. With the expected refurbishment and new construction of pressure-tube–heavy-water reactors (PT-HWRs) within the international community, there is a good opportunity to gain experience with thorium-based fuels and to start the transition phase toward use of thorium as part of the nuclear fuel cycle. Previous studies have shown that in the near term (10 to 15 years), small amounts of thorium could be introduced into conventional 37-element fuel bundles made with natural uranium (NU) or slightly enriched uranium for use in PT-HWRs, which could help improve performance and safety characteristics relative to the use of NU alone. This work is an evaluation of various types of thorium-based oxide fuels in 35-element bundles that could potentially be implemented in current PT-HWRs to enable the transition toward the thorium fuel cycle to harness the energy potential from thorium and to conserve uranium resources. Such fuel concepts could have additional benefits, including enhanced reactor performance and safety characteristics, a reduction in the production of fissile plutonium, and the ability to consume stockpiles of plutonium. Thorium (in the form of ThO2) is mixed with a fissile driver of either reactor-grade plutonium (67 wt% Pu-fissile/Pu, 3.5 to 4.5 wt% PuO2), low-enriched uranium (LEU) (5 wt% 235U/U, 40 to 50 wt% LEUO2) or 233U (1.8 wt% 233UO2). The 35-element fuel bundle has a central graphite displacer rod to help maintain a low coolant void reactivity (CVR), while also flattening the radial power distribution in the fuel bundle due to supplementary neutron moderation. Analysis suggests that these thorium-based fuel types have lower fuel temperature coefficients (FTCs) and CVR values and higher fissile utilization (FU) compared to conventional 37-element natural uranium oxide (NUO2) fuel used currently in PT-HWRs. Lower values of CVR and FTCs will lead to lower and more negative values of the power coefficient of reactivity (PCR), which will help increase the safety margin for PT-HWRs further. Achieving higher FU will help better conserve nuclear fuel resources (both uranium and thorium), and should lead to lower fuel costs in a once-through fuel cycle. The use of a higher–fissile content fuel (in the form Pu, LEU, or 233U mixed with thorium) enables both higher fuel burnups and FU relative to NU, while also breeding and burning 233U. The low-burnup (LEU,Th)O2 core configuration shows the most promise as a feasible and practical means of producing energy from thorium (up to 33% of total energy produced) while also generating stockpiles of 233U. PT-HWR cores using (Pu,Th)O2 fuels may have some challenges in regard to the power balance in the bundle and reactor operation. Without reactivity control devices, some (Pu,Th)O2 cores may require power reductions to avoid exceeding bundle power limits, if a maximum linear element rating of 54 kW/m (which applies to uranium-based fuels) is assumed. PT-HWR cores fueled with (Pu,Th)O2 also demonstrate significant benefits as a means of consuming stockpiles of fissile plutonium, reducing inventories by 84% to 90% in a single pass through a PT-HWR core, while also producing substantial quantities of 233U (up to 314 kg/year/reactor) in the spent fuel for potential reprocessing and recycling in a closed fuel cycle.

Use of Er Burnable Absorber for Improvement of CANDU6 Safety Parameters

This paper reports on the improvement of the power coefficient of reactivity (PCR) and minimization of the coolant void reactivity (CVR) of a CANDU6 reactor. A burnable absorber of Er2O3 (erbia) was mixed homogeneously with UO2 fuel in the central fuel element to maximize the Doppler broadening and minimize the CVR of the CANDU6 reactor. In this study, recovered uranium (RU) with 0.9 wt% 235U enrichment was utilized in the advanced CANFLEX fuel bundle instead of natural uranium (NU). First, the optimal loading of erbia was investigated through lattice-based analysis, and its impact on the lattice characteristics was examined. In particular, both the fuel Doppler effect and CVR were evaluated for the RU-loaded lattice. For a more reliable analysis, a three-dimensional (3-D) equilibrium core was determined based on the standard time-average methods for erbia-loaded CANDU6 cores using the Serpent-COREDAX/CANDU code system. The core analysis was based on a hybrid two-step method in which the lattice analysis was performed by the Serpent Monte Carlo code, and the 3-D whole-core analysis was done using a diffusion theory–based nodal code named COREDAX. For the derived equilibrium cores, the core performances were evaluated in terms of the fuel burnup and power profile. Additionally, the safety parameters, including the PCR and CVR, were evaluated for the equilibrium core conditions. The safety parameters of the 3-D whole core were compared with those obtained with simple lattice-based analysis. It was observed in the analysis that Er-loaded CANFLEX-RU fuel provides a 60% more negative fuel temperature coefficient than standard CANDU-NU fuel.

