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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...
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... A typical pin-wise peaking factor for oxide fuel is around 1.11. The scope of the BWR analysis was limited; neutronic and thermal hydraulic analyses were not consistently coupled, fuel rod vibration analysis was not performed in detail, while transient and fuel rod mechanical integrity analysis were not performed. Moreover, core hydrodynamic stability performance was not explicitly calculated; rather, the susceptibility of the cores analyzed to instability phenomena was qualitatively limited by constraining the core average exit quality (in the whole-core analysis) and the hot bundle average exit quality (in the single-bundle analysis) to the values of the reference BWR. This does not guarantee avoidance of instability phenomena; however, it prevents operation at high-quality conditions, where BWRs are more susceptible to hydrodynamic instabilities. The neutronic and thermal hydraulic analyses were coupled indirectly as follows: (1) Core axial and radial power distributions were assumed to be the same for oxide and hydride cores and equal to those of a typical oxide-fueled BWR. It appears reasonable that hydride-fueled BWRs can be designed to have similar power distribution as of oxide cores. (2) The in-bundle pin power distribution was calculated using accurate 3D neutronic analysis. (3) The void fraction axial distribution of a typical oxide-fueled BWR was used for the neutronic analysis of both oxide and hydride cores. Again, this appears a reasonable assumption because the hydride- fueled core can be designed to have a similar axial void distribution. These assumptions enabled comparison of the power density and other characteristics expected from the hydride-fueled BWR relative to those of the oxide-fueled BWR based on the comparison of single fuel bundles of identical outer dimensions. Nevertheless, the conclusions of the BWR study are only indications of possible per- formance gains; more detailed consistent analysis needs to be done before firm conclusions can be drawn. The BWR work consisted of neutronic, thermal hydraulic, and economic analyses. The objective of the neutronic analysis was to identify the acceptable combinations of fuel rod outer diameter, D , and the square lattice pitch to diameter ratio, P / D – referred to as “geometry”, of hydride as well as oxide fuels and to quan- tify the attainable discharge burnup. To be acceptable a geometry must have negative fuel and CTCs of reactivity as well as negative void reactivity feedback throughout the core life. The objective of the thermal hydraulic analysis was to estimate the maximum power density attainable using different geometries, both oxide- and hydride-fueled, subjected to a number of design constraints. The objective of the economic analysis was to use the results from the neutronic and thermal hydraulic analyses to estimate the COE attainable from BWRs designed with hydride fuel versus oxide- fueled BWRs. The neutronic, thermal hydraulic, and economic analyses, which are discussed in detail in the specific papers (Fratoni et al. (this issue); Ferroni et al. (this issue); Ganda et al. (this issue-b)), are summarized in, respectively, Section 7.4, Section 7.5, and Section 7.6. Section 7.3 provides a brief description of the BWR core chosen as reference. The BWR/5 and the 9 × 9 fuel bundle of Fig. 28 are used as the reference reactor and reference bundle, respectively. Table 14 summarizes key parameters of the reference reactor (Ferroni et al., this issue). The fact that the power density of the oxide fuel bundle and core selected as reference is low relative to more advanced BWR designs (loaded with 10 × 10 bundles) does not affect the comparison between hydride and oxide fuels performed in this work, as we are searching for maximum power density oxide and hydride bundle designs using the same set of assumptions, constraints, and methodology. The approach adopted for this study is to first estimate an upper bound to the possible power density gain relative to the reference oxide fuel bundle design shown in Fig. 28. This is done by exam- ining the “Idealized” bundle design shown in Fig. 29. It features the minimum feasible space in-between the fuel bundles and the most uniform hydride fuel bundle concept possible. 