In-phase and out-of-phase modes stability analysis with LAPUR5 code for Kuosheng

A study of the in-phase (global, core-wide) and out-of-phase (regional) modes of stability with LAPUR5 code for Kuosheng NPP is presented. Comparisons made with vendor’s STAIF results indicated close agreements, within the acceptable 20% deviation in decay ratios, for Kuosheng NPP Unit 1 Cycle 18 reloads design. Parametric studies for stability have shown that the inlet loss coefficient, exposure, control rod position, and axial power shape are the most important parameters that might affect the accuracy of decay ratios for in-phase and out-of-phase modes of stability analysis. A decrease in the inlet loss coefficient or increase in the control rods insertion causes an increase in the global and regional decay ratios. An increase in the exposure causes an increase in the global decay ratio, the maximum value of the regional decay ratio occurs in MOC condition, and the regional decay ratio is higher than global decay ratio in the BOC condition.     

Effect of parametric sensitivity on stability boundary of Chinshan NPP

Boiling water reactors are susceptible to instabilities under certain condition and require stability boundary to prevent instability. On this stability boundary, any perturbation or oscillation may easily induce unstable events. All of the parameters that may dominate global mode instability must be analyzed and evaluated. In each core reload design, a decay ratio analysis must be conducted to ensure system stability. LAPUR5.2 and SIMULATE-3 codes were employed to establish a process to conduct such analysis. This investigation contributes to the parametric sensitivity effect, which is a variation of the decay ratio with power/flow operating points along the stability boundary. The results indicate that fractional changes of decay ratios with variation in certain important parameters under different power/flow points. The density reactivity coefficient, gap conductance, recirculation loop gain and system pressure, at the points of high power/flow, all are associate with larger fractional changes in the decay ratio than the points of low power/flow.     

Transient flow analysis in reactor coolant pump systems during flow coastdown period

The rapid flow transient calculation in reactor coolant pump system is important in the safety analysis of a nuclear reactor. An accurate transient analysis of flow coastdown is also important and necessary for the design and manufacture of a reactor coolant pump. Only under the reliable work of a reactor coolant pump the safety of a nuclear power plant can be guaranteed. A mathematical model is developed for solving flow rate transient and pump speed transient during flow coastdown period. The detailed information of the centrifugal pump characteristics is not required. The flow rate and pump speed are solved analytically. The analytic solution of non-dimensional flow rate indicates that non-dimensional flow rate is determined by energy ratio β. The kinetic energy of the loop coolant fluid and the kinetic energy stored in the rotating parts are two important parameters in form of β. When the steady-state flow rate and pump speed are constant, the inertia of primary loop fluid and the pump moment of inertia are also two important parameters in flow transient analysis. For the condition all pump shafts are seized, the flow decay depends on the inertia of primary loop fluid. For the case that pump inertia is very large, the flow decay is determined by the pump inertia. The calculated non-dimensional flow rate and non-dimensional pump speed using the model are compared with published experimental data of two nuclear power plants and a reactor model test on flow coastdown transients. The comparison results show a good agreement. As the flow rate approaches to zero, the increase difference between experimental and calculated value is due to the effect of the mechanical friction loss.     

Azcatl-CRP: An ant colony-based system for searching full power control rod patterns in BWRs

We show a new system named AZCATL-CRP to design full power control rod patterns in BWRs. Azcatl-CRP uses an ant colony system and a reactor core simulator for this purpose. Transition and equilibrium cycles of Laguna Verde Nuclear Power Plant (LVNPP) reactor core in Mexico were used to test Azcatl-CRP. LVNPP has 109 control rods grouped in four sequences and currently uses control cell core (CCC) strategy in its fuel reload design. With CCC method only one sequence is employed for reactivity control at full power operation. Several operation scenarios are considered, including core water flow variation throughout the cycle, target different axial power distributions and Haling conditions. Azcatl-CRP designs control rod patterns (CRP) taking into account safety aspects such as k_eff core value and thermal limits. Axial power distributions are also adjusted to a predetermined power shape.     

