Impacts of Burnup-Dependent Swelling of Metallic Fuel on the Performance of a Compact Breed-and-Burn Fast Reactor

Hartanto, Donny,Heo, Woong,Kim, Chihyung,Kim, Yonghee Korean Nuclear Society 2016 Nuclear Engineering and Technology Vol.48 No.2

The U-Zr or U-TRU-Zr cylindrical metallic fuel slug used in fast reactors is known to swell significantly and to grow during irradiation. In neutronics simulations of metallic-fueled fast reactors, it is assumed that the slug has swollen and contacted cladding, and the bonding sodium has been removed from the fuel region. In this research, a realistic burnup-dependent fuel-swelling simulation was performed using Monte Carlo code McCARD for a single-batch compact sodium-cooled breed-and-burn reactor by considering the fuel-swelling behavior reported from the irradiation test results in EBR-II. The impacts of the realistic burnup-dependent fuel swelling are identified in terms of the reactor neutronics performance, such as core lifetime, conversion ratio, axial power distribution, and local burnup distributions. It was found that axial fuel growth significantly deteriorated the neutron economy of a breed-and-burn reactor and consequently impaired its neutronics performance. The bonding sodium also impaired neutron economy, because it stayed longer in the blanket region until the fuel slug reached 2% burnup.

Alternative reflectors for a compact sodium-cooled breed-and-burn fast reactor

Hartanto, Donny,Kim, Yonghee Elsevier 2015 Annals of nuclear energy Vol.76 No.-

<P><B>Abstract</B></P> <P>The breed-and-burn fast reactor (B&BR) is a unique fast reactor concept that offers attractive characteristics in terms of core performance and non-proliferation aspects. The B&BR has the ability to breed its own fuel and use it in situ to achieve an extremely long life. In order to achieve the breed-and-burn condition, the neutron economy should be very good. In this work, a compact sodium-cooled B&BR is investigated from the physics point of view. In a compact size B&BR which usually has a higher neutron leakage, a good neutron reflector is essential to maintain the neutron economy. In the conventional sodium-cooled B&BRs, a steel reflector such as HT-9 is usually adopted as reflector material. In the current work, various alternative reflector materials such as pure lead, lead–bismuth eutectic (LBE), lead–magnesium eutectic (LME), PbO, MgO, Ni, and HT-9 are investigated from the neutronics perspectives. Several characterizations including the reflecting performance, core neutron spectrum, core leakage, power distribution, and sodium coolant void reactivity (CVR) have been performed to understand the behavior of each reflector material. The impact of the coarse lattice reflector configuration on the reflector performance and CVR has also been analyzed. In addition, the impact of the reflector material on the gas expansion module (GEM) worth has been investigated, too. Finally, it was concluded that lead-based reflectors such as LBE, LME, and PbO, are promising alternative reflector candidates for a compact B&BR. The calculations were all done by using a continuous energy Monte Carlo code.</P> <P><B>Highlights</B></P> <P> <UL> <LI> A good neutron reflector is necessary in a small breed-and-burn fast reactor (B&BR) to improve the neutron economy. </LI> <LI> Impacts of various alternative reflectors on a sodium-cooled B&BR have been investigated from the neutronics perspectives. </LI> <LI> LME (lead–magnesium eutectic) and PbO are recommended as promising alternative reflector materials for the small B&BR. </LI> </UL> </P>

Analysis of the first core of the Indonesian multipurpose research reactor RSG-GAS using the Serpent Monte Carlo code and the ENDF/B-VIII.0 nuclear data library

Hartanto, Donny,Liem, Peng Hong Korean Nuclear Society 2020 Nuclear Engineering and Technology Vol.52 No.12

This paper presents the neutronics benchmark analysis of the first core of the Indonesian multipurpose research reactor RSG-GAS (Reaktor Serba Guna G.A. Siwabessy) calculated by the Serpent Monte Carlo code and the newly released ENDF/B-VIII.0 nuclear data library. RSG-GAS is a 30 MWth pool-type material testing research reactor loaded with plate-type low-enriched uranium fuel using light water as a coolant and moderator and beryllium as a reflector. Two groups of critical benchmark problems are derived on the basis of the criticality and control rod calibration experiments of the first core of RSG-GAS. The calculated results, such as the neutron effective multiplication factor (k) value and the control rod worth are compared with the experimental data. Moreover, additional calculated results, including the neutron spectra in the core, fission rate distribution, burnup calculation, sensitivity coefficients, and kinetics parameters of the first core will be compared with the previous nuclear data libraries (interlibrary comparison) such as ENDF/B-VII.1 and JENDL-4.0. The C/E values of ENDF/B-VIII.0 tend to be slightly higher compared with other nuclear data libraries. Furthermore, the neutron reaction cross-sections of ¹⁶O, ⁹Be, ²³⁵U, ²³⁸U, and S(𝛼,𝛽) of ¹H in H₂O from ENDF/B-VIII.0 have substantial updates; hence, the k sensitivities against these cross-section changes are relatively higher than other isotopes in RSG-GAS. Other important neutronics parameters such as kinetics parameters, control rod worth, and fission rate distribution are similar and consistent among the nuclear data libraries.

