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Parks Geoff - One of the best experts on this subject based on the ideXlab platform.
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Lattice benchmarking of deterministic, Monte Carlo and hybrid Monte Carlo Reactor Physics codes for the soluble-boron-free SMR cores
'Organisation for Economic Co-Operation and Development (OECD)', 2020Co-Authors: Alam Syed, Kumar D, Almutairi B, Goodwin C, Ridwan Tuhfatur, Parks GeoffAbstract:Since the use of deterministic transport code WIMS can significantly reduce the computational time compared to the Monte Carlo (MC) code Serpent and hybrid MC code MONK, one of the major objectives of this study is to observe whether deterministic code WIMS can provide accuracy in Reactor Physics calculations while comparing Serpent and MONK. Therefore, numerical benchmark calculations for a soluble-boron-free (SBF) small modular Reactor (SMR) assembly have been performed using the WIMS, Serpent and MONK. Although computationally different in nature, these codes can solve the neutronic transport equations and calculate the required neutronic parameters. A comparison in neutronic parameters between the three codes has been carried out using two types of candidate fuels: 15%235U enriched homogeneously mixed all-UO2 fuel and 18%235U enriched micro-heterogeneous ThO2-UO2 duplex fuel in a 2D fuel assembly model using a 13×13 arrangement. The eigenvalue/reactivity (k∞) and 2D assembly pin power distribution at different burnup states in the assembly depletion are compared using three candidate nuclear data files: ENDF/B- VII, JEF2.2 and JEF3.1. A good agreement in k∞ values was observed among the codes for both the candidate fuels. The differences in k∞ between the codes are ∼200 pcm when cross-sections based on the same nuclear data file are used. A higher difference (up to ∼450 pcm) in the k∞ values is observed among the codes using cross-sections based on different data files. Finally, it can be concluded from this study that the good agreement in the results between the codes found provides enhanced confidence that modeling of SBF, SMR propulsion core systems with micro-heterogeneous duplex fuel can be performed reliably using deterministic neutronics code WIMS, offering the advantage of less expensive computation than that of the MC Serpent and hybrid MC MONK codes
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Small modular Reactor core design for civil marine propulsion using micro-heterogeneous duplex fuel. Part II: whole-core analysis
'Organisation for Economic Co-Operation and Development (OECD)', 2019Co-Authors: Sb Alam, Kumar D, Almutairi B, Goodwin C, Ridwan Tuhfatur, Parks GeoffAbstract:In an e↵ort to de-carbonise commercial freight shipping, there is growing interest in the possibility of using nuclear propulsion systems. In this Reactor Physics study, we seek to design a soluble-boron-free (SBF) and low-enriched uranium (LEU) (
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Reactor Physics modelling of accident tolerant fuel for LWRs using answers codes
European Physical Journal N, 2016Co-Authors: B A Lindley, Parks Geoff, Kotlyar D, Jn Lillington, Petrovic BAbstract:The majority of nuclear Reactors operating in the world today and similarly the majority of near-term new build Reactors will be LWRs. These currently accommodate traditional Zr clad UO2/ PuO2 fuel designs which have an excellent performance record for normal operation and most transients. However, the events at Fukushima culminated in significant hydrogen production and hydrogen explosions, resulting from high temperature Zr/steam interaction following core uncovering for an extended period. These events have resulted in increased emphasis towards developing more accident tolerant fuels (ATFs)-clad systems, particularly for current and near-term build LWRs. R&D programmes are underway in the US and elsewhere to develop ATFs and the UK is engaging in these international programmes. Candidate advanced fuel materials include uranium nitride (UN) and uranium silicide (U3Si2). Candidate cladding materials include advanced stainless steel (FeCrAl) and silicon carbide. The UK has a long history in industrial fuel manufacture and fabrication for a wide range of Reactor systems including LWRs. This is supported by a national infrastructure to perform experimental and theoretical R&D in fuel performance, fuel transient behaviour and Reactor Physics. In this paper, an analysis of the Integral Inherently Safe LW R design (I2S-LWR), a Reactor concept developed by an international collaboration led by the Georgia Institute of Technology, within a U.S. DOE Nuclear Energy University Program (NEUP) Integrated Research Project (IRP) is considered. The analysis is performed using the ANSWERS Reactor Physics code WIMS and the EDF Energy core simulator PANTHER by researchers at the University of Cambridge. The I2S-LWR is an advanced 2850 MWt integral PWR with inherent safety features. In order to enhance the safety features, the baseline fuel and cladding materials that were chosen for the I2S- LWR design are U3Si2 and advanced stainless steel respectively. In addition, the I S-LWR design adopts an integral configuration and a fully passive emergency decay heat removal system to provide indefinite cooling capability for a class of accidents. This paper presents the equilibrium cycle core design and Reactor Physics behaviour of the I2S-LWR with U3Si2 and the advanced steel cladding. The results were obtained using the traditional two-stage approach, in which homogenized macroscopic cross-section sets were generated by WIMS and applied in a full 3D core solution with PANTHER. The results obtained with WIMS/PANTHER were compared against the Monte Carlo Serpent code developed by VTT and previously reported results for the I2S-LWR. The results were found to be in a good agreement (e.g. < 200 pcm in reactivity) among the compared codes, giving confidence that the WIMS/PANTHER Reactor Physics package can be reliably used in modelling LWRs with ATFs.This is the final version of the article. It first appeared from Springer via http://dx.doi.org/10.1051/epjn/201601
Go Chiba - One of the best experts on this subject based on the ideXlab platform.
