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1 Advanced Fission Systems ABS-1


1)Bahçeşehir University, Faculty of Engineering and Natural Sciences, Beşiktaş, İstanbul, TÜRKİYE
Email: sumer.sahin[at]

2)Near East University, LefkoÅŸa/KKTC, Turkish Republic of Northern Cyprus, Mersin 10, TÃœRKÄ°YE


The Fixed Bed Nuclear Reactor (FBNR) is being developed under the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on Small Reactors (70-120 MWel) without On-site Refueling. The FBNR is adequate for developing countries with small electric grids and limited investment capabilities, as well as with the weakness of manpower for development of nuclear power plants. The FBNR fuel element consists of 500 microns in diameter UO2 micro spheres covered by 25 microns thick zirconium cladding embedded in a spherical zirconium matrix with 300 μm thick Zircaloy-4 cladding to form a 15 mm diameter fuel element. Approximately 1.62 million spherical fuel elements are suspended in a core height of 200 cm and diameter of 170 cm by upwards flowing light water coolant with 280 oC entry and 340 oC exit temperatures. The unit lattice geometry is dodecahedron with fuel element in center and surrounded with light water coolant-moderator.
FBNR can operate both with conventional nuclear fuel 235U as well as with 233U and make use of the worldwide abundant thorium on that way. Utilization potential of alternative fuels, such as thorium, reactor grade plutonium and minor actinides enables FBNR to have a high level of sustainability.
In the present work, 9 % enriched UO2 fuel is used. Calculations are conducted with the MCNPX 2.7 code. As, it is not practical to tackle millions of fuel elements in the same run, at first unit fuel cell calculations are performed in spherical geometry for one single lattice consisting of separate fuel, cladding and moderator regions. In the second run, the fuel cell is homogenized. Infinite cell calculations have been executed for variable moderator/fuel (H/235U) ratios in order to determine under- and over-moderated criticality values. The criticality values for the homogenized cell geometry are slightly lower and remains <5 % range, i.e., on the conservative side, which is acceptable for the purpose of this study. This allows us to homogenize the entire core for the full 3-D reactor with core and reflector regions. Reactor criticality increases with increasing H/235U ratios in the, under moderated region, where H/235U = 20 is selected for further investigations. Temporal variation of the reactor criticality is pursued for a reactor power of 400 MWth (~ 120 MWel net) with keff = 1.311 at the beginning of life down to keff = 1.044 for a full power operation time of 570 days, leading to a fuel burn up of 44 GWd/MTU.

Keywords: nuclear energy; small modular reactor; fixed bed nuclear reactor; uranium

PermaLink | Plain Format | Corresponding Author (Sumer Sahin)

2 Advanced Fission Systems ABS-2

Muhammad Ilham (a*), Helen Raflis (a), Zaki Suud (a)

Nuclear Physics and Biophysics Research Division, Physics Department,
Faculty of Mathematics and Natural Science, Bandung Institute of Technology
Jalan Ganesha 10, Bandung 40132, INDONESIA
E-mail: muhilham99[at]


This thesis research discusses the development of program code that coupling the ORIGEN2.2 for burn-up analysis with the Monte Carlo program, OpenMC, for neutron analysis program called OpenMC-ORIGEN (Op-OR). The results of this program have been compared with benchmark results from previously published results and MCNP6 program. The acquired results show a good agreement with benchmark results. This program is written using the Python3 program language. This linkage program codes perform well for designing a new advanced reactor and analyzing the neutronic parameter and burnup/depletion calculation.

Keywords: OpenMC, Monte Carlo, ORIGEN, Burnup, Parallelization, Op-OR

PermaLink | Plain Format | Corresponding Author (Muhammad Ilham)

3 Advanced Fission Systems ABS-3

Performance Analysis of Reflectors Material for Core Design of Modular Gas-cooled Fast Reactor (GFR) using OpenMC
Helen Raflis(a,b), M. Ilham(a), Zaki Su’ud(a), Abdul Waris(a), and Dwi Irwanto(a)

a)Nuclear Physics and Biophysics Research Division, Physics Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology, Jalan Ganesha 10 Bandung 40132, Indonesia.
b)Center for Regulatory Assessment of Nuclear Installation and Materials, Nuclear Energy Regulatory Agency (BAPETEN), Jalan Gadjah Mada no. 8 Jakarta Pusat 10120 Indonesia

e-mail: helenraflis98[at]


The selection of reflector material for core design of modular Gas-cooled Fast Reactor (GFR) that has the potential for use actinide recycling and closed fuel cycle using helium gas as the main coolant, high working temperature and low void reactivity effect has been concerned. The modular GFR has the ability to maintain the criticality to achieve a long life of fast reactor. The neutron economy should be very good to improve the neutron economy via reduced neutron leakage so that the long life condition achieved. In this research, selection of potential reflector in modular GFR investigated from physics phenomenon and criticality condition. The good neutron reflector is essential to maintain the neutron economy which fast reactor design usually has a higher neutron leakage. The various alternative reflector materials such as pure lead, pure nickel, pure magnesium, pure bismuth, Ba2Pb, BeO, SiC, PbO, and Zr3Si2 are calculated investigated from the neutronics perspectives. The Monte Carlo method has advantages in full-scale and heterogeneous three-dimensional (3D) geometry modeling using Evaluated Nuclear Data File (ENDF/B-VII.b5) nuclear data and continuous energy of OpenMC code. The physics parameters characterized including the value of keff value profile, reflecting performance, core neutron spectrum, core leakage, power distribution, and neutron flux profiles to understand the behavior of each reflector material. The most important neutronic parameters of GFR core design is determined for beginning of life (BOL) conditions. Finally, we got the result that lead-based reflectors and zirconium-based-reflectors are good reflector candidates for a design of modular GFR.

Keywords: Modular Gas-cooled Fast Reactor, Reflector, Monte Carlo Method, Core Leakage.

PermaLink | Plain Format | Corresponding Author (Helen Raflis)

4 Advanced Fission Systems ABS-7

Comparative Study on Neutronic Characteristics of VHTR core
Odmaa Sambuu(a,b*), Khukhsuvd Batsaikhan (b), Jamiyansuren Terbish (b) Munkhbat Byambajav (a,b)

a) Department of Chemical and Biological Engineering, School of Engineering and Applied Sciences, National University of Mongolia, Ikh surguuliin gudamj 3, Sukhbaatar district, Ulaanbaatar 14201, Mongolia (*odmaa[at]
b) Nuclear Research Center, National University of Mongolia, Peace Avenue 122, Bayanzurkh District, Ulaanbaatar 14201, Mongolia