Monte Carlo simulation of Feynman-α and Rossi-α techniques for calculation of kinetic parameters of Tehran Research Reactor

Noise analysis techniques including Feynman-α (variance-to-mean) and Rossi-α (correlation) have been simulated by MCNP computer code to calculate the prompt neutron decay constant (α₀), effective delayed neutron fraction (β_eff) and neutron generation time (Λ) in a subcritical condition for the first operating core configuration of Tehran Research Reactor (TRR). The reactor core is considered to be in zero power (reactor power is less than 1 W) in the entire simulation process. The effect of some key parameters such as detector efficiency, detector position and its dead time on the results of simulation has been discussed as well. The results of proposed method in the current study are validated against both the experimental data and the results of MTR_PC computer code.     

Comparison of TSSD results obtained by differential scanning calorimetry and neutron diffraction

Due to their low absorption cross-section for neutrons, Zr alloys are used for reactor core components. The terminal solid solubility (TSS) for hydrogen in these alloys is very low - in Zr-2.5 wt% Nb, used to fabricate pressure tubes for CANDU (CANDU-CANada Deuterium Uranium is a registered trademark of Atomic Energy of Canada Ltd.) power reactors, the TSS is ∼0.7 at.% H at 300 °C. The mechanical properties of the components may deteriorate when their hydrogen concentration exceeds TSS. Therefore, accurate values of the TSS are needed to assess the operating and end-of-life behaviours of these components. Differential scanning calorimetry (DSC) is used to measure the TSS of hydrogen in Zr alloys. Three distinct features are marked on a typical DSC heat flow curve when the material is being heated and the hydrides are dissolving; ‘peak temperature’, ‘maximum slope temperature’ and ‘completion temperature’. Usually, the maximum slope temperature, being about the average of the three temperatures, is interpreted as the TSS temperature for hydride dissolution (T_TSSD). A set of coordinated DSC and neutron diffraction measurements have been carried out to identify the features of the heat flow signal that closely correspond to the T_TSSD. Neutron diffraction was chosen because hydrides generate distinctive diffraction peaks whose intensities approach zero at the transition temperature - an unambiguous indication of dissolution. Neutron diffraction shows that the temperature of hydride dissolution correlates closely with the DSC peak temperature.     

An assessment of the BWR fuel box-lattice-cell homogenization procedures with the Tarapur BWR incore and excore measurements

A code system has been developed to provide the incorefuel-management guidelines to the Tarapur BWR reactors. Constant checking of the design calculational methods is rendered possible by the steady flow of operating data from the Tarapur units over the last few cycles. The operating data include cold/hot criticals and detailed flux/power maps. Besides these, the burnups and isotopic composition of a few irradiated fuel pins obtained by mass-spectrometric analyses, are also available for validation of the BWR core and lattice-cell modelling.The calculated eigen values for different power levels and at different core average burnups are found to have a spread of less than 0.25% ΔK. Analyses of a number of TIP measurements show that the core power distribution can be predicted in a satisfactory manner for uncontrolled fuel bundles and non-peripheral fuel assemblies (<10%). For prediction of cold-criticals the void-history effects are found to be unimportant.The pin burnups and isotopic densities of important U and Pu isotopes relative to ²³⁸U have been compared with mass-spectrometric measurements. The pin-burnup profile comparison is found to be good for fuel pins, which are not near water gaps. Deviation histograms of various isotopes are presented in this paper. ²³⁵U is predicted within ± 3% (r.m.s.). The total Pu is overpredicted by 5–8%, while the quality of Pu is predicted within ± 1.0% (r.m.s.).     