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BURNUP SENSITIVITY CALCULATIONS WITH CBZ FOR LIGHT WATER Reactor ASSEMBLY PROBLEMS
'EDP Sciences', 2021Co-Authors: Go ChibaAbstract:Sensitivities of k∞ and nuclides number densities during nuclear fuel burnup with respect to nuclear data are calculated with a Reactor Physics code system CBZ. Sensitivity calculations are carried out with the depletion perturbation theory applicable to nuclear fuel assemblies including burnable absorbers. Numerical results are presented both for BWR and PWR assemblies, and those demonstrate usefulness and effectiveness of burnup sensitivity calculation capabilities for LWR fuel assemblies
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research activities on nuclear Reactor Physics and thermal hydraulics in japan after fukushima daiichi accident
Journal of Nuclear Science and Technology, 2018Co-Authors: Shuichiro Miwa, Yasunori Yamamoto, Go ChibaAbstract:Research and development in nuclear Reactor Physics and thermal-hydraulics continue to be vital parts of nuclear science and technology in Japan. The Fukushima accident not only brought tremendous change in public attitudes towards nuclear engineering and technology, but also had huge influence towards the research and development culture of scientific communities in Japan. After the Fukushima accident, thorough accident reviews were completed by independent committees, namely, Tokyo Electric Power Company (TEPCO), the Japanese government, the Diet of Japan, the Rebuild Japan Initiative Foundation, and the Nuclear and Industrial Safety Agency. Reactor Physics and thermal-hydraulics divisions of Atomic Energy Society of Japan (AESJ) also issued the roadmaps after the accident. As a result, lessons learned from the accident were made clear, and a number of new research activities were initiated. The present paper reviews ongoing nuclear engineering research activities in Japanese institutes, univers...
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advanced bondarenko method for resonance self shielding calculations in deterministic Reactor Physics code system cbz
Annals of Nuclear Energy, 2016Co-Authors: Go Chiba, Tadashi NarabayashiAbstract:Abstract The advanced Bondarenko method for resonance self-shielding calculations is devised and proposed. This method is based on three numerical methods; the Bell factor optimization for accurate fuel escape probability representation, extension of resonance interference factors and correction factors for current-weighted total cross sections. A 107-group library for light water Reactor applications based on the advanced Bondarenko method is generated for a Reactor Physics code system CBZ. Performance of the CBZ code with this 107-group library is examined against a suit of light water Reactor cell problems. The infinite neutron multiplication factors calculated with CBZ agree with reference continuous-energy Monte Carlo solutions within 0.15 % Δ k / kk ′ differences, and no significant biases on fuel compositions and geometrical specifications are observed. Energy-averaged cross sections are also examined. Numerical tests reveal that significant accuracy improvements in resonance self-shielding calculations are realized by adopting the advanced Bondarenko method without any significant increase of computational burden.
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development of a fuel depletion sensitivity calculation module for multi cell problems in a deterministic Reactor Physics code system cbz
Annals of Nuclear Energy, 2016Co-Authors: Go Chiba, Yosuke Kawamoto, Tadashi NarabayashiAbstract:Abstract A new functionality of fuel depletion sensitivity calculations is developed as one module in a deterministic Reactor Physics code system CBZ. This is based on the generalized perturbation theory for fuel depletion problems. The theory for fuel depletion problems with a multi-layer depletion step division scheme is described in detail. Numerical techniques employed in actual implementation are also provided. Verification calculations are carried out for a 3 × 3 multi-cell problem consisting of two different types of fuel pins. It is shown that the sensitivities of nuclide number densities after fuel depletion with respect to the nuclear data calculated by the new module agree well with reference sensitivities calculated by direct numerical differentiation. To demonstrate the usefulness of the new module, fuel depletion sensitivities in different multi-cell arrangements are compared and non-negligible differences are observed. Nuclear data-induced uncertainties of nuclide number densities obtained with the calculated sensitivities are also compared.