The solid and annular cylindrical prismatic core designs of VHTR which were fuelled with an advanced TRISO particle fuel with additional ZrC were proposed and the preliminary neutronic analyses were carried out previously [1]. The preliminary neutronic results were compared between the cores operating at temperature of 850oC with and without ZrC additional layer and it showed that the effective neutron multiplication factor in BOC and discharged burnup was increased, while core lifetime was reduced due to existence of the ZrC layer. Neutronic feature of an annular prismatic VHTR core with ZrC-containing TRISO fuel was improved for long term operation as higher fuel burnup in effect of inner reflector.
In this study, a comparative study has been performed to evaluate the capability of the alternative fuel kernel for annular prismatic VHTR core with TRIZO fuel. A representative VHTR core with different types of fuel was the same as 100 MWt at 850oC operating temperature. (U,Th)O2 and UCO kernels, as representatives of thorium, carbide fuels, are loaded into annular VHTR core. Uranium enrichment contents are adjusted to 20% for these fuel cases respectively, aiming at the comparison of oxide fuel of our previous research [1]. The neutronic analyses are performed using continuous energy Monte Carlo code MVP2.0 [2] and MVPBURN [3] with JENDL4.0 nuclear data library [4].
Based on neutronic analyses results of the present study and comparing to other fuel cases, carbide fuel case was proved to be able to provide the longest margin to the working limits, concerning the effective neutron multiplication factor at BOC, core operating lifetime and fuel burnup. This enables the carbide fuel to be competitive candidate for the VHTR usage.
1. Sambuu Odmaa. Batsaikhan Khukhsuvd, Terbish Jamiyansuren, Byambajav Munkhbat and Nanzad Norov, Preliminary neutronic analyses on VHTR core design, International Journal of Nuclear Safety and Simulation, Vol.9, No.2 December 2018.
2. Yasunobu Nagaya, Keisuke Okumura, Takamasa Mori, Masayuki Nakagawa, MVP/GMVP II: General Purpose Monte-Carlo Code for Neutron and Photon Transport Calculations Based on Continuous Energy and Multigroup Methods, JAERI-1348, Japan: Japan Atomic Energy Research Institute, 2005.
3. Keisuke Okumura, Yasunobu Nagaya, Takamasa Mori, MVP-BURN Users Manual, Japan: Atomic Energy Agency, 2005.
4. Keiichi Shibata, Osamu Iwamoto, Tsuneo Nakagawa, Nobuyuki Iwamoto et al, Japanese Evaluated Nuclear Data Library-JENDL-4.0, A New Library for Nuclear Science and Engineering, J.Nucl.Sci.Tech, 2011, 48(1):1-30.

Keywords: VHTR, advanced TRISO fuel, thorium oxide fuel, carbide fuel, neutronic analysis

PermaLink | Plain Format | Corresponding Author (Khukhsuvd Batsaikhan)

5 Advanced Fission Systems ABS-13

R&D Activities of Neutronics and Advanced Nuclear Systems at INEST • FDS Team
Yican Wu*, FDS Team

Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031, China


Advanced nuclear energy systems have attracted more and more attention all over the world for its great superiority of sustainability, safety, and economics. The Institute of Nuclear Safety Technology, Chinese Academic of Sciences • FDS Team (INEST • FDS Team) has focused on the fundamental and applied research in the area of neutronics and advanced lead-based reactors.
Neutronics is the key basis for the innovative development of nuclear systems and nuclear safety. INEST • FDS Team have been focused on the research of neutron transport physics and technology, including the development of advanced theories and software for neutron transport, technologies for neutron control and measurement, and the nuclear design and safety evaluation of advanced nuclear systems, etc. In this contribution, the latest development of Super Multi-functional Calculation Program for Nuclear Design and Safety Evaluation (SuperMC) and High Intensity D-T Fusion Neutron Generator (HINEG) are introduced.
Lead-based reactor is one of the most promising nuclear systems for Generation-IV reactor, SMR(Small modular reactor) and Accelerator Driven subcritical System (ADS). INEST • FDS Team have placed more emphases on the design and R&D of China LEAd-based Reactor (CLEAR) for more than 30 years. In this contribution, the latest progress on the designs and R&D activities for CLEAR series reactors are introduced.

Keywords: Advanced nuclear systems, Neutronics, Lead-based reactor, CLEAR

PermaLink | Plain Format | Corresponding Author (Xiaoliang Zou)

6 Advanced Fission Systems ABS-16

The Influence of the Volume Ratio of Moderator and Fuel on the Core Size of Small and Long-life Reactor
Yanting Sun *, Qi Yang, Jun Gao, FDS Team

Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031, China


As one of the key physical parameters for the core design of small and long-life reactor, the neutron spectrum affects the core size and the reactivity temperature coefficients directly. From the view of core size optimization design, designs based on the thermal neutron spectrum and fast neutron spectrum have their advantages and disadvantages.
Based on the proposed core model with annular channel fuel elements, the influence of the volume ratio of moderator and fuel on the core size was studied by using Super Multi-function Calculation Program for Nuclear Design and Safety Evaluation (SuperMC). Also the optimum volume ratio of moderator and fuel which corresponds to the minimum core size under different energy output demands (power × refueling period) was explored. Moreover, the influence of the volume ratio of moderator and fuel on the reactivity feedback was analyzed. The results show that when the reactor total energy output demand is greater than 275 MWt·Year, the core designed with a fast neutron spectrum (without moderator) has minimum size. When the reactor total energy output demand is less than 275 MWt·Year, the core designed with a thermal neutron spectrum (with moderator) has the minimum size. The less the energy output demand is, the higher volume ratio of moderator and fuel for minimum core size should be. At the same time, the volume ratio of moderator and fuel should be less than 3.18 to ensure the negative reactivity temperature coefficients. The research results in this paper will provide references for the physical design of small and long-life reactor.

Keywords: Small and long-life reactor, Neutronics, Neutron spectrum, Volume ratio of moderator and fuel

PermaLink | Plain Format | Corresponding Author (Yanting Sun)

7 Advanced Fission Systems ABS-18

A New Design Concept of Burnable Poison for PWR Core – BP Attached to GT
Aiman Dandi and Myung Hyun Kim*

Kyung Hee University, 1732 Deogyeong-daero, Giheung-gu, Yongin-si, Gyeonggi-do, Korea, 17104


A longer cycle option in Pressurized Water Reactor (PWR) core offers many benefits both in economics and waste concerns. However, there is a requirement to have stronger burnable poison rods in order to compensate increased excess reactivity. In this study, a new BP design concept called Burnable poison Attached to Guide tube (BAG) is presented. Despite similar concept is proposed in the previous studies, BAG design is unique in application.
16×16 Combustion Engineering (CE) design model is chosen as a reference for fuel assembly design. For the feasibility study of the longer cycle PWR, fuel enrichment were increased to 6.96w/o for base fuel pins and 4.10w/o for zoning pins. In this study, assembly calculation was done by DeCART-2D code without checking impact to the core design.
Although BAG design cannot reduce the initial excess reactivity into reasonable level by itself, its good properties make it a very good option to give support to any conventional BP by combining it with them. Even though BAG+Erbia case have about half of the Erbia pins than that in the Erbia only case, both cases control the excess reactivity with almost the same performance. These two cases have the ability to control the excess reactivity longer than any other cases. The residual reactivity penalty of BAG+Erbia case is very low due to the fully depletion of B-10 in BAG design and a very low number of Erbia pins. The last two points are very important to design PWR core with longer operation cycle. Lastly, BAG+Erbia case provides reasonable values of power peaking factor and moderator temperature coefficient (MTC).

