A prototype of a high temperature gas-cooled reactor for electric and cogeneration application, called Reaktor Daya Eksperimental (RDE) is being developed at BATAN. RDE is developed to fulfill the ...huge yet distributed Indonesian energy demand, in particular for the eastern part of Indonesia. As a moving fuel core, analysis of the fuel management of PBR can be divided into the analysis of the start-up phase where the core composition, neutron flux, and power density profile still changing and equilibrium core. Equilibrium core design usually represent the general performance over the entire range of reactor operation. As part of the design development, current paper presents the equilibrium core design of RDE and depressurized-loss-of-forced-cooling (DLOFC) accident analysis. PEBBED code is utilized for the equilibrium core analysis. Results of this study include several possible equilibrium designs for RDE which then the design with 123 fuel pebble recirculated per day with equilibrium keff 1.02571, discharge burnup of 79.36 and maximum power generation 0.66kW per pebble fuel is chosen. Maximum fuel temperature of this design under the DLOFC accident is 1015.9°C which occurred 8.5h after the initiation of the accident. Results of DLOFC accident shows the strong passive safety features of RDE.
Abstract
Several studies related to simplifying the modeling of pebble bed High-Temperature Reactor core (HTR) has been developed before. From some calculation on several MCNP models with a fueled ...pebble to dummy ratio 57:43, using a combination of several types of TRISO (TRi-structural ISOtropic particle fuel) unit and Pebble unit is modeled to achieve its first criticality. In this paper, some MCNP model that uses 27000 pebbles with a 57:43 ratio and 100% fueled pebble is created to be used on burnup calculation and to compare its k-eff and nuclide inventory. From this burnup calculation, it could be seen that SC (Simple Cubic) TRISO unit has faster calculation time followed by the HCP (Hexagonal Close Packed) TRISO unit and then the FCC (Face-Centered Cubic) TRISO unit. The BCC (Body-Centered Cubic) pebble unit had some consistent deviation from another pebble unit, and it still needs more study to know more about the reason behind it. It could be seen that if there are some dummy pebbles inside the reactor, then the deviation would be higher than if there is just fueled pebble inside the reactor. On the 57:43 ratio, the absolute average deviation of k-eff on burnup calculation is lower than 2% and 10% for nuclide inventory (mass). On 100% fueled pebble, it’s below 0.15% on k-eff absolute deviation and below 8% on nuclide inventory deviation.
Specific energy demand of Indonesia with its distributed energy demand required a special energy supply solution. The energy supply solution should be able to fulfil the electric demand of huge and ...growing population which spread in many different islands also to fulfil the energy demand to process abundances yet distributed natural sources. This paper proposes a pebble bed high temperature reactor as an energy supply in the form electric and/or cogeneration application. A parametric survey of the core geometry of a 150MWt Pebble Bed Reactor called PeLUIt is performed, follow by an equilibrium analysis of the optimized design. An accident analysis of depressurized loss of forced cooling (DLOFC) is performed to show the strong passive safety features of the design. Finally the heat-mass balance of the plant is also presented.
In experimental power reactor (reaktor daya eksperimental, RDE), a full MOX fueled core can be accommodated in its operation without significant modifications on the reactor core design, but the MOX ...core tends to produce less favorable safety features and deliver transient behavior into unwanted fatal accidents. This paper aimed to investigate the MOX core characteristics of RDE through a series of calculations with MCNP6 code and ENDF/B-VII library. The calculation results show that MOX core with 235U enrichment of above 5 % can reach the criticality condition, maintain and run the reactor during the operation cycle. Utilizing MOX fuel in RDE with lower 235U enrichment will have a more negative impact on temperature coefficient of reactivity. The lower βeff makes the MOX core is more difficult to control, especially with low 235U enrichment. These results conclude that the selection of 235U enrichment in the MOX core must be carefully considered because it is one of the strategies to ensure the reactor safety criteria are fulfilled.
Abstract
One of the main critical issues on a nuclear reactor is safety and control system. The control rod worth plays an important role in the safety and control of nuclear reactors. The control ...rods worth calculation is used to specify the safety margin of the reactor. The main objective of this work is to investigate impact of different nuclear data libraries on calculating the control rod reactivity worth on small pebble bed reactor. Calculation of the control rod reactivity worth in small high temperature gas cooled reactor has been conducted using the Monte Carlo N-Particle 6 (MCNP6) code coupled with a different nuclear data library. Famous evaluated nuclear data libraries such as JENDL-40u, ENDF/B-VII.1 and JEFF-3.2 continuous cross section-energy data libraries were used. The overall calculation results of integral control rod worth show that the ENDF/B-VII.1, JENDL-40u and JEFF-3.2 files give values of - 17.814%☐k/k, -18.0204 %☐k/k and -18.0267%☐k/k, respectively. Calculations using ENDF/B-VII.1 give a slightly lower value than the others, while the JENDL-4.0u file gives results that are close to JEFF-3.2 file. The different nuclear data libraries have a relatively small impact on the control rod worth of small pebble bed reactor. Accurate prediction by simulation of control rod worth is very important for the safety operation of all reactor types, especially for new reactor designs.
A computer code for analyzing the safety feature of triso coated particles for high temperature reactor has been developed based on PANAMA code. This python based code were separated into several ...modules to perform the main TRIAC-BATAN and linear interpolation calculation, to read the input data file, and to control the sequence of all TRIAC-BATAN calculation. In this initial development phase, TRIAC-BATAN can produce similar pattern as PANAMA calculation in the case of failure fraction of triso particles, either for Depressurized Loss Of Forced Cooling (DLOFC), 100°C higher than DLOFC temperature and constant accident temperature at 1600°C. By using the euclidean distance, TRIAC-BATAN calculation and PANAMA for DLOFC condition are separated in around 3.23 ⋅ 10−7.