Jaakko Leppänen - One of the best experts on this subject based on the ideXlab platform.
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on the use of delta tracking and the collision flux estimator in the serpent 2 monte carlo particle transport code
Annals of Nuclear Energy, 2017Co-Authors: Jaakko LeppänenAbstract:Abstract The Serpent Monte Carlo code was originally developed for the purpose of spatial homogenization and other computational problems encountered in the field of Reactor Physics. However, during the past few years the implementation of new methodologies has allowed expanding the scope of applications to new fields, including radiation transport and fusion neutronics. These applications pose new challenges for the tracking routines and result estimators, originally developed for a very specific task. The purpose of this paper is to explain how the basic collision estimator based cell flux tally in Serpent 2 is implemented, and how it is applied for calculating integral reaction rates. The methodology and its limitations are demonstrated by an example, in which the tally is applied for calculating collision rates in a problem with very low physical collision density. It is concluded that Serpent has a lot of potential to expand its scope of applications beyond Reactor Physics, but in order to be applied for such problems it is important that the code users understand the underlying methods and their limitations.
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The Serpent Monte Carlo code: Status, development and applications in 2013
Annals of Nuclear Energy, 2015Co-Authors: Jaakko Leppänen, Maria Pusa, Ville Valtavirta, Tuomas Viitanen, Toni KaltiaisenahoAbstract:Abstract The Serpent Monte Carlo Reactor Physics burnup calculation code has been developed at VTT Technical Research Centre of Finland since 2004, and is currently used in over 100 universities and research organizations around the world. This paper presents the brief history of the project, together with the currently available methods and capabilities and plans for future work. Typical user applications are introduced in the form of a summary review on Serpent-related publications over the past few years.
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preventing xenon oscillations in monte carlo burnup calculations by enforcing equilibrium xenon distribution
Annals of Nuclear Energy, 2013Co-Authors: Aarno Isotalo, Jaakko Leppänen, Jan DufekAbstract:Existing Monte Carlo burnup codes suffer from instabilities caused by spatial xenon oscillations. These oscillations can be prevented by forcing equilibrium between the neutron flux and saturated xenon distribution. The equilibrium calculation can be integrated to Monte Carlo neutronics, which provides a simple and lightweight solution that can be used with any of the existing burnup calculation algorithms. The stabilizing effect of this approach, as well as its limitations are demonstrated using the Reactor Physics code Serpent.
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performance of woodcock delta tracking in lattice Physics applications using the serpent monte carlo Reactor Physics burnup calculation code
Annals of Nuclear Energy, 2010Co-Authors: Jaakko LeppänenAbstract:Abstract This paper presents the delta-tracking based geometry routine used in the Serpent Monte Carlo Reactor Physics burnup calculation code. The method is considered a fast and efficient alternative to the conventional surface-to-surface ray-tracing, and well suited to the lattice Physics applications for which the code is mainly intended. The advantages and limitations of the routine are discussed and the applicability put to test in four example cases. It is concluded that the method performs well in LWR lattice applications, but really shows its efficiency when modeling HTGR particle fuels.
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htgr Reactor Physics and burnup calculations using the serpent monte carlo code
Transactions of the American Nuclear Society, 2009Co-Authors: Jaakko Leppänen, Mark D DehartAbstract:aA new explicit particle fuel model was developed to account for the heterogeneity effects. The results are compared to other Monte Carlo and deterministic transport codes and the study also serves as a test case for the modules and methods in SCALE 6.[2]
Almutairi B - One of the best experts on this subject based on the ideXlab platform.