Keywords: Burnable Poison, PWR, Design Concept, attached to Guide Tube

PermaLink | Plain Format | Corresponding Author (Aiman Moftah A Dandi)

8 Advanced Fission Systems ABS-23

Tung Dong Cao Nguyen, Jiwon Choe, Xianan Du, Sooyoung Choi and Deokjung Lee*

Department of Nuclear Engineering, Ulsan National Institute of Science and Technology, 50 UNIST-gil, Ulju-gun, Ulsan, 44919, Republic of Korea
tungnguyen[at], chi91023[at], dxndxnwww[at], schoi[at], deokjung[at]*


A conceptual development for the long-life small modular lead-bismuth eutectic fast reactor (SMLFR) has been presented in this work. The key design constraint for this fast reactor is the transportation capability in spent nuclear fuel (SNF) cask so as to be able to use as a single or cluster power propulsion for icebreakers. Another innovated feature of this suggested SMLFR is all the core components are included within a small reactor vessel, which can be immediately transferred into the SNF cask after its entire operation time. The thermal power of the SMLFR is 37.5 MW with an assumption of 40% thermal efficiency by using an advanced energy conversion system based on supercritical carbon dioxide (S-CO2) as the working fluid. It is also designed to target more than 30 years of cycle length without refueling and a small reactivity swing by adopting a breed and burn concept. For such a long-life, small and portable reactor, an excellent neutron economy is a vital requirement. A recent study has been reported that the LBE cooled fast reactor demonstrates a better performance in neutron economy, burn-up reactivity swing, and void coefficient rather than sodium fast reactor (SFR). In addition, uranium nitride (UN) with a high thermal conductivity and a high-concentrated amount of fissile fuel is chosen as one of the primary fuel candidates for LFR due to better compatibility with the LBE coolant and providing an immense improvement in neutron economy compared to uranium oxide fuel. The core inlet and outlet temperatures are 300oC and 400oC, respectively. The 15-15Ti stabilized steel is selected as cladding and structure material due to its excellent swelling resistance and stability in LBE. The performance in design and analyses of this core are conducted with the fast reactor analysis code system MC2-3/TWODANT/REBUS-3 developed by Argonne National Laboratory (ANL) and the UNIST in-house Monte Carlo code MCS with ENDF/B-VII.0 cross-section library. It is confirmed through depletion calculations that the designed reactor is capable to operate for more than 40 years without refueling and a reactivity swing less than 500 pcm. In addition, core performance features are analyzed for criticality, radial and axial power profiles and thermal-hydraulic (T/H) calculation. A preliminary T/H calculation is achieved by a T/H one-dimensional module using single-phase closed-channel model. Pin-by-pin temperature profiles are obtained as receiving the pin-wise power profiles from MCS. It is basically confirmed the outlet coolant and maximum fuel temperatures and the coolant flow velocity are within the acceptance criteria. The SMLFR core is also evaluated in view of various significant safety parameters, including control rod worth, fuel temperature coefficient, and coolant density coefficient.

Keywords: small modular reactor, fast reactor, long-cycle, icebreaker, breed and burn.

PermaLink | Plain Format | Corresponding Author (Dong Cao Tung Nguyen)

9 Advanced Fission Systems ABS-26

V. E. Moiseenko, S.V. Chernitskiy

National Science Center Kharkiv Institute of Physics and Technology, Kharkiv, Ukraine


A uranium based nuclear fuel and closed fuel cycle are proposed for energy production. The fuel composition is chosen so that during reactor operation the amount of each transuranic component remains unchanged since the production rate and nuclear reaction rate are balanced. In such a balanced fuel only 238U content has a tendency to decrease and, to be kept constant, must be sustained by continuous supply. The spent nuclear fuel of such a reactor should be reprocessed and used again after separation of fission products and adding depleted uranium. In the fuel balance calculations 9 isotopes of uranium, neptunium, plutonium and americium are used. The model accounts for fission, neutron capture and decays. Using MCNPX numerical code the neutron calculations are performed for the reactor of industrial nuclear power plant size with MOX fuel and for a small reactor with metallic fuel. The calculation results indicate that it is possible to achieve criticality of the reactor in both cases and that production and consuming rates are balanced for the transuranic fuel components.

Keywords: fast reactor, MCNPX calculations, neutron spectrum, reaction rates

PermaLink | Plain Format | Corresponding Author (Vladimir Moiseenko)

10 Advanced Fission Systems ABS-28

MCS Monte Carlo multi-physics depletion analysis of an OPR-1000 reactor
Vutheam Dos (a), Hyunsuk Lee (a), Jiwon Choe (a), Matthieu Lemaire (a), Ho Cheol Shin (b), Hwan Soo Lee (b), and Deokjung Lee (a*)

a) Department of Nuclear Engineering, Ulsan National Institute of Science and Technology
50 UNIST-gil, Ulsan 44919, Republic of Korea
Vutheam Dos (a): dosvutheam8[at]
Deokjung Lee (a*): deokjung[at]
b) Korea Hydro & Nuclear Power Central Research Institute (KHNP-CRI)
Daejeon 34101, Republic of Korea


The Monte Carlo code MCS is developed at Ulsan National Institute of Science and Technology for the purpose of multi-physics high-fidelity analysis – neutron transport coupled with thermal/hydraulic and fuel performance (N/TH/FP) solvers - of large-scale pressurized water reactors (PWRs). In this work, the 3-D whole-core pin-wise depletion analysis of the OPR-1000 reactor system is conducted with two multi-physics coupled systems: Monte Carlo MCS with 1-D closed-channel T/H feedback (MCS/TH1D) and Monte Carlo MCS coupled with the fuel performance code FRAPCON (MCS/FRAPCON). The multi-physics coupling capability and accuracy of MCS is demonstrated by comparison of the calculated results against measured data for several neutronic parameters of interest: the critical boron concentration (boron letdown curve) and axial/radial power distributions. The solutions obtained by the MCS-based coupled systems show good agreement with the measurement data. This study therefore brings validation elements of the capability of MCS to perform the high-fidelity multi-physics simulation of a practical PWR core.