Molecular dynamics evaluation of the impact of Ga, He, and vacancy concentration on the mechanical properties of Ga-stabilized δ-Pu

The effects of self-irradiation on the mechanical properties of Ga-stabilized δ-Pu are investigated by means of classical molecular dynamics simulations using a modified embedded-atom method interatomic potential. The impact that variations in the concentration of Ga, He, and vacancies have on the elastic properties (Young’s modulus, Poisson’s ratio, shear modulus, and bulk modulus) of Ga-stabilized δ-Pu is determined through CMD simulation. The effect that variations in these concentrations have on plastic flow is assessed at strain rates of 10⁷-10¹⁰ s⁻¹. Comparison is made to experimental data.     

Thermal hydraulic analysis for grid supported pressurized water reactor cores

This paper presents the methodology and results for thermal hydraulic analysis of grid supported pressurized water reactor cores using U(45% wt)-ZrH_1.6 hydride fuel in square arrays. The same methodology is applied to the design of UO₂ oxide fueled cores to provide a fair comparison of the achievable power between the two fuel types. Steady-state and transient design limits are considered. Steady-state limits include: fuel bundle pressure drop, departure from nucleate boiling ratio, fuel temperature (average for UO₂ and centerline/peak for U-ZrH_1.6), and fuel rod vibrations and wear. Transient limits are derived from consideration of the loss of flow and loss of coolant accidents, and an overpower transient.In general, the thermal hydraulic performance of U-ZrH_1.6 and UO₂ fuels is very similar. Slight power differences exist between the two fuel types for designs limited by rod vibrations and wear, because these limits are fuel dependent. Large power increases are achievable for both fuels when compared to the reference core power output of 3800 MW_th. In general, these higher power designs have smaller rod diameters and larger pitch-to-diameter ratios than the reference core geometry. If the pressure drop across new core designs is limited to the pressure drop across the reference core, power increases of ∼400 MW_th may be realized. If the primary coolant pumps and core internals could be designed to accommodate a core pressure drop equal to twice the reference core pressure drop, power increases of ∼1000 MW_th may be feasible.     

Fuel performance analysis for PWR cores

Attainable discharge burnups for oxide and hydride fuels in PWR cores were investigated using the TRANSURANUS fuel performance code. Allowable average linear heat rates and coolant mass fluxes for a set of fuel designs with different fuel rod diameters and pitch-to-diameter ratios were obtained by VIPRE and adopted in the fuel code as boundary conditions. TRANSURANUS yielded the maximum rod discharge burnups of the several design combinations, under the condition that specific thermal-mechanical fuel rod constraints were not violated. The study shows that independent of the fuel form (oxide or hydride) rods with (a) small diameters and moderate P/Ds or (b) large diameters and small P/Ds give the highest permissible burnups limited by the rod thermal-mechanical constraints. TRANSURANUS predicts that burnups of ∼74 MWd/kg U and ∼163 MWd/kg U (or ∼65.2 MWd/kg U oxide-equivalent) could be achieved for UO₂ and UZrH_x fuels, respectively. Furthermore, for each fuel type, changing the enrichment has only a negligible effect on the permissible burnup. The oxide rod performance is limited by internal pressure due to fission gas release, while the hydride fuel can be limited by excessive clad deformation in tension due to fuel swelling, unless the fuel rods will be designed to have a wider liquid metal filled gap. The analysis also indicates that designs featuring a relatively large number of fuel rods of relatively small diameters can achieve maximum burnup and provide maximum core power density because they allow the fuel rods to operate at moderate to low linear heat rates.     

On the space cross-correlation of noise sources in boiling water reactors

In this paper, an attempt has been made to investigate the space cross-correlation of boiling noise in boiling water reactors by assuming the slip ratio to be unity and power profile to be constant along the core height of the reactor. Two cases have been considered: One, no flow fluctuations (ΔV = 0) caused by boiling noise and two, finite flow fluctuations (ΔV ≠ 0) caused by the boiling noise. It has been found that finite range space cross-correlation of noise sources exists only for ΔV ≠ 0 case and not for ΔV = 0 case. Auto power spectral density of steam content fluctuations, Δα have a break frequency which is directly related with the flow velocity and attenuation coefficient, μ of the cross-correlation of noise sources in BWRs. Normalized root mean square value of Δα is more sensitive to μ in the upper half of the core for μ ≤ 10. For $\text{μ ? 20}$, it is more sensitive to μ in the lower half of the core.     