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Lattice benchmarking of deterministic, Monte Carlo and hybrid Monte Carlo Reactor Physics codes for the soluble-boron-free SMR cores
'Organisation for Economic Co-Operation and Development (OECD)', 2020Co-Authors: Alam Syed, Kumar D, Almutairi B, Goodwin C, Ridwan Tuhfatur, Parks GeoffAbstract:Since the use of deterministic transport code WIMS can significantly reduce the computational time compared to the Monte Carlo (MC) code Serpent and hybrid MC code MONK, one of the major objectives of this study is to observe whether deterministic code WIMS can provide accuracy in Reactor Physics calculations while comparing Serpent and MONK. Therefore, numerical benchmark calculations for a soluble-boron-free (SBF) small modular Reactor (SMR) assembly have been performed using the WIMS, Serpent and MONK. Although computationally different in nature, these codes can solve the neutronic transport equations and calculate the required neutronic parameters. A comparison in neutronic parameters between the three codes has been carried out using two types of candidate fuels: 15%235U enriched homogeneously mixed all-UO2 fuel and 18%235U enriched micro-heterogeneous ThO2-UO2 duplex fuel in a 2D fuel assembly model using a 13×13 arrangement. The eigenvalue/reactivity (k∞) and 2D assembly pin power distribution at different burnup states in the assembly depletion are compared using three candidate nuclear data files: ENDF/B- VII, JEF2.2 and JEF3.1. A good agreement in k∞ values was observed among the codes for both the candidate fuels. The differences in k∞ between the codes are ∼200 pcm when cross-sections based on the same nuclear data file are used. A higher difference (up to ∼450 pcm) in the k∞ values is observed among the codes using cross-sections based on different data files. Finally, it can be concluded from this study that the good agreement in the results between the codes found provides enhanced confidence that modeling of SBF, SMR propulsion core systems with micro-heterogeneous duplex fuel can be performed reliably using deterministic neutronics code WIMS, offering the advantage of less expensive computation than that of the MC Serpent and hybrid MC MONK codes
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Lattice benchmarking of deterministic, Monte Carlo and hybrid Monte Carlo Reactor Physics codes for the soluble-boron-free SMR cores
2020Co-Authors: Sb Alam, Kumar D, Almutairi B, Goodwin C, Ridwan T, Gt ParksAbstract:© 2019 Elsevier B.V. Since the use of deterministic transport code WIMS can significantly reduce the computational time compared to the Monte Carlo (MC) code Serpent and hybrid MC code MONK, one of the major objectives of this study is to observe whether deterministic code WIMS can provide accuracy in Reactor Physics calculations while comparing Serpent and MONK. Therefore, numerical benchmark calculations for a soluble-boron-free (SBF) small modular Reactor (SMR) assembly have been performed using the WIMS, Serpent and MONK. Although computationally different in nature, these codes can solve the neutronic transport equations and calculate the required neutronic parameters. A comparison in neutronic parameters between the three codes has been carried out using two types of candidate fuels: 15% 235U enriched homogeneously mixed all-UO2 fuel and 18% 235U enriched micro–heterogeneous ThO2-UO2 duplex fuel in a 2D fuel assembly model using a 13×13 arrangement. The eigenvalue/reactivity (k∞) and 2D assembly pin power distribution at different burnup states in the assembly depletion are compared using three candidate nuclear data files: ENDF/B-VII, JEF2.2 and JEF3.1. A good agreement in k∞ values was observed among the codes for both the candidate fuels. The differences in k∞ between the codes are ~200 pcm when cross-sections based on the same nuclear data file are used. A higher difference (up to ~450 pcm) in the k∞ values is observed among the codes using cross-sections based on different data files. Finally, it can be concluded from this study that the good agreement in the results between the codes found provides enhanced confidence that modeling of SBF, SMR propulsion core systems with micro-heterogeneous duplex fuel can be performed reliably using deterministic neutronics code WIMS, offering the advantage of less expensive computation than that of the MC Serpent and hybrid MC MONK codes
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Small modular Reactor core design for civil marine propulsion using micro-heterogeneous duplex fuel. Part I: Assembly-level analysis
'Elsevier BV', 2019Co-Authors: Bahauddin Alam S, Kumar D, Almutairi B, Goodwin C, Pk Bhowmik, Gt ParksAbstract:© 2019 Elsevier B.V. In an effort to de-carbonise commercial freight shipping, there is growing interest in the possibility of using nuclear propulsion systems. In this Reactor Physics study, we seek to design a soluble-boron-free (SBF) and low-enriched uranium (LEU) (