An optimization study on the excess reactivity in a linear breed-and-burn fast reactor (B&BR)

Hartanto, Donny,Kim, Chihyung,Kim, Yonghee Elsevier 2016 Annals of nuclear energy Vol.94 No.-

<P><B>Abstract</B></P> <P>This work is concerned with minimization of the excess reactivity in a compact linear breed-and-burn fast reactor (B&BR). Neutronic studies are performed to reduce the core excess reactivity below 1$ in order to prevent possibility of the prompt criticality accident. A reference B&BR core design is introduced as a starting point for the optimization, which is loaded with vented annular metallic (U–Zr) fuels. The initial criticality of the B&BR core is achieved by using an LEU (low-enriched U) fuel and a spent nuclear fuel is reutilized as the blanket fuels. In order to minimize the transitional excess reactivity of the B&BR core, this study proposes a splitting of the Zr content in the U–Zr metallic fuel in combination with a geometrical optimization of the initial LEU core to flatten the radial power distribution simultaneously. A Zr-zoning is first applied to the initial LEU core to reduce excess reactivity and then a concave initial LEU core is also adopted to achieve both small excess reactivity and long lifetime of the B&BR core. In addition, a similar Zr-zoning strategy is also applied to the blanket region to optimize the excess reactivity behavior and flatten the radial power as well. It is demonstrated that a Zr-zoning combined with a concave core configuration can provides a promising neutronic performances. Finally, a recommended core has been characterized in terms of the burnup reactivity change, conversion ratio, power profiles, and reactivity coefficients.</P> <P><B>Highlights</B></P> <P>A small B&BR core design adopting vented annular metallic fuel is presented.Excess reactivity can be reduced below 1$ via Zr-zoning in the metallic fuels.A Zr-zoned and pan-shaped core leads to a small excess reactivity and long lifetime.The radial power profile is also flattened by the Zr-zoning.</P>

Uranium Enrichment Reduction in the Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) with PBO Reflector

Kim, Chihyung,Hartanto, Donny,Kim, Yonghee Korean Nuclear Society 2016 Nuclear Engineering and Technology Vol.48 No.2

The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR) is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

Neutronics feasibility of simple and dry recycling technologies for a self-sustainable breed-and-burn fast reactor

Kim, Chihyung,Hartanto, Donny,Kim, Yonghee Elsevier 2017 Annals of nuclear energy Vol.110 No.-

<P><B>Abstract</B></P> <P>This paper is concerned with the neutronics analysis of extremely simplified recycling technologies of spent fuels in a small breed-and-burn fast reactor (B&BR). The discharged fuels of the first generation B&BR, which achieved an average burnup of 160 GWd/MTHM, were used to construct a second generation B&BR core. Two types of high proliferation resistant recycling technologies, melt refining and the newly suggested super-simplified melt and treatment, were applied to process and treat the discharged fuels. Because the burnup profile of discharged fuels varies largely depending on its position in the core, the recycling of the discharged fuels was also carried out by grouping them into recycling regions including 1, 3, and 6 recycling regions. In this study, the core performance of the 2nd generation B&BR loaded with the recycled fuel, which was produced by different recycling technologies and recycling regions, was analyzed and compared. An optimum fuel loading scheme was also adopted to maximize the performance of the 2nd generation B&BR in terms of the burnup reactivity change, core lifetime, and power profiles.</P> <P><B>Highlights</B></P> <P> <UL> <LI> The preliminary study on self-sustainability of a B&BR fuel cycle was investigated. </LI> <LI> A proliferation-resistant and simple fuel recycling technology is used. </LI> <LI> The 2nd generation B&BR can be designed using the discharged fuel from the 1st B&BR. </LI> <LI> The importance of region-wise fuel recycling in nuclear fuel reprocessing has been confirmed. </LI> </UL> </P>

Static and transient analyses of Advanced Power Reactor 1400 (APR1400) initial core using open-source nodal core simulator KOMODO

Jwaher Alnaqbi,Donny Hartanto,Reem Alnuaimi,Muhammad Imron,Victor Gillette 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.2