11 The reactivity control is provided in this idealized design using control rods inside the bundle. Then two practical hydride fuel bundle designs are studied; the corner cruciform control rods (CCCR) design and the control blade (CB) design. They were conceived to minimize the space occupied by the water gaps while providing space for instrumentation tubes and avoiding the design challenge of control elements insertion inside the bundle. The CCCR design features truncated cruciform-shaped control rods at the bundle corners outside the bundle box (Fig. 30), while the CB design uses more conventional cruciform CBs in-between the fuel bundles; however, the water gap between bundles is minimized on two sides of the bundle (Fig. 31). Both the CCCR and the CB designs are intended for newly built BWRs. The bundle pitch of all hydride bundles considered is the same as of the reference oxide bundle shown in Fig. 28. Relative to the 71 effective full-length fuel rods of the reference 9 × 9 oxide fuel bundle, the hydride fuel bundles examined have, in the order presented, 96, 93, and 100 full-length fuel rods. The first two hydride fuel bundles have a similar, while the third bundle has a slightly smaller fuel rod diameter than of the reference oxide fuel bundle. The performance of these hydride fuel bundles was also compared against that of a 10 × 10 oxide fuel bundle that has thinner fuel rods. The neutronic feasibility study consists of three parts, all involv- ing a 3D fuel bundle analysis. The first part is a scoping analysis that covers a limited number of fuel rod outer diameters, D , for a given pitch, P . The objective of this scoping analysis is to identify the geometry, i.e., D–P combination that offers the maximum achievable burnup. The second part of the study is a detailed neutronic analysis of this maximum burnup fuel bundle as well as of a bundle offering a larger power level identified in the companion study (Ferroni et al., this issue). All the above studies examined the “Idealized” hydride fuel bundle concept. The last part of the study examines the two alternative (CCCR and CB) hydride fuel bundle designs that are more practical to implement. The 3D neutronic analysis was performed using the MOCUP code system accounting for a typical axial water density distribution. Twenty-four depletion zones were considered for the reference oxide bundle, corresponding to eight groups of fuel rods and three average axial enrichments per group. Being of significantly more uniform design, only nine depletion zones were considered for hydride fuel bundles – three equal length axial and three radial zones. A four-batch fuel management scheme was assumed for estimating the discharge burnup, core average, k , and reactivity coefficients. The statistical uncertainty in calculating k was <5 × 10 − 4 such that, after propagation through the k averaging procedure, the uncertainty in the core average k was <2 × 10 − 3 . The thermal hydraulic study consisted of two independent analyses: a whole core analysis, performed for both oxide and hydride fuel over 400 geometries, i.e., 400 D – P / D combinations, and a single bundle analysis, performed in greater detail on a limited set of oxide and hydride fuel bundles. Matlab and the VIPRE-EPRI code were coupled to perform the whole-core analysis while the single- bundle analysis needed use of the VIPRE code only. The whole-core analysis was mainly a scoping study, since the large range of bundle geometries examined required that simplifying assumptions be made. Among them, the assumption of the same bundle-wise pin power peaking factor, regardless of the fuel type and bundle geometry, is the most conservative. While the power density results derived from this analysis are consequently approximated, this analysis gave important insight about the range of bundle geometries that promise maximum power. In the single-bundle analysis, in which the bundle modeling was performed with greater detail of geometric and rod power distribution than that characterizing the whole core analysis, the performance of two oxide fuel bundles were compared to those of four hydride fuel bundles three of which were introduced in Section 7.4. Table 15 summarizes key geometric characteristics of the six bundles examined. The oxide bundles were the reference 9 × 9 bundle of Fig. 28, representing the GE11 design, and a 10 × 10 bundle resembling the GE14 design. Both are for a Backfit core configuration. The hydride bundles examined were: For Backfit core configuration: - 9 × 9 bundle: it represents the least challenging way to retrofit the reference core since D and P are the same as in the reference GE11 ...
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Citations
... The proposed Hyperion Power Reactor [16] and MARVEL Research Microreactor designed by Idaho National Laboratory, USA [11] also use hydride fuel. Several studies have proposed the use of uranium hydride as an LWR fuel [17,18]. The use of hydride fuel in nuclear power reactors poses another interesting characteristic. ...