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Small modular Reactor core design for civil marine propulsion using micro-heterogeneous duplex fuel. Part II: whole-core analysis
'Organisation for Economic Co-Operation and Development (OECD)', 2019Co-Authors: Sb Alam, Kumar D, Almutairi B, Goodwin C, Ridwan Tuhfatur, Parks GeoffAbstract:In an e↵ort to de-carbonise commercial freight shipping, there is growing interest in the possibility of using nuclear propulsion systems. In this Reactor Physics study, we seek to design a soluble-boron-free (SBF) and low-enriched uranium (LEU) (
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Reactor Physics analysis of thorium-based fuel for long-life SMR cores using seed-blanket fuel concept
2019Co-Authors: Almutairi B, Sb Alam, Kumar D, Ridwan T, Parks G, Cs Goodwin, Usman SAbstract:This study examines the feasibility of thorium-based seedblanket (SB) fuel for small modular Reactor (SMR) core derived from a previous study [1]. The study aims to develop a Reactor Physics core design for SMR where net breeding of fuel can be achieved with higher end-of-life (EOL) fissile content than the initial loading. SB fuel assembly and core design studies were performed to design a core that can operate over a 20 effective full-power-years life at 333 MWth, utilizing the improved cumulative energy share of U-233. A heterogeneous fuel assembly with a uniform distribution of fissile nuclides was modified to create a heterogeneous two fuel zone configuration: the “seed" region containing 19% U-235 enriched UO2 and a fertile “blanket" region of 100% ThO2. The UO2 rich seed is the supplier of neutrons, and the ThO2 blanket generates new fuel (U-233) from Th-232 through neutron capture. We use WIMS to develop subassembly designs and PANTHER to examine whole-core arrangements. We compare designs with all-UO2 fuel with SB fuel in 13X13 assemblies. It is observed that core life is 5% higher than UO2 at EOL due to enhanced neutron absorption in thorium at the beginning-of-life which promotes breeding. Furthermore, it is seen that the fertile-capture-to-fissile-absorption ratio of the SB fuel is significantly higher than the UO2 fuel. This higher CR of duplex fuel compared to UO2 fuel enables a long core life, which is a desirable feature for our SMR core. From the Reactor Physics standpoint, a higher “fertile capture-to-fissile absorption ratio" is advantageous for achieving better fissile accumulation potential
Kumar D - One of the best experts on this subject based on the ideXlab platform.
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Lattice benchmarking of deterministic, Monte Carlo and hybrid Monte Carlo Reactor Physics codes for the soluble-boron-free SMR cores
'Organisation for Economic Co-Operation and Development (OECD)', 2020Co-Authors: Alam Syed, Kumar D, Almutairi B, Goodwin C, Ridwan Tuhfatur, Parks GeoffAbstract:Since the use of deterministic transport code WIMS can significantly reduce the computational time compared to the Monte Carlo (MC) code Serpent and hybrid MC code MONK, one of the major objectives of this study is to observe whether deterministic code WIMS can provide accuracy in Reactor Physics calculations while comparing Serpent and MONK. Therefore, numerical benchmark calculations for a soluble-boron-free (SBF) small modular Reactor (SMR) assembly have been performed using the WIMS, Serpent and MONK. Although computationally different in nature, these codes can solve the neutronic transport equations and calculate the required neutronic parameters. A comparison in neutronic parameters between the three codes has been carried out using two types of candidate fuels: 15%235U enriched homogeneously mixed all-UO2 fuel and 18%235U enriched micro-heterogeneous ThO2-UO2 duplex fuel in a 2D fuel assembly model using a 13×13 arrangement. The eigenvalue/reactivity (k∞) and 2D assembly pin power distribution at different burnup states in the assembly depletion are compared using three candidate nuclear data files: ENDF/B- VII, JEF2.2 and JEF3.1. A good agreement in k∞ values was observed among the codes for both the candidate fuels. The differences in k∞ between the codes are ∼200 pcm when cross-sections based on the same nuclear data file are used. A higher difference (up to ∼450 pcm) in the k∞ values is observed among the codes using cross-sections based on different data files. Finally, it can be concluded from this study that the good agreement in the results between the codes found provides enhanced confidence that modeling of SBF, SMR propulsion core systems with micro-heterogeneous duplex fuel can be performed reliably using deterministic neutronics code WIMS, offering the advantage of less expensive computation than that of the MC Serpent and hybrid MC MONK codes
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Lattice benchmarking of deterministic, Monte Carlo and hybrid Monte Carlo Reactor Physics codes for the soluble-boron-free SMR cores