Reaktor Daya Eksperimental (RDE) is a high temperature gas-cooled reactor (HTGR) producing a 10 MW thermal with a pebble bed fuels which being developed by BATAN. The purpose of this paper is to ...study the burnup distribution and characteristic in the equilibrium core for different multipass recirculation method. Understanding the reactor physics, in particular the burnup, is important for optimum design and safety analysis of nuclear reactor. Related with the design approval phase of RDE, this study provides a design data which needed for the RDE's safety analysis report, e.g. fuel pebble performance analysis. Analysis in this study was performed using PEBBED diffusion code. PEBBED is designed to solve the neutronics and thermalhydraulics parameter cases for high temperature pebble bed reactors for different fuel recirculation including once-through-then-out (OTTO) and multipass scheme. The reactor core calculation was performed by applying multipass scheme with variation from 5 passes up to 15 passes. With energy and power as inputs, the calculations produce the burnup fraction at the end of the cycle and the fast and thermal neutron fluxes in 8 energy groups. The calculation results showed that the lowest 5 passes fuel recirculation pattern has the highest and lowest minimum discharged burnup value. This is related to the average power distribution in the core which means more passes will flatter the burnup distribution including the power distribution and also seen in the value of flux and power peaking factor produced. power peaking factor produced.
Simplification on modeling of high temperature reactor core (HTR) of pebble bed type has been widely developed before. The last calculation that writer develop on modeling and simplification for HTR ...modelling has been done on TRISO and single pebble, especially on it is KINF and burnup calculation. From that calculation, it is known that TRISO modeling that using fuel kernels and cover (homogenized buffer, PyC and SiC) gives K-INF values that are not (much) different than the complete TRISO modeling (fuel kernel, buffer, PyC and SiC on each different cell). Then the purpose of this study was to analyze the effect on KEFF and calculation time of combined TRISO particle arrangement modeling in the HTR fuel pebble and this pebble arrangement inside reactor core using MCNPX. The modeling variations that will be carried out include modeling of HTR fuel pebble using simplified TRISO (fuel kernel and homogenized TRISO coating layers) that uniformly dispersed inside pebble using HCP, SC, FCC and BCC lattice, that arranged inside reactor core (also) using HCP, SC, FCC and BCC lattice. For all variations mentioned above, K-EFF calculations has been done for each fuel height variation from 90-190 cm inside reactor core, with 50% fuel pebble to fuel-moderator pebble ratio, in 61% pebble volume fraction ratio except Simple Cubic (SC) lattice (max. 52.36%). Whole modelling gives K-EFF that have not much deviation except all modelling using SC pebble arrangement inside reactor core, because its low packing fraction. SC-type TRISO dispersion mode inside pebble always need more calculation time than other model. The modeling using HCP TRISO unit inside HCP pebble unit inside reactor core gives consistently short calculation time, same as almost all calculation using FCC TRISO unit inside pebble always have shortest calculation time compare to other TRISO unit. FCC and BCC pebble unit inside reactor core using less calculation time compare to HCP at most calculation.
The 10 MW Indonesia's Reaktor Daya Eksperimental (RDE) is a Pebble Bed High Temperature Reactor-type and to be constructed in Serpong Nuclear Zone, Puspiptek, Indonesia. The reactor applies helium ...gas coolant, graphite moderator, and low uranium enriched fuel (17%). A water-ingress accident may take place in the RDE core and such the accident is caused by a pipe break in the steam generator system of the reactor and hence the accident influences the reactivity of the core. To begin with, an initial core of the RDE should be modelled and the model contains the identification of material zones in the reactor core. To estimate the reactivity change due to the water-ingress accident in the initial RDE core, a void core condition from any coolant should be firstly deliberated. The core reactivity change is then estimated based on effective multiplication factor (keff) as a function of water density which comes into void cells among pebble containing helium coolant in the RDE core and that of the variation of 235U enrichment. The data library of thermal scattering S(α,β) as a function of temperature has been applied. The data base of thermal scattering for graphite was also utilized to comprehend the validity calculation using MCNPX. The ENDF/B-VII continuous energy nuclear data library was also smeared for all calculations at temperature of 900 0K. For the calculation using MCNPX, two options of KCODE and KSRC were taken into account to estimate keff of the RDE core. To achieve the accuracy of keff in the RDE reactor core, the amount of 20,000 neutrons per cycle for 150 non-active cycles and 650 active cycles was taken into account. The results showed that water-ingress accident in the RDE core is a dynamical behaviour which may result in reactivity core change of the RDE core. Finally, from all caclulated results taking into account the varietes of water density and 235U fuel enrichment, the RDE reactor core is totally safe in the event of water-ingress accident ensuing during the RDE reactor operation.
Abstract
This paper shows the effort to implement the RDE and its challenges from 2013 to 2018. RDE was a program to introduce nuclear power plants by building non-commercial power reactors. The RDE ...program was also used to prove that Indonesian engineers can design a reactor that will later supply electricity and steam for industry. The technology used is a high-temperature gas-cooled reactor. This RDE program is a very strategic intermediate target for energy security and national sovereignty. The development of RDE-based nuclear power plants, in the long run, is expected to have implications for reducing reliance on fossil fuels, more self-sufficiency in energy supply increases national industrial capacity and competitiveness in the global economic order, as well as enhance energy and political diplomacy. Also, RDE can be a reference installation for PeLUIt (Power and Steam Generators for Industry) power plants for small and medium enterprises to meet the demand for electricity and industrial heat in an area’s needs. However, many challenges occurred to implement this program; among them were cost estimation and cost-benefit analysis. Although the program has not been realized, mainly for financial reasons, many positive things have been obtained from these activities.
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