The United Arab Emirates is currently building and operating four units of the APR-1400 developed by aSouth Korean vendor, Korea Electric Power Corporation (KEPCO). This paper attempts to perform APR1400 reactor core analysis by using the well-known two-step method. The two-step method wasapplied to the APR-1400 first cycle using the open-source nodal diffusion code, KOMODO. In this study,the group constants were generated using CASMO-4 fuel transport lattice code. The simulation wasperformed in Hot Zero Power (HZP) at steady-state and transient conditions. Some typical parametersnecessary for the Nuclear Design Report (NDR) were evaluated in this paper, such as effective neutronmultiplication factor, control rod worth, and critical boron concentration for steady-state analysis. Otherparameters such as reactivity insertion, power, and fuel temperature changes during the ReactivityInsertion Accident (RIA) simulation were evaluated as well. The results from KOMODO were verifiedusing PARCS and SIMULATE-3 nodal core simulators. It was found that KOMODO gives an excellentagreement

Neutronics study on small power ADS loaded with recycled inert matrix fuel for transuranic elements transmutation using Serpent code

Vu, Thanh Mai,Hartanto, Donny,Ha, Pham Nhu Viet Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.7

A small power ADS design using thorium oxide and diluent matrix reprocessed fuel is proposed for a high transmutation rate, small reactivity swing, and strong safety features. Two fuel matrices (CERCER and CERMET) and different recycled fuel compositions recovered from UO₂ spent fuels with 45 GWd/tU and 60 GWd/tU burnup were investigated to determine the suitable fuel for the ADS. It was found that the transmutation of each isotope depends on TRU initial loading amount. After examining the cores, the results show that CERCER fueled ADS has a negative coolant void reactivity (CVR) and a smaller radiotoxicity at discharge compared to that of CERMET core. It implies that CERCER fuel has enhanced safety features and more flavor in terms of radiotoxicity management. To increase fuel utilization and core operation efficiency, a simple assembly shuffling pattern for the CERCER fueled ADS is also proposed. Eigenvalue and burnup calculations were conducted using Serpent 2 with ENDF/B-VII.0 library in both kcode and external source modes, and it indicates that the results of transmutation analyses obtained by kcode only is reliable to discuss the transmutation potential of ADS. Burnup calculation with the fixed-source mode is essential to be used for more practical results of the transmutation by ADS.

On the Particle Swarm Optimization of Cask Shielding Design for a prototype Sodium-cooled Fast Reactor

임동원,이철우,임재용,Donny Hartanto 한국원자력학회 2019 Nuclear Engineering and Technology Vol.51 No.1

For the continuous operation of a nuclear reactor, burnt fuel needs to be replaced with fresh fuel, whereappropriate (ex-vessel) fuel handling is required. Particularly for the Sodium-cooled Fast Reactor (SFR)refueling, its process has unique challenges due to liquid sodium coolant. The ex-vessel spent fueltransportation should concern several design features such as the radiation shielding, decay-heatremoval, and inert space separated from air. This paper proposes a new design optimization methodologyof cask shielding to transport the spent fuel assembly in a prototype SFR for the first time. TheParticle Swarm Optimization (PSO) algorithm had been applied to design trade-offs between shieldingand cask weight. The cask is designed as a double-cylinder structure to block an inert sodium region fromthe air-cooling space. The PSO process yielded the optimum shielding thickness of 26 cm, considering theweight as well. To confirm the shielding performance, the radiation dose of spent fuel removed at itspeak burnup and after 1-year cooling was calculated. Two different fuel positions located duringtransportation were also investigated to consider a functional disorder in a cask drive system. This studyconcludes the current cask design in normal operations is satisfactory in accordance with regulatoryrules

Verification of Reduced Order Modeling Based Uncertainty/Sensitivity Estimator (ROMUSE)

Bassam Khuwaileh,Brian Williams,Paul Turinsky,Donny Hartanto 한국원자력학회 2019 Nuclear Engineering and Technology Vol.51 No.4

This paper presents a number of verification case studies for a recently developed sensitivity/uncertaintycode package. The code package, ROMUSE (Reduced Order Modeling based Uncertainty/SensitivityEstimator) is an effort to provide an analysis tool to be used in conjunction with reactor core simulators,in particular the Virtual Environment for Reactor Applications (VERA) core simulator. ROMUSE has beenwritten in Cþþ and is currently capable of performing various types of parameter perturbations andassociated sensitivity analysis, uncertainty quantification, surrogate model construction and subspaceanalysis. The current version 2.0 has the capability to interface with the Design Analysis Kit for Optimizationand Terascale Applications (DAKOTA) code, which gives ROMUSE access to the various algorithmsimplemented within DAKOTA, most importantly model calibration. The verification study isperformed via two basic problems and two reactor physics models. The first problem is used to verify theROMUSE single physics gradient-based range finding algorithm capability using an abstract quadraticmodel. The second problem is the Brusselator problem, which is a coupled problem representative ofmulti-physics problems. This problem is used to test the capability of constructing surrogates viaROMUSE-DAKOTA. Finally, light water reactor pin cell and sodium-cooled fast reactor fuel assemblyproblems are simulated via SCALE 6.1 to test ROMUSE capability for uncertainty quantification andsensitivity analysis purposes