Keywords: Multi-Physics; MCS; whole-core pin-wise depletion; OPR-1000 PWR Core

PermaLink | Plain Format | Corresponding Author (VUTHEAM DOS)

11 Advanced Fission Systems ABS-31

Recent Advances in Control of Nuclear Power Plants
Areai Nuerlan and Rizwan-uddin

School of Nuclear Science and Technology,
Xi’an Jiaotong University, No.28, Xianning West Road, Xian, 710049, P.R. China

Department of Nuclear, Plasma, and Radiological Engineering
University of Illinois at Urbana-Champaign, 104 S Wright StUrbana, IL 61801, USA


Several new nuclear reactor designs with enhanced safety features are in either design stage or the first of the kind are being constructed. These include AP1000, CPR1000 and multi-module plants with more than one small or medium sized power units. The safety features of these have received increased attention since the Fukushima Daiichi accident. In addition to new types of cores, new control systems for core and U-tube steam generators (UTSG) are also being developed for some of these reactor designs. These designs are also expected to have better load following capabilities than GEN-II reactors. Control systems regulate the core power as well as axial power distribution (namely axial offset), possibly under load following conditions. Steam generator water level is another important variable that needs an effective control mechanism. The water level on the secondary side of UTSG must be maintained at the desirable value to ensure proper heat transfer from the reactor coolant to the secondary side, thus ensuring satisfactory operation of steam-drying equipment. Due to the shrink and swell phenomena taking place in the UTSG, the dynamic model of the UTSG is very critical for studying steam generator behavior, and thus for the design of feedwater controller. In this paper, we will review some of the recent advances made in the modeling and control of reactor cores and steam generators. Modeling approaches—such as the point reactor model, nodal reactor core model, and lumped parameter dynamic model for reactor core, and two phase flow model for the UTSG—will be reviewed. Power control methods for reactor cores, such as the feedback control with a state observer, will be reviewed. Load following control techniques for reactor cores—such as Mode A, Mode G, Mode T, Mechanical Shim and advanced multivariable frequency control methods—will also be discussed. As for UTSG level control, methods such as intelligent virtual reference feedback tuning and output feedback dissipation will be reviewed. Suggestions will be made for novel control techniques for new reactor designs.

Keywords: NPP, control, load following

PermaLink | Plain Format | Corresponding Author (Rizwan Uddin)

12 Advanced Fission Systems ABS-34

Fuzzy Logic-Based Feedwater Controller for the Super Fast Reactor
Sutanto, Marili Santi, Anggun Dwi Lestari, Ayu Jati Puspitasari

Department of Nuclear Technophysics,
Polytechnic Institute of Nuclear Technology, National Nuclear Energy Agency, Indonesia
Babarsari Street, PO BOX 6101 YKBB, Yogyakarta, Indonesia


A Generation IV reactor of supercritical-pressure light water fast reactor is expected to have significant improvements on the efficiency, safety and economy. Absence of boiling phenomenon with high heat capacity of the coolant at supercritical pressure results a compact plant system with low coolant flow rate and high outlet temperature, giving an increase of its efficiency up to 44%. The high outlet temperature with low coolant flow, however, leads the plant to be more sensitive to a perturbation of power to flow ratio. A change of outlet coolant temperature must be minimized to maintain the structure integrity of the reactor, such as the outlet nozzles and the main steam line. A fuzzy logic-based feedwater controller was applied to suppress the outlet coolant temperature deviation during an experience of the perturbation. Two signals of the temperature and the power changes were taken as the inputs of the fuzzy controller, and change of the feedwater coolant flow was taken as the control output. Fuzzifications used triangular member functions with Tsukamoto fuzzy model for inferencing process. Two control parameters of member function spread and input values which lead to fuzzy degree of 1 were optimized to satisfy the criteria of allowable outlet coolant temperature and stability. Performance of the control system showed that the outlet coolant temperature deviation could be kept the outlet coolant temperature within the criteria when perturbations of power to flow ratio were applied.

Keywords: Generation IV reactor, supercritical pressure, fast reactor, fuzzy control, outlet temperature

PermaLink | Plain Format | Corresponding Author (Sutanto Sutanto)

13 Advanced Fission Systems ABS-42

ThorCon Design Status
Jack Devanney(a*), Robert Hargraves(b), Lars Jorgensen(a), Ralph Moir(c)

(a) ThorCon, 242 Lyons Rd, Stevenson WA 98648 USA
(b) Osher Institute at Dartmouth College, Hanover NH 03755 USA
(c) Vallecitos Molten Salt Research, 607 E. Vallecitos Rd, Livermore CA 94550 CA USA


Thorium. ThorCon is a thorium converter, a denatured molten salt reactor that converts some thorium to uranium-233 and fissions it in situ. Fissile fuel is 19.75% LEU supplemented by the thorium that generates a quarter of the power, proportionately reducing costs for U-235. Thorium strengthens nonproliferation, substantially diluting chemically similar plutonium. Thorium additions are used to control reactivity via neutron absorption.

Design status. ThorConIsle is a complete basic design of a fission power plant generating 500 MWe from two 557 MWt reactors. It is integrated within a large ship hull, to be towed to a shallow-water site, ballasted down, then connected to the power grid. Such a hull resting on the seabed creates new opportunities to deal with seismic motions.

Power flows through fuel salt, clean salt, solar salt, then steam loops to a supercritical steam-turbine/generator. Fuel salt 704°C temperature enables 45% power conversion efficiency. ThorCon relies on existing materials and technologies to avoid R&D delays and minimize costs. It is designed in 150 to 500 ton blocks for economic shipyard manufacturing and assembly. World shipyards have sufficient capacity to manufacture 100 1-GWe power plants annually.

Safety. ThorCon’s negative thermal coefficient of reactivity exceeds 4.5 pcm/°K throughout the fuel salt life. The molten salt thermal mass and a 700°C margin to boiling enable effective radiative cooling from the fuel salt drain tank. Independent computational models show loss of primary heat path and simultaneous failure of all three shutdown rods results only in acceptable creep of primary loop materials. ThorCon’s design incorporates three, separate-technology, decay-heat removal paths. ThorCon implements passive safety though physical principles rather than auxiliary safety-grade electronic or electrical control systems. At least three barriers protect against radioactive material releases from equipment failures, power blackouts, control system errors, or deliberate operator malfeasance.

Our paper will describe the current status of the design and discuss ongoing multi-physics and transient modeling efforts, including seismic response of the hull, fission reactor, molten salt loops, and steam turbine-generator platform.

Keywords: molten salt, thorium, passive safety, transcient modeling, seismic response

PermaLink | Plain Format | Corresponding Author (Robert Hargraves)

14 Advanced Fission Systems ABS-43

Initial Core Design of Modified CANDLE Reactor with Pb 208-Bi eutectic as a coolant
Nina Widiawati1,a), Zaki Su’ud1,b), Dwi Irwanto1,c) and Sidik Permana2,d)

1Nuclear and Biophysics Department, Institut Teknologi Bandung, Indonesia.
2Nuclear science and engineering, Institut Teknologi Bandung, Indonesia.


Modified CANDLE conceptual design reactors can directly consume natural uranium as a fuel input without enrichment. At first, the core was divided into several regions with the same volume. Region 1 contains natural uranium, after ten years of burnup then moved on to region 2, fuel in region 2 moved on to region 3, and so on. Whereas fuel in region 10 was removed from the core. This scheme can also make a reactor long-lived with a 10-year refuelling period. The use of Pb208-Bi as coolant also contributed to the increase in k-eff value because the cross-section absorption neutron of Pb208 is the lowest compared to the other isotopes and natural Pb itself. Some of the benefits obtained from the conceptual design reactor led to the need to prepare a Modified CANDLE initial core with easily available material. In this study, a neutronic study will be carried out on the initial core of Modified CANDLE conceptual design. Calculations were carried out using SRAC 2006. The PIJ module was used to calculate the fuel pin burnup while the CITATION module was used to calculate the multigroup diffusion of the reactor core. The nuclear library data used is JENDL 4.0.

