Activities in Divertor Reflector and Linear Plates Using WCLL and HCPB Breeding Blanket Concepts

In fusion devices, such as European Demonstration Fusion Power Reactor (EU DEMO), primary neutrons can cause material activation due to the interaction between the source particles and the targeting material. Subsequently, the reactor’s inner components become activated. For safety and safe performance purposes, it is necessary to evaluate neutron-induced activities. Activities results from divertor reflector and liner plates are presented in this work. The purpose of liner shielding plates is to protect the vacuum vessel and magnet coils from neutrons. As for reflector plates, the function is to shield the cooling components under plasma-facing components from alpha particles, thermal effects, and impurities. Plates are made of Eurofer with a 3 mm layer of tungsten, while the water is used for cooling purposes. The calculations were performed using two EU DEMO MCNP (Monte Carlo N-Particles) models with different breeding blanket configurations: helium-cooled pebble bed (HCPB) and water-cooled lithium lead (WCLL). The TENDL–2017 nuclear data library has been used for activation reactions cross-sections and nuclear reactions. Activation calculations were performed using the FISPACT-II code at the end of irradiation for cooling times of 0 s–1000 years. Radionuclide analysis of divertor liner and reflector plates is also presented in this paper. The main radionuclides, with at least 1% contribution to the total value of activation characteristics, were identified for the previously mentioned cooling times.

Neutronic Analysis of The SMART Modular Reactor Fuel Using SRAC 2006

Neutronic analysis of The SMART modular reactor fuel using SRAC 2006 has been carried out. Electrical energy is important today because the need is increasing along with the increase in human population, advanced technology and the economy. On the other hand, there are demands from the community for the clean, efficient and consistent energy. This is the reason why nuclear power plants are considered as one of the candidates for electrical energy suppliers in Indonesia in particular. This study evaluates a SMART reactor with Gadolinium as the burnable absorber material. The two kinds of fuel assembly were analyzed using the SRAC 2006 code system with the JENDL 4.0 as nuclear data library. This study aims to observe the neutronic characteristics of the fuel assembly designs according to the reference used. The results of the study show that of all types of fuel assemblies used can reach criticality at the beginning of the operating cycle and last up to 3 till 5 years when it finally reaches subcritical condition. Another parameter observed is the conversion ratio value, which from this study is in accordance with the characteristics of the conversion ratio for thermal reactors.

Preliminary Neutronic Analysis of 150 MWt High-Temperature Gas Reactor to Produced Electricity in Small Area

There are small areas in Indonesia with insufficient electricity. High-Temperature Gas Reactor (HTGR) is a promising nuclear power plant that can be used in such areas as its capability to produce electricity and co-generation applications. A preliminary study on the neutronic aspect of the 150 MWt HTGR design is performed in this research. High Temperature Engineering Test Reactor (HTTR) is used as a basic model. The calculation was performed by Standard Thermal Reactor Analysis Code (SRAC) code, and Japanese Evaluated Nuclear Data Library (JENDL) 4.0 as nuclear data library. As a result, by increasing HTTR fuel assembly geometry to 1.5 times its original and using higher uranium enrichment, the reactor can be operated for five years.

Neutronics performances of gas-cooled fast reactor for 300-600 MWt Output Power with Modified CANDLE burn-up scheme in radial direction

Gas-Cooled Fast Reactor-GFR is a Generation IV reactor that is helium-cooled and has a closed fuel cycle. Due to the target operation on 2022-2030, this reactor type still needs further research and development technologies. We investigated the neutronics performances of a GFR balance type core with some modification of CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy production) burn-up scheme in the radial direction. The output power varied from 300 to 600 MWt. The neutronics calculation was performed using SRAC 2002 with JENDL 4.0 nuclear data library. The analysis indicate the reactor could operate critically for ten years without refueling with burn-up level 20% HM.

Impact of Different Nuclear Data Library on Control Rod Reactivity Worth Calculation of Small Pebble Bed Reactor

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.

Nuclear Data Sensitivity Study for the EBR-II Fast Reactor Benchmark Using SCALE with ENDF/B-VII.1 and ENDF/B-VIII.0

The EBR-II benchmark, which was recently included in the International Handbook of Evaluated Reactor Physics Benchmark Experiments, served as a basis for assessing the performance of the SCALE code system for fast reactor analyses. A reference SCALE model was developed based on the benchmark specifications. Great agreement was observed between the eigenvalue calculated with this SCALE model and the benchmark eigenvalue. To identify potential gaps and uncertainties of nuclear data for the simulation of various quantities of interest in fast spectrum systems, sensitivity and uncertainty analyses were performed for the eigenvalue, reactivity effects, and the radial power profile of EBR-II using the two most recent ENDF/B nuclear data library releases. While the nominal results are consistent between the calculations with the different libraries, the uncertainties due to nuclear data vary significantly. The major driver of observed uncertainties is the uncertainty of the 235U (n,γ) reaction. Since the uncertainty of this reaction is significantly reduced in the ENDF/B-VIII.0 library compared to ENDF/B-VII.1, the obtained output uncertainties tend to be smaller in ENDF/B-VIII.0 calculations, although the decrease is partially compensated by increased uncertainties in 235U fission and ν¯.

Generation of Proton- and Alpha-Induced Nuclear Cross-Section Data via Random Forest Algorithm: Production of Radionuclide 111In

We investigated the generation of proton- and alpha-induced nuclear cross-section data in the production of Indium-111 (111In) for application in nuclear medicine. Here, we are interested in three reaction channels, which are 109Ag (α, 2n), 111Cd (p, n) and 112Cd (p, 2n), in the production of 111In. A random forest algorithm was used to generate nuclear cross-section data by using an experimental nuclear cross-section from the Experimental Nuclear Reaction Data (EXFOR) database as input. Hence, reasonably accurate regression curves of nuclear cross-section data could be produced with the evaluated nuclear data library ENDF/B-VII.0 set as the benchmark.