The GAMA Microreactor is a low-power nuclear power plant dedicated for remote area with a power output of 300 kWe. The primary characteristics of the GAMA Microreactor are its system compactness, design simplicity, and safe operation without moving parts. The fuel is uranium hydride powder, which works simultaneously as a moderator, contained within a stainless-steel vessel and surrounded by a graphite reflector. Liquid sodium-filled heat pipes are embedded in the fuel powder to extract fission-generated heat via passive coolant flow due to capillary action. As a part of the design process, neutronic parameters of the GAMA Microreactor have been determined using SCALE6.2 and OpenMC codes. Among the analyzed parameters are the excess reactivity as a function of burn-up, shutdown margin, control rod worth, temperature coefficient of reactivity (TCR), fuel density coefficient of reactivity (DCR), and reactivity coefficient due to hydrogen dissociation. From the calculation results, the GAMA Microreactor has a maximum excess reactivity of 0.049 at the operational temperature (700 °C) and 0.0788 at ambient temperature. The TCR value is negative (−2.267 pcm/°C) at the beginning of the cycle (BOL) and −4.9618 pcm/°C at the end of the cycle (EOC). Hydrogen dissociation imposes negative reactivity on the reactor both at the BOL (−0.2631 Δk/k-%dissociation) and EOC (−0.3524 Δk/k-%dissociation). Therefore, the GAMA Microreactor maintains its inherent safety characteristics over a considerably long operation time.
... The U-Zr hydride structure used in the TRIGA fuel has 1.6 hydrogen atoms ratio over Zr atom, i.e., it is U-ZrH1.6. Greenspan el ta [19,20] were proposed a new combination of hydride fuel of U-PuH2-ZrH1.6 (PUZH fuel) compared MOX fuel and the results showed that PUZH achieved more than double higher of average discharge burnup than MOX fuel. Four reactivity coefficients of PUZH has been calculated and were all found to be negative. ...
... The feasibility of using solid hydrides fuel to improve the performance of LWRs core has been reported [20,19] where the primary hydride fuel U-ZrH1.6 TRIGA type fuel having 45 wt% U and proposing a new combination of hydride fuel of U-PuH2-ZrH1.6 (PUZH fuel) are considered to investigate compared with MOX fuel for PWR and BWR system. Hydrogen in hydride fuel needed for neutron moderation within fuel volume allows to achieve optimal neutron spectrum while using smaller water volume. ...
... Hydrogen in hydride fuel needed for neutron moderation within fuel volume allows to achieve optimal neutron spectrum while using smaller water volume. This feature allows to design core with optimal moderation, in term of the attainable discharge burnup and to have a large number of fuel rods per unit core volume than a LWR core that uses oxide fuel rods [20]. Greenspan el ta [20] acknowledged that using hydride fuel would provide the two unique safety-related fuel temperature reactivity feedback mechanism; ...
Due to widely increasing of nuclear technology, the exposed to ionizing radiation is inevitable. Therefore, a shield is needed to absorb and reduce the intensity of any kind of radiation. To achieve the enhanced shielding capability, composite and multilayer have been proposed. In this study, multilayer configuration is designed by utilizing the metal hydride in the front layer and heavy Z-number metal to shield neutron and gamma-ray simultaneously. The multilayer configuration is designed from 2-, 4-, 6-, and 8-layer. To assessment the multilayer configuration analysis, the single layer analysis of metal hydride and heavy metal is conducted against the particle energy and material thickness change in order to select the potential candidate material for metal hydride and heavy Z-number metal. To analyze and obtain the particle flux, MCNPs is chosen and adopted for the whole simulation. The results of single layer analysis are suggested three main points, (1) metal hydrides contain the highest hydrogen atoms worked effectively against neutron at higher energy, (2) heavy metals have the high density worked effectively against gamma-rays, (3) displacement per atom is analyzed to estimate the lifetime of shielding material.
For multilayer configuration analysis results show that TaH5 coupled with either W, Ta, Hf, Pb give a better performance than other selected metal hydrides. However, only TaH5 coupled with either W or Ta in multiplayer configuration against neutron enhanced better shielding capability when the paired increases up to 6-layers. The optimization quantification of simulated annealing is also carried out to quantify shielding results in different multilayer configuration models.
In conclusion, the application of metal hydride coupled with heavy metal in multilayer configuration could be absolutely used for shielding application in the nuclear industry. The application such as shielding transportation dry storage cask of nuclear spent fuel can be applicable
to ensure the safety of any leakage radiation from the cask system which will be leave for future works.