Assessment of Neutronic Characteristics of Accident-Tolerant Fuel and Claddings for CANDU Reactors

The objective of this study was to evaluate accident-tolerant fuel (ATF) concepts being considered for CANDU reactors. Several concepts, including uranium dioxide/silicon carbide (UO2-SiC) composite fuel, dense fuels, microencapsulated fuels, and ATF cladding, were modelled in Serpent 2 to obtain reactor physics parameters, including important feedback parameters such as coolant void reactivity and fuel temperature coefficient. In addition, fuel heat transfer was modelled, and a simple accident model was tested on several ATF cases to compare with UO2. Overall, several concepts would require enrichment of uranium to avoid significant burnup penalties, particularly uranium-molybdenum (U-Mo) and fully ceramic microencapsulated (FCM) fuels. In addition, none of the fuel types have a significant advantage over UO2 in terms of overall accident response or coping time, though U-9Mo fuel melts significantly sooner due to its low melting point. Instead, the different ATF concepts appear to have more modest advantages, such as reduced fission product release upon cladding failure, or reduced hydrogen generation, though a proper risk assessment would be required to determine the magnitude of these advantages to weigh against economic disadvantages. The use of uranium nitride (UN) enriched in N15 would increase exit burnup for natural uranium, providing a possible economic advantage depending on fuel manufacturing costs.

An Advanced Sodium-Cooled Fast Reactor Core Concept Using Uranium-Free Metallic Fuels for Maximizing TRU Burning Rate

In this paper, we designed and analyzed advanced sodium-cooled fast reactor cores using uranium-free metallic fuels for maximizing burning rate of transuranics (TRU) nuclides from PWR spent fuels. It is well known that the removal of fertile nuclides such as 238U from fuels in liquid metal cooled fast reactor leads to the degradation of important safety parameters such as the Doppler coefficient, coolant void worth, and delayed neutron fraction. To resolve the degradation of the Doppler coefficient, we considered adding resonant nuclides to the uranium-free metallic fuels. The analysis results showed that the cores using uranium-free fuels loaded with tungsten instead of uranium have a significantly lower burnup reactivity swing and more negative Doppler coefficients than the core using uranium-free fuels without resonant nuclides. In addition, we considered the use of axially central B4C absorber region and moderator rods to further improve safety parameters such as sodium void worth, burnup reactivity swing, and the Doppler coefficient. The results of the analysis showed that the final design core can consume ~353 kg per cycle and satisfies self-controllability under unprotected accidents. The fuel cycle analysis showed that the PWR–SFR coupling fuel cycle option drastically reduces the amount of waste going to repository and the SFR burner can consume the amount of TRUs discharged from 3.72 PWRs generating the same electricity.

Nuclear data uncertainty propagation on a sodium fast reactor

Uncertainties on criticality parameters and reactivity effects in a SFR (Sodium Fast Reactor) were calculated and characterized. An SFR model loaded with ≈10% MAs (Minor Actinides) was chosen as test case of a burner reactor, its core geometry was implemented in MCNP-5 according to benchmarking specifications. Monte Carlo random sampling on the ENDF/B-VII.1 neutron-induced nuclear data was performed using the SANDY code developed at the Belgian nuclear research center SCK•CEN. SANDY was proved to be a successful tool for uncertainty quantification in the model case of a MA burner SFR. The results show that despite the high content in MAs, the contribution of U-238 data is still dominant in most instances, especially the inelastic scattering cross section data. Minor actinides together with Pu make up to 33% of the total calculated keff uncertainty, which amounts to ±1345pcm of reactivity. Relative uncertainty greater than ±3.47% was estimated for βeff with significant contribution from the prompt multiplicity data of U-238 and Pu-239. Relative uncertainty on the Doppler reactivity worth was found to be at least ±12.6%. Coolant voiding in different core zones yields reactivity worth with relative uncertainties whose lower bounds range from ±6.37% to ±22.47% depending on the voiding conditions, with important contributions due to the Na-23 nuclear data uncertainty.

ON THE USE OF A CENTRAL THORIUM FUEL ELEMENT IN PRESSURE-TUBE HEAVY-WATER REACTOR FUEL BUNDLES

An incremental approach to introducing thorium to the conventional pressure-tube heavy-water reactor natural uranium fuel cycle is investigated. The approach involves the replacement of the centre fuel element of the bundle with an element of thorium dioxide. Increasing the operating margin of a key safety parameter, the coolant void reactivity, is a prime motivating factor. The analyses showed that the simple use of a single pin of thorium is unlikely to be economically advantageous due to a large burnup penalty and increased fuel costs. However, a slight reduction in the void reactivity is observed, and this approach does allow the exploitation of the energy potential available in thorium as an alternative nuclear fuel resource through the development of a U-233 resource. This bundle concept may also be advantageous from a fuel disposal point of view, as the fuel requires less time in storage before emplacement in a deep geological repository.