Economic analysis of grid and wire wrap supported hydride and oxide fueled pressurized water reactors

An economic analysis is performed to calculate the levelized unit cost of electricity (COE) for a pressurized water reactor (PWR) retrofitted with a range of potential U (45 wt.%)-ZrH_1.6 hydride and UO₂ oxide fueled geometries (i.e., combinations of rod diameter and pitch) supported by traditional grid spacers (square array) and wire wrap spacers (hexagonal array). The time frame considered in computing the COE is the remaining plant life, beginning at the time of retrofit. The goals of the analysis are twofold: (1) comparing the economic performance of UO₂ and U-ZrH_1.6 fuels for a range of retrofitted geometries supported by grid and wire wrap spacers; and (2) investigating the potential economic benefits for nuclear utilities considering retrofitting new fuels and/or geometries into existing PWR pressure vessels. Fuel cycle, operations and maintenance (O & M), and capital costs are considered.The economic performance of U-ZrH_1.6 and UO₂ fuels is found to be similar, with UO₂ fueled designs providing a slight advantage when supported by grid spacers, and U-ZrH_1.6 providing a slight advantage when supported by wire wrap spacers. These small differences in cost, however, are within the bounds of uncertainty of this study and are not believed to provide a strong economic argument for the use of one fuel type over the other.To demonstrate the potential economic benefits of retrofitted designs to nuclear utilities, two different comparisons are made. The first compares the COE for retrofitted designs with the COE for a reference PWR, assumed to have operated long enough to recuperate its initial capital investment. The costs for this reference PWR reflect the “do-nothing” case for current plant owners whose primary expenditures are fuel cycle and O & M costs. The second comparison introduces a different reference PWR that includes the costs to operate an existing unit and the cost to purchase power from a newly constructed PWR, for comparison with retrofitted designs which offer increased power relative to existing commercial PWRs.For the first comparison, no grid supported designs and only one wire wrap supported design (i.e., U-ZrH_1.6 Stretch Case) provide a lower levelized unit cost of electricity than the reference “do-nothing” PWR. The primary cause of this conclusion is the capital costs incurred by retrofitted designs to change the core geometry and, for many designs, to upgrade primary and secondary loop components for operation at higher power than the reference PWR. The reference “do-nothing” PWR cost in this first comparison includes only operations and maintenance as well as fuel cycle costs but does not include a capital component. For the second comparison, significant cost savings are demonstrated for both grid (15-19% savings) and wire wrap (30-40% savings) supported designs using U-ZrH_1.6 and UO₂ fuels. These cost savings are enabled by enhancing the pumping capacity of the primary system and, for wire wrap supported designs, by taking advantage of enhanced critical heat flux performance. The optimal geometry for retrofitted UO₂ and U-ZrH_1.6 fueled PWR cores supported by grid spacers is D_rod = 6.5 mm and P/D = 1.39. The cost savings over the second case reference PWR are ∼19 and 15%, respectively. The cost savings for retrofitted PWRs that incorporate wire wrap spacing are even larger because of operation at even higher power. Cost savings over the reference PWR range between 30 and 40% for the U-ZrH_1.6 and UO₂ Achievable and Stretch Cases. The optimal geometries for the U-ZrH_1.6 Achievable and Stretch Cases are D_rod = 8.08 mm, P/D = 1.41 and D_rod = 8.71 mm, P/D = 1.39, respectively. The optimal geometries for the UO₂ Achievable and Stretch Cases are D_rod = 7.13 mm, P/D = 1.42 and D_rod = 9.34 mm, P/D = 1.27, respectively. Utilities seeking to meet rising demand by expanding capacity may therefore strongly benefit from retrofitting existing PWRs with either U-ZrH_1.6 or UO₂ fueled designs. These new designs have different geometries than are currently used by commercial plants. A conclusion on which fuel type to use, however, could not be reached in this analysis as both offer similar economic performance.     

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.     

Neutronic analysis of hydride fueled PWR cores

The neutronic properties of U-ZrH_1.6 fuelled PWR cores are investigated and compared against those of the currently used UO₂ fuelled cores. In the first part of this work a parametric study is performed to quantify the neutronically achievable burnup for both hydride and oxide fuels at a number of enrichment levels and for a large number of geometries covering a wide design space of fuel rod outer diameter, D, and lattice pitch, P. The fuel temperature and coolant temperature reactivity coefficients as well as the small and large void reactivity coefficients are calculated for hydride fuel with 5% and 12.5% enriched uranium. For this purpose a simplified procedure was developed that can, using single unit cell or assembly calculations, (1) account for non-linear burnup dependent k_∞ and thus to adequately predict the discharge burnup; (2) estimate the burnup dependent soluble boron concentration and; (3) estimate the reactivity coefficients; all of the above for a multi-batch core. In the second part of this work a detailed neutronic analysis is carried out for the six most economical geometries of both oxide and hydride fuels, with the purpose of designing the U-ZrH_1.6 fueled PWR cores to have negative reactivity coefficients. The preferred design found is replacement of 25 ^v/_o of the ZrH_1.6 by thorium hydride, along with addition of some IFBA burnable poison. It is also found that the conversion from oxide to hydride fueled PWR cores could be done without modifications in the control system.     