A sensitivity study of BWR instability over global and regional modes at different exposures

Boiling water reactors have unique mechanisms coupling between neutronic and two-phase flow thermalhydraulic behaviors, and may exhibit in-phase (global mode) instability and out-of-phase (regional mode) instability. In some observation modes, the regional mode instability is associated with an increase in power in one half of the core and a simultaneous decrease in power in the other half, such that the average power remains essentially constant. Yet in practice, sometimes the real situation is hidden, the neutron flux may oscillate more vigorously than expected. To investigate the stability behavior at the stability boundary from BOC (beginning of cycle) to EOC (end of cycle), fractional changes of the decay ratio are used to evaluate the parametric sensitivity of the global mode and the regional mode at different exposures. Decay ratios for regional mode oscillations are much less than those under core-wide conditions. Current studies demonstrated that for some of the parameters under particular conditions, the variation in the regional mode decay ratio exceeded that in the global mode. In this work, the thermalhydraulic parameters (such as flow rate and system pressure) exhibit a more sensitive regional variation than global. Moreover, some parameters (density reactivity coefficient and delayed-neutron fraction, for example) depend on the shape of the axial power shape; for the bottom peak axial power shape, the regional mode decay ratio variation is more sensitive than global; for the top peak axial power shape, the opposite is true.     

Thermal-fluid characteristics of the transuranics fuel in a deep-burn HTR core

The Deep Burn Project is evaluating the feasibility of the DB-HTR (Deep Burn High Temperature Reactor) to achieve a very high utilization of transuranics (TRU) derived from the recycle of LWR spent fuel. This study intends to evaluate the thermal-fluid and safety characteristics of TRU fuel in a DB-HTR core using GAMMA+ code. The key design characteristics of the DB-HTR core are more fuel rings (five fuel-rings), less central reflectors (three rings) and decay power curves due to the TRU fuel compositions that are different from the UO₂ fuel. This study considered three types of TRU kernel compositions such as 100%(PuO₂ + NpO₂ + Am), 99.8%(PuO_1.8, NpO₂) + 0.2%UO₂ + 0.6 mole SiC getter, and 70%(PuO_1.8, NpO₂) + 30%UO₂ + 0.6 mole SiC getter. The first fuel type of TRU kernel produces higher decay power than the UO₂ kernel. For the second and the third fuel types, removing the initial Am isotopes and reducing the volumetric packing fraction of TRISO particles will reduce the decay power. The flow distribution, core temperature and TRISO temperature profiles at the steady state were examined. As a safety performance, this study mainly evaluated the peak fuel temperature during LPCC (low pressure conduction cooling) event with considering the impact of decay power, the annealing effect of the irradiated thermal conductivity of graphite, and the impact of the FB (fuel block) end-flux-peaking. For the 600 MW_th DB-HTR core, the peak fuel temperature of 100%(PuO₂ + NpO₂ + Am) TRU was found to be much higher than the transient fuel design limit of 1600 °C due to the lack of heat absorber volume in the central reflector as well as to the increased decay power of the TRU fuel compositions. For a 0.2%UO₂ mixed or a 30%UO₂ mixed TRU, the peak fuel temperature was decreased due to the reduced decay power, however, it was still higher than 1600 °C due to the lack of heat absorber volume in the central reflector.     