Keywords: Metal hydride, Multilayer configuration, Shielding, Neutron, Gamma, Optimization, Simulated annealing, MCNP, DPA
... U-ZrH 1.6 fuel developed by Dr. Massoud Simnad is being successfully used for over 50 years in TRIGA type research reactors around the world with no safety problems [18][19][20]. Six hydride-fueled space reactors were built and operated, and one was placed in earth orbit for the Space Nuclear Auxiliary Power (SNAP) project [21]. ...
... The BWR reference assembly model is based on the GE14 10Â10 design. Key parameters are summarized in Table 2 ( Greenspan et al., 2009;Fensin, 2004;ABWR, 2006). Such assembly contains 78 full length fuel rods, 14 partial length fuel rods (PLFR), and 2 water rods ( Greenspan et al., 2009). ...
... Key parameters are summarized in Table 2 ( Greenspan et al., 2009;Fensin, 2004;ABWR, 2006). Such assembly contains 78 full length fuel rods, 14 partial length fuel rods (PLFR), and 2 water rods ( Greenspan et al., 2009). Four different axial regions were used to model uranium enrichment and gadolinium distri- butions ( Maldonado et al., 2010): (1) . ...
In severe accident conditions with loss of active cooling in the core, zirconium alloys, used as fuel cladding materials for current light water reactors (LWR), undergo a rapid oxidation by high temperature steam with consequent hydrogen generation. Novel fuel technologies, named accident tolerant fuels (ATF), seek to improve the endurance of severe accident conditions in LWRs by eliminating or at least mitigating such detrimental steam-cladding interaction. Most ATF concepts are expected to work within the design framework of current and future light water reactors, and for that reason they must match or exceed the performance of conventional fuel in normal conditions. This study analyzed the neutronic performance of ATF when employed in both pressurized and boiling water reactors. Two concepts were evaluated: (1) coating the exterior of zirconium-alloy cladding with thin ceramics to limit the zirconium available for reaction with higherature steam; (2) replacing zirconium alloys with alternative materials possessing slower oxidation kinetics and reduced hydrogen production. Findings show that ceramic coatings should remain 10-30 μm thick to limit the neutronic penalty. Alternative cladding materials, with the exception of SiC, enhance neutron loss compared to zirconium-alloys. An extensive parametric analysis concluded that reference performance metrics can be met by employing 300-μm or less thick cladding or increasing fuel enrichment by up to 1.74% depending on material and geometry.
... General Atomics Technologies Inc. has developed for its TRIGA research reactor the hydride fuel U-ZrH 1.6 which has been successfully used for several years [2]. A recent US study [3,4] has investigated the feasibility of the use of hydride fuels instead of commonly used oxide fuels in light water reactors to improve performance and safety. The study found that hydride fuels could safely operate in LWRs and among others could be employed to reduce undermoderation in BWR. ...
... The steady-state fuel temperature for normal operation is limited to 750 ∘ C [2] which could be a serious limitation for use in HPLWR. The recent US study [3] has established that the TRIGA type fuel U-ZrH 1.6 can safely operate in LWR at comparable linear heat generation rate level (∼50 kW/m) to that attainable with oxide fuel and higher discharge burnup (∼ 80 MWd/kg). In general terms hydride fuel application needs to be verified in the operating conditions of HPLWR. ...
Hydride fuels have features which could make their use attractive in future advanced power reactors. The potential benefit of use of hydride fuel in HPLWR without introducing significant modification in the current core design concept of the high-performance light water reactor (HPLWR) has been evaluated. Neutronics and thermal hydraulic analyses were performed for a single assembly model of HPLWR with oxide and hydride fuels. The hydride assembly shows higher moderation with softer neutron spectrum and slightly more uniform axial power distribution. It achieves a cycle length of 18 months with sufficient excess reactivity. At Beginning of Cycle the fuel temperature coefficient of the hydride assembly is higher whereas the moderator and void coefficients are lower. The thermal hydraulic results show that the achievable fuel temperature in the hydride assembly is well below the design limits. The potential benefits of the use of hydride fuel in the current design of the HPLWR with the achieved improvements in the core neutronics characteristics are not sufficient to justify the replacement of the oxide fuel. Therefore for a final evaluation of the use of hydride fuels in HPLWR concepts additional studies which include modification of subassembly and core layout designs are required.