Thermal hydraulic design of a hydride-fueled inverted PWR core

An inverted PWR core design utilizing U(45%, w/o)ZrH_1.6 fuel (here referred to as U-ZrH_1.6) is proposed and its thermal hydraulic performance is compared to that of a standard rod bundle core design also fueled with U-ZrH_1.6. The inverted design features circular cooling channels surrounded by prisms of fuel. Hence the relative position of coolant and fuel is inverted with respect to the standard rod bundle design. Inverted core designs with and without twisted tape inserts, used to enhance critical heat flux, were analyzed. It was found that higher power and longer cycle length can be concurrently achieved by the inverted core with twisted tape relative to the optimal standard core, provided that higher core pressure drop can be accommodated. The optimal power of the inverted design with twisted tape is 6869 MW_t, which is 135% of the optimally powered standard design (5080 MW_t—determined herein). Uncertainties in this design regarding fuel and clad dimensions needed to accommodate mechanical loads and fuel swelling are presented. If mechanical and neutronic feasibility of these designs can be confirmed, these thermal assessments imply significant economic advantages for inverted core designs.     

CFD analysis on subcooled boiling phenomena in PWR coolant channel

Eulerian two-fluid model coupled with wall boiling model was employed to calculate the three dimensional flow field and heat transfer characteristics in a hot channel with vaned spacer grid in PWR. The heat transfer from pellet-gap-cladding to coolant was also taken into account by a system coupled code MpCCI. The wall boiling model utilized in this study was validated by Bartolomei experiment data, and a good agreement can be observed. By solving the governing equation in a two-way coupled method, the distribution of temperature in the pellet-gap-cladding region and the distribution of temperature, void fraction and velocity of two-phase flow in coolant channel can be obtained. The influences of spacer grid and mixing vane on the thermal-hydraulic characteristics were analyzed. The heat transfer capacity was strongly improved by the spacer grid and mixing vane, while the flow resistance was also enlarged. Localized volume fraction of vapor phase decreased due to mixing vane, which will decrease the possibility of the departure from nucleate boiling (DNB) and increase the critical heat flux (CHF). By analyzing the temperature and void fraction at cladding outer surface, the critical regions where hot spot may occur were determined.     

Flow Characteristics in Spacer Grids Measured by Rod-embedded Fiber Laser Doppler Velocimetry

Precise measurement of velocity in fuel bundles is required to improve the thermal-hydraulic properties of Pressurerized Water Reactor (PWR) spacer grids. To better understand the cross-flow characteristics in rod bundles for developing spacer grids, we used the rod-embedded fiber laser Doppler velocimetry (rod LDV) to measure the flow velocities inside the spacer grid flow channels. As the result of measurement, we found that the flow distribution inside the spacer grid depends on the local flow resistance of the grid straps and is clearly affected by the presence of a mixing vane. We also clarified the relationship between cross-flow velocity in the fuel bundle downstream of the spacer grid and the axial flow inside the spacer grid.     

A small long-cycle PWR core design concept using fully ceramic micro-encapsulated (FCM) and UO2–ThO2 fuels for burning of TRU

In this paper, a new small pressurized water reactor (PWR) core design concept using fully ceramic micro-encapsulated (FCM) particle fuels and UO₂–ThO₂ fuels was studied for effective burning of transuranics from a view point of core neutronics. The core of this concept rate is 100 MWe. The core designs use the current PWR-proven technologies except for a mixed use of the FCM and UO₂–ThO₂ fuel pins of low-enriched uranium. The significant burning of TRU is achieved with tri-isotropic particle fuels of FCM fuel pins, and the ThO₂–UO₂ fuel pins are employed to achieve long-cycle length of ～4 EFPYs (effective full-power year). Also, the effects of several candidate materials for reflector are analyzed in terms of core neutronics because the small core size leads to high sensitivity of reflector material on the cycle length. The final cores having 10 w/o SS303 and 90 w/o graphite reflector are shown to have high TRU burning rates of 33%–35% in FCM pins and significant net burning rates of 24%–25% in the total core with negative reactivity coefficients, low power peaking factors, and sufficient shutdown margins of control rods.     