A neutronics analysis of long-life, sectored compact reactor concepts for lunar surface power

Neutronics analysis is performed of low temperature (≤900 K) Sectored Compact Reactor (SCoRe-N_x) concepts for operational lives >20 years. They are cooled with circulating liquid Nak-78. The reactor vessel and the UN fuel pins cladding are made of stainless steel. The power systems with the SCoRe-N_x concepts have no moving parts and employ SiGe thermoelectric elements for converting the reactor’s thermal power to electricity at a load voltage >300 VDC. Separate SiGe elements power the electromagnetic pumps that circulate the liquid NaK-78 in the reactor and 6 pairs of primary and secondary loops. Analysis confirmed that the submerged bare reactors in wet sand and flooded with seawater, in the unlikely event of a postulated launch abort accident, are sufficiently subcritical. Results show that the SCoRe-N₅-S concept is capable of operating for 62 full power years at 200 kW_th, while requiring 75% of the control drums to shutdown the reactor at BOL. With only 50% of the control drums required at BOL to shutdown the reactor, its operational life decreases to 39 full power years.     

Thermal hydraulic challenges of Gas Cooled Fast Reactors with passive safety features

Transient response of a Gas Cooled Fast Reactor (GFR) coupled to a recompression supercritical CO₂ (S-CO₂) power conversion system (PCS) in a direct cycle to a Loss of Coolant Accident (LOCA) and a Loss of Generator Load Accident is analyzed using RELAP5-3D. A number of thermal hydraulic challenges for GFR design are pointed out as the designers strive to accommodate cooling of the high power density core of a fast reactor by a gas with its inherently low heat transfer capability, in particular under post-LOCA events when system pressure is lost and when reliance on passive decay heat removal (DHR) is emphasized. Although it is possible to design a S-CO₂ cooled GFR that can survive LOCA by cooling the core through natural circulating loops between the core and elevated emergency cooling heat exchangers, it is not an attractive approach because of various bypass paths that can, depending on break location, degrade core cooling. Moreover, natural circulation gas loops can operate in deteriorated heat transfer regimes with substantial reduction of heat transfer coefficient: as low as 30% of forced convection values, and data and correlations in these regimes carry large uncertainties. Therefore, reliable battery powered blowers for post-LOCA decay heat removal that provide flow in well defined regimes with low uncertainty, and can be easily overdesigned to accommodate bypass flows were selected. The results confirm that a GFR with such a DHR system and negative coolant void worth can withstand LOCA with and without scram as well as loss of electrical load without exceeding core temperature and turbomachinery overspeed limits.     

Safety aspects of the 200 MWth Siemens Nuclear District Heating Plant, including accident management measures

Specific requirements of district heating lead to a simple and safe reactor design with small absolute power, low temperatures and pressures, and modest load following behaviour. Plant safety is essentially guaranteed by inherent features. Shutdown is assured by gravity drop of the control rods, decay heat is removed from the core by means of natural circulation via dedicated intermediate circuits to external aircoolers.The accident analysis and probabilistic assessment attest the heating reactor a well balanced safety concept and a probability for Accident Management (AM) measures for all conceivable events less than 10⁻⁷ a⁻¹. Provisions for AM measures for decay heat removal, water supply into primary system and electrical power supply can be achieved using existing systems and components. The possibilities are manifold and can be repeated independent from each other at any time. Therefore large grace periods are available and can be extended by above mentioned measures thus it is justified to state that core melting can be precluded.     

Decay heat in fast reactors with transuranic fuels

In this work, we performed an evaluation of decay heat power of advanced, fast spectrum, lead and molten salt-cooled reactors, with flexible conversion ratio. The decay heat power was calculated using the BGCore computer code, which explicitly tracks over 1700 isotopes in the fuel throughout its burnup and subsequent decay. In the first stage, the capability of the BGCore code to accurately predict the decay heat power was verified by performing a benchmark calculation for a typical UO₂ fuel in a Pressurized Water Reactor environment against the (ANSI/ANS-5.1-2005, “Decay Heat Power in Light Water Reactors,” American National Standard) standard. Very good agreement (within 5%) between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power for fast reactors with different coolants and conversion ratios, for which no standard procedure is currently available. Notable differences were observed for the decay power of the advanced reactor as compared with the conventional UO₂ LWR. The importance of the observed differences was demonstrated by performing a simulation of a Station Blackout transient with the RELAP5 computer code for a lead-cooled fast reactor. The simulation was performed twice: using the code-default ANS-79 decay heat curve and using the curve calculated specifically for the studied core by BGCore code. The differences in the decay heat power resulted in failure to meet maximum cladding temperature limit criteria by ∼100 °C in the latter case, while in the transient simulation with the ANS-79 decay heat curve, all safety limits were satisfied. The results of this study show that the design of new reactor safety systems must be based on decay power curves specific to each individual case in order to assure the desired performance of these systems.     