... U-ZrH 1.6 fuel developed by Dr. Massoud Simnad has being successfully used for over 40 years in TRIGA type research reactors around the world with no safety problems [7][8][9]. Six hydride-fueled space reactors were built and operated, and one was placed in earth orbit for the Space Nuclear Auxiliary Power (SNAP) project [10]. It has been suggested that U-ThH 2 fuel is even more stable than U-ZrH 1.6 fuel and can operate at higher temperatures [11]. ...
... The findings of the neutronics and thermal hydraulics analyses of BWR cores loaded with hydride fuel are reported in the companion papers [4, 5], respectively. Properties of the UZrH 1.6 hydride fuel examined in these studies are summarized in Ref. [6]. The present paper summarizes an analysis of the economic implications of substituting hydride-based fuels for conventional oxide fuels in BWR reactors. ...
This paper presents a design of an advanced boiling water reactor (ABWR) equilibrium bundle using UZrH 1.6 as a fuel. Monte Carlo N-Particle Transport (MCNPX) code extended has been used for design advanced boiling water reactor model. MCNPX code based on Mont Carlo Method is used to design a three dimensional model for a BWR fuel assembly in a typical operating temperature and pressure conditions. A test case was compared with a benchmark problem, and good agreement was found. This model is used to calculate the distribution of pin by pin power and flux inside the assembly, calculate the axial thermal neutron flux and the axial normalized power across the fuel rod at different insertion of control rod, and to study the effect of increasing the fuel enrichment on the axial thermal neutron flux and the axial normalized power.
... Among these are uranium silicide, uranium nitride and uranium carbide for which some irradiation experience exists, but not in light water reactors. Greenspan et al. (2009) studied the suitability of hydride fuel in the DOE NERI Program sponsored project '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 and boiling water reactor cores by using solid hydride fuels instead of the commonly used oxide fuel. ...
Fuel assemblies and their components are subjected to the most harsh conditions existing in a nuclear reactor. They are designed and manufactured to satisfy stringent functional and safety requirements for normal operation and transient conditions. With an emphasis on boiling and pressurised light water reactors and going from general principles to details, the chapter describes the main components of a fuel assembly and how design and functionality are related to operating conditions, safety criteria and reactor physics. Phenomena affecting fuel rod endurance are addressed and illustrated.
... B-11 as well as carbide is known as efficient moderator for neutrons. Uranium zirconium hydride (UZrH 1.6 ) is used in TRIGA reactors [ 17] and is currently under investigation for the use as advanced LWR fuel [ 18]. There is large operational experience with this material in low power use. ...
This work shows the effect of the use of moderating layers on the sodium void effect in sodium cooled fast breeder reactors. The moderating layers consisting of either boron carbide B4C or uranium–zirconium hydride UZrH cause a strong reduction of the sodium void effect. Additionally these layers improve the fuel temperature effect and the coolant effect of the system. The use of the UZrH is significantly more effective for the reduction of the sodium void effect as well as for the improvement of the fuel temperature and the coolant effect. All changes cause by the insertion of the UZrH layer cause a significantly increased stability of the fast reactor system against transients. The moderating layers have only a small influence on the breeding effect and on the production of minor actinides.
... Although nuclear power plants have been used for many years, scientists are seeking methods to produce optimized and economic power in nuclear power plants. Therefore, a group of scientists from University of California, and Massachusetts Institute of Technology, with financial and scientific support from Westinghouse Company, carried out a comprehensive research on hydride fuel, and decided to establish power nuclear reactors and research reactors using this type of fuel [2][3][4]. ...
... In order to decrease central temperature in hydride fuel rods, helium has been replaced by the liquid metal (Pb-Bi-Sn) in the gaps. For more information on hydride fuel properties, see [1,2]. ...
This research presents neutronic calculations for heavy water research reactor core substituting hydride fuel for uranium dioxide fuel. The aim of this research is feasibility analysis of reactor utilization with its original design using a new proposed fuel and changing the coolant and moderator circuit to light water. The required group constants for the CITATION code will be calculated using WIMSD-4 code. Neutronic calculations such as multiplication factors, radial and axial power peaking factor and fuel burn-up calculations are carried out by the CITATION code.