ADAGIO technique: From UO2 fuels to MOX fuels

The amount of gas at the grain boundaries plays an important role in the fuel transient behaviour during accident conditions, such as a loss-of-coolant accident (LOCA) or a reactivity-initiated accident (RIA). Direct experimental determination of the grain boundary gas inventory has been performed for MOX fuel irradiated in an EDF pressurised water reactor (PWR) using the ADAGIO technique (ADAGIO is a French acronym meaning ‘Discriminatory Analysis of Accumulated Inter-granular and Occluded Gas’). The ADAGIO protocol applied to a MOX MIMAS fuel produced inter-granular gas fraction results that were consistent with those reached with other methods of evaluation i.e. electron probe microanalysis (EPMA). Furthermore, a new methodology for the numerical treatment of ⁸⁵Kr release kinetics which was developed for UO₂ was applied to MOX fuels. The corresponding results evidenced two types of release kinetics. These kinetics were attributed to the inter-granular bubbles of the UO₂ matrix and the bubbles located in the restructured zones, i.e. Pu agglomerates.     

Experimental research progress on critical heat flux of Chinese PWR

One of the most important requirements in the design of pressurized water reactor (PWR) is to avoid the occurrence of critical heat flux (CHF). The design criteria for PWR specify that they must be operated at a certain percentage below CHF at all times and locations so as to the cladding temperature of fuel element at safe values. So in the process of safety assessment, CHF is one of important thermal-hydraulic parameters limiting the available power, whose size directly affects safety and economy of PWR nuclear power plant. This paper deals with a summary of experimental research progress on CHF of Chinese PWR. It mainly presents CHF experimental researches of Φ10 fuel assembly, CHF experimental researches of standard fuel assembly, and CHF experimental progress of non-uniform heated rod bundles. It should be emphasized that it also presents experimental research programs on CHF of Chinese advanced fuel assembly with self-reliance copyright. All CHF data obtained will be used for design improvement of Chinese PWR and R&D program of New Generation 1000 MWe PWR.     

Single-phase CFD applicability for estimating fluid hot-spot locations in a 5 × 5 fuel rod bundle

High-thermal performance PWR spacer grids require both of low pressure loss and high critical heat flux (CHF) properties. Therefore, a numerical study using computational fluid dynamics (CFD) was carried out to estimate pressure loss in strap and mixing vane structures. Moreover, a CFD simulation under single-phase flow condition was conducted for one specific condition in a water departure from nucleate boiling (DNB) test to examine the applicability of the CFD model for predicting the CHF rod position. Energy flux around the rod surface in a water DNB test is the sum of the intrinsic energy flux from a rod and the extrinsic energy flux from other rods, and increments of the enthalpy and decrements of flow velocity near the rod surface are assumed to affect CHF performance. CFD makes it possible to model the complicated flow field consisting of a spacer grid and a rod bundle and evaluate the local velocity and enthalpy distribution around the rod surface, which are assumed to determine the initial conditions for the two-phase structure. The results of this study indicate that single-phase CFD can play a significant role in designing PWR spacer grids for improved CHF performance.     

压水堆棒束定位格架两相搅混特性数值研究

本文提出一种CFD方法用于评价压水堆燃料棒束定位格架两相搅混特性。针对两种典型的定位格架,采用CFX12.0进行了空气-水两相流动的数值模拟,并与采用氟里昂工质开展的临界热流密度（CHF）实验进行对比。结果表明,CFD方法可初步应用于评价格架下游汽泡的分布特性。     

燃料组件精细化定位格架模型开发及评价

为提高燃料组件子通道内两相局部参数预测的准确性,本文基于分布式阻力方法建立精细化定位格架模型,选用合适的摩擦阻力表达式,对格架上的交混翼进行精细化建模,采用Carlucci湍流交混模型计算湍流交混速率,引入阻塞因子计算由定位格架引起的湍流交混效应,并将建立的精细化定位格架模型植入子通道分析程序（ATHAS）,对压水堆子通道和棒束实验（PSBT）基准题进行计算分析。结果表明,本文开发的精细化定位格架模型能够提高燃料组件子通道内空泡份额和温度分布的预测准确性,为棒束通道流场、焓场计算和临界热流密度（CHF）预测奠定了基础。      

Core loss during a severe accident (COLOSS)