Keywords: Modified CANDLE, Initial core, Pb208-Bi Eutectic, Natural uranium

PermaLink | Plain Format | Corresponding Author (Nina Widiawati)

15 Advanced Fission Systems ABS-47

Preliminary Neutronic Analysis of Small MSFR with Th-U Fuel
Abdul Waris1*, Cici Wulandari2 , Robi Dany Riupassa2, Dwi Irwanto1 , and Asril Pramutadi AM1

1Department of Physics and Department of Nuclear Science & Engineering,
Faculty of Mathematics and Natural Sciences,
Institut Teknologi Bandung,INDONESIA
2Department of Physics, Faculty of Mathematics and Natural Sciences,
Institut Teknologi Bandung,INDONESIA


Molten Salt Reactor (MSR) is one of the six Generation IV of nuclear reactor systems. This reactor system MSR has many merits, such as better safety features, high thermal efficiency, and capability for waste. It was believed that the Generation IV Small Modular Reactors (SMR) are very promising for electricity generation in the middle part and the east part of Indonesia. In this study, the neutronic analysis of the Small Molten Salt Fast Reactor (MSFR) with the fuel composition is LiF-BeF2-ThF4-UF4 has been conducted. Neutronic calculations were performed by employing PIJ and CITATION modules of SRAC 2006 code with JENDL 4.0 nuclear data library.

Keywords: Small Modular Reactor, MSFR, CITATION, SRAC, JENDL 4.0

PermaLink | Plain Format | Corresponding Author (Abdul Waris)

16 Advanced Fission Systems ABS-48

Neutronic Analysis of HTTR 10 MWth with (Th, U) Fuel and CO2 Coolant
Abdul Waris1*, Fauzan G. Anshari2, Anni Nuril Hidayati2, Zaki Su’ud1

1Department of Physics and Department of Nuclear Science & Engineering,
Faculty of Mathematics and Natural Sciences,
Institut Teknologi Bandung,INDONESIA
2Department of Physics, Faculty of Mathematics and Natural Sciences,
Institut Teknologi Bandung,INDONESIA


Original HTTR (high temperature test reactor) is a 30 MWth HTGR (High Temperature Gas Reactor) with a graphite moderator, helium gas coolant, UO2 fuel with outlet coolant temperature of about 900oC. In this study, we have performed the neutronic analysis of 10 MWth HTTR with (Th, U) fuel and CO2 gas coolant. The burnup period is 1100 days, which corresponds to 3 years of fuel cycle length. The reactor calculation was performed by employing PIJ and CITATION modules of SRAC 2006 code, with the nuclear data library was derived from JENDL4.0. Several results on neutronics aspects will be presented in full paper and conference presentation.

Keywords: HTTR, Thorium, Uranium, CO2, SRAC, JENDL 4.0

PermaLink | Plain Format | Corresponding Author (Abdul Waris)

17 Advanced Fission Systems ABS-53

Investigation of the use of different Cladding Materials on the Neutronic and Thermal-hydraulic Parameters of Tehran Research Reactor
Farhad Salari, Mohammad Reza Nematollahi *, Mohsen Ebrahimian

Department of Nuclear Engineering, School of Mechanical Engineering, Shiraz University, 71936-16536 Shiraz, Iran


In this study, the effect of cladding material on the neutronic and thermal-hydraulic parameters of the Tehran research reactor (TRR) is investigated. First, calculations are done for the first core configuration of TRR with aluminum fuel cladding. Then, using stainless steel 316, zircaloy 4, SiC, and the alloy composition of FeCrAl for the cladding calculations are performed. The considered neutronic parameters are multiplication factor, average power distribution, thermal and fast neutron flux distribution, power peaking factor, excess reactivity, core shutdown margin, safety reactivity factor and temperature reactivity coefficient have been calculated and analyzed. The effect of cladding material on thermal-hydraulic parameters such as the temperature distribution of different parts of the cell and coolant are also analyzed. In view of the results, SiC cladding can be introduced as a more appropriate choice for this reactor compared with other materials.

Keywords: Cladding, Neutronic Parameters, Thermal-hydraulic Parameters, Tehran Research Reactor

PermaLink | Plain Format | Corresponding Author (Mohammadreza Nematollahi)

18 Advanced Fission Systems ABS-54

Morphology of Y-Ti Nano-oxides in ODS Alloys Irradiated with High Energy Heavy Ions
J.H. OConnell (a*), V.A. Skuratov (b), A.S. Sohatsky (b), K. Kornieieva (b), A.D. Volkov (c), M. Zdorovets (d)

a) CHRTEM, NMU, Port Elizabeth, South Africa
b) FLNR, JINR, Dubna, Russia
c) JSC VNIINM, Moscow, Russia
d) Institute of Nuclear Physics, Nur-Sultan, Kazakhstan


Oxide dispersion strengthened ferritic martensitic steels (ODS) are considered as candidates for fuel claddings for Gen IV nuclear reactors. The radiation tolerance of ODS steels is considered to be due to trapping of lattice defects and helium atoms by oxide particles and fine grain boundaries. When used as fuel cladding, these materials will be in close proximity to fissile fuel and exposed to fission fragment irradiation. Recent experiments demonstrated that heavy ions of fission fragment energy may induce amorphous latent tracks in Y-Ti oxides and at present there is no experimental data demonstrating that amorphized nanoparticles in ODS materials will assure the same properties and the same excellent radiation resistance as observed for steels containing crystalline nanoparticles. A lot of data related to swift heavy ion induced changes in morphology and properties of metal and semiconductor nanoparticles in oxide matrices are known from the literature. However, almost nothing is known about oxide nanoparticles in metallic matrices irradiated with high energy heavy ions.
The aim of this report is to summarize recent experimental results on the morphology of swift (167 and 220 MeV) Xe ion induced latent tracks in Y2Ti2O7 nanoparticles within ODS alloys during post-irradiation heat treatment and after irradiation at different temperatures.