2020Co-Authors: Sb Alam, Kumar D, Almutairi B, Goodwin C, Ridwan T, Gt ParksAbstract:© 2019 Elsevier B.V. Since the use of deterministic transport code WIMS can significantly reduce the computational time compared to the Monte Carlo (MC) code Serpent and hybrid MC code MONK, one of the major objectives of this study is to observe whether deterministic code WIMS can provide accuracy in Reactor Physics calculations while comparing Serpent and MONK. Therefore, numerical benchmark calculations for a soluble-boron-free (SBF) small modular Reactor (SMR) assembly have been performed using the WIMS, Serpent and MONK. Although computationally different in nature, these codes can solve the neutronic transport equations and calculate the required neutronic parameters. A comparison in neutronic parameters between the three codes has been carried out using two types of candidate fuels: 15% 235U enriched homogeneously mixed all-UO2 fuel and 18% 235U enriched micro–heterogeneous ThO2-UO2 duplex fuel in a 2D fuel assembly model using a 13×13 arrangement. The eigenvalue/reactivity (k∞) and 2D assembly pin power distribution at different burnup states in the assembly depletion are compared using three candidate nuclear data files: ENDF/B-VII, JEF2.2 and JEF3.1. A good agreement in k∞ values was observed among the codes for both the candidate fuels. The differences in k∞ between the codes are ~200 pcm when cross-sections based on the same nuclear data file are used. A higher difference (up to ~450 pcm) in the k∞ values is observed among the codes using cross-sections based on different data files. Finally, it can be concluded from this study that the good agreement in the results between the codes found provides enhanced confidence that modeling of SBF, SMR propulsion core systems with micro-heterogeneous duplex fuel can be performed reliably using deterministic neutronics code WIMS, offering the advantage of less expensive computation than that of the MC Serpent and hybrid MC MONK codes
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Small modular Reactor core design for civil marine propulsion using micro-heterogeneous duplex fuel. Part I: Assembly-level analysis
'Elsevier BV', 2019Co-Authors: Bahauddin Alam S, Kumar D, Almutairi B, Goodwin C, Pk Bhowmik, Gt ParksAbstract:© 2019 Elsevier B.V. In an effort to de-carbonise commercial freight shipping, there is growing interest in the possibility of using nuclear propulsion systems. In this Reactor Physics study, we seek to design a soluble-boron-free (SBF) and low-enriched uranium (LEU) (
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Small modular Reactor core design for civil marine propulsion using micro-heterogeneous duplex fuel. Part II: whole-core analysis
'Organisation for Economic Co-Operation and Development (OECD)', 2019Co-Authors: Sb Alam, Kumar D, Almutairi B, Goodwin C, Ridwan Tuhfatur, Parks GeoffAbstract:In an e↵ort to de-carbonise commercial freight shipping, there is growing interest in the possibility of using nuclear propulsion systems. In this Reactor Physics study, we seek to design a soluble-boron-free (SBF) and low-enriched uranium (LEU) (
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Reactor Physics analysis of thorium-based fuel for long-life SMR cores using seed-blanket fuel concept
2019Co-Authors: Almutairi B, Sb Alam, Kumar D, Ridwan T, Parks G, Cs Goodwin, Usman SAbstract:This study examines the feasibility of thorium-based seedblanket (SB) fuel for small modular Reactor (SMR) core derived from a previous study [1]. The study aims to develop a Reactor Physics core design for SMR where net breeding of fuel can be achieved with higher end-of-life (EOL) fissile content than the initial loading. SB fuel assembly and core design studies were performed to design a core that can operate over a 20 effective full-power-years life at 333 MWth, utilizing the improved cumulative energy share of U-233. A heterogeneous fuel assembly with a uniform distribution of fissile nuclides was modified to create a heterogeneous two fuel zone configuration: the “seed" region containing 19% U-235 enriched UO2 and a fertile “blanket" region of 100% ThO2. The UO2 rich seed is the supplier of neutrons, and the ThO2 blanket generates new fuel (U-233) from Th-232 through neutron capture. We use WIMS to develop subassembly designs and PANTHER to examine whole-core arrangements. We compare designs with all-UO2 fuel with SB fuel in 13X13 assemblies. It is observed that core life is 5% higher than UO2 at EOL due to enhanced neutron absorption in thorium at the beginning-of-life which promotes breeding. Furthermore, it is seen that the fertile-capture-to-fissile-absorption ratio of the SB fuel is significantly higher than the UO2 fuel. This higher CR of duplex fuel compared to UO2 fuel enables a long core life, which is a desirable feature for our SMR core. From the Reactor Physics standpoint, a higher “fertile capture-to-fissile absorption ratio" is advantageous for achieving better fissile accumulation potential