The COLOSS project was a 3-year shared-cost action, which started in February 2000. The work-programme performed by 19 partners was shaped around complementary activities aimed at improving severe accident codes. Unresolved risk-relevant issues regarding H₂ production, melt generation and the source term were studied through a large number of experiments such as (a) dissolution of fresh and high burn-up UO₂ and MOX by molten Zircaloy, (b) simultaneous dissolution of UO₂ and ZrO₂, (c) oxidation of U–O–Zr mixtures, (d) degradation–oxidation of B₄C control rods.Corresponding models were developed and implemented in severe accident computer codes. Upgraded codes were then used to apply results in plant calculations and evaluate their consequences on key severe accident sequences in different plants involving B₄C control rods and in the TMI-2 accident.Significant results have been produced from separate-effects, semi-global and large-scale tests on COLOSS topics enabling the development and validation of models and the improvement of some severe accident codes. Break-throughs were achieved on some issues for which more data are needed for consolidation of the modelling in particular on burn-up effects on UO₂ and MOX dissolution and oxidation of U–O–Zr and B₄C–metal mixtures. There was experimental evidence that the oxidation of these mixtures can contribute significantly to the large H₂ production observed during the reflooding of degraded cores under severe accident conditions.The plant calculation activity enabled (a) the assessment of codes to calculate core degradation with the identification of main uncertainties and needs for short-term developments and (b) the identification of safety implications of new results.Main results and recommendations for future R&D activities are summarized in this paper.     

Modeling for Void Reactivity Effect by Subcooled Boiling under Reactivity Initiated Accident Condition of BWR

A simple mechanistic model is presented to evaluate the subcooled void reactivity effect under a Reactivity Initiated Accident (RIA) at cold critical condition of BWR. This model consists of a drift flux model for vapor velocity and a vapor mass conservation model with a term of vapor source on a heated wall, and it was incorporated into a homogeneous and equilibrium thermal-hydraulic code EUREKA-JINS. A sample analysis by this model showed that the subcooled void reactivity effect leads to reduction of the maximum fuel enthalpy by about 20 cal/g UO₂ in the case of RIA at cold critical condition. Though the reduced value is dependent on the reactor core condition, this result indicates the significance of subcooled void reactivity effect in the accident, while the effect can be neglected in the hot stand-by case where, at most, only 4 cal/g UO₂ is reduced for the maximum fuel enthalpy.     

Development of critical heat flux correlation for In-vessel retention

ABSTRACT

In-vessel retention (IVR) is a strategy for severe accident management in which the lower head of the reactor vessel is submerged in a water-flooded reactor cavity. Critical heat flux (CHF) data for IVR are important for estimating cooling capacity of the reactor vessel. The existing CHF data for IVR which were obtained for the specific geometries and thermal-hydraulic conditions of actual plants are difficult to be applied to plants with other specifications. Hence, the purpose of this study is to develop CHF correlations applicable to various pressurized water reactor plants in a wide range of thermal outputs based on newly obtained CHF data. A rectangular test section with a cross-section of 150 mm × 150 mm and length of 600 mm was used for simulating a cooling channel. The thermal-hydraulic conditions expected in actual plants were studied, and the results were used in the experiment. The effects of parameters such as pressure, mass flux, thermodynamic quality, and angle on CHF were investigated . Based on these results, we developed a CHF correlation formula that can be applied to a wider range than previously, up to a maximum heat flux of 3000 kW/m², and that predicts CHF with an error of ± 10%.     

Development of enriched Gd-155 and Gd-157 burnable poison designs for a PWR core

In this study, a genetic algorithm developed by the authors was applied to design the optimal enriched Gd-155 and Gd-157 burnable poisons in a reference PWR TMI-1 core. The CASMO-4/TABLES/SIMULATE-3 package calculated the neutronic performance of the enriched UO₂/Gd₂O₃ fuel pin configurations. These configurations included different fractions of neutron absorbing isotopes Gd-155 and Gd-157, and 100 w/o enriched Gd-155 designs. Fuel cost analysis was performed to evaluate the economical benefits of these optimized enriched gadolinium designs. The break-even point for unit Gd-155 enrichment cost was determined to be around ∼$30/gram-Gd-155 with current unit cost scenario. The projected savings were 3.13% in gross and 2.08% in net compared to total fuel cycle cost of a reference TMI-1 core loading, if all of the 68 feed assemblies would be replaced with the optimized designs.     