Keywords: ODS, fission fragments, cladding

PermaLink | Plain Format | Corresponding Author (Jacques O Connell)

19 Advanced Fission Systems ABS-56

Felipe M. G. Pereira, Renato V. A. Marques, Márcia S. Santos, Carlos. E. Velasquez and Claubia Pereira

Departamento de Engenharia Nuclear - Universidade Federal de Minas Gerais
Av. Antonio Carlos, 6627 campus UFMG
31.270-901, Belo Horizonte, MG


Several different nuclear codes have been used to perform depletion and criticality calculations, already widespread among worldwide researchers. The neutron transport and depletion codes have their particularities such as the number of energy groups and multigroup cross section data included for each code. Therefore, this work aims to validate the model and cross sections data generated at DEN/UFMG using NJOY99 adopting a thorium fuel pin benchmark performed by MIT, INEEL and Czech Technical University, and using different computational nuclear codes. The validation consists in comparing results from codes and reference using benchmark methodology in criticality and depletion situations. To perform criticality at steady state and depletion calculations are used MCNPX, MCNP5, SERPENT, SCALE6.0, and MONTEBURNS. Besides that, an extension of the benchmark calculations is performed and nuclear reactor safety parameters are calculated for developed model. In this work are evaluated quantities such as the effective delayed neutron fraction, fuel temperatures coefficients and production and transmutation rates for each code considering fresh fuel and depletion situations. It is achieved effective delayed neutron fractions that decreased responding to changes in fuel composition and infinite multiplication factors that began simulation with lower differences than the ones obtained at burnup end, both results are a reflection of production and transmutation rates considered by each code.

Keywords: Thorium;Nuclear codes;Validation;Criticality calculation;Cross sections data;Depletion;Infinite multiplication factor; Effective delayed neutron fraction

PermaLink | Plain Format | Corresponding Author (Claubia Pereira)

20 Advanced Fission Systems ABS-59

A. A. P. Macedo, M. Gilbert, A. L. Vieira, A. A. Cunha, G. H. P. Dias, M. C. Ramos, M. E. Scari, F. C. Silva, P. A. L. Reis, C. A. M. Silva, A. L. Costa, M. A. F. Veloso, C.E. Velasquez and C. Pereira

Departamento de Engenharia Nuclear
Universidade Federal de Minas Gerais
Av. Antônio Carlos, 6627, Campus UFMG PAC 1 – Anexo Engenharia, Pampulha,
31270-901, Belo Horizonte, MG, Brasil


There is growing understanding that nuclear power plants are needed to complement intermittent energy sources and to decrease the emissions of carbon dioxide. However, the next generation reactors will need to incorporate innovative solutions regarding the radioactive wastes, safety improvements, proliferation-resistance, sustainability, efficiency, and cost. In the last decades the nuclear power industry has been developing and improving reactor technology as the projects of the next generation of nuclear power reactors. The advanced reactors concepts offer significant potential benefits because they can provide reliable, safe, clean energy as part of a mix with current nuclear technology. In this scope, the Nuclear Engineering Department of the Universidade Federal de Minas Gerais - Brazil (DEN-UFMG) has studied advanced nuclear systems, in the last years. This paper presents the results of advanced nuclear reactors concepts as the Very High Temperature Reactor (VHTR), Molten Salt Reactor (MSR), Sodium Fast Reactor (SFR), Gas-cooled Fast Reactor (GFR) and Advanced CANDU Reactor (ACR). These studies evaluate the nuclear system at steady state and during burnup using models developed at DEN-UFMG. Moreover, studies about the thermal hydraulic behavior of advanced nuclear systems as the HTTR (High Temperature Engineering Test Reactor), HTR-10 (High Temperature Test Module Reactor) e LS-VHTR (Liquid-Salt-Cooled Very-High-Temperature Reactor), developed at the DEN-UFMG, are also presented. The nuclear code systems such as SCALE 6.0 (KENO-VI/ORIGENS), MCNPX 2.6.0, MCNP5, RELAP5-3D, WIMS-D and others, have been used in the investigations, mainly related to the safe operation. In this way, the aim of this paper is to present an overview about the research activities developed by the Department of Nuclear Energy at the Universidade Federal de Minas Gerais on Advanced Nuclear Systems.

Keywords: advanced nuclear reactor; VHTR; MSR; SFR; GFR

PermaLink | Plain Format | Corresponding Author (Claubia Pereira)

21 Advanced Fission Systems ABS-67

Neutronic Analysis of Sodium-Cooled Fast Reactor (SFR) Design with Various Fuel Types Using Shuffling Strategy
Mohammad Ali Shafii(a*), Revina Septi (a), Feriska Handayani (a), Artoto Arkundato (b), Zaki Su’ud (c)

a) Department of Physics, Andalas University Padang Indonesia
b) Department of Physics, Jember University, Jember Indonesia
c) Nuclear and Biophysics Laboratory, Bandung Institute of Technology Bandung Indonesia


Neutronic analysis of Sodium-Cooled Fast Reactor (SFR) design with variations of fuel types using radial fuel suffling strategy has been investigated. One type of generation IV reactor that currently being researched for commercial implementation is SFR. In this research, the SFR design utilizes natural uranium as a fuel input. The reactor core is designed in the form of two-dimensional cylindrical geometry for various type of fuel such as MOX, UN-PuN, and U-Zr. Radial suffling strategy in the direction of R- Z axis is applied to SFR to manage the fuel burn up for long life reactor with natural circulation as a fuel cycle input. The burn up process is follows the fuel region movement scheme. The designed reactor core is divided into 10 regions, representing the reactor is operated for 10 year without refueling, where the volume in each region is made equal to one another. At beginning, the first region of reactor core is filled with natural uranium fuel as a input and it is called by first fuel cycle. The scheme just needs natural uranium as a fuel cycle input every beginning of 10 year of cycle. Furthermore, the fuel movement scheme is carried out for several types of fuel. The global neutronic parameters such as multiplication factor (keff) and burn up analysis are observed and optimized. Overall, in the output power of 550 MWt, the results indicate that U-Zr is the most optimal fuel to be applied and a greater chance of being operated for SFR.

Keywords: SFR, fuel type, multiplication factor, burn up, shuffling strategy

PermaLink | Plain Format | Corresponding Author (Mohammad Ali Shafii)

22 Advanced Fission Systems ABS-68

Simulation of different density mixing pebble flow in a two-dimension circulating packed bed system
Dwi Irwanto, Sparisoma Viridi

Department of Physics, Institut Teknologi Bandung, Bandung 40132, Indonesia


Using molecular dynamics method simulation of pebble flow in a two-dimension circulating packed bed system, where the bed are mixture of three different densities, is reported in this work. Basis of the system is real pebble bed reactor HTR-10, which is simplified to a two-dimension simulation system from its real three-dimesion simulation sytem (Wu et al., 2019). Two types of force are considered in this work. The first is normal contact force in the form of linear spring-dashpot (Schaefer et al., 1996) and the second is earth gravitation force. Friction force is neglected for simplicty, where it can induce rotation of the spherical fuel elements. Each type of element with different density enter the system in different radial position. Variations of density area performed in order to observe their influence to element radial positiion as it circulated in to (at the top) and out from (at the bottom) the system. It is observed that there is a weak relation between element density and its stable radial position.

Keywords: molecular dynamics, HTGR, circulating bed, pebble flow

PermaLink | Plain Format | Corresponding Author (Sparisoma Viridi)

23 Advanced Fission Systems ABS-72

The conceptual design of thorium-based molten salt energy amplifier
Yangpu(a,b), Wan Weishi(c), Yu Xiaohan(a,b), Cai Xiangzhou(a,b), Dai Zhimin(a,b), Lin Zuokang(a,b*)

a) Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800, China
b) CAS Center for Innovation in Advanced Nuclear Energy, Shanghai 201800, China
c) ShanghaiTech University, Shanghai 201210, China


From the previous research, we know that thorium-uranium fuel cycle system, i.e. U233 for fissile fuel and Th232 for breeding material, is hardly to achieve self-sustaining state in the moderated molten salt reactor without the complex online post-processing system. In our research, we use a proton accelerator to drive a thorium-based fast neutron molten salt subcritical reactor that improves the neutron efficiency in the system. The research results show that the molten salt energy amplifier driven by the proton accelerator we designed can achieve a long-term stable state, more than 10 years, under a rated power and a stabilizing k value without any online post-processing system and online replenishment of fuel. A physical design of the most simplified single loop molten salt energy amplifier was accomplished. Through the burnup calculation, a rated power 300 MWth molten salt energy amplifier will continue to run for 30 years without any online processing but inputting a 1 GeV proton beam within 4 mA during the whole operation period. And the temperature coefficient of the molten salt reactor is totally negative in the whole period.

Keywords: energy amplifier, molten salt, self-sustaining, proton accelerator

PermaLink | Plain Format | Corresponding Author (Zuokang Lin)

24 Advanced Fission Systems ABS-74

Analysis of integrated target in the thorium-based molten salt energy amplifier
Yangpu(a,b), Wan Weishi(c), Yu Xiaohan(a,b), Cai Xiangzhou(a,b), Dai Zhimin(a,b), Lin Zuokang(a,b*)

a) Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800, China
b) CAS Center for Innovation in Advanced Nuclear Energy, Shanghai 201800, China
c) ShanghaiTech University, Shanghai 201210, China


Conventional neutron targets in Accelerator Driven Sub-critical System (ADS) are based on the design using heavy metal like lead. Considering high density of energy deposition on the target, it generally needs a separated loop for the target cooling. In the thorium based molten salt energy amplifier, an integrated target is conceived. Based on liquid fuel properties of the core and neutron productivity of the molten salt fuel by bombarding with the proton, the proton beam is direct led in the molten salt core without a separated single loop for the neutron target. The analysis of the design is presented, including the neutron productivity calculation and the thermal hydraulics simulation. Besides, evaluation of different layouts of the proton beam introducer in the core is presented to compare the influence of the neutron utilization efficiency in the system, depend on which we can choose different types of the beam introducer, with or without a beam window.

Keywords: integrated target, energy amplifier, molten salt fuel, proton beam introducer

PermaLink | Plain Format | Corresponding Author (Zuokang Lin)

25 Advanced Fission Systems ABS-75

Sumer ŞAHİN (a), Hacı Mehmet ŞAHİN (b*), Özgür Erol (c)

(a) Bahcesehir University, Departmen of Energy System Engineering, İstanbul, TURKİYE

(b) University of Karabük, Departmen of Energy System Engineering, Karabük, TURKİYE

(c) Baskent University, Departmen of Mechanical Engineering, Ankara, TURKİYE


In this study, Gas Turbine - Modular Helium Reactor (GT-MHR), one of the new generation reactors has been investigated because it has many advantages. Very high efficiencies can be achieved due to the gas coolant, and additionally alternative fuels can be utilized in this reactor. It has also lower waste quantity and higher safety margins. Moreover, one of the most important characteristics that possibility of usage of weapons-grade plutonium (WGrPu), reactor grade plutonium (RGrPu) and minor actinide (MA) with fertile fuel types (natural uranium and thorium) is a good option to the efficient usage as an alternative fuel mixture in a GT-MHR.
In this purpose, utilization of natural uranium (nat-U) and thorium as fertile fuels has been analyzed using alternative fuels (WGrPu, RGrPu and MA) as driver fuel. Then, possibility of utilization of the alternative fuels/fertile fuels mixture was investigated and an optimum mixture ratio was determined. Therefore, a neutronic analysis for the full core reactor was performed by using MCNP5 with ENDF/B-VI cross-section library. Different mixture ratios were tested in order to find the appropriate mixture ratio of fertile and fissile fuel particles that gives a comparable keff value of the reference uranium fuel. Time dependent calculations were performed by using MONTEBURN2.0 with ORIGEN2.2 for each selected mixture. Calculations showed that, a GT-MHR type reactor, which is using the original TRISO fuel particle mixture of 20% enriched uranium + natural uranium (original fuel) has an effective multiplication factor (keff) of 1.27. Corresponding to this keff value the alternative fuels/fertile fuels mixture was found as ratio of percent. In the final analysis, different parameters (operation time, burnup value, fissile isotope change, etc.) were subject of performance comparison.

Keywords: Thorium, Alternative Fuels, GT-MHR, MCNP

PermaLink | Plain Format | Corresponding Author (Haci Sahin)

26 Advanced Fission Systems ABS-91

Modified CANDLE Analsysis Using Microscopic Cross Section From SLAROM Code for Detail Analysis of Multi Scenario Modified CANDLE Scheme
Zaki Su’ud, Fitria Miftasani, Feriska H. Irka, Nina Widiawati, Helen Raflis, H Sekimoto, Sumer Sahin, Mehmed Sahin, Zuhair

1Nuclear and Biophysics Research Divisions, Bandung Institute of Technology
2Emeritus Professor, Tokyo Institute of Technology, Japan
3Bachcisehir University, Turkey
4National Nuclear Energy Agency ( BATAN), Indonesia
Email: zakisuud[at]


: Implementation of Modified CANDLE burnup scheme based on microscopic cross section from SLAROM code has been implemented using two dan three dimensional analysis including iterative multi-group diffusion and burn-up analysis. In the previous calculation based on SRAC, the burn-up analysis generally performed for each region or sub region bases, while in this study the burn-up calculation is performed for each individual mesh. Therefore more flexible model of Modified CANDLE can be implemented including pure axial shuffling, pure radial shuffling, and also various combination of axial-radial shuffling can be implemented with or without assumption of special adjustment process in the pin level to minimized burn-up peaking. The general simulation scheme, initially microscopic cross section is generated by SLAROM code system to generate sets of microscopic crosss sections. Then multigroup diffusion calculation is performed and then continued by burnup analysis for every 10 years of period. After 10 years of burnup the fuel material in the core are shifted according to the detail model of Modified Burn-up scheme (axial shuffling, radial shuffling or combined axial-radial shuffling). Some calculation results show that in general for the same conditions, the calculation using this system agreed with those in the previous model. More detail results will be discussed in the conference including comparison of CANDLE, precious Modified CANCLE model and current Modified CANDLE models.

Keywords: Modified CANDLE, Microscopic cross section, iterative scheme, shuffling strategy

PermaLink | Plain Format | Corresponding Author (Zaki Suud)

27 Advanced Fission Systems ABS-92

Preliminary Study of Nuclear Safety System Analysis and Simulation for Molten Salt Reactor
Muhammad Ilham (a), Cici Wulandari (a), Putranto Ilham Yazid (a), Sidik Permana (a)

a) Nuclear Physics and Biophysics Research Division, Physics Department,
Faculty of Mathematics and Natural Science, Bandung Institute of Technology
Jalan Ganesha 10, Bandung 40132, INDONESIA


The development of technology and Molten Salt Reactor (MSR) research in the world has increased in the 2000s. For safety factor, MSR operates at low pressure which reducing the risk of pipe rupture or leak when an accident occurs. If the temperature increase over the design limit due to accident, there is an emergency cooling dump underneath the core as a safety where all the melting fuel will fall into the dump. MSR has large negative feedback reactivity when there is an expansion in fuel volume due to the temperature exceeding the design limits. In this study, point kinetic modeling is performed using reactor transient conditions where the fuel salt is circulating in core and loop. An analysis was conducted on the FUJI-12 type when there was a positive reactivity added in to evaluate the safety system. The calculation is verified and give a similar trend result with the previous study. The results obtained will serve as the minimum limits as the benchmark to design the MSR safety system.

Keywords: MSR, Point kinetic, safety system, simulation

PermaLink | Plain Format | Corresponding Author (Muhammad Ilham)

28 Advanced Fission Systems ABS-93

Design Development of PeLUIt-40 a Small Cogeneration Nuclear Power Plant
Topan Setiadipura(a*), Dwi Irwanto(b), Hery Adrial(a), Suwoto(a), Zuhair(a)

(a)Centre for Nuclear Reactor Technology and Safety – National Nuclear Energy Agency, Puspiptek Area Building No. 80, Serpong, South Tangerang, Indonesia 15310
(b)2Nuclear Physics and Biophysics Research Division, Physics Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology Indonesia, Jalan Ganesha 10 Bandung 40132, INDONESIA


Economic performance is among important aspects for the success of a nuclear reactor design to be applied. A follow-up design based on Reaktor Daya Eksperimental (RDE) is developed in this study from the initial 10MWt power of RDE, current design will be upscaled to class of 30MWt without any update on the fuel and core geometry. Current design is called PeLUIt-30. This triplet power improvement hopely will increase the economic feasibility of the small modular pebble bed reactor. The important parmeter as the constrain of the design upscaling is the maximum temperature of the fuel in equilibrium condition also in the depressurized loss of forced-cooling (DLOFC) which assumed to be the severest accident hyptothetical scenario. PEBBED code is utilized for the equilibrium core analysis including the neutronic and thermal hydraulic module. In current study the material composition of the pebble fuel is maintained to assure the direct practical application of this design as the commercial follow-up of the RDE. Results of this study show that improving the power to 40MWt with a discharge burnup of 80 MWd/Kg-HM will have a maximum fuel temperature at equilibrium and DLOFC of 1037.4°C and 1562°C, respectively. These results show that PeLUIt-40 able to maintained the sound passive safety of the RDE while improving its commercial feasibility with higher power output.

Keywords: pebble bed reactor, high temperature gas-cooled reactor, cogeneration

PermaLink | Plain Format | Corresponding Author (Topan Setiadipura)

29 Advanced Technology and Other Issues ABS-5

preliminary calculation on the containment external cooling effect for FLEX strategy using containment analysis codes

KHNP Central Research Insitute


The external containment cooling strategy is involved in the FLEX Support Guideline (FSG). According to this document, the external containment cooling strategy will be most effective if the steel containmnet vessel itself can be sprayed with cool water. additionally, this cooling strategy should be evaluated for plant-specific containment building design. In case of Korean nuclear power plant, the material of containment building is pre-stressed concrete. Therefore, it should be checked that the external cooling strategy which is specified in FSG has an effect on the depressurazation of containment building. In this paper, the containment external cooling effect was anlyzed using GOTHIC and CAP code. In order to invertigate the influence of external cooling, an Extended Loss of All AC Power (ELAP) condition which is one of the entry condition for FSG-12 was applied. Additionally, the maximum RCP leakage was also assumed. As a calculation results, it was showed that the external spray cooling effect using portable pump have little effect on depressurization of containment building after 48 hours. And, the containment pressure is low sufficiently to maintain the containment integrity and implement other mitigation strategies.

Keywords: Containment, FSG, External Cooling, GOTHIC, CAP

PermaLink | Plain Format | Corresponding Author (KYUNGHO NAM)

30 Advanced Technology and Other Issues ABS-6

Jose Rubens Maiorino

Federal University of ABC


Just after the discovery of thorium by Jöns Borzelius, a Sweden in 1829., the exploitation of thorium from monazite sand in Brazil date back to 1886, when Englishman John Gordon began exporting to Europe the ore mined in the municipality of Prado, Bahia State to use in lighting (incandescent gas lamps). In the late of 19th and early of 20th century, the interest in monazite increased owing to the use of thorium nitrate by gas mantle industries. Later, the use of lanthanide elements turned monazite into a much more important commodity than it was in pre-war years. The commercial exploitation of monazite sand starts in 1948, by a private company called ORQUINA in São Paulo city, and later the production and purification of thorium compounds was carried out at IPEN, a Research Institute in Sao Paulo, for about 18 years. The raw materials used were some thorium concentrates obtained from the industrialization of monazite sands, a process carried out in Sao Paulo between 1948 and 1994 on an industrial scale by the company ORQUIMA, later by a state company NUCLEMON (acronym for Nuclear Monazite), which operates up today.
The first national program to use thorium was conducted during the 60’ by a research group from a Brazilian State, Minas Gerais, very rich in mineral resources, including thorium. This research group was called the “Thorium Group”, and in the framework of a cooperation agreement with the French CEA aimed at the development of a thorium fueled PHWR with a concept of a pre stressed concrete reactor vessel. Also, in the beginning of the seventies, in the frame work of a cooperation agreement of IPEN, in São Paulo, with the USA General Atomic (GA), several activities, theoretical and experimental, were developed on thorium technology and utilization mainly for the HTGR concept. However, it was in the framework of the Brazilian German agreement that the biggest R&D program on thorium utilization was developed with the incentive of “International Nuclear Fuel Cycle Evaluation”. This program was conducted by the Brazilian Center for Development of Nuclear Technology (CDTN), in that time the R&D branch of the former holding, NUCLEBRAS, and the Germans KFA- Jûlich, Siemens A.G-KWU, and NUKEN.).
This paper, besides the historical overview, will discuss the natural resources of thorium in Brazil, the technological capability to produce nuclear thorium, and the main results obtained in the previous national programs, as well as, the academic researchers on going at the Brazilian Universities. As main conclusion, given the huge reserves available in Brazil, the government should support researchers on thorium to be on line with this important energetic in the international community.

Keywords: Thorium, Monazite, Nuclear Reactors, Brazil

PermaLink | Plain Format | Corresponding Author (Jose Rubens Maiorino)

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