Hydride fuel for LWRs—Project overview

This special issue of Nuclear Engineering and Design consists of a dozen papers that summarize the research accomplished in the DOE NERI Program sponsored project NERI 02-189 entitled “Use of Solid Hydride Fuel for Improved Long-Life LWR Core Designs”. The primary objective of this project was to assess the feasibility of improving the performance of pressurised water reactor (PWR) and boiling water reactor (BWR) cores by using solid hydride fuels instead of the commonly used oxide fuel. The primary measure of performance considered is the cost of electricity (COE). Additional performance measures considered are attainable power density, fuel bundle design simplicity, in particular for BWRs, safety, attainable discharge burnup, and plutonium (Pu) transmutation capability.Collaborating on this project were the University of California at Berkeley Nuclear Engineering Department (UCB), Massachusetts Institute of Technology Nuclear Science and Engineering Department (MIT), and Westinghouse Electric Company Science and Technology Department. Disciplines considered include neutronics, thermal hydraulics, fuel rod vibration and mechanical integrity, and economics.It was found that hydride fuel can safely operate in PWRs and BWRs having comparable or higher power density relative to typical oxide-fueled LWRs. A number of promising applications of hydride fuel in PWRs and BWRs were identified: (1) Recycling Pu in PWRs more effectively than is possible with oxide fuel by virtue of a number of unique features of hydride fuel-reduced inventory of ²³⁸U and increased inventory of hydrogen. As a result, the hydride-fueled core achieves nearly double the average discharge burnup and the fraction of the loaded Pu it fissions in one pass is double that of the MOX fuel. (2) Eliminating dedicated water moderator volumes in BWR cores, thus enabling significant increase of the cooled fuel rod surface area as well as the coolant flow cross-section area in a given fuel bundle volume while reducing the heterogeneity of BWR fuel bundles, thus achieving flatter pin-by-pin power distribution. The net result is an increase in the core power density and a reduction of the COE.A number of promising oxide-fueled PWR core designs were also found in this study: (1) The optimal oxide-fueled PWR core design features a smaller fuel rod diameter (D) of 6.5 mm and a larger pitch to rod diameter (P/D) ratio of 1.39 than that presently practiced by industry of 9.5 mm and 1.326. This optimal design can provide a 27% increase in the power density and a 19% reduction in the COE provided the PWR can be designed to have the coolant pressure drop across the core increased from the reference 0.20 MPa (29 psi) to 0.414 MPa (60 psi). Under the set of constraints assumed in this work, hydride fuel was found to offer comparable power density and economics as oxide fuel in PWR cores when using fuel assembly designs featuring square lattice and grid spacers. This is because pressure drop constraints prevented achieving sufficiently high power using hydride fuel with a relatively small P/D ratio of around 1.2 or less, where it offers the highest reactivity and a higher heavy metal (HM) loading. (2) Using wire-wrapped oxide fuel rods in hexagonal fuel assemblies, it is possible to design PWR cores to operate at ∼50% higher power density than the reference PWR design that uses grid spacers and a square lattice, provided 0.414 MPa coolant pressure drop across the core could be accommodated. Uprating existing PWRs to use such cores could result in up to 40% reduction in the COE. The optimal lattice geometry is D = 9.34 mm and P/D = 1.37. The most notable advantages of wire-wraps over grid spacers are their significantly lower pressure drop, higher critical heat flux, and improved vibration characteristics.The achievement of the highest power gains claimed in this study is possible as long as mechanical components like assembly hold-down devices (both in PWRs and in BWRs) and steam dryers (only in BWRs) are appropriately upgraded to accommodate the higher coolant pressure drop and flow velocities required for the high-performance LWR designs. The compatibility of hydride fuel with Zircaloy clad and with PWR and BWR coolants need yet be experimentally demonstrated. Additional recommendations are given for future studies that need to be undertaken before the commercial benefits from use of hydride fuel could be reliably quantified.     

Incorporating reliability analysis into the design of passive cooling systems with an application to a gas-cooled reactor

A time-dependent reliability evaluation of a two-loop passive decay heat removal (DHR) system was performed as part of the iterative design process for a helium-cooled fast reactor. The system was modeled using RELAP5-3D. The uncertainties in input parameters were assessed and were propagated through the model using Latin hypercube sampling. An important finding was the discovery that the smaller pressure loss through the DHR heat exchanger than through the core would make the flow to bypass the core through one DHR loop, if two loops operated in parallel. This finding is a warning against modeling only one lumped DHR loop and assuming that n of them will remove n times the decay power. Sensitivity analyses revealed that there are values of some input parameters for which failures are very unlikely. The calculated conditional (i.e., given the LOCA) failure probability was deemed to be too high leading to the identification of several design changes to improve system reliability. This study is an example of the kinds of insights that can be obtained by including a reliability assessment in the design process. It is different from the usual use of PSA in design, which compares different system configurations, because it focuses on the thermal–hydraulic performance of a safety function.     

Steady state thermal-hydraulic analysis of hydride-fueled grid-supported BWRs

A steady state thermal-hydraulic analysis was performed to estimate the power density attainable with hydride-fueled boiling water reactor (BWR) cores with respect to that of an existing oxide BWR core chosen as reference. The power-limiting constraints taken into account were the minimum critical power ratio (MCPR), core pressure drop, fuel average and centerline temperature, cladding outer temperature, flow-induced vibrations and power/flow ratio.The study consisted of two independent analyses: a whole core analysis and a single bundle analysis. The whole core analysis was performed, with a fixed core volume, for both hydride and oxide fuel over hundreds of combinations of rod diameter-rod pitch, referred to as “geometries”, in the ranges 0.6 ≤ D ≤ 1.6 cm and 1.1 ≤ P/D ≤ 1.6. For each geometry, the maximum achievable steady state core power was calculated. Preliminary neutronics results derived from a companion neutronic study were then overlaid on the whole-core thermal-hydraulic results to estimate the reduction in maximum achievable power caused by the application of neutronic constraints. The single bundle analysis was performed to compare in greater detail the thermal-hydraulic performance of a limited number of hydride and oxide fuel bundles having D and P values similar to those of the reference oxide bundle, and for which the compliance with neutronic constraints was demonstrated in a companion neutronic study.The study concluded that, if the core pressure drop is not allowed to increase above the reference core value, the power density increase attainable with hydride fuel is estimated to be in the range 0-15%. If the pressure drop is allowed to increase up to a value 50% higher than the reference core value, the power density increase is estimated to be in the range 25-45%. These power density increases, which are defined with respect to the reference oxide core, decrease about 10% if the comparison is made with respect to oxide designs resembling the most recent commercial high-performance oxide cores.The power gain capability of hydride fuel is primarily due to the possibility of: (1) replacing volumes occupied by water rods and water gaps in oxide fuel cores with fuel rods, thus increasing the heat transfer area per core volume, and (2) flattening the bundle pin-by-pin power distribution.The actual achievement of the above-mentioned power density increase is however conditioned to the compliance of hydride-fueled cores to safety requirements related to core behavior during transients, hydrodynamic stability and steam dryer performance, which are fields of study not addressed in this work. A potential 25-45% power density increase justifies however interest for further investigation on this alternative fuel.