A critical heat flux approach for square rod bundles using the 1995 Groeneveld CHF table and bundle data of heat transfer research facility

The critical heat flux (CHF) approach using CHF look-up tables has become a widely accepted CHF prediction technique. In these approaches, the CHF tables are developed based mostly on the data bank for flow in circular tubes. A set of correction factors was proposed by Groeneveld et al. [Groeneveld, D.C., Cheng, S.C., Doan, T., 1986. 1986 AECL-UO Critical Heat Flux lookup table. Heat Transf. Eng. 7(1–2), 46] to extend the application of the CHF table to other flow situations including flow in rod bundles. The proposed correction factors are based on a limited amount of data not specified in the original paper. The CHF approach of Groeneveld and co-workers is extensively used in the thermal hydraulic analysis of nuclear reactors. In 1996, Groeneveld et al. proposed a new CHF table to predict CHF in circular tubes [Groeneveld, D.C., et al., 1996. The 1995 look-up table for Critical Heat Flux. Nucl. Eng. Des. 163(1), 23]. In the present study, a set of correction factors is developed to extend the applicability of the new CHF table to flow in rod bundles of square array. The correction factors are developed by minimizing the statistical parameters of the ratio of the measured and predicted bundle CHF data from the Heat Transfer Research Facility. The proposed correction factors include: the hydraulic diameter factor (K_hy), the bundle factor (K_bf), the heated length factor (K_hl), the grid spacer factor (K_sp), the axial flux distribution factors (K_nu), the cold wall factor (K_cw) and the radial power distribution factor (K_rp). The value of constants in these correction factors is different when the heat balance method (HBM) and direct substitution method (DSM) are adopted to predict the experimental results of HTRF. With the 1995 Groeneveld CHF Table and the proposed correction factors, the average relative error is 0.1 and 0.0% for HBM and DSM, respectively, and the root mean square (RMS) error is 31.7% in DSM and 17.7% in HBM for 9852 square array data points of HTRF.     

Analysis of Mixed Oxide Fuel Critical Experiments with Neutronics ANalysis Codes for Boiling Water Reactors

Critical experiments of UO₂ and full mixed oxide (MOX) fuel cores conducted at the Tank-type Critical Assembly (TCA) were aNalyzed using BWR design-purpose codes HINES and CERES with ENDF/B files and Monte Carlo fine analysis codes VMONT and MVP with the JENDL-3.2 library.

The averaged values of the multiplication factors calculated with HINES/CERES, VMONT and MVP agreed with those of experiments within 0.3%ΔAk. The values by the design-purpose codes showed a small difference of 0.1%Δk between UO₂ and MOX cores. Monte Carlo code results showed that the JENDL-3.2 library had a tendency to overestimate the multiplication factors of UO₂ cores by about 0.3%Δk compared with those values of MOX cores. The root mean square errors of calculated power distributions were less than 1% for HINES/CERES and VMONT.

These results showed that (1) the accuracy of these codes when applied to full MOX cores was almost the same as their accuracy for UO₂ cores, which confirmed the accuracy of present core design codes for full MOX cores; and (2) the accuracy of the 190-energy-group Monte Carlo calculation code VMONT was almost the same as that of the continuous-energy Monte Carlo calculation code MVP.     

Neutronic and thermal-hydraulic analysis of alternative ceramic fuels in the next-generation of light water reactors

The Burn-Up enlargement is one of the most important issues in the nuclear reactor core fuel management. In recent years some reactor design companies have focused on the reactor cycle length enlargement in next generation of pressurized water reactors. An increased cycle length results in an increased fuel burn-up which directly leads to low electricity costs and more efficiency. One of the promising issues is to change the chemical state of fuel that is on the agenda of the Mitsubishi Company as US-APWR nuclear power plant designer. In the present study, the neutronic as well as thermal-hydraulic analysis of some commercial ceramic fuels such as UN, UC, and UN¹⁵ instead of conventional UO₂ have been studied. The sub-channel analysis approach has been selected for these investigations. In this regard, a US-APWR fuel assembly was modelled using MCNPX2.6 Monte Carlo code by considering the periodic boundary condition in X–Y directions. It was found that the use of UC and UN¹⁵ instead of UO₂ has a deep effect on the reactor cycle length such that the power plant operational time was increased by a factor of 1.5. The COBRA-EN code with modified MATPRO subroutine has been used in thermal-hydraulic tasks. Since the thermal conductivity of these selected fuels is six times greater than UO₂, the thermal-hydraulic analysis of candidate fuels was led to outstanding results. It was found that the fuel centerline temperature in UN¹⁵ and UC cases are about half of UO₂ one, which is drastically beneficial. In summary the thermal power of next generation of pressurized water reactors could be increased considerably by using the candidate ceramic fuels instead of conventional UO₂ one